atw - International Journal for Nuclear Power | 06.2021

Ever since its first issue in 1956, the atw – International Journal for Nuclear Power has been a publisher of specialist articles, background reports, interviews and news about developments and trends from all important sectors of nuclear energy, nuclear technology and the energy industry. Internationally current and competent, the professional journal atw is a valuable source of information. www.nucmag.com

Ever since its first issue in 1956, the atw – International Journal for Nuclear Power has been a publisher of specialist articles, background reports, interviews and news about developments and trends from all important sectors of nuclear energy, nuclear technology and the energy industry. Internationally current and competent, the professional journal atw is a valuable source of information.


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nucmag.com<br />

2021<br />

6<br />

ISSN · 1431-5254<br />

32.50 €<br />

First Repatriation of<br />

Vitrified Reprocessing<br />

Waste from Sellafield<br />

Interview with<br />

Massimo Garribba<br />

Practical Response<br />

to a Dirty Bomb

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

<strong>Nuclear</strong> Energy: A Path to Reason<br />

3<br />

Dear Reader, in the year 10 after Fukushima and the political reaction in Germany, unique in the world at that time,<br />

with the immediate closure of 8 of the 17 nuclear power plants in operation at that time and less than a year be<strong>for</strong>e the<br />

last three plants are to be shut down by 31 December 2022 at the latest, the facts seem to be catching up with this policy<br />

and the mistakes of the national nuclear phase-out are clearly revealing themselves. The facts are not based on political<br />

decisions, political wishful thinking or visions, but simply on our natural living conditions. We heat or cool because<br />

nature only offers pleasant weather throughout the year in a few places in the world. We need “mechanical energy” to<br />

move ourselves, sometimes over long distances, because there are no alternatives to sensible transport, neither <strong>for</strong><br />

ourselves nor <strong>for</strong> the things we need every day or <strong>for</strong> what we want. We use a variety of energy sources and technologies<br />

because man as an energy source with an overall efficiency in the low single-digit range is the worst of all alternatives<br />

<strong>for</strong> energy supply – just ask yourself why battery-powered bicycles are currently booming when they originally relied on<br />

muscle power?<br />


Germany is currently confronted with two facts more than<br />

ever:<br />

1. Politics in Germany has set ambitious targets <strong>for</strong> the<br />

reduction of climate-impacting emissions, and is<br />

currently even aiming <strong>for</strong> so-called climate neutrality<br />

by 2050 at the latest or earlier. At the same time, greenhouse<br />

gas emissions in Germany have been significantly<br />

reduced since 1990. Total emissions converted into<br />

carbon dioxide equivalents (excluding carbon dioxide<br />

emissions from land use, land use change and <strong>for</strong>estry)<br />

fell by around 439 million tonnes (mt) or 35.1 % by<br />

2019. These reductions were largely achieved by the<br />

energy industry, which is often berated in public on this<br />

issue, mainly through the expansion of low-emission<br />

renewables and the switch from coal to natural gas. But<br />

this balance could be much better, because in total, the<br />

nuclear power plants in operation in Germany be<strong>for</strong>e<br />

Fukushima had avoided around 140 million tonnes of<br />

CO 2 emissions year after year – emissions equal, <strong>for</strong><br />

example, to those of all German car traffic. This<br />

“ shortfall”, moreover provided by highly flexible and<br />

highly available nuclear power plants, i.e. important <strong>for</strong><br />

balancing volatile generation from renewables, is<br />

missing and will be missing.<br />

2. A development in the prices of fossil fuels and biomass<br />

that is almost unique in terms of speed and scope is<br />

burdening all consumers. Expressed in figures and<br />

using the example of heating oil, this means an increase<br />

from € 38 per 100 l at the beginning of November 2020<br />

to € 91 per 100 l now, in mid-October 2021, i.e. an<br />

increase of 139 %! The last time crude oil and crude oil<br />

product prices reached this level was be<strong>for</strong>e the banking<br />

and financial crisis in 2008.<br />

However, while politicians in Germany persist in their<br />

rigid course, almost in rigidity, looking at the development<br />

of energy prices, other countries and regions have long<br />

since taken other paths. Within two decades, China has<br />

built up an efficient nuclear industry. 50 high-per<strong>for</strong>mance<br />

nuclear power plants are an important part of the<br />

electricity supply without fear of contact with renewables,<br />

and in the coming years four to eight units can be added –<br />

every year. Moreover, China is open to new reactor<br />

concepts. High-temperature reactor technology with<br />

pebble fuel elements, which originated in Germany, as<br />

well as sodium-cooled reactors and now even “molten salt<br />

reactors” are being designed, built and commissioned.<br />

In addition, China is also slowly pushing into the worldwide<br />

export of its plants. The same is true <strong>for</strong> nuclear technology<br />

from Russia. In recent years, the Russian nuclear industry<br />

has impressively shown that it can build plants without<br />

major delays even in “green field” countries, i.e. in nuclear<br />

newcomer countries without their own power reactors<br />

so far.<br />

And now Great Britain: <strong>Nuclear</strong> energy is to play a<br />

central role here on the way to climate neutrality with<br />

simultaneously high supply security at af<strong>for</strong>dable prices.<br />

To this end, investments in nuclear energy are to be<br />

promoted and secured by the state, and reactor projects of<br />

small and medium capacity (Small Modular Reactors,<br />

SMR) are to be supported.<br />

Last but not least, there is also France, where President<br />

Emmanuel Macron announced a 30 billion innovation<br />

programme at the beginning of October, with SMR reactors<br />

as an important infrastructure component.<br />

In the end, the German “Energiewende” will not be able<br />

to escape the laws of nature and the facts. It will then be<br />

interesting to see with what verbal acrobatics possible<br />

future problem developments will be conveyed by the<br />

actors who in Germany are pursuing the end of nuclear<br />

energy, but noticeably also an end to all other proven<br />

power plants and a secure energy supply.<br />

Worldwide, on the other hand, it is still true and<br />

rein<strong>for</strong>ced that the energy supply will ultimately be based<br />

on nuclear energy in combination with all other energy<br />

sources – a path of reason.<br />

Christopher Weßelmann<br />

– Editor in Chief –<br />

Editorial<br />

<strong>Nuclear</strong> Energy: A Path to Reason

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

4<br />


Issue 6<br />

2021<br />

November<br />

Contents<br />

Editorial<br />

<strong>Nuclear</strong> Energy: A Path to Reason 3<br />

Inside <strong>Nuclear</strong> with NucNet<br />

How UAE’s Barakah <strong>Nuclear</strong> Project has Set Standard<br />

<strong>for</strong> Newcomer Countries 6<br />

Did you know? 7<br />

Calendar 8<br />

Feature | Decommissioning and Waste Management<br />

First Repatriation of Vitrified Reprocessing Waste from Sellafield 9<br />

Marco Wilmsmeier and Michael Köbl<br />

Interview with Massimo Garribba<br />

“EU Member States Can Choose Their Energy Sources and<br />

Can Include <strong>Nuclear</strong> in Their Energy Mix as Part of Their Ef<strong>for</strong>t<br />

to Achieve Decarbonisation and Carbon Neutrality by 2050.” 14<br />

Decommissioning and Waste Management<br />

Investigations of the Tailskin Seal During the Retrieval<br />

Concept ‘Shield Tunnelling with Partial-face Excavation’<br />

in the Asse II Mine 18<br />

Birte Froebus<br />

Environment and Safety<br />

Practical Response to a Dirty Bomb 22<br />

James Conca<br />

Equipment Selection Methodology of Seismic Probability<br />

Safety Assessment <strong>for</strong> <strong>Nuclear</strong> <strong>Power</strong> Plant 27<br />

Junghyun Ryu and Moosung Jae<br />

Research and Innovation<br />

Experimental Study of Convective Heat Transfer<br />

Through Fuel Pins of a <strong>Nuclear</strong> <strong>Power</strong> Plant 37<br />

Atif Mehmood, Ajmal Shah, Mazhar Iqbal, Ali Riaz, Muhammad Ahsan Kaleem and Abdul Quddus<br />

Preliminary CFD Analysis of Innovative Decay Heat Removal<br />

System <strong>for</strong> the European Sodium Fast Reactor Concept 41<br />

Aleksander Grah, Haileyesus Tsige-Tamirat, Joel Guidez, Antoine Gerschenfeld, Konstantin Mikityuk,<br />

Janos Bodi and Enrico Girardi<br />

Cover:<br />

Return transport of vitrified high-level radioactive<br />

waste from reprocessing of German fuel elements<br />

in Sellafield in November 2020 (Courtesy of GNS<br />

Gesellschaft für Nuklear Service mbH).<br />

Study on Verification of SPACE Code Based on an<br />

MSGTR Experiment at the ATLAS-PAFS Facility 49<br />

Kyungho Nam<br />

News 57<br />

<strong>Nuclear</strong> Today<br />

For the Sake of <strong>Nuclear</strong> Safety and Security,<br />

it’s Time <strong>for</strong> Rogue Actors to Leave the Stage 62<br />

Imprint 36<br />


<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Feature<br />

Decommissioning and<br />

Waste Management<br />

9 First Repatriation of Vitrified Reprocessing Waste<br />

from Sellafield<br />

5<br />


Marco Wilmsmeier and Michael Köbl<br />

Interview with Massimo Garribba<br />

14 “EU Member States Can Choose Their Energy Sources and<br />

Can Include <strong>Nuclear</strong> in Their Energy Mix as Part of Their Ef<strong>for</strong>t<br />

to Achieve Decarbonisation and Carbon Neutrality by 2050.”<br />

Decommissioning and Waste Management<br />

18 Investigations of the Tailskin Seal During the Retrieval Concept ‘Shield<br />

Tunnelling with Partial-face Excavation’ in the Asse II Mine<br />

Birte Froebus<br />

Environment and Safety<br />

22 Practical Response to a Dirty Bomb<br />

James Conca<br />

27 Equipment Selection Methodology of<br />

Seismic Probability Safety Assessment <strong>for</strong> <strong>Nuclear</strong> <strong>Power</strong> Plant<br />

Junghyun Ryu and Moosung Jae<br />

Research and Innovation<br />

49 Study on Verification of SPACE Code Based on an<br />

MSGTR Experiment at the ATLAS-PAFS Facility<br />

Kyungho Nam<br />


<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

6<br />

How UAE’s Barakah <strong>Nuclear</strong> Project has Set Standard<br />

<strong>for</strong> Newcomer Countries<br />

‘They did it in a way that is measurable and can be looked at by the rest of the international<br />

community’<br />


When Unit 2 of the four-unit Barakah nuclear power station<br />

in the United Arab Emirates was connected to the grid earlier<br />

this year, delivering its first megawatts of electricity, project<br />

owner Emirates <strong>Nuclear</strong> Energy Corporation (ENEC) said<br />

the milestone took it another step closer to its goal of using<br />

nuclear to supply up to a quarter of the country’s electricity<br />

needs 24/7, while driving reductions in carbon emissions –<br />

the leading cause of climate change.<br />

Unit 2’s grid connection came five months after Unit 1<br />

became the first commercial nuclear power reactor in the<br />

Arab World to begin commercial operation.<br />

For ENEC, these recent operational milestones show the<br />

world that the UAE is serious about ef<strong>for</strong>ts to use nuclear<br />

energy to further climate change mitigation ef<strong>for</strong>ts through<br />

the decarbonization of the electricity sector. But the benefits<br />

of the Barakah project – one of the largest nuclear energy<br />

stations in the world – go deeper than that, with the facility<br />

providing energy security and powering social and economic<br />

growth by providing high-value jobs and supporting an<br />

entirely new industry.<br />

More than 2,000 UAE-based companies are part of the<br />

supply chain <strong>for</strong> Barakah and local contracts have been<br />

awarded to a value of $ 4.8 bn. “The local supply chain is,<br />

and will continue to be essential <strong>for</strong> the operations and<br />

maintenance of Barakah over the coming 60 years,” ENEC<br />

told NucNet. “It will help ensure we have the localized<br />

services and components we need.”<br />

Unit 1 at Barakah is already the largest single generator<br />

connected to the UAE grid and adding Unit 2 in the coming<br />

months will drive significant energy security, enabling the<br />

electrification of key industries.<br />

“The dual role of decarbonization and electrification<br />

offers a proven solution to cutting carbon emissions,” ENEC<br />

said. “It is the culmination of more than a decade of policy<br />

development, planning and program management.”<br />

Once fully operational, the four units at Barakah will<br />

prevent about 21 million tonnes of carbon emissions<br />

annually and potentially drive other clean energy<br />

technologies such as green hydrogen. In June, Enec signed<br />

an initial agreement with French state- controlled power<br />

group EDF to eventually cooperate on research and<br />

development in the nuclear energy sector – research that<br />

could include exploring the production of green hydrogen<br />

powered by nuclear energy.<br />

ENEC chief executive officer Mohamed al-Hammadi<br />

told the Atlantic Council Global Energy Forum earlier this<br />

year that the right environment exists <strong>for</strong> commercial<br />

nuclear power plants to produce green hydrogen with<br />

capital costs making it viable after falling 50 % in the past<br />

five years.<br />

So the benefits go far beyond clean electricity. The<br />

creation of a new local nuclear energy sector, along with<br />

thousands of skilled jobs and career paths <strong>for</strong> UAE nationals,<br />

and the national knowledge and expertise that has been<br />

developed, mean that Barakah is powering social and<br />

economic growth. ENEC believes nuclear energy can also<br />

support innovation in related sectors, such as deep space<br />

exploration and agriculture.<br />

Construction at Barakah began in 2012. Units 3 and 4 are<br />

in the final stages of commissioning at 94 % and 90 %<br />

complete respectively and the development of the station as<br />

a whole is now more than 96 % complete.<br />

For the UAE, as a newcomer country to nuclear, capacity<br />

development was vital. From the start, ENEC was training<br />

UAE nationals to ensure qualified professionals, working<br />

alongside international experts, could operate and maintain<br />

the facility. The UAE now has multiple academic institutions<br />

offering qualifications in nuclear sciences, and vocational<br />

training in related applications. The aim was always to<br />

ensure that the UAE has the educational facilities to support<br />

generations of staff entering the nuclear sector.<br />

Almost 60 % of the work<strong>for</strong>ce are UAE nationals and more<br />

than 20 % are women, one of the highest percentages of<br />

female nuclear energy professionals globally. By collaborating<br />

with universities and schools across the UAE, Enec has<br />

trained and qualified more than 500 young engineers.<br />

The other key lesson <strong>for</strong> newcomer countries is that clear<br />

policies and full government support are needed from the<br />

beginning – not just in terms of a commitment to new-build<br />

and its financing, but also in areas such as safety, quality,<br />

transparency and security.<br />

The UAE worked directly with the <strong>International</strong> Atomic<br />

Energy Agency (IAEA), using the agency’s milestones<br />

approach to the construction of new nuclear power plants to<br />

establish a comprehensive and systematic guide that can be<br />

followed by other new comer countries planning commercial<br />

reactors.<br />

Hamad al Kaabi, the UAE’s permanent represen tative to<br />

the IAEA, told a webinar to discuss the UAE’s commercial<br />

nuclear energy programme that the IAEA was instrumental<br />

in helping the UEA prepare infrastructure, establish nuclear<br />

laws and oversee the development, construction and startup<br />

phases at Barakah.<br />

He said the task now is <strong>for</strong> the UAE to share its experience<br />

with other IAEA member states. “We now have a case study,<br />

not just theoretical guidelines,” Mr al Kaabi said. “Other<br />

countries can use it to develop nuclear infrastructure that s<br />

in line with best practices.”<br />

IAEA director-general Mariano Grossi said the UAE’s<br />

journey to nuclear power “is already being looked at as<br />

an exemplary one”. He said: “They did it in a way that is<br />

measur able and can be looked at by the rest of the international<br />

community.”<br />

Mr Grossi said that because of what the UAE had achieved<br />

“we now have a complete, successful example, a logical<br />

sequence of steps that can be followed by those who would<br />

like to build nuclear plants.<br />

Lessons learned during construction meant experience<br />

from the development of each unit could be applied to the<br />

next one. This has brought significant benefits in terms of<br />

manpower reduction, efficiency and sustainability.<br />

According to ENEC, Barakah provides a successful case<br />

study <strong>for</strong> the development of new nuclear projects,<br />

particularly <strong>for</strong> newcomer countries. It demonstrates that it<br />

is possible <strong>for</strong> a country to establish a nuclear energy<br />

industry as part of ef<strong>for</strong>ts to ensure energy security, to<br />

diversify its energy mix and decarbonize electricity<br />

generation. The Barakah model, says ENEC, stands as an<br />

example <strong>for</strong> others to follow.<br />

Authors<br />

By Kamen Kraev and David Dalton<br />

Inside <strong>Nuclear</strong> with NucNet<br />

How UAE’s Barakah <strong>Nuclear</strong> Project has Set Standard <strong>for</strong> Newcomer Countries

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Net-Zero Germany 2045 – Contradictory McKinsey-Study<br />

Recently McKinsey & Company published the study “Net-Zero Deutschland –<br />

Chancen und Heraus<strong>for</strong>derungen auf dem Weg zur Klimaneutralität bis<br />

2045” (Net-Zero Germany – Opportunities and Challenges on the Way to<br />

Climate Neutrality 2045) which presented necessary development paths,<br />

obstacles and alternative routes to reach the recent government goal of<br />

climate policy in the different sectors of the economy, energy generation,<br />

industry, mobility, buildings, agriculture and banking.<br />

The tenor of the report as expressed in the summary claims the feasibility<br />

of climate neutrality by 2045 and makes the point that it would not cost very<br />

much in net terms because of the 6,000 billions Euro of investment in real<br />

assets deemed necessary only 1,000 billions are additional but 5,000 billions<br />

are needed anyway and just will have to be redirected in green technologies<br />

and measures. Thereby the study concludes that under optimal circumstances<br />

the entire social cost over the entire period and balanced over all sectors<br />

could be zero. Which would enable to reach net zero without compromising<br />

the quality of life and without social and economic hardship.<br />

While the stated zero-cost conditions already indicate a perspective from<br />

a very high ground at a high level of abstraction, looking into the sectorspecific<br />

analysis shows that the zero-cost outcome is not very likely in the<br />

real world. Below you can find a graph detailing the factual emission<br />

reductions in previous decades and the necessary reduction <strong>for</strong> the time<br />

frames 2019 – 2030 and 2030 – 2045 which shows that particularly the<br />

energy sector, industry and mobility have to increase the pace in emission<br />

reductions drastically.<br />

A closer look at the energy sector that makes up 32 per cent of German<br />

green house gas emissions shows that the pace of installation of new<br />

renewable energy (REN) generation capacity has to be almost tripled <strong>for</strong> the<br />

coming decade as compared to preceding years from 6,5 GW p.a. to some<br />

18 GW p.a. which is considerably more than in the years of the German REN<br />

bonanza 2010 to 2012 (see Graph 2). Also, the transport grid will have to be<br />

expanded by 12,700 kilometers by 2040 instead of 6,100 in previous<br />

assumptions and the distribution grid will need strengthening or expansion<br />

in the order of 400,000 kilometers. Not surprisingly there is more skepticism<br />

articulated about current developments here. The study points out that the<br />

energy trans<strong>for</strong>mation up to now just used the existing (large) safety margins<br />

in the system, but that REN deployment is currently insufficient to move<br />

beyond this point. Consequently, investment in REN and the grid have to be<br />

increased considerably.<br />

If this does not happen there is an important risk of rising electricity prices<br />

due to scarcity and of under coverage of demand that might not in the short<br />

term be replaceable with imports. Hence the study considers the possibility<br />

of load shedding in the future. Part of the proposed solution is load management,<br />

but the discussion about limiting the capacity <strong>for</strong> e-vehicles charging<br />

early this year showed, that there is not much acceptance <strong>for</strong> this kind of<br />

measures in Germany which indeed look strikingly similar to the load<br />

shedding that they are intended to avoid. The proposed alternative paths <strong>for</strong><br />

carbon emissions reduction are CCU (Carbon Capture and Utilization) and<br />

CCS (Carbon Capture and Storage) and the import of green hydrogen. The<br />

study asserts, that the viable trans<strong>for</strong>mation path is still not decided and<br />

planned. The risk exists, that the energy trans<strong>for</strong>mation as it is running now<br />

will lead to a situation where neither the energy needs of households and<br />

enterprises can be met nor the climate targets will be reached.<br />

Another sector that shows contradictions in the study is buildings. While the<br />

study realistically assumes that in 2050 still some 80 per cent of buildings will<br />

be of a construction date prior to 2011, it is still supposed that 50 per cent of<br />

heating will be done with heat pumps and that oil and gas boilers will have<br />

been phased-out by 2040 and 2050 respectively. No account is made of the<br />

fact that older buildings are not appropriate <strong>for</strong> heat pumps and that changing<br />

this would require extremely cumbersome and expensive reconstruction<br />

measures. Without these, the assumed gains in efficiency like those that exist<br />

in electrifying the mobility sector, cannot be realised. The subsequent<br />

enormous increase in electrical power consumption and particularly the<br />

extreme peaking effect of this measure in a period of low solar radiation is not<br />

accounted <strong>for</strong> either. The obvious and much more viable alternative of<br />

importing green gases is not taken into consideration <strong>for</strong> the building sector.<br />

There also is a macroeconomic weak spot in the reasoning behind the<br />

proposed net zero path: <strong>for</strong> some sectors measures are proposed that are<br />

comparatively labor intensive. What might sound good at first glance and<br />

could provide some relief <strong>for</strong> employees e.g. in the automotive industry<br />

which will face serious pressure on employment due to the much simpler<br />

manufacturing of electric vehicles compared to vehicles with internal<br />

combustion engines, does not play well on the macro level. In the context of<br />

demographic change with a considerably shrinking work <strong>for</strong>ce in Germany<br />

and many European countries it seems ill fated to propose development<br />

paths and try to generate growth with labor intensive and relatively low<br />

productivity activities such as energetic refurbishment of buildings or<br />

massively accelerated deployment of renewable power.<br />

Not only to prevent load shedding, to avoid high volatility in energy prices<br />

and to assure true and profound decarbonization of the energy sector, but<br />

also in terms of labor productivity and effective use of this scarce resource of<br />

the future, nuclear power appears to be the more reasonable and viable<br />

instrument to reach net zero.<br />

7Did you know?<br />


Average Annual Reduction of Emissions per Sector<br />

in Million Ton CO 2 -equiv.<br />

Energy<br />

Industry<br />

Mobility<br />

Buildings<br />

Agriculture<br />

0<br />

0.7<br />

1.1<br />

3.3<br />

5.7<br />

4.5<br />

5.1<br />

3.0<br />

3.7<br />

6.3<br />

7.2<br />

7.2<br />

7.2<br />

7.9<br />

p 1990–2019<br />

p 2019–2030<br />

p 2030–2045<br />

13.6<br />

Annual Net Increase in Installed Capacity of REN<br />

<strong>for</strong> Electricity Generation in GW<br />

18.0<br />

p Annual Capacity Addition in GW<br />

14.0<br />

10.9 10.7<br />

9.3<br />

5.6 6.6 7.5 8.1<br />

6.6 6.6 6.2<br />

6.7<br />

2010<br />

2011<br />

2012<br />

2013<br />

2014<br />

2015<br />

2016<br />

2017<br />

2018<br />

2019<br />

2020<br />

2021 –2030<br />

2030 –2045<br />

Source:<br />

Net-Zero Deutschland –<br />

Chancen und Heraus<strong>for</strong>derungen<br />

auf dem<br />

Weg zur Klimaneutralität<br />

bis 2045, McKinsey &<br />

Company, September<br />

2021<br />

For further details<br />

please contact:<br />

Nicolas Wendler<br />

KernD<br />

Robert-Koch-Platz 4<br />

10115 Berlin<br />

Germany<br />

E-mail: presse@<br />

KernD.de<br />

www.KernD.de<br />

Did you know?

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

8<br />

Calendar<br />


2021<br />

31.10. – 12.11.2021<br />

COP26 – UN Climate Change Conference.<br />

Glascow, Scotland, www.ukcop26.org<br />

01.11. – 05.11.2021<br />

<strong>International</strong> Conference on<br />

Radioactive Waste Management:<br />

Solutions <strong>for</strong> a Sustainable Future.<br />

IAEA, Vienna, Austria,<br />

https://www.iaea.org/events/<br />

international-conference-on-radioactive-wastemanagement-2021<br />

Online Conference04.11.2021<br />

Small and Advanced Reactors.<br />

PMI-Live,<br />

https://registration.pmi-live.com/<br />

tc-events/small-and-advanced-reactor/<br />

Online Conference 07.11. – 12.11.2021<br />

PSA 2021 – <strong>International</strong> Topical Meeting on<br />

Probabilistic Safety Assessment and Analysis.<br />

ANS, Columbus, OH, USA,<br />

http://psa.ans.org/2021<br />

Online Conference 15.11. – 16.11.2021<br />

<strong>Nuclear</strong> New Builds 2021.<br />

Prospero Events Group,<br />

https://www.prosperoevents.com/event/<br />

nuclear-new-builds/<br />

Hybrid Conference 15.11. – 17.11.2021<br />

NESTet2021 – <strong>Nuclear</strong> Education and Training.<br />

ENS, Brussels, Belgium,<br />

https://ens.eventsair.com/<br />

nuclear-education-and-training<br />

30.11. – 02.12.2021<br />

Enlit (<strong>for</strong>mer European Utility Week<br />

and POWERGEN Europe).<br />

Milano, Italy,<br />

www.enlit-europe.com<br />

30.11. – 02.12.2021<br />

WNE2021 – World <strong>Nuclear</strong> Exhibition.<br />

Paris, France, Gifen,<br />

www.world-nuclear-exhibition.com<br />

2022<br />

26.01. – 28.01.2022<br />

<strong>Power</strong>Gen <strong>International</strong>.<br />

Clarion Events, Dallas, TX, USA,<br />

www.powergen.com<br />

06.03. – 10.03.2022<br />

WM2022 - Waste Management Conference<br />

X-CD Technologies, Phoenix, AZ, USA<br />

https://www.wmsym.org/<br />

06.03. – 11.03.2022<br />

NURETH19 – 19 th <strong>International</strong> Topical<br />

Meeting on <strong>Nuclear</strong> Reactor Thermal<br />

Hydraulics. SCK·CEN, Brussels, Belgium,<br />

https://www.ans.org/meetings/view-334/<br />

24.03. – 25.03.2022<br />

<strong>Nuclear</strong> Innovation Conference 2022.<br />

NRG, Amsterdam, The Netherlands,<br />

https://www.nuclearinnovationconference.eu/<br />

29.03. – 30.03.2022<br />

KERNTECHNIK 2022.<br />

Leipzig, Germany, KernD and KTG,<br />

www.kerntechnik.com<br />

04.04. – 08.04.2022<br />

<strong>International</strong> Conference<br />

on Geological Repositories.<br />

Helsinki, Finland, EURAD,<br />

https://www.ejp-eurad.eu/events/<br />

international-conference-geological-repositories<br />

Postponed to Spring 2022<br />

4 th CORDEL Regional Workshop –<br />

Harmonization to support the operation<br />

and new build of NPPs including SMR.<br />

Lyon, France, World <strong>Nuclear</strong> Association,<br />

https://events.<strong>for</strong>atom.org<br />

04.05. – 06.05.2022<br />

NUWCEM 2022 – 4 th <strong>International</strong><br />

Symposium on Cement-Based Materials<br />

<strong>for</strong> <strong>Nuclear</strong> Wastes.<br />

Sfen, Avignon, France,<br />

https://new.sfen.org/evenement/nuwcem-2022<br />

15.05. – 20.05.2022<br />

PHYSOR 2022 – <strong>International</strong> Conference<br />

on Physics of Reactors 2022.<br />

ANS, Pittsburgh, PA, USA,<br />

www.ans.org<br />

22.05. – 25.05.2022<br />

NURER 2022 – 7 th <strong>International</strong> Conference<br />

on <strong>Nuclear</strong> and Renewable Energy Resources.<br />

ANS, Ankara, Turkey,<br />

www.ans.org<br />

Postponed to 30.05. – 03.06.2022<br />

20 th WCNDT – World Conference<br />

on Non-Destructive Testing.<br />

Incheon, Korea, The Korean Society<br />

of Nondestructive Testing,<br />

https://20thwcndt.com/<br />

10.07. – 15.07.2022<br />

SMiRT 26 – 26 th <strong>International</strong> Conference on<br />

Structural Mechanics in Reactor Technology.<br />

German Society <strong>for</strong> Non-Destructive Testing,<br />

Berlin/Potsdam, Germany,<br />

www.smirt26.com<br />

04.09. – 09.09.2022<br />

NUTHOS-13 – 13 th <strong>International</strong> Topical<br />

Meeting on <strong>Nuclear</strong> Reactor Thermal<br />

Hydraulics, Operation and Safety.<br />

ANS, Taichung, Taiwan,<br />

www.ans.org<br />

NUCLEAR 2021.<br />

NIA, London, United Kingdom,<br />

https://events.<strong>for</strong>atom.org/calendar/<br />

nuclear-2021/<br />

02.12.2021<br />

05.04. – 07.04.2022<br />

GLOBAL 2022 – <strong>International</strong> Conference<br />

on <strong>Nuclear</strong> Fuel Cycle.<br />

Sfen, Reims, France,<br />

https://new.sfen.org/evenement/global-2022/<br />

This is not a full list and may be subject to change.<br />


<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

First Repatriation of Vitrified<br />

Reprocessing Waste from Sellafield<br />

Managing challenges with transport organization<br />

and radiation protection<br />

Marco Wilmsmeier and Michael Köbl<br />

Repatriation from the United Kingdom Reprocessing contracts which the operators of German nuclear<br />

power plants (EVU) had concluded with the <strong>Nuclear</strong> Decommissioning Authority (NDA, <strong>for</strong>mer BNFL) <strong>for</strong>med the basis<br />

<strong>for</strong> the transport of spent fuel elements to the reprocessing plant at Sellafield, where they were reprocessed. Until 1994,<br />

the reprocessing of irradiated fuel elements from nuclear power plant operation was mandatory in Germany pursuant<br />

to § 9 AtG (Atomic Energy Act). Since a national reprocessing concept was dispensed with in 1989, the waste had to be<br />

transported abroad <strong>for</strong> reprocessing. From 1994 onwards, an amendment to the Atomic Energy Act made it possible to<br />

also transport nuclear waste directly to a repository in parallel with reprocessing as a waste management concept. As of<br />

July 1, 2005, the transfer of irradiated fuel elements to a reprocessing plant is banned. There is an obligation to take<br />

back the amount of heavy metals produced during reprocessing in accordance with national agreements and the<br />

corresponding exchange of notes between Germany and the UK. <strong>International</strong> law states that an amount equivalent to<br />

the mass of 768 t of heavy metal delivered to Sellafield must be returned to Germany.<br />

| Fig. 1.<br />

HLW canister.<br />

The mass to be returned<br />

equates to 560 HLW canisters,<br />

whose return and<br />

packaging is in turn contractually<br />

agreed between<br />

the EVU and the NDA. The<br />

HLW canister is a waste<br />

product from the reprocessing<br />

and takes the <strong>for</strong>m<br />

of a sealed stainless steel<br />

cylinder which is filled with<br />

a glass matrix in which<br />

dissolved high-level radioactive<br />

fission products are<br />

embedded.<br />

Preparations <strong>for</strong> the first transport<br />

from Sellafield<br />

In preparation <strong>for</strong> the return of the stream of high-level<br />

radioactive waste from Sellafield, the design and statutory<br />

approval of the CASTOR® HAW28M transport and storage<br />

cask, and the loading planning, were undertaken on the<br />

basis of the specification of the Sellafield canister and the<br />

inventory specification in compliance with international<br />

and national regulations.<br />

Furthermore, the return of the HLW canisters from the<br />

UK required the infrastructure <strong>for</strong> the transport, loading,<br />

and dispatch to be created, a process which was successfully<br />

verified by carrying out a cold trial of the cask loading<br />

and handling operation <strong>for</strong> the CASTOR® HAW28M cask in<br />

Sellafield in 2013.<br />

The first six of a total of 20 empty casks of the design<br />

CASTOR® HAW28M which were needed – each capable of<br />

holding 28 canisters – were transported by rail and ship<br />

from the GNS production plant in Mülheim to Sellafield in<br />

June 2018. In Sellafield, each was loaded with a total of<br />

28 HLW canisters in the loading facility of Sellafield<br />

Limited between December 2018 and November 2019. The<br />

first return transport from Sellafield to Germany and the<br />

storage facility of BGZ Gesellschaft für Zwischenlagerung<br />

mbH at the nuclear power plant site in Biblis (BZB) took<br />

place in November 2020.<br />

Challenges with the transport<br />

organization<br />

The companies and organizations<br />

involved in the return transport have<br />

many years of experience in the<br />

transport of radioactive materials<br />

and waste. The means of transport<br />

employed and the equipment used<br />

have already been successfully utilized<br />

several times in comparable projects.<br />

Until 2011, casks of the type CASTOR®<br />

HAW28M were already being transported<br />

by rail from La Hague to<br />

Gorleben. The same type of cask was<br />

also used in 2016 to transport vitrified<br />

waste from Sellafield to the ZWILAG<br />

in Switzerland.<br />

Nevertheless, several firsts at the<br />

same time presented special challenges:<br />

The transport from Sellafield<br />

to Biblis was the first repatriation<br />

from the UK to Germany. For the first<br />

time in nine years, loaded CASTOR®<br />

HAW28M casks were transported<br />

within Germany, and <strong>for</strong> the first time<br />

ever, their destination was not the<br />

central interim storage facility in<br />

Gorleben, but an interim storage<br />

facility of a nuclear power plant.<br />

Coronavirus pandemic<br />

Whereas the logistical challenges<br />

were largely <strong>for</strong>e seeable, manageable,<br />

and plannable, the coronavirus pandemic<br />

brought completely new types of complications: The<br />

transport was originally planned <strong>for</strong> spring 2020 and all<br />

necessary preparations had been completed on time. On<br />

February 27, 2020, the GNS issued the so-called “ Clearance<br />

<strong>for</strong> Shipment”.<br />

But on March 12, 2020, less than a week be<strong>for</strong>e loading<br />

was scheduled to commence in Sellafield, the Federal<br />

Police regional headquarters in Hanover, which was<br />

| Fig. 2.<br />

Cask of the type CASTOR® HAW28M (shown<br />

in cross section, without shock absorbers).<br />

9<br />


Feature<br />

First Repatriation of Vitrified Reprocessing Waste from Sellafield ı Marco Wilmsmeier and Michael Köbl

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


Previous repatriations from France<br />

Until 2005, spent fuel elements from the operation of German<br />

nuclear power plants were transported to the United Kingdom and<br />

France <strong>for</strong> reprocessing. The radioactive waste produced during<br />

reprocessing has to be returned to Germany. The GNS Gesellschaft<br />

für Nuklear-Service mbH has been commissioned by the German<br />

nuclear power plant operators to prepare and undertake the return<br />

of this waste to German interim storage facilities. Between 1996 and<br />

2011, GNS already undertook twelve transports to the central interim<br />

storage facility in Gorleben, Lower Saxony; these transports involved<br />

a total of 108 large casks filled with vitrified high-level radioactive<br />

waste from the reprocessing of German fuel elements in the French<br />

reprocessing plant in La Hague. The sometimes sizeable protests<br />

against the transports, extensive discussions on the political level and<br />

within society, and days of media coverage, resulted in these transports,<br />

which were generally known as “Castor transports”, becoming<br />

a symbol of the resistance against nuclear energy.<br />

responsible <strong>for</strong> the security of the transport, gave notification<br />

that the security measures were not justifiable given<br />

the spread of coronavirus at that time, and hence the<br />

planning and realization of the transport was suspended<br />

with immediate effect.<br />

In the months that followed, extensive coordination<br />

with all stakeholders, and the police <strong>for</strong>ces involved in<br />

particular, was necessary to draw up a new transport<br />

schedule – but still with the proviso that the coronavirus<br />

risks were manageable, and under much more difficult<br />

conditions. The pandemic situation led to a new transport<br />

schedule being agreed in July 2020 <strong>for</strong> late fall. It was thus<br />

possible to avoid additional complications resulting from<br />

the completion of BREXIT at the end of the year, and repeat<br />

tests which would otherwise have had to be carried out in<br />

Sellafield on the casks which were already loaded, and the<br />

shock absorbers which were already mounted.<br />

In addition to the extensive preparations which had<br />

already been undertaken <strong>for</strong> the transport in spring 2020,<br />

the pandemic brought completely new types of problems:<br />

A continuous, unbroken hygiene concept had to be<br />

compiled <strong>for</strong> the whole transport route from Sellafield<br />

to the port of Barrow-in-Furness, the sea crossing to<br />

Nordenham, and the subsequent rail transport to Biblis,<br />

and agreed with the authorities in both countries. In<br />

addition to the hygiene measures themselves, which have<br />

meanwhile become established, the contact between all<br />

those involved had especially to be kept to a minimum.<br />

One consequence was there<strong>for</strong>e that German representatives<br />

were not allowed to be present at the operations in<br />

the UK, <strong>for</strong> example. We had to fall back on local experts<br />

instead. And the pandemic meant that the police <strong>for</strong>ces in<br />

particular also had to overcome considerable challenges<br />

posed by the operational planning along the transport<br />

route.<br />

Transport schedule<br />

On October 2, 2020, GNS was again able to issue the<br />

“Clearance <strong>for</strong> Shipment” and this time it was final. Shortly<br />

after 6 a.m. GMT on October 26, 2020, the first three casks<br />

left the Sellafield plant heading <strong>for</strong> the port of Barrow- in-<br />

Furness. On October 26 and 27, 2020, the casks were<br />

transferred from the railroad car and onto the transport<br />

ship, the MV Pacific Grebe. After a sea crossing lasting<br />

several days, the Pacific Grebe arrived in the port of<br />

Nordenham on November 2, 2020, under tight security.<br />

On the very same day, November 2, and the following day,<br />

| Fig. 3.<br />

Transferring a CASTOR® HAW28M from the MV Pacific Grebe<br />

onto a railroad car at the Port of Nordenham.<br />

November 3, the six casks were again loaded onto railroad<br />

cars <strong>for</strong> the last stage of their journey to Biblis. The train<br />

with its special cars left Nordenham on the evening of<br />

November 3 and arrived in Biblis without any major<br />

incidents on the morning of November 4.<br />

While the transport was in progress, comprehensive<br />

measurements and tests were carried out to verify the<br />

proofs which had been provided in advance to obtain the<br />

authorizations.<br />

Temperature measurements during the transport<br />

A condition <strong>for</strong> the German Federal Institute <strong>for</strong> Materials<br />

Research and Testing (BAM) agreeing to the use of the<br />

transport ship, the MV Pacific Grebe, was proof that heat<br />

could be safely removed from the closed holds. The<br />

verification was there<strong>for</strong>e undertaken by continuously<br />

recording the surface temperatures of the casks during the<br />

several days they were at sea. A thermal imaging camera<br />

was used to verify that the maximum temperature on the<br />

container surface was 41 °C, and the exhaust air<br />

temperature was a maximum of 19 °C during the whole<br />

170 hours of the measurement.<br />

The safe removal of heat from the holds of the MV<br />

Pacific Grebe was there<strong>for</strong>e guaranteed at all times. The<br />

maximum temperature on the package was below the<br />

Feature<br />

First Repatriation of Vitrified Reprocessing Waste from Sellafield ı Marco Wilmsmeier and Michael Köbl

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

| Fig. 4.<br />

Thermographic image of a CASTOR® HAW28M in the hold<br />

of the MV Pacific Grebe.<br />

value calculated in advance and thus significantly below<br />

the maximum permissible surface temperature of 96 °C.<br />

Acceleration measurements<br />

during the various transport stages<br />

A further measurement requirement originates from the<br />

9 th amended storage license of Spent Fuel Interim Storage<br />

Biblis (BZB) regarding the storage of the casks with the<br />

canisters from Sellafield. To ensure that the leak tightness<br />

of the sealing barrier <strong>for</strong> the primary lid complied with the<br />

specification, it was necessary to verify that the routine<br />

transport conditions (RBB) were complied with. To this<br />

end, the acceleration values <strong>for</strong> each individual cask on<br />

each stage of the transport were recorded by means of data<br />

loggers (Moni Log).<br />

Here as well, it was possible to verify that all stages of<br />

the transport complied with the RBB, since the acceleration<br />

of the casks during the transport did not exceed the<br />

maximum permissible value of 2g in each spatial direction.<br />

The highest acceleration value on all transport stages was<br />

0.85 g.<br />

Radiological Tests<br />

By far the most comprehensive measuring program<br />

involved the radiological tests be<strong>for</strong>e and during the<br />

transport. In Sellafield, the ports of Barrow-in-Furness and<br />

Nordenham, and in Biblis, too, comprehensive radiological<br />

tests were carried out during the complete transport cycle<br />

of the six transport and storage casks to verify compliance<br />

with the requirements of transport legislation and storage<br />

legislation.<br />

To safeguard the protection of humans and, as far as the<br />

long-term protection of human health is concerned,<br />

the environment against the harmful effects of ionizing<br />

radiation during the transport and storage of the loaded<br />

CASTOR® HAW28M, compliance with the relevant limit<br />

values <strong>for</strong> the dose rate and surface contamination is also<br />

checked. The limit values originate from the ADR,<br />

RID [1, 2] regulations, and the IMDG Code [3]. The<br />

specifications of § 58 StrlSchV [4] apply to the relocation<br />

of equipment from nuclear facilities in Germany<br />

( controlled area), i.e. the (surface) contamination is<br />

< 0.04 Bq/cm 2 <strong>for</strong> α-emitters and < 0.4 Bq/cm 2 <strong>for</strong><br />

β/g-emitters. The limit values under storage legislation<br />

are specified in the technical acceptance conditions of the<br />

accepting interim storage facility of a nuclear power plant.<br />

The appropriate limit values which have to be applied to<br />

the cask or the package (cask with shock absorbers), the<br />

particular means of transport, and the handling equipment,<br />

are laid down in the GNS test specifications.<br />

Execution of the radiological tests<br />

To document all handling and test steps while the empty<br />

casks are being transported from Mülheim to Sellafield,<br />

while they are being loaded and dispatched in Sellafield,<br />

and also while the loaded casks are being transported from<br />

Sellafield to Biblis, cask-specific sequence plans including<br />

test and measurement logs which have to be signed were<br />

used.<br />

Prior to carrying out the radiological tests, the calibration<br />

certificates <strong>for</strong> the measuring equipment which had<br />

already been submitted were checked to make sure they<br />

had not expired. At the start of each radiological test, care<br />

was taken that the measuring range of the equipment used<br />

to measure the dose rate was set correctly, and the contamination<br />

measuring devices were functioning correctly.<br />

The tests were carried out by qualified radiation protection<br />

staff.<br />

Contamination tests:<br />

All contamination tests were carried out in compliance<br />

with the limit values specified under transport legislation.<br />

Protocols which identify representative measuring points,<br />

and which were laid down on the basis of experience<br />

gained during the cold trails or previous use of the<br />

equipment, were used <strong>for</strong> the measurements.<br />

The contamination tests per<strong>for</strong>med while the empty<br />

casks were being transported, and <strong>for</strong> relocating equipment<br />

out of the controlled area, served to verify they were contamination<br />

free according to the limit values under transport<br />

legislation of < 0.04 Bq/cm 2 <strong>for</strong> α-emitters or < 0.4<br />

Bq/cm 2 <strong>for</strong> β/g-emitters.<br />

Further contamination tests were carried out on the<br />

loaded cask and equipment while the cask was being<br />

loaded and dispatched, and while it was being transported,<br />

to verify it was contamination free according to the<br />

limit values under transport legislation of < 0.4 Bq/cm 2<br />

<strong>for</strong> α-emitters or < 4 Bq/cm 2 <strong>for</strong> β/g-emitters.<br />

While the empty CASTOR® HAW28M casks were<br />

being transported from the GNS production plant in<br />

Mülheim to Sellafield, while they were being loaded and<br />

dispatched in Sellafield, and during the return transport<br />

to Biblis until the handover to the BZB, the first test<br />

method used was the wipe test <strong>for</strong> non-fixed surface<br />

contamination. A preliminary screening test, if provided<br />

<strong>for</strong>, served as an indicative test on the railroad car, <strong>for</strong><br />

example, to prevent contaminated surfaces from being<br />

14 Screening tests<br />

29 Wipe tests<br />

Direct measurements<br />

at all accessible points<br />

| Fig. 5.<br />

Plan of the measuring points <strong>for</strong> contamination measurements<br />

on the railroad car be<strong>for</strong>e loading/after unloading.<br />


Feature<br />

First Repatriation of Vitrified Reprocessing Waste from Sellafield ı Marco Wilmsmeier and Michael Köbl

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


overlooked. Afterward, at each operational site, the<br />

equipment coming into direct or indirect contact with the<br />

cask, such as crane traverses, transport and storage racks,<br />

as well as railroad cars and the decks of the ship, were<br />

subjected to a direct measurement at all accessible points<br />

prior to use and checked <strong>for</strong> fixed contamination.<br />

While the casks were being loaded and dispatched in<br />

Sellafield, and during the return transport to Biblis until<br />

the handover to the BZB, wipe tests were carried out on<br />

the particular package as well to check <strong>for</strong> non-fixed<br />

contamination.<br />

A total of 270 wipe tests were conducted on the<br />

casks alone and the contamination measurements were<br />

assessed.<br />

Dose rate measurements<br />

After loading each cask, the g and neutron dose rate was<br />

measured on the vertical cask at twelve predefined<br />

measuring points which are highlighted in color and<br />

spread over the outer surface of the cask. Afterward, the<br />

average dose rates were calculated taking into account the<br />

limit values <strong>for</strong> neutrons of ≤ 250 mSv/h and g+neutrons<br />

of ≤ 350 mSv/h stipulated in the storage legislation.<br />

After the cask had been transferred onto a means<br />

of transport, dose rate measurements were carried out on<br />

the package as a contact measurement on the basis<br />

of a predefined plan of the measuring points with<br />

13 measuring points, and the maximum value was determined.<br />

The limit value under transport legislation <strong>for</strong> g+neutrons<br />

of ≤ 2 mSv/h had to be verified here.<br />

After the particular cask had been transferred within<br />

the site from the loading facility to a store to prepare it <strong>for</strong><br />

transport in Sellafield, the g and neutron dose rate was<br />

measured at ten measuring points on the outer surface of<br />

the package at a distance of one meter to determine the<br />

transport index. The transport index is required to rate the<br />

package category, <strong>for</strong> the hazardous goods labeling among<br />

other things.<br />

Be<strong>for</strong>e the loaded casks were transported from the<br />

dispatching facility in Sellafield in October 2020, further<br />

dose rate measurements were carried out on at least four<br />

measuring points in each case at a distance of two meters<br />

from the outer surface of the transport vehicle after they<br />

had been transferred onto a means of transport, and after<br />

changing the mode of transport in Nordenham, to verify<br />

compliance with the limit value of 100 mSv/h under<br />

transport legislation.<br />

While the casks were being loaded and transported,<br />

540 dose rate measurements (contact, 1 m and 2 m) were<br />

carried out and assessed.<br />

Assessment of the measurements<br />

As expected, the contamination values measured while the<br />

empty casks were being transported were below the limit<br />

values under transport legislation, so that this transport<br />

could be dealt with as a conventional transport.<br />

The contamination values measured <strong>for</strong> relocating<br />

equipment were also below the limit values under<br />

transport legislation, so that the equipment could be<br />

returned to its conventional use.<br />

The contamination measurements on casks and<br />

equipment during loading, dispatch, return transport until<br />

the handover to the BZB, showed that the measured values<br />

were less than 10% of the limit values specified under<br />

transport legislation of < 0.4 Bq/cm 2 <strong>for</strong> α-emitters or<br />

< 4 Bq/cm 2 <strong>for</strong> β/g-emitters.<br />

The average g and neutron dose rate was determined<br />

from the results of the measurement on the loaded casks<br />

| Fig. 6.<br />

Dosage measurement at a distance of 2 m from the transport vehicle<br />

in Nordenham.<br />

Requirement / result<br />

Package CASTOR® HAW28M-<br />

01 02 03 04 05 06<br />

Proportion of limit value<br />

(in relation to max. value)<br />

in %<br />

Dose rate on surface of cask in acc.<br />

with storage legislation requirements<br />

Limit value: g+neutrons < 0.35 μSv/h 0.098 0.099 0.102 0.112 0.108 0.109 32.0<br />

Dose rate on surface of package in acc.<br />

with transport legislation requirements<br />

Limit value: g+neutrons < 2 μSv/h 0.384 0.275 0.342 0.318 0.299 0.383 19.2<br />

Dose rate at distance of 2 m from outside<br />

of vehicle in acc. with transport legislation<br />

Limit value: g+neutrons < 0.1 μSv/h 0.023 0.029 0.025 0.026 0.026 0.028 29.0<br />

Total activity of cask content in 10 15 Bq<br />

Maximum permissible: 1,270 × 10 15 Bq 309 325 309 316 324 379 29.8<br />

| Tab. 1.<br />

Dose rates and activities of the casks <strong>for</strong> the transport from Sellafield to Biblis.<br />

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<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

required by storage legislation, and recorded. During the<br />

subsequent check of the calculated dose rates, it was<br />

confirmed in the presence of experts and GNS that all the<br />

casks complied with the requirements under storage<br />

legislation. It was ascertained that the values did not<br />

exceed 35 % of the maximum permissible limit values.<br />

The dose rate measurements as a contact measurement<br />

on the packages showed that the maximum value was<br />

approx. 20 % of the limit value <strong>for</strong> g+neutrons of ≤ 2 mSv/h<br />

under transport legislation.<br />

The verification of the limit values <strong>for</strong> the dose rate<br />

measurements under transport legislation was conducted<br />

at a distance of two meters from the packages in both<br />

Sellafield and Nordenham. It was ascertained that the<br />

maximum value was approx. 30 % of the limit value of<br />

≤ 100 mSv/h under transport legislation.<br />

The fact that the values were only a fraction of the limit<br />

values under transport legislation and storage legislation<br />

corresponds with the total activity of the HLW canister<br />

inventory in the respective CASTOR® HAW28M casks<br />

which were loaded in Sellafield. This is also due to the<br />

design of the cask shielding and an optimized loading<br />

plan.<br />

Conclusion and outlook<br />

All radiological tests carried out while the empty casks<br />

were being transported, while the casks were being loaded<br />

and dispatched in Sellafield, and also while the loaded<br />

casks were being transported from Sellafield to Biblis until<br />

the handover to the BZB, prove with the aid of the test<br />

results determined that during the first HLW return<br />

transport from the UK to Germany, the limit values under<br />

transport and storage legislation were safely complied<br />

with or nowhere near reached, at all locations. The<br />

measurements of the temperature and the acceleration of<br />

the casks, which were conducted in addition, likewise<br />

verified the reliable compliance with all specifications.<br />

Although the pandemic meant that the measures required<br />

were much more complex, it was possible to conduct the<br />

whole transport safely, reliably, and on schedule. The<br />

experience and insights gained while transporting the<br />

casks to Biblis <strong>for</strong>m an optimum basis <strong>for</strong> the two still<br />

outstanding return transports, both of which involve<br />

transporting seven federal casks of the type CASTOR®<br />

HAW28M from Sellafield to the interim storage facilities at<br />

the Isar and Brokdorf nuclear power plants.<br />

References<br />

[1] ADR: European Agreement concerning the <strong>International</strong> Carriage of Dangerous Goods by Road<br />

(Accord relatif au transport international des marchandises Dangereuses par Route), in the<br />

version valid as of January 1, 2021<br />

[2] RID: Regulation concerning the <strong>International</strong> Carriage of Dangerous Goods by Rail (Règlement<br />

concernant le transport <strong>International</strong> ferroviaire des marchandises Dangereuses), in the version<br />

valid as of January 1, 2021<br />

[3] IMDG Code: <strong>International</strong> Maritime Code <strong>for</strong> Dangerous Goods, 2020 edition including<br />

Amendment 40-20<br />

[4] § 58 StSchV – Radiation Protection Act (Verordnung zum Schutz vor der schädlichen Wirkung<br />

ionisierender Strahlung – Strahlenschutzverordnung): Verlassen von und Herausbringen aus<br />

Strahlenschutzbereichen, Artikel 1 V. v. 29.11.2018 BGBl. I S. 2034, 2036 (Nr. 41);<br />

zuletzt geändert durch Artikel 6 G. v. 20.05.2021 BGBl. I S. 1194<br />

Authors<br />

Marco Wilmsmeier<br />

GNS Gesellschaft für Nuklear-Service mbH, Germany<br />


Marco Wilmsmeier studied bilaterally at the Gelsenkirchen University of Applied<br />

Sciences and Sheffield Hallam University, UK, waste management technology and<br />

environmental engineering. He graduated from Prof. Malcom Denman in Sheffield<br />

in 2000. Until mid-2001 he was employed as technical operations manager at SCA<br />

Packaging in Hövelhof be<strong>for</strong>e he started as Project Manager <strong>for</strong> loading service and<br />

reprocessing transports at GNS. In 2004 he changed to the department of recycling<br />

of reprocessing waste within the GNS and since then has been responsible as<br />

Project Manager <strong>for</strong> HAW recycling from the UK.<br />

Michael Köbl<br />

GNS Gesellschaft für Nuklear-Service mbH, Germany<br />

Michael.Koebl@gns.de<br />

| Fig. 7.<br />

Arrival of the MV Pacific Grebe at the Port of Nordenham. Transferring a<br />

CASTOR® HAW28M from the MV Pacific Grebe onto a railroad car.<br />

After completing his business administration studies, Michael Köbl started in 2005<br />

as Head of Internal Communication at Grundig AG in Fürth. In 2008 he joined the<br />

Public Relations department at Dynamit Nobel AG in Troisdorf, which he was<br />

leading since 2014. Since 2006 he is employed at GNS in Essen, initially as a press<br />

and public relations officer and since 2009 as Head of Communications.<br />

Feature<br />

First Repatriation of Vitrified Reprocessing Waste from Sellafield ı Marco Wilmsmeier and Michael Köbl

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

14<br />


“EU Member States Can Choose Their<br />

Energy Sources and Can Include<br />

<strong>Nuclear</strong> in Their Energy Mix as Part of<br />

Their Ef<strong>for</strong>t to Achieve Decarbonisation<br />

and Carbon Neutrality by 2050.”<br />

Interview with Massimo Garribba<br />

ı Deputy Director-General of DG Energy<br />

Massimo Garribba<br />

Deputy Director-General of DG Energy<br />

Originally from Padova (Italy), Dr Massimo Garribba is a<br />

qualified electronic engineer with a PhD in In<strong>for</strong>matics and<br />

Industrial Electronics. After spending seven years on fusion<br />

research, he joined the European Commission in 1995,<br />

originally working on digital issues. In 2004 he moved to the<br />

Commission’s Directorate-General <strong>for</strong> Energy, working on<br />

Euratom coordination and international relations, first as<br />

Head of Unit, then as Director with a broader remit. In July<br />

2020, he was promoted to Deputy Director-General<br />

responsible <strong>for</strong> the coordination of all aspects of EURATOM<br />

policy.<br />

The issue of climate change and the reduction of<br />

greenhouse gases has become a very prominent<br />

political priority <strong>for</strong> the EU, the member states and<br />

many other nations. At the same time, the EU is<br />

developing a sustainable investment framework as<br />

a front-runner among the major economies. Has<br />

the Commission finally reached a position on<br />

nuclear and the taxonomy after more than two<br />

years of controversial debates?<br />

Not yet. The Taxonomy Regulation reflects a delicate<br />

compromise on the question of whether or not to include<br />

nuclear energy in the EU taxonomy. While nuclear energy<br />

is consistently acknowledged as a low-carbon energy<br />

source, opinions differ notably on the potential impact on<br />

other environmental objectives, such<br />

as the environmental impact of<br />

nuclear waste.<br />

The Commission considers that the<br />

credibility of this assessment is<br />

crucial. It has requested the Joint<br />

Research Centre (the Commission’s<br />

internal scientific service) to draft a<br />

technical report on the ‘do no<br />

significant harm’ aspects of nuclear<br />

energy.<br />

This is one step in the process. The JRC report was<br />

reviewed by experts on radiation protection and waste<br />

management under Article 31 of the Treaty establishing<br />

the European Atomic Energy Community (Euratom<br />

Treaty), as well as by experts on environmental impacts<br />

from the Scientific Committee on Health, Environ mental<br />

and Emerging Risks. The experts’ reviews are published on<br />

the Commission’s website. 1<br />

The Commission is carefully analysing the findings of<br />

the report, the reviews by the experts, as well as all of the<br />

other extensive feedback submitted by other interested<br />

stakeholders.<br />

The Commission will complete its assessment in prompt<br />

fashion and will follow up in accordance with the steps laid<br />

out in the Commission’s Communication on a Strategy<br />

<strong>for</strong> Financing the Transition to a Sustain able Economy 2<br />

published in July 2021.<br />

As stated in the a<strong>for</strong>e-mentioned Communication,<br />

the Commission will adopt a complementary Climate<br />

Taxonomy Delegated Act covering activities not yet covered<br />

in the first EU Taxonomy Climate Delegated Act, notably<br />

certain energy sectors, in line with the requirements of the<br />

Taxonomy Regulation.<br />

The complementary<br />

Delegated Act will also<br />

cover nuclear energy activities,<br />

subject to and consistent<br />

with the specific expert review<br />

process that the Commission<br />

set out <strong>for</strong> this purpose.<br />

The complementary Delegated<br />

Act will also cover nuclear energy<br />

activities, subject to and consistent<br />

with the specific expert review<br />

process that the Commission set out<br />

<strong>for</strong> this purpose. The Commission<br />

will adopt the complementary<br />

Delegated Act as soon as possible<br />

after the end of the specific review<br />

process focused on nuclear.<br />

1 EU taxonomy <strong>for</strong> sustainable activities | European Commission (europa.eu).<br />

2 https://ec.europa.eu/info/publications/210706-sustainable-finance-strategy_en<br />

Interview<br />

“EU Member States Can Choose Their Energy Sources and Can Include <strong>Nuclear</strong> in Their Energy Mix as Part of Their Ef<strong>for</strong>t to Achieve Decarbonisation and Carbon Neutrality by 2050.” ı Massimo Garribba

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Setting aside the Commission proposal in a<br />

complementary Delegated Act, what will in your<br />

opinion be the consequences <strong>for</strong> the European<br />

nuclear industry in either case, i. e. inclusion in or<br />

exclusion from the taxonomy?<br />

It will have an influence on the range of (private) financing<br />

options available to nuclear operators to finance their<br />

projects.<br />

An activity that does not qualify as a green economic<br />

activity under the EU Taxonomy is not necessarily<br />

unsustainable, given the need to make a ‘substantial<br />

contribution’ to one of the six objectives and do no<br />

significant harm (DNSH) to the other five.<br />

The EU Taxonomy is designed to guide market<br />

participants in their investment decisions, and it does not<br />

prohibit investment in any activity. There is no obli gation<br />

<strong>for</strong> companies to be Taxonomy- aligned.<br />

Financial market participants can choose to invest<br />

in companies that carry out activities that have different<br />

degrees of environmental per<strong>for</strong>mance, including<br />

activities that do not comply with the EU Taxonomy<br />

criteria.<br />

In addition, the EU Taxonomy is a dynamic tool and<br />

it will evolve as technologies advance and as our<br />

understanding of solutions progresses. The Taxonomy<br />

Regulation is open to review every three years and may allow<br />

the inclusion of activities not covered initially, provided<br />

technological progress allows <strong>for</strong> market entry in the near<br />

future.<br />

If nuclear in the end would not be considered as a<br />

sustainable investment, could there be and should<br />

there be a compensation mechanism in the Euratom<br />

framework <strong>for</strong> those member states that want to<br />

pursue nuclear power <strong>for</strong> decarbonization, say in<br />

<strong>for</strong>m of financial assistance?<br />

EU Member States can choose their energy sources and<br />

can include nuclear in their energy mix as part of their<br />

ef<strong>for</strong>t to achieve decarbonisation and carbon neutrality by<br />

2050.<br />

The European Commission, in line with the<br />

Euratom Treaty, supports actions to improve the safety of<br />

nuclear installations, including<br />

research on safety, security, waste<br />

management, and assistance on<br />

nuclear decommis sioning. However,<br />

it does not provide financial<br />

support <strong>for</strong> the construction of<br />

new nuclear fission power plants.<br />

Member States opting <strong>for</strong> nuclear<br />

energy will of course underline that<br />

nuclear is a cleaner, more af<strong>for</strong>dable<br />

and much more reliable source of<br />

energy than imported fossil fuels.<br />

Apart from the well-known trenches between the<br />

ardent opponents and supporters of nuclear power<br />

among the member states, there are remarkable<br />

developments going on in countries that have no<br />

or a low profile on nuclear. Are there Euratom<br />

policies to eventually support newcomer countries<br />

like Poland or countries that might endeavour a<br />

major expansion of their nuclear sector such as<br />

possibly the Netherlands?<br />

The EU Treaties leave the choice of the energy mix to the<br />

individual Member States.<br />

Based on this and the Euratom Treaty, it is the Member<br />

State which takes the decision to introduce nuclear power<br />

in its energy mix and bears the ultimate responsibility <strong>for</strong><br />

its safety and security.<br />

Once such decision is taken, the Member State must<br />

fully comply with all relevant provisions of EU rules, i.e. all<br />

primary and secondary legislation on inter alia nuclear<br />

safety, radiation protection, management of spent fuel and<br />

radioactive waste, and non-proliferation.<br />

The European Commission can certainly advise<br />

Member States in navigating the different requirements.<br />

The Commission’s primary role is to ensure that all<br />

Member States effectively fulfil their obligations. To<br />

this end, the Commission will readily provide guidance to<br />

any interested Member State in ensuring compliance<br />

with EU legislation from the first preparative steps at<br />

national level.<br />

Since January, we can observe a major surge in<br />

energy prices, gas, coal, electricity and the carbon<br />

price all over Europe. The issue did not pop up in<br />

the German election campaign, but features<br />

prominently e.g. in Italy, France and also outside<br />

the EU in the UK. How might this impact the debate<br />

on climate policy, energy and nuclear power?<br />

The rise in wholesale energy prices in the EU since the<br />

summer is a big issue <strong>for</strong> the Commission, as it is also<br />

impacting consumers and companies at this sensitive<br />

moment of recovery after the Covid-19 pandemic.<br />

The Commission has published a ‘toolbox’ 3 document<br />

on 13 October, highlighting the various short and<br />

medium-term options available to national governments<br />

to ease the burden on end-users – in particular the most<br />

vulnerable consumers.<br />

The surge in prices is driven by an un<strong>for</strong>tunate<br />

constellation on the gas market – increased global demand<br />

at a time of limited supply. But latest indications are that<br />

the effects will be relatively short-term (with the market<br />

ex pected to be much more balanced by the spring).<br />

One key message from the Commission in this<br />

context is that this should not in any way put the transition<br />

to clean energy in question. On the contrary, greater<br />

investment in renewables and energy efficiency measures<br />

will reduce the EU’s dependence on imported fossil fuels<br />

and enhance our energy security, and there<strong>for</strong>e limit the<br />

chances of such a market spike being repeated.<br />

Member States opting <strong>for</strong><br />

nuclear energy will of course<br />

underline that nuclear is a cleaner,<br />

more af<strong>for</strong>dable and much more<br />

reliable source of energy than<br />

imported fossil fuels. And I note<br />

the declaration to this effect by<br />

10 Member States in early October,<br />

which in the context of rising energy prices have<br />

once again called to include nuclear energy in the EU<br />

Taxonomy to ensure energy inde pendence and security of<br />

supply.<br />

As regards the competitiveness of nuclear energy<br />

in the longer term, it is dependent on several<br />

factors, from technological development to policy<br />

and regulatory frameworks, sustain able supply chains,<br />

but also market design, business models, financing<br />

instruments, etc.<br />

Innovation will certainly be key, if nuclear is to<br />

make a larger contribution to our carbon- neutral<br />

future – both in making it competitive and keeping on<br />

implementing the con tinuous safety improvement<br />

principle.<br />

INTERVIEW 15<br />

3 https://eur-lex.europa.eu/legal-content/EN/TXT/?uri=COM%3A2021%3A660%3AFIN&qid=1634215984101<br />

Interview<br />

“EU Member States Can Choose Their Energy Sources and Can Include <strong>Nuclear</strong> in Their Energy Mix as Part of Their Ef<strong>for</strong>t to Achieve Decarbonisation and Carbon Neutrality by 2050.” ı Massimo Garribba

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

INTERVIEW 16<br />

Building a hydrogen ecosystem<br />

Participation in the Clean Hydrogen Alliance<br />

500 companies<br />

2020<br />

Focus on renewable hydrogen<br />

6 GW clean hydrogen<br />

2024<br />

1000 companies<br />

2024<br />

2000 companies<br />

2050<br />

40 GW (EU) + 40 GW (non-EU) clean hydrogen<br />

2030<br />

would be part of a carbon-free European power system<br />

largely based on renewables.<br />

In presenting on 13 October the Commission Toolbox<br />

of possible measures to address the energy price<br />

spike, Commissioner Simson underlined that we need to<br />

ensure a well-functioning and more resilient gas<br />

market framework. She underlined that gas has a<br />

role in the transition, but at the same time its<br />

contribution is bound to change in the long term. It will<br />

have to become green. And to promote this process,<br />

the Commission will present be<strong>for</strong>e the end of the year a<br />

comprehensive legislative package to decarbonise our<br />

gas and hydrogen markets by 2050. As part of this<br />

package, we will be addressing the issues of security of<br />

supply and storage.<br />

Finally, the Commission will launch new actions to<br />

improve the resilience of critical energy infrastructure to<br />

new evolving threats. This will include new rules <strong>for</strong> the<br />

cybersecurity of cross-border electricity to be published<br />

next year.<br />

Emission reduction potential in industries<br />

min. 9 million tons per year<br />

2024<br />

Investment needs in renewable hydrogen electrolysers:<br />

€ 5-9 billion<br />

2024<br />

Next to the price issue and its impact on industrial<br />

competitiveness there is also the aspect of security<br />

of supply. Many states are pursuing a coal phaseout,<br />

Germany and Belgium are phasing out nuclear,<br />

France, Sweden and Switzerland reduced nuclear<br />

capacity, the Netherlands end domestic gas production<br />

and gas storage is running low in Europe<br />

and particularly Germany <strong>for</strong> the coming winter. Is<br />

it time to complement European climate policy<br />

with a security of energy supply policy beyond<br />

current Energy Union policies?<br />

It is true that within the EU we see different policies among<br />

Member States as regards the role of nuclear power in their<br />

energy mix. Whereas<br />

The analyses conducted as part of the<br />

Commission's 2050 Long-Term Strategy<br />

confirm that in practically all models and<br />

under all scenarios nuclear energy would<br />

be part of a carbon-free European power<br />

system largely based on renewables.<br />

min. 90 million tons per year<br />

2030<br />

€ 26-44 billion<br />

2030<br />

several Member States<br />

have expressed their<br />

long-term commitment<br />

towards nuclear energy<br />

<strong>for</strong> ensuring security of<br />

energy supply and<br />

meeting climate targets,<br />

others have decided<br />

or are considering<br />

phase-out or cut-down policies of their nuclear<br />

programmes in the coming decades.<br />

The analyses conducted as part of the Commission’s<br />

2050 Long-Term Strategy confirm that in practically<br />

all models and under all scenarios nuclear energy<br />

Apart from sustainable finance there are other<br />

European policies that impact the nuclear sector,<br />

such as the Energy System Integration and the<br />

Hydrogen Strategies, the Guidelines on State aid<br />

<strong>for</strong> environmental protection and energy and the<br />

Industrial Strategy. What will we see in Commission<br />

initiatives here in the next 18 months and<br />

what will be at stake <strong>for</strong> the nuclear sector in these<br />

policies?<br />

The main energy policy initiatives that the Commission<br />

will be pursuing in the next 18 months are related to the<br />

overall ambition of the European Green Deal, following<br />

on from the proposals to revise energy efficiency and<br />

renew able energy rules already tabled in July and energy<br />

infrastructure (TEN-E) where the inter-institutional<br />

negotiations are very<br />

much advanced and<br />

agreement could be<br />

reached be<strong>for</strong>e the end<br />

of this year. Be<strong>for</strong>e the<br />

end of the year there<br />

will also be proposals<br />

to look at decarbonising<br />

Hydrogen production can<br />

also be an important<br />

element <strong>for</strong> countries<br />

considering nuclear<br />

energy in sector coupling.<br />

the gas market and establishing a hydrogen market, and a<br />

revision of the Energy Per<strong>for</strong>mance of Buildings Directive.<br />

The Commission will also publish legislative proposals <strong>for</strong><br />

reducing methane emissions. The priority next year will<br />

primarily be negotiating these proposals with the Member<br />

States and the European Parliament.<br />

While recognising that the electricity system of the<br />

future will be largely based on renewables, the Energy<br />

System Integration Strategy does not close the door to the<br />

contribution of other zero emission generation options,<br />

such as nuclear – recognising the preferences and<br />

specificities of Member States in this regard.<br />

<strong>Nuclear</strong> energy can complement renew able energy<br />

sources in the integrated energy systems.<br />

Hydrogen production can also be an important<br />

element <strong>for</strong> countries considering nuclear energy in<br />

sector coupling.<br />

As mentioned be<strong>for</strong>e, the Commission’s over arching<br />

priority is to ensure that Member States choosing to use<br />

nuclear energy do so within the applicable Euro pean legal<br />

framework, meeting the highest standards on nuclear<br />

safety, on safe and responsible radioactive waste management<br />

and on radiation protection.<br />

Interview<br />

“EU Member States Can Choose Their Energy Sources and Can Include <strong>Nuclear</strong> in Their Energy Mix as Part of Their Ef<strong>for</strong>t to Achieve Decarbonisation and Carbon Neutrality by 2050.” ı Massimo Garribba

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

The path towards a European hydrogen eco-system step by step :<br />

Today - 2024 2025 - 2030<br />

2030 -<br />

H 2<br />

INTERVIEW 17<br />

From now to 2024, we will<br />

support the installation of<br />

at least 6GW of renewable<br />

hydrogen electrolysers in<br />

the EU, and the production<br />

of up to 1 million tonnes of<br />

renewable hydrogen.<br />

From 2025 to 2030,<br />

hydrogen needs to become<br />

an intrinsic part of our<br />

integrated energy<br />

system, with at least 40GW<br />

of renewable hydrogen<br />

electrolysers and the<br />

production of up to<br />

10 million tonnes of<br />

renewable<br />

hydrogen in the EU.<br />

From 2030 onwards,<br />

renewable<br />

hydrogen will be<br />

deployed at a large<br />

scale across all<br />

hard-to-decarbonise<br />

sectors.<br />

The most immediate energy project is the Fit <strong>for</strong><br />

55 package on implementing the Green Deal till<br />

2030. Despite the long and short term decarbonization<br />

ambitions of the EU and member states we<br />

see practical policies running in the opposite<br />

direction such as the German phase-out of nuclear<br />

power with insufficient compensation by low carbon<br />

power sources and the new Belgian policy of<br />

replacing nuclear with gas fired power plants. Will<br />

we see some kind of Maastricht mechanism on<br />

climate policy or a European carbon semester in the<br />

future?<br />

I can’t immediately see an appetite within the Commission<br />

– or among EU leaders – to re-open the EU treaties and<br />

redefine Article 194 of the treaty.<br />

Already with the National<br />

<strong>Nuclear</strong> energy<br />

can complement<br />

renewable energy<br />

sources in the<br />

integrated energy<br />

systems.<br />

Energy and Climate Plans<br />

(NECP), which were introduced<br />

in the Clean Energy <strong>for</strong><br />

all Europeans package, the<br />

EU has introduced a level of<br />

transparent <strong>for</strong>ward planning<br />

that has not previously<br />

been seen.<br />

The key point here is<br />

that our absolute priority is to reduce greenhouse gas<br />

emissions by 55% by 2030. And this is being addressed<br />

in the Fit <strong>for</strong> 55 package. The more ambitious targets<br />

proposed <strong>for</strong> the new Directives on Energy Efficiency and<br />

Renewable Energy will require legally-binding commitments<br />

from Member States – and a monitoring process.<br />

But there is no one-size-fits-all approach.<br />

The Commission can also play a role in accompanying<br />

the EU nuclear industry in improving its competitiveness<br />

and better integrating the EU energy system of the future,<br />

by securing the application of the highest safety standards<br />

and supporting the regulatory processes in EU Member<br />

States opting <strong>for</strong> nuclear energy.<br />

Author<br />

Nicolas Wendler<br />

Head of Media Relations and Political Affairs<br />

KernD (Kerntechnik Deutschland e.V.)<br />

nicolas.wendler@kernd.de<br />

Interview<br />

“EU Member States Can Choose Their Energy Sources and Can Include <strong>Nuclear</strong> in Their Energy Mix as Part of Their Ef<strong>for</strong>t to Achieve Decarbonisation and Carbon Neutrality by 2050.” ı Massimo Garribba

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

18<br />


Investigations of the Tailskin Seal<br />

During the Retrieval Concept ‘Shield<br />

Tunnelling with Partial-face Excavation’<br />

in the Asse II Mine<br />

Birte Froebus<br />

Introduction The Asse II mine is a <strong>for</strong>mer salt mine near Remlingen where potash salt was mined from 1909 to<br />

1925 and rock salt from 1916 to 1964. After salt mining ceased, about 126,000 containers of low- and intermediate- level<br />

radioactive waste were stored between 1967 and 1978. Due to its high degree of excavation, the mine workings were<br />

damaged by pressure from the overburden. Saturated access water is penetrating through the cracks that have<br />

developed. In addition, the stability of the mine is deteriorating. The radionuclides from the present damaged waste<br />

packages could be transported into the environment via resulting pathways and transport media.<br />

Due to the situation described above,<br />

long-term safety can only be guaranteed<br />

by retrieving the radioactive<br />

waste according to the current state of<br />

knowledge, which is why this has<br />

been stipulated by law since 2013. The<br />

adopted law ‚Lex Asse‘ creates an<br />

important legal basis <strong>for</strong> the retrieval<br />

of radioactive waste. Through simplified<br />

procedures and the possibility of<br />

carrying out work in parallel, the law<br />

makes it possible to accelerate the<br />

work. In addition, the public’s right<br />

to comprehensive in<strong>for</strong>mation is<br />

strengthened. Despite all this, it<br />

should be mentioned that no practical<br />

experience has been gained so far <strong>for</strong><br />

such retrieval from deep geological<br />

<strong>for</strong>mations. In the event that the<br />

de<strong>for</strong>mations caused by the rock<br />

pressure and the resulting solution<br />

influx exceed a manageable level and<br />

the mine has to be abandoned, there<br />

are explicit emergency measures. In<br />

this scenario, among other things, a<br />

targeted counterflooding of the entire<br />

caverns is initiated – in this case, the<br />

radioactive waste must remain in the<br />

mine and could not be retrieved. [1]<br />

The radioactive waste in the Asse II<br />

mine is placed in a total of 13 ‘emplacement<br />

chambers’ (German: Einlagerungskammer<br />

= ELK) on three<br />

levels at depths of 511 m, 725 m and<br />

750 m, which are shown in red in<br />

Figure 1. ELK 8a/511 is located on the<br />

511-m-level and ELK 7/725 on the<br />

725-m-level. The other eleven emplacement<br />

chambers 1/750, 2/750,<br />

2/750 Na2, 4/750, 5/750, 6/750,<br />

7/750, 8/750, 10/750, 11/750 and<br />

12/750 are located on the 750-mlevel.<br />

[1]<br />

In this article, several terms related<br />

to mining are used to describe the<br />

mine. The emplacement chambers are<br />

located on several levels, so to say the<br />

| Fig. 1.<br />

Emplacement chambers on different levels in the mine workings of the Asse II mine [1].<br />

floors of the mine. Between the<br />

chambers there are so-called pillars,<br />

i.e. the walls in the vertical direction,<br />

and the so-called roofs, i.e. the mighty<br />

ceilings in the horizontal direction.<br />

The pillars and the roofs are the<br />

remaining rock after the chambers<br />

have been created, which now<br />

represents the load-bearing structure<br />

in the mine building. Analogous to<br />

tunnelling, within the chambers the<br />

bottom is called the ‚floor‘ and the<br />

ceiling is called the ‚roof‘.<br />

Each of the three levels offers a different<br />

spatial situation, resulting in<br />

fundamentally different strategies <strong>for</strong><br />

retrieval:<br />

p ELK 7/725 is still accessible and is<br />

currently used as a storage facility<br />

<strong>for</strong> radioactive waste produced<br />

during operations. In addition,<br />

low-level radio active material is<br />

stored there and the chamber is<br />

continuously ventilated. There<strong>for</strong>e,<br />

the retrieval of this single<br />

emplacement chamber is to be<br />

started as planned. For this<br />

purpose, a ‚Long-front construction<br />

method with vertical excavation<br />

direction‘ is aimed at, which<br />

is based on mining extraction<br />

methods. For the recovery of the<br />

casks, i.e. <strong>for</strong> un covering, loosening<br />

and loading; floor-operated<br />

technology is to be used. Transport,<br />

on the other hand, is to be<br />

carried out using decoupled<br />

ceiling-operated technology. [1]<br />

p In ELK 8a/511, almost exclusively<br />

casks with intermediate-level<br />

radio active contents were stored.<br />

This is also the reason <strong>for</strong> the<br />

special feature of this emplacement<br />

chamber: a charging<br />

chamber on the 490-m-level above<br />

the emplacement chamber, from<br />

which the casks were lowered<br />

into the emplacement chamber.<br />

Assuming that the emplacement<br />

chamber can be driven on the floor,<br />

only floor-based technology is to be<br />

used. [1]<br />

p The 750-m-level contains the<br />

majority volume of the radioactive<br />

Decommissioning and Waste Management<br />

Investigations of the Tailskin Seal During the Retrieval Concept ‘Shield Tunnelling with Partial-face Excavation’ in the Asse II Mine ı Birte Froebus

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

waste, with eleven emplacement<br />

chambers containing low-level<br />

radioactive waste (see Figure 2).<br />

Due to the different rockmechanical<br />

boundary conditions, the<br />

chambers are divided into three<br />

groups here. Here, the roofs in<br />

particular play a decisive role. In<br />

the central part of the mine<br />

buildings lies ELK 2/750 Na2,<br />

directly below ELK 7/725 and a<br />

roof of 6 m thickness. This roof is<br />

considered intact <strong>for</strong> retrieval, as<br />

ELK 7/725 will have been cleared<br />

and backfilled by that time. The<br />

roofs of the emplacement chambers<br />

of the chamber group ‚East‘<br />

(ELK 1/750, 2/750, 12/750) are<br />

considered intact, as there are no<br />

voids above them on the 725-mlevel.<br />

The remaining seven<br />

emplacement chambers of the<br />

chamber group ‚South‘ are closer<br />

to the overburden of the southern<br />

flank and probably have unstable<br />

roofs. In contrast to the 511-m- and<br />

725-m-levels, there is no final<br />

planning <strong>for</strong> the 750-m-level,<br />

instead there are several different<br />

approaches. On one hand, there<br />

are various possibilities to<br />

approach the chambers individually,<br />

on the other hand, there is a<br />

variant to approach the chambers<br />

one after the other along a curve.<br />

In the latter variant, only the<br />

chamber groups ‚South‘ and ‚East‘<br />

can be taken into account, and a<br />

study has already been carried out<br />

on this issue. [1] In the further<br />

course of the article, only aspects<br />

of this retrieval concept will be<br />

dealt with.<br />

Retrieval of the chamber<br />

groups ‚South‘ and ‚East‘<br />

(750-m-level) by means of<br />

‘Shield tunnelling with<br />

partial-face excavation’<br />

In the ‘Study on the Suitability and<br />

Development Needs of Equipment<br />

and Tools <strong>for</strong> Use in the Asse II Mine’<br />

in 2015, shield tunnelling with<br />

partial- face excavation was already<br />

examined as a possible retrieval<br />

method. Shield machines are normally<br />

used in mechanised tunnel<br />

construction and are characterised by<br />

the fact that the entire processes<br />

take place in the protection of an<br />

enveloping shield. In Figures 3, 4 and<br />

5, the shield is shown in green, while<br />

the rear part of the shield, the<br />

so-called tailskin, is shown in grey.<br />

Partial-face excavation in particular<br />

is used <strong>for</strong> very easily detachable rock<br />

in stable rock, short heading lengths<br />

| Fig. 2.<br />

Ground plan of the construction method in relation to the emplacement chambers of the 750-m-level [2].<br />

| Fig. 3.<br />

Conveying route of the transport container from the tunnel face, to the excavation chamber and to the<br />

disposal tunnel [2].<br />

and non-circular cross-sections. One<br />

advantage of partial-face excavation<br />

is the good accessibility to the tunnel<br />

face. The tunnel face describes the<br />

material in front of the excavation<br />

tools of the machine and is located on<br />

the left in Figures 3 and 4. This<br />

accessi bility is required when there is<br />

a high probability of encountering<br />

obstacles. [3]<br />

It can be seen in the site plan of<br />

Figure 2, the emplacement chambers<br />

of the chamber groups ‚South‘ and<br />

‚East‘ are located next to each other<br />

along a curve on the 750-m-level. For<br />

this reason, five parallel tunnelling<br />

routes were prioritised in the concept<br />

‚Shield tunnelling with partial-face<br />

excavation‘, which encompasses the<br />

different widths of the emplacement<br />

chambers and the pillars in between<br />

in their total width and run along this<br />

curve in their length. Five partial-face<br />

excavation machines can drive up the<br />

drifts with a time delay over different<br />

distances from east to west and<br />

completely clear the similarly high<br />

emplacement chambers by taking into<br />

account an overprofile. The overprofile<br />

is necessary to completely<br />

cover and remove the damaged and<br />

possibly contaminated inner sides of<br />

the emplacement chambers. Assembly<br />

and disassembly caverns must be<br />

constructed at the beginning and end<br />

of each driveway.<br />

A partial-face excavation machine<br />

there<strong>for</strong>e alternately passes the pillars<br />

between the emplacement chambers,<br />

which are made of salt rock, and the<br />

contents of the emplacement chambers<br />

themselves. The contents of the<br />

emplacement chambers and their<br />

structure vary from chamber to<br />

chamber.<br />

The waste packages were tipped as<br />

well as stacked horizontally or vertically,<br />

depending on the storage or<br />

discharge technique. Salt was blown,<br />

tipped or dumped to bed and shield<br />

the radioactive waste packages. Both<br />

partial and full backfills were produced<br />

with the crushed salt, also to<br />

stabilise the surrounding rock. Due to<br />

convergences as well as post-fractures<br />

of the surrounding salt rock, the<br />

solution influxes that occurred over<br />

the past decades and the resulting<br />

corrosion phenomena on the containers,<br />

the conditions within the<br />

emplacement chambers are unclear.<br />

The corroded casks themselves may<br />


Decommissioning and Waste Management<br />

Investigations of the Tailskin Seal During the Retrieval Concept ‘Shield Tunnelling with Partial-face Excavation’ in the Asse II Mine ı Birte Froebus

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


also be de<strong>for</strong>med, burst as well as<br />

shattered, and it is assumed that there<br />

is a matrix of (possibly solidified)<br />

crushed salt around the casks. [2]<br />

Partial-face excavation<br />

machine<br />

At the tunnel face (left in Figure 3)<br />

the salt rock of the pillars or the<br />

packages inside the emplacement<br />

chambers can be loosened and lifted<br />

with the help of the tools. A vacuum<br />

can be created between the tunnel<br />

face and the machine to prevent<br />

contamination from being carried<br />

into the interior of the machine via<br />

radioactive particles in the air. [2]<br />

The driving of the emplacement<br />

chambers and the pillars creates a<br />

cavity that has to be backfilled behind<br />

the machine. For this purpose,<br />

sorel concrete is placed both as<br />

shotcrete and as extruded concrete<br />

(see Figure 5). In this way, the rock<br />

can be stabilised and at the same time<br />

a structure can be created on which<br />

the partial-face excavation machine<br />

can support itself <strong>for</strong> further travel.<br />

In this so-called ‚cavity filling‘ (see<br />

| Fig. 4.<br />

Rear view of the partial-face excavation machine with cavity filling [2].<br />

Figure 4), two tunnel tubes are cut<br />

out <strong>for</strong> the period of retrieval. After<br />

suitable repacking <strong>for</strong> transport, the<br />

materials can be removed through the<br />

larger tunnel tube (disposal tunnel)<br />

(see also Figure 3), the smaller tube<br />

serves as access <strong>for</strong> personnel and <strong>for</strong><br />

the transport of working materials<br />

( passenger and utility tunnel). After<br />

successful retrieval and salvage of the<br />

partial-face excavation machines, the<br />

two tubes can be backfilled. Both the<br />

cavity filling and the backfilling of<br />

the tunnel tubes (transportation<br />

tunnels) are made of sorel concrete.<br />

[2] Sorel concrete has already been<br />

used <strong>for</strong> the production of flow<br />

barriers and various backfillings in the<br />

Asse II mine. The main components<br />

of sorel concrete are magnesium<br />

oxide as a binder and crushed salt as<br />

aggregate, which are mixed with<br />

magnesium chloride solution. [4]<br />

Once the tunnel face has been<br />

completed, the machine can support<br />

itself against the cavity filling by<br />

means of thrust cylinders and move<br />

<strong>for</strong>ward. When advancing, the<br />

annular gap that becomes free behind<br />

| Fig. 5.<br />

Internals and machine equipment of the partial-face excavation machine, e.g. thrust cylinders and<br />

( inner) tailskin seal [2].<br />

the tailskin can be filled with grout<br />

immediately. At the end of the shield<br />

shell, an (inner) tailskin seal is provided,<br />

which seals the interior of the<br />

machine against the annular gap. [2]<br />

In Figures 3, 4 and 5, the thrust<br />

cylinders in the rear part of the<br />

machine (tailskin) are shown in red,<br />

the tailskin seal in blue.<br />

In the course of the in-depth<br />

investigations, an outer tailskin seal<br />

was added to the partial-face excavation<br />

machine designed in the study.<br />

This seals the steering gap against the<br />

annular gap, and prevents material<br />

from the annular gap from getting<br />

around the shield or up to the tunnel<br />

face (see Figure 6). This ensures that<br />

the annular gap is filled conclusively<br />

under pressure. This is particularly<br />

important in the area of the pillars<br />

between the emplacement chambers,<br />

as this minimises contamination<br />

carry-over between the chambers and<br />

the barrier effect of the pillars can be<br />

maintained to the maximum.<br />

Differentiated consideration<br />

of the tailskin seal<br />

Due to the technical and legal<br />

framework conditions existing in a<br />

repository, it is necessary to mirror all<br />

components and systems of a shield<br />

machine against the resulting special<br />

requirements, to evaluate their<br />

applicability and functioning, and to<br />

make adjustments if necessary.<br />

For the investigations of the two<br />

tailskin seals, a comprehensive catalogue<br />

was compiled in each case, in<br />

which various loads play an essential<br />

role. The mechanical loads on both<br />

seals are exceptionally high due to the<br />

square cross-sectional shape of the<br />

shield machine. This is intensified by<br />

damage in the roofs, which can lead to<br />

loosening on the shield skin of the<br />

machine, and thus also on tailskin<br />

seals. In addition, chemical stresses<br />

occur, mainly from the liquid sorel<br />

concrete and saturated sodium<br />

chloride solution. The influence of<br />

ionising radiation on the tailskin seals<br />

is considered negligible due to the low<br />

activity inventory.<br />

Based on this catalogue, conventionally<br />

used tailskin seals were<br />

evaluated with regard to their<br />

suitability <strong>for</strong> shield tunnelling in the<br />

Asse II mine. Different aspects have to<br />

be given priority in the requirements<br />

<strong>for</strong> the respective tailskin seal. The<br />

inner tailskin seal must primarily fulfil<br />

a barrier effect towards the trailing<br />

annular gap, usually by building up a<br />

high pressure difference between the<br />

areas to be separated. The reason <strong>for</strong><br />

Decommissioning and Waste Management<br />

Investigations of the Tailskin Seal During the Retrieval Concept ‘Shield Tunnelling with Partial-face Excavation’ in the Asse II Mine ı Birte Froebus

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

| Fig. 6.<br />

Schematic of an inner and outer tailskin seal in partial-face excavation machine with description of the<br />

steering gap and annular gap [own figure].<br />

this is the classification of radiation<br />

protection areas according to §52 of<br />

the Radiation Protection Ordinance<br />

(German: Strahlenschutzverordnung<br />

= StrlSchV), which defines the<br />

annular gap as a ‚exclusion area‘,<br />

while the interior of the machine<br />

represents a ‚controlled area‘ (see<br />

Figure 6). For the outer tailskin seal<br />

the focus is on high wear resistance to<br />

obstacles outside the machine.<br />

As a result of the investigations, an<br />

inner tailskin seal was first designed<br />

which combines various conventional<br />

design elements. The basis <strong>for</strong> this is a<br />

wire brush seal with external spring<br />

plates and three sealing chambers,<br />

which is modified with a lubricant<br />

compression line and enlarged sealing<br />

chambers. With the help of the<br />

pressurised tail seal compound, this<br />

can meet the high demands on the<br />

barrier effect. In addition, the spring<br />

plates at the rear of the annular gap,<br />

in combination with the lubricant<br />

used, make it easier to start up the<br />

machine, as brushes in contact with<br />

the liquid backfilling (grout inside the<br />

annular gap) would stick together and<br />

be damaged over time. The lubricant<br />

itself also protects against the aggressive<br />

chemical environment in the<br />

adjacent annular gap. The enlarged<br />

sealing chambers increase the protection<br />

of the sealing system with a<br />

kind of spring effect – both in case of<br />

post-fractures from the rock mass and<br />

in case of occurring convergences.<br />

Subsequently, the outer tailskin<br />

seal was designed from three rows of<br />

spring plates, which are also modified<br />

with lubricant injection lines. The<br />

arrangement of three seal rows<br />

ensures sufficient redundancy, which<br />

is why the lubricant supply was also<br />

designed redundantly, so that the<br />

outermost seal row always has an<br />

associated supply line. In order to<br />

prevent the spring plates from being<br />

torn off by obstacles – such as barrel<br />

bundles from the emplacement<br />

chambers or rock fragments from the<br />

surrounding salt rock – their bearing<br />

was embedded in the shield casing<br />

and thus protected. The number of<br />

seals in a row also ensures a certain<br />

redundancy here.<br />

Conclusion<br />

Within the scope of the investigations,<br />

comprehensive findings were<br />

obtained <strong>for</strong> the problems occurring<br />

during shield tunnelling in the Asse II<br />

mine. So far, not all components have<br />

been investigated in a differentiated<br />

manner.<br />

In further investigations, <strong>for</strong><br />

example, aspects such as the interactions<br />

between the shield machine<br />

and the surrounding salt rock or the<br />

synergies and disadvantages of the<br />

parallel tunnel routes could be<br />

determined. It has been shown that<br />

the various situations of the partialface<br />

excavation machine with<br />

changing framings – such as salt rock,<br />

already hardened sorel concrete from<br />

preceding machines or contents of the<br />

emplacement chambers – represent a<br />

special challenge. Especially when<br />

feeding from cavity filling, the<br />

stability of the lateral walls along the<br />

travel distances plays an important<br />

role. Against this background, <strong>for</strong><br />

example, the sequence of the partialface<br />

excavation machines would have<br />

to be adjusted so that the first machine<br />

starts in the direction of travel on the<br />

left in order to be able to support itself<br />

against the salt rock in the south –<br />

especially in the area of the emplacement<br />

chambers (see Figure 2). The<br />

grouting (of the annular gap) in the<br />

area of the emplacement chambers<br />

could also be considered in more<br />

in-depth investigations. A further<br />

aspect would be the compatibility of<br />

the tunnelling concept with the<br />

emergency planning <strong>for</strong> a possible<br />

flooding of the Asse mine. At present,<br />

the reduced barrier effect of the<br />

drilled-through pillars between the<br />

emplacement chambers as well as the<br />

volume of the five shield machines<br />

and their underground infrastructure<br />

represent significant disadvantages of<br />

shield tunnelling with partial-face<br />

excavation.<br />

The retrieval of radioactive waste<br />

from the Asse II mine is scheduled to<br />

begin in 2033, in this sense: Glück<br />

auf!<br />

References<br />

[1] Plan zur Rückholung der radioaktiven Abfälle aus der<br />

Schachtanlage Asse II – Rückholplan, Verfasser:<br />

Bundesgesellschaft für Endlagerung mbH Peine/Remlingen,<br />

Stand: 19.02.2020<br />

[2] Machbarkeitsstudie für die Methode „Schildvortrieb mit<br />

Teilflächenabbau“ – Studie zur Eignungsfähigkeit und zum<br />

Entwicklungsbedarf von Gerätschaften/ Werkzeugen für den<br />

Einsatz in der Schachtanlage Asse II, Herrenknecht AG im<br />

Auftrag des Karlsruher Instituts für Technologie (KIT),<br />

Technologie und Management des Rückbaus kerntechnischer<br />

Anlagen (TMRK), Schwanau, Karlsruhe; Stand: 13.05.2015<br />

[3] Maidl, B., Herrenknecht, M., Maidl, U., Wehrmeyer, G. 2011.<br />

Maschineller Tunnelbau im Schildvortrieb. Ernst & Sohn<br />

[4] Schachtanlage Asse II: Nachweis der Langzeitbeständigkeit<br />

für den Sorelbaustoff der Rezeptur A1; ERCOSPLAN, TU BAF<br />

( IFAC), IFG; Stand: 10.08.2018<br />

Author<br />

Birte Froebus<br />

Research Assistant at<br />

Karlsruhe Institute of<br />

Technology, Karlsruhe,<br />

Germany<br />

birte.froebus@kit.edu<br />

Birte Froebus studied civil engineering at the<br />

Karlsruhe Institute of Technology (KIT), specialising in<br />

geotechnical engineering. In her master thesis, she<br />

took the opportunity to combine mechanised tunnel<br />

construction with the retrieval of radioactive waste,<br />

<strong>for</strong> which she was awarded the WiN-Germany<br />

Prize 2020/21. She then started working as a<br />

research assistant at the Institute <strong>for</strong> Technology<br />

and Management in Construction (TMB) in<br />

the ’ Deconstruction and Decommissioning of<br />

Conventional and <strong>Nuclear</strong> Buildings‘ department.<br />


Decommissioning and Waste Management<br />

Investigations of the Tailskin Seal During the Retrieval Concept ‘Shield Tunnelling with Partial-face Excavation’ in the Asse II Mine ı Birte Froebus

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

22<br />


Practical Response to a Dirty Bomb<br />

James Conca<br />

Introduction The global community has recognized that a class of weapons known as radiation dispersal devices<br />

(RDDs), or dirty bombs, pose a grave threat to the United States and the European Union. Dirty bombs use conventional<br />

methods, such as a car bomb, to disperse radioactive materials in a populated economic district to cause great economic<br />

and social disruption disproportionate to their actual radiological effects and well beyond the physical destruction from<br />

their conventional bomb components.<br />

They take many <strong>for</strong>ms, from containers<br />

of radioactive materials<br />

wrapped with conventional explosives,<br />

to aerosolized materials sprayed<br />

by conventional equipment, to manual<br />

dispersion of fine powders into the<br />

environment [1]. Also included are<br />

radiation-exposure devices (REDs),<br />

used to expose people to dangerous<br />

beams or particles of radiation. RDD<br />

attacks can produce general panic,<br />

immediate death and long-term<br />

increases in cancer incidence, longterm<br />

loss of property use, disruption<br />

of services, and costly remediation of<br />

property and facilities.<br />

The risk is more than just theo retical.<br />

Several credible designs and<br />

plans <strong>for</strong> a dirty bomb attack against<br />

the United States have been found in<br />

Al Qaeda records. About 20 years ago,<br />

two actual dirty bombs were deployed<br />

by Chechen separatists; one was foiled<br />

and the other failed. In the mid-2000s,<br />

38 Alazan shoulder-fired missiles outfitted<br />

with 137 Cs dirty bomb warheads<br />

were <strong>for</strong> sale from a Moldovan arms<br />

dealer.<br />

Fortunately, few people will likely<br />

die from the radiological effects of a<br />

dirty bomb although tens to hundreds<br />

could die from the conventional blast.<br />

Un<strong>for</strong>tunately, with no precedence,<br />

we are struggling with generalizations<br />

| Fig. 1.<br />

Radiation dispersal devices (RDDs), or dirty<br />

bombs, are devices that disperse radioactive<br />

materials by any means possible. The most<br />

effective means of dispersal from a terrorist<br />

standpoint is the car bomb with several<br />

pounds of highly radioactive material that is<br />

easily dispersable, e.g., 137 CsCl powder in an<br />

Oklahoma-type ANFO car bomb. (Photo:<br />

Courtesy of Middle East Intelligence Bulletin)<br />

and how to prepare and respond to the<br />

first event. The following structures<br />

the problem and provides guidelines<br />

that are fluid and which should be <strong>for</strong>malized<br />

by the Department of Homeland<br />

Security (DHS) in conjunction<br />

with the Department of Energy (DOE,<br />

the <strong>Nuclear</strong> Regulatory Commission<br />

(NRC), the <strong>International</strong> Atomic<br />

Energy Agency (IAEA) and other cognizant<br />

agencies around the world.<br />

Rules of Thumb<br />

The challenge to training anyone<br />

outside the fields of radiation, radiochemistry<br />

and nuclear science is to<br />

provide enough in<strong>for</strong>mation to be useful<br />

without creating confusion. First<br />

responders know all too well how<br />

dangerous it is to oversimplify. However,<br />

it is not possible to make all<br />

responders radiochemists or health<br />

physicists. So some simplifying concepts<br />

are in order:<br />

1) the logistical difficulty in successfully<br />

carrying out a significant dirty<br />

bomb attack is roughly the same as<br />

that of the 9-11 attacks,<br />

2) the most likely device will be a<br />

137 CsCl car bomb and the level of<br />

response and danger to responders<br />

is of the order of a 20 alarm fire,<br />

3) defining the hot zone is the most<br />

important first response and a<br />

simple alarming dosimeter is the<br />

most useful piece of equipment <strong>for</strong><br />

a dirty bomb attack,<br />

4) following protocols, it is difficult to<br />

obtain a significant radiation dose<br />

in the first hours of response to an<br />

attack without actually handling<br />

radioactive material,<br />

5) the greater the dispersion, the<br />

greater the affected area, but the<br />

lower the dose,<br />

6) the scene will be a war zone, not a<br />

superfund or environmental cleanup<br />

site,<br />

7) it is possible to quickly triage most<br />

victims without significant harm to<br />

the responder,<br />

8) persons with no significant physical<br />

injuries should not be significantly<br />

contaminated,<br />

9) while not effective against g-radiation,<br />

the PPE of a firefighter<br />

(uni<strong>for</strong>m, gloves, goggles and/or air<br />

purifying respirator) will provide<br />

complete protection against a- and<br />

b-radiation, and almost complete<br />

protection against ingestion and<br />

inhalation of radioactive material<br />

dispersed by the attack, including<br />

g-emitting radioactive material, and<br />

10) removing clothing and washing<br />

with soap and water is effective<br />

at removing radioactive contamination.<br />

The greater the training and experience<br />

with radioactive materials, the<br />

better able the responder will be to<br />

evaluating the usefulness of such<br />

generalizations.<br />

Dirty Bomb Materials<br />

Radioactive materials are used in<br />

many fields in almost all countries<br />

around the world, particularly <strong>for</strong><br />

research, medical, and industrial<br />

applications. Dozens of radiological<br />

source producers and suppliers are<br />

found on six continents, and about a<br />

billion sources exist worldwide<br />

although most, like household smoke<br />

detectors, have such low specific<br />

activities that they pose no threat.<br />

With the increase of radioisotope<br />

applications in nuclear medicine<br />

diagnostics, therapeutics, sterilization<br />

and the food irradiation, the<br />

radio logical source production and<br />

fabri cation industry is an emerging<br />

growth industry in several countries,<br />

parti cularly in areas with depressed<br />

eco nomies. The rise in the number of<br />

terrorist acts during the last ten years<br />

has raised concerns about these<br />

radiological sources being used in<br />

RDDs. Because the general public is so<br />

frightened about anything radioactive,<br />

panic must be anticipated even<br />

if there is no likely health threat<br />

from the radioactive component. The<br />

degenerate case of a phantom RDD,<br />

where no radioactive material is used<br />

but an implication or anonymous tip<br />

indicates there is, could still cause<br />

considerable panic with large economic<br />

consequences, particularly in<br />

this era of fake news.<br />

The serious radiological threats<br />

come from large RDDs containing<br />

Environment and Safety<br />

Practical Response to a Dirty Bomb ı James Conca

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

| Fig. 2.<br />

The primary uses, specific isotope and activity levels of radiological source materials. Note that the<br />

largest sources use only 60 Co, 137 Cs or 90 Sr (Courtesy of Greg van Tuyle).<br />

thousands to hundreds of thousands<br />

of curies (Ci) of activity. These<br />

could cause significant and lasting<br />

health and contamination problems.<br />

Although many variables determine<br />

the effectiveness of an RDD attack,<br />

the key factor is the quantity and type<br />

of radiological source material that is<br />

dispersed. Although it has been<br />

difficult to quantify, globally there are<br />

about 10,000 sources that exceed<br />

1000 Ci, and perhaps a thousand that<br />

exceed 100,000 Ci.<br />

Briefly, the differences in sources<br />

relate to their specific activity (the<br />

type and amount of radiation<br />

emitted), and its chemical <strong>for</strong>m<br />

(whether it is a powder, and nonmetal<br />

solid or a metal). Gamma radiation<br />

(g) can penetrate great distances<br />

and, depending upon the energies,<br />

requires shielding of about 7 inches<br />

(18 cm) of lead (Pb) or about 3-feet<br />

(1 meter) of rein<strong>for</strong>ced concrete. Beta<br />

radiation (b) can only penetrate a<br />

short distance and the personal<br />

protective gear of a firefighter can<br />

block much of the dose. Alpha radiation<br />

(a) is the least penetrating of all<br />

and can be stopped by a piece of paper<br />

or ordinary clothing.<br />

The most important pathway of<br />

accumulating dose from a or b sources<br />

is ingestion, or particularly inhalation,<br />

where the emitter is directly adjacent<br />

to tissue <strong>for</strong> long periods of time. For g<br />

sources, mere proximity is all that is<br />

required <strong>for</strong> significant doses. For<br />

RDD discussions, isotopes of Pu, Am<br />

and U are primarily a emitters, 60 Co<br />

and 137 Cs are g emitters, and 90 Sr is a b<br />

emitter. 60 Co usually occurs as a metal<br />

(either pellets or small rods), 137 Cs is<br />

as a powder, 90 Sr is as a ceramic<br />

titanate, and Pu, Am and U are various<br />

oxides, metals, salts and non-metal<br />

solids.<br />

Although the public generally<br />

thinks of Pu and enriched-U when<br />

hearing the word radioactive, these<br />

are not considered RDD materials of<br />

choice because they are primarily a<br />

emitters, are costly, cannot be<br />

obtained in large amounts, are welltracked<br />

and secured, and are more<br />

useful to terrorists in the production<br />

of actual nuclear weapons than in<br />

being wasted in an RDD. In this sense,<br />

137 CsCl powder is much more effective<br />

as an RDD material.<br />

Although inclusion of any radioactivity,<br />

no matter how small, in a<br />

dirty bomb will cause disruption at<br />

some level, the real health and economic<br />

threat resides in large sources<br />

| Fig. 3.<br />

137 CsCl powder, presently used in the irradiation<br />

industry, is the dirty bomb material of<br />

choice. 137 Cs is inexpensive (

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


an automobile, working in the fossil<br />

fuel industry, working in the construction<br />

industry, and working in the<br />

nuclear power industry. Almost every<br />

respondent ranked working in the<br />

nuclear power industry as either the<br />

most dangerous, or second most<br />

dangerous, activity.<br />

In reality, working in a nuclear<br />

power plant is the safest job in the<br />

world. In the last five years, no one has<br />

died in the nuclear power industry,<br />

whereas over 2 million Americans<br />

have died in the other activities:<br />

smoking > 700,000; consuming<br />

alcohol > 500,000; driving an automobile<br />

> 250,000; working in the<br />

fossil fuel industry > 100,000;<br />

working in the construction industry<br />

> 5,000; working in the nuclear<br />

power industry – 1. This misperception<br />

of how dangerous radiation is<br />

constitutes the number one issue<br />

concerning the effectiveness of a dirty<br />

bomb because how responders and<br />

the public respond to a dirty bomb<br />

attack will determine whether the<br />

attack is successful or not. Of course,<br />

there is no comparison between a<br />

dirty bomb attack and anything within<br />

our past experience.<br />

There<strong>for</strong>e, short of a multi-year<br />

national public education initiative on<br />

radiation, how can we prepare and<br />

respond to this type of attack? The<br />

answer is to develop a simplified,<br />

practical guidance <strong>for</strong> responding to a<br />

radiological attack that does not<br />

depend upon in-depth understanding<br />

of radiation, that dispels the fear and<br />

panic that comes from misperceptions,<br />

and that fits into the NIMS/ICS<br />

incident command framework so that<br />

it can actually be implemented during<br />

a crisis. At the same time, we must<br />

provide sufficient resources in the<br />

<strong>for</strong>m of training, documents, web<br />

sites, and expert consultation that<br />

allows the responder, if desired, to<br />

obtain a greater depth of understanding<br />

over a reasonably short time<br />

period.<br />

The Problem <strong>for</strong> Responders<br />

Another way to think of the problem<br />

of providing first responders sufficient<br />

understanding of radiation is to<br />

reverse the scenario. Two firefighters<br />

pull a handline from an engine, open<br />

the nozzle checking <strong>for</strong> water flow<br />

and eliminating air in the line, and<br />

charging into the burning structure<br />

make an interior, offensive attack.<br />

Moments later they emerge stating<br />

that the fire has been knocked down. A<br />

passerby observing this activity would<br />

question the sanity of people rushing<br />

into a burning structure under a<br />

canopy of rolling fire just inches above<br />

their heads. The passerby does<br />

not know the numerous evaluations<br />

resulting from great experience that<br />

occurred in the firefighters’ minds<br />

prior to making the decision that it<br />

was relatively safe to enter the structure<br />

in search of the seat of the fire.<br />

The passerby would probably not<br />

have rushed into the structure, and<br />

probably would have no clue of how to<br />

find his way to the seat of the fire in<br />

zero visibility. Just how well firefighters<br />

could teach these techniques<br />

in a three-day course to someone not<br />

in the field and let them loose in<br />

that situation is open to debate.<br />

Fortunately, no one would dream of<br />

trying it.<br />

Whether a firefighter is teaching a<br />

scientist how to make an interior fire<br />

attack, or a health physicist is teaching<br />

a firefighter how to handle various<br />

radioactive materials, the challenge is<br />

<strong>for</strong> the trainer to instill in the trainee<br />

not just the knowledge of how to<br />

function, but also an understanding of<br />

why to function that way, and most<br />

importantly, the wisdom that comes<br />

from experience with risks versus<br />

benefits of doing it in different ways,<br />

resulting in good decisions. In the<br />

classroom, laboratory, and field setting,<br />

knowledge and understanding<br />

can be taught by a skilled teacher.<br />

The student must have trust in the<br />

teacher’s wisdom while building his<br />

own bank of wisdom from the outcomes<br />

of his own decisions, which<br />

makes hands-on training, regional<br />

drills, field exercises, and full-scale<br />

events such as Homeland Security’s<br />

TOPOFF training, so important.<br />

Resources <strong>for</strong> the Responder<br />

Many resources have appeared,<br />

particularly online, to fill gaps in<br />

knowledge on radiological terrorism<br />

<strong>for</strong> first responders.<br />

DOE’s national laboratories provide<br />

basic radiation training and<br />

resources, e.g., Sandia National Lab<br />

www.sandia.gov/mission/homeland/<br />

solutions/emergency/index.html,<br />

and Pacific Northwest National<br />

Lab www.hammertraining.com/, and<br />

there are many sites linked to DHS<br />

(www.dhs.gov/dhspublic/) and various<br />

private and quasi-private sites<br />

such as www.homelandresponse.<br />

org/, and the Responder Knowledge<br />

Base at http://www.rkb.mipt.org/.<br />

Sites sponsored by the federal<br />

government are built specifically to<br />

serve the needs of emergency<br />

responders and contain in<strong>for</strong>mation<br />

on currently available products, along<br />

with related in<strong>for</strong>mation such as<br />

standards, training, and grants.<br />

The Responder Knowledge Base is<br />

anticipating posting the Radiation<br />

Community Preparedness Resource<br />

(RadCPR) data base, a comprehensive<br />

data base on radiation and radiological<br />

incidents prepared by a team at<br />

Los Alamos National Laboratory.<br />

Because these resources consist of<br />

thousands of pages of in<strong>for</strong>mation,<br />

they are best utilized periodically by<br />

the first responder as ongoing education.<br />

However, <strong>for</strong> the heat of the<br />

moment, in the immediate aftermath<br />

of a radiological attack, the first<br />

responder needs a simple set of basic<br />

principles that are useful but not<br />

daunting. An excellent guide <strong>for</strong> a<br />

simplified approach recently appeared<br />

as a result of many years of experiments<br />

at Sandia (S. V. Musolino and<br />

F. T. Harper in the April 2006 issue of<br />

Health Physics vol. 90(4), p. 377-385).<br />

This approach has been incorporated<br />

into many training programs.<br />

Training <strong>for</strong> the Responder<br />

Many training programs exist <strong>for</strong> the<br />

first responder with respect to radiological<br />

incidents. Because a dirty<br />

bomb attack is closer to a radwaste<br />

spill than to any other event, some<br />

training programs have logically<br />

adapted the DOE or the NRC equivalent<br />

of RadWorker II that have been<br />

used <strong>for</strong> decades to train radiation<br />

workers in the nuclear, clean-up and<br />

disposal industries. However, the first<br />

responder is not a radiation worker,<br />

and may not need most of the in<strong>for</strong>mation<br />

contained in these trainings.<br />

What is needed is a combination<br />

and streamlining of three classes of<br />

training: RadWorker II, MERRTT,<br />

and NIMS/ICS. MERRTT (Modular<br />

Emergency Response Radiological<br />

Transportation Training Program) is<br />

the only nationally recognized first<br />

responder training <strong>for</strong> handling radiological<br />

transportation incidents, and<br />

NIMS/ICS is the general incident<br />

management system utilized by the<br />

nation’s response communities.<br />

At a minimum, any course should<br />

include equal time in classroom,<br />

laboratory and field exercises in areas<br />

including:<br />

p basic concepts of radiation physics<br />

and chemistry<br />

p biological effects of radiation<br />

p hazard recognition<br />

p characteristics of RDDs: the source<br />

and the explosives<br />

p how dirty bombs are packaged and<br />

dispersed<br />

Environment and Safety<br />

Practical Response to a Dirty Bomb ı James Conca

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

p initial response actions<br />

p incident control and command<br />

p handling the walking worried and<br />

the terrified<br />

p pre-hospital practices<br />

p situation board drills<br />

p radiological instrumentation and<br />

dosimetry devices<br />

p Department of Homeland Security<br />

Guidelines <strong>for</strong> RDD events and<br />

what that means to first responders<br />

p clean-up and ways to mitigate the<br />

effects of RDDs<br />

p site <strong>for</strong>ensics and preservation of<br />

evidence<br />

p decontamination, disposal and<br />

documentation<br />

p when to return to work and living<br />

spaces after a dirty bomb attack<br />

What differentiates these courses<br />

from most is the inclusion of first<br />

responders on the faculty, including<br />

fire chiefs, state police detectives,<br />

National Guard CST personnel, EMTs<br />

and <strong>for</strong>ensic scientists, as well as the<br />

radiochemists and health physicists.<br />

These courses are not necessarily<br />

meant <strong>for</strong> advanced radiation event<br />

responders such as the National<br />

Guard Civil Support or DOE RAP<br />

teams that will arrive to assist local<br />

responders within 12 hours of the<br />

event, but are meant to provide local<br />

responders with sufficient in<strong>for</strong>mation<br />

to provide incident control<br />

and command, determine the effected<br />

areas, address immediate concerns<br />

such as fire, priority rescue, aid<br />

citizens and medical personnel, provide<br />

support to the CST and RAP<br />

teams after they arrive, and generally<br />

keep panic to a minimum. This last<br />

point is critical; if the responders first<br />

on the scene do not understand what<br />

a dirty bomb event entails, that<br />

uncertainty will be communicated to<br />

the civilians in the area and attempts<br />

to contain the situation and the<br />

contamination may fail. Because there<br />

are over 200,000 first responders in<br />

the top 100 target areas in the United<br />

States, many such programs are<br />

needed.<br />

So what can a first responder do?<br />

The following summarizes a simplified<br />

12-point guidance (variations<br />

exist and some <strong>for</strong>m will be<br />

standardized in the near future).<br />

1. Assume all explosions or areas,<br />

particularly car/truck explosions,<br />

are dirty. <br />

2. If no dose or activity readings are<br />

available, set up an affected or<br />

exclusion zone boundary at 500 m<br />

from ground zero. If readings are<br />

available, set the full exclusion<br />

zone (around ground zero) outer<br />

boundary as about 1 rem/hr<br />

(10 mSv/hr). This boundary<br />

will also be the hot zone inner<br />

boundary. Set the hot zone outer<br />

boundary as about 0.1 rem/hr<br />

(1 mSv/hr). Within this zone,<br />

essential personnel can operate <strong>for</strong><br />

several hours without accumulating<br />

significant dose. Exact<br />

adherence may not be feasible<br />

because of logistical or geometric<br />

issues and plus or minus a factor<br />

of 2 can be expected. Set the outer<br />

boundary of the warm zone<br />

( affected area) to about 2 mrem/hr<br />

(20 μSv/hr) depending upon<br />

operability. Local decisions may<br />

warrant establishing boundaries at<br />

2x or 4x background, but these<br />

may be miles from ground zero. <br />

3. All personnel in the hot zone<br />

should wear full PPE (turnout or<br />

bunker gear) with a particulate full<br />

face mask and have an updating,<br />

alarming cumulative dosimeter<br />

that can be used to track total dose.<br />

Take any precaution necessary to<br />

avoid inhaling or ingesting dust<br />

and particulates. Radioactivity will<br />

be in particulate <strong>for</strong>m. <br />

4. When it is determined the situation<br />

is radiological, immediately<br />

alert the appropriate secondary<br />

response teams: CST, RAP and FBI,<br />

as advised in the unified command<br />

protocols <strong>for</strong> your region. If<br />

necessary in the U.S., call:<br />

p National Response Center<br />

1-800-424-8802<br />

p National Guard CST<br />

1-800-343-6701<br />

p FBI (ATF bomb)<br />

1-888-283-2662<br />

p DOE (RAP Coordinator)<br />

1-505-845-4667<br />

p NRC1-301-816-5100<br />

p DHS 1-202-727-6161<br />

p FEMA 1-202-586-8100<br />

In the European Union, contact<br />

the IAEA. Specific agencies are<br />

not yet assembled EU-wide <strong>for</strong> this<br />

purpose, although coordinated<br />

planning is underway. Each<br />

Member State has its own radiation-monitoring<br />

system or network<br />

which usually covers its whole<br />

territory. The density of radioactivity<br />

measuring points is<br />

variable – from several, uni<strong>for</strong>mly<br />

distributed over the territory to<br />

hundreds of measuring stations,<br />

with increased density close to<br />

nuclear installations. EURDEP<br />

makes radiological monitoring<br />

data from most European countries<br />

available on a routine daily<br />

basis – and in close to real-time in<br />

emergencies. To achieve this,<br />

EU Member States, and other<br />

European countries which are<br />

members of EURDEP, send their<br />

data to EURDEP on a daily basis<br />

from at least one territorial radiation-monitoring<br />

network (some<br />

countries have more than one).<br />

The EURDEP system makes this<br />

radiological data available via a<br />

web page.<br />

5. Occupancy time outside the hot<br />

zone but within the warm zone is<br />

unrestricted <strong>for</strong> essential personnel<br />

during initial response (days –<br />

weeks). Establish Incident Command<br />

upwind of ground zero at<br />

the closest point outside the<br />

affected zone. Have alternative<br />

positions ready in case of change in<br />

wind direction. <br />

6. Evacuate all people from the<br />

affected area (> 2 mrem/hr) and<br />

exclude non-essential personnel<br />

thereafter. Expect self-evacuation<br />

<strong>for</strong> large populations of uninjured<br />

persons and provide them with<br />

safe designated routes out of the<br />

affected area (work with building<br />

managers to establish subterranean<br />

routes). Try to establish<br />

quick dose-rate screening, or<br />

radiological monitors, to determine<br />

those relatively few needing<br />

decontamination, but do not<br />

attempt mass decon of large populations.<br />

Instead, advise removal of<br />

external clothing, bag if possible,<br />

avoid eating, drinking or touching<br />

facial region, go directly home,<br />

shower with warm water and soap,<br />

and do not use hair conditioner,<br />

hair color, or other fixative hygiene<br />

products. Local decisions may<br />

warrant establishing large fire hose<br />

wash down curtains along decon<br />

corridors <strong>for</strong> rapid decon of<br />

evacuees and equipment, however,<br />

in large urban settings this will not<br />

be feasible. <br />

7. Do not decontaminate vehicles or<br />

structures during the initial<br />

response phase. Do not try to<br />

contain contaminated water, but<br />

allow, or even encourage, it to<br />

enter the municipal stormwater<br />

drainage system. Alert City<br />

Manager or wastewater treatment<br />

facility manager <strong>for</strong> possible<br />

diversion strategies. <br />

8. For those heavily contaminated<br />

persons, e.g., where there is<br />

obvious surface radioactive<br />

material or where they are heavily<br />

injured from the blast, establish<br />

decon areas and decon corridors<br />

connecting the hot zone to the<br />


Environment and Safety<br />

Practical Response to a Dirty Bomb ı James Conca

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


boundary of the warm zone or<br />

affected area. Provide those with<br />

heavy external contamination of<br />

the upper body with follow-up<br />

exams to determine possible<br />

contaminant inhalation or<br />

ingestion. Countermeasures, e.g.,<br />

Prussian Blue, should be evaluated<br />

promptly. <br />

9. Separate persons needing immediate<br />

medical attention and<br />

remove outer garments, survey <strong>for</strong><br />

surface contamination, decon if<br />

necessary and possible, wrap in<br />

clean blankets in decon zone and<br />

evacuate. In<strong>for</strong>m the receiving<br />

medical facility that the person has<br />

little or no surface contamination<br />

or they may deny admittance.<br />

10. Commence mapping the affected<br />

area to obtain a rough dose profile<br />

of the area, marking hot and cold<br />

spots to assist in avoiding large<br />

doses during operations, and to<br />

assess the magnitude of the situation.<br />

<br />

11. Essential personnel within the<br />

affected area should record cumulative<br />

dose, if possible, and not<br />

exceed about 5 rem (50 mSv) total<br />

unless protection of critical infrastructure<br />

is deemed imperative<br />

and no alternative exists. Do not<br />

exceed about 10 rem (100 mSv)<br />

except to save lives and protect<br />

critical infrastructure. Note: no<br />

health effects ever observed <strong>for</strong><br />

doses less than 10 rem. Do not<br />

exceed about 25 rem (250 mSv)<br />

unless the responder decides<br />

voluntarily, and with full<br />

knowledge of the risks, to save<br />

large numbers of lives and protect<br />

critical infrastructure that may<br />

harm large populations if not<br />

secured. Do not exceed about<br />

50 rem (500 mSv). <br />

12. Sheltering in place is advisable if<br />

the population is aware of the<br />

radiological nature ahead of the<br />

plume, unlikely in most<br />

cases. Evacuate buildings along<br />

determined safe routes away from<br />

the hot zone. Do not shut down<br />

building ventilation systems.<br />

Modern ventilation systems will<br />

filter most radioactive particulates<br />

and shut down may cause chimney<br />

effects. <br />

Once an attack has happened, there<br />

are limited options <strong>for</strong> mitigation.<br />

Spray-on fixatives are being investigated<br />

to prevent secondary migration<br />

and to make subsequent clean up<br />

easier. These may be ideal <strong>for</strong> heavily<br />

affected areas such as the immediate<br />

blast area, and <strong>for</strong> specific source<br />

materials such as a-emitters, but the<br />

best option may result from the fact<br />

that 137CsCl is so soluble that it can<br />

be washed off of surfaces with water.<br />

However, wash down must be done<br />

quickly and completely, within days<br />

of the event, to preclude further<br />

effects such as diffusion into building<br />

materials, secondary migration, and<br />

cumulative dose effects.<br />

Diffusion rates are primarily a<br />

function of moisture content and<br />

depend strongly upon weather conditions<br />

and porosity of the materials.<br />

If the weather remains dry and sunny,<br />

little diffusion will occur, but if the<br />

surfaces become wet (but not enough<br />

to wash off the Cs), or if surfaces are<br />

wet during deposition, then significant<br />

diffusion can occur quickly. On a<br />

wet surface, 137Cs can diffuse into<br />

concrete more than a quarter of an<br />

inch each week, but it would not<br />

diffuse that much into the granite,<br />

glass or metal even after several years,<br />

no matter what the conditions.<br />

There is considerable debate over<br />

the wash down approach, but it is<br />

unlikely any other strategy can be<br />

implemented rapidly enough to be<br />

affective. Although a 10 by 10 block<br />

area in downtown Manhattan has<br />

approximately one billion square feet<br />

of surface area, one hundred fire<br />

hydrants operating <strong>for</strong> 24 hours<br />

delivers about one hundred million<br />

gallons of water, sufficient to wash off<br />

large areas and wash most of the Cs<br />

into the stormwater drainage system<br />

where it will be sufficiently diluted<br />

and deposited in areas of lesser consequence.<br />

Alternatively, the wash water<br />

can be treated at the outflow points<br />

using inexpensive materials such as<br />

gabions of zeolitic gravel ($80/ton),<br />

that are extremely specific <strong>for</strong> Cs and<br />

other radionuclides.<br />

It is essential that the United States<br />

and the European Union responds<br />

quickly and efficiently should an RDD<br />

attack occur in order to minimize the<br />

long-term effects and deter future<br />

RDD attacks by showing a high-cost/<br />

low-benefit to terrorist groups <strong>for</strong> such<br />

weapons. But this can be accomplished<br />

only if emergency response agencies<br />

are well trained and have suitable<br />

plans that can be executed within a<br />

comprehensive interagency command<br />

structure.<br />

Author<br />

Dr. James Conca<br />

Senior Scientist<br />

UFA Ventures, Inc.<br />

Richland, USA<br />

jim@ufaventures.com<br />

Geochemist and Energy scientist, speaker and author<br />

Dr. James Conca is Senior Scientist <strong>for</strong> UFA Ventures,<br />

Inc. in the Tri-Cities, Washington, a Trustee of the<br />

Herbert M. Parker Foundation, an Adjunct Professor<br />

at Washington State University in the School of the<br />

Environment, an Affiliate Scientist at Los Alamos<br />

National Laboratory and a Science Contributor to<br />

Forbes on energy and nuclear issues. Conca obtained<br />

a Ph.D. in Geochemistry from the Cali<strong>for</strong>nia Institute<br />

of Technology in 1985, an MS in Planetary Science in<br />

1981, and a Bachelors in Geology and Biochemistry<br />

from Brown University in 1979.<br />

Environment and Safety<br />

Practical Response to a Dirty Bomb ı James Conca

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Equipment Selection Methodology<br />

of Seismic Probability Safety Assessment<br />

<strong>for</strong> <strong>Nuclear</strong> <strong>Power</strong> Plant<br />

Junghyun Ryu and Moosung Jae<br />

1 Introduction A typical seismic probability safety assessment (SPSA) requires geological, structural, and<br />

systems analysts. In the case of probabilistic seismic hazard analysis (PSHA), geological experts can independently<br />

per<strong>for</strong>m the analysis, but structural experts per<strong>for</strong>ming seismic fragility analysis (SFA). Probability safety assessment<br />

(PSA) model analysts per<strong>for</strong>ming system analysis, cannot per<strong>for</strong>m the analysis independently. Only when the analysis<br />

is per<strong>for</strong>med through mutual co-operation between the two fields can a reasonable result be achieved. This is a unique<br />

characteristic of SPSA.<br />

For SPSA, a seismic equipment list<br />

(SEL) is first prepared, and reviews of<br />

the design, construction installation<br />

data, walkdown and system are<br />

conducted to analyze structures,<br />

systems and components (SSCs) that<br />

have little or no effect on the safety of<br />

the entire power plant. The final<br />

selected SSCs are reflected in the<br />

PSA model to per<strong>for</strong>m SPSA. The<br />

following are the considerations in<br />

selecting a general SEL [1]:<br />

p Identify SSCs that are important to<br />

safe shutdown from full-power<br />

PRA models.<br />

p Identify SSCs from a review of<br />

seismic evaluation per<strong>for</strong>med <strong>for</strong><br />

the IPEEE (Individual Plant Examination<br />

<strong>for</strong> External Events).<br />

p Identify structures and passive<br />

components that are important to<br />

the seismic response.<br />

p Identify additional SSCs from a<br />

plant walkdown.<br />

As can be seen from the above SEL<br />

selection criteria, equipment is required<br />

systematically <strong>for</strong> the safe shutdown<br />

of a nuclear power plant, and<br />

critical equipment vulnerable to earthquakes<br />

is selected through seismic<br />

response analysis and site surveys. In<br />

this paper, we propose the equipment<br />

selection methodology that should be<br />

essential to SPSA using the characteristics<br />

of individual plants and internal<br />

events PSA <strong>for</strong> existing qualitative and<br />

ambiguous selection criteria.<br />

2 Practices <strong>for</strong> seismic<br />

equipment selection and<br />

screening criteria<br />

In this paper, we review the practice of<br />

SEL selection and characteristics<br />

based on cases applied in Korea and<br />

the United States. In Korea, when an<br />

SPSA is per<strong>for</strong>med, the SEL is selected<br />

in consideration of the general points<br />

mentioned in Chapter 1. After preparing<br />

the SEL, special screening<br />

criteria are set, which will be reviewed<br />

in detail here. Quantitative screening<br />

means that SSCs, which generally<br />

establish and meet appropriate<br />

screening criteria that are expected<br />

not to affect overall plant safety, are<br />

excluded from the SPSA model, and<br />

reflect and analyze only equipment<br />

and buildings that do not meet the<br />

screening criteria of the SPSA model.<br />

For example, there are two<br />

screening criteria: the first is the<br />

specific median acceleration capacity<br />

(A m ), and the second is the high<br />

confidence of a low probability of<br />

failure (HCLPF). First, if A m is set as<br />

the screening criteria, it should not<br />

contribute to plant safety risk regardless<br />

of the results of the site’s PSHA, so<br />

a relatively high value should be set as<br />

the screening criteria, which can be an<br />

overly conservative analysis. Moreover,<br />

there is a weakness in that there<br />

is no difference in core damage<br />

frequency (CDF), which is a risk metric<br />

<strong>for</strong> SPSA between a power plant with a<br />

high probability of seismic occurrence<br />

on the site and a power plant with a<br />

low probability of seismic occurrence.<br />

Second, when HCLPF is set as the<br />

screening criteria, its value is the result<br />

of the convolution of the hazard curve<br />

derived by PSHA and fragility of a<br />

single piece of equipment. As an<br />

example, if 1.00E-05/yr, the convolution<br />

value computed in between<br />

the fragility of specific equipment and<br />

the site hazard curve is set as the<br />

screening criterion HCLPF, even if all<br />

of this equipment directly causes core<br />

damage, its CDF is 5.00E-07/yr,<br />

meaning less impact. In other words,<br />

HCLPF corresponds to a probability of<br />

equipment failure of 0.05 based on<br />

95 % reliability; so conservatively, if<br />

the corresponding earthquake occurrence<br />

frequency is 1.00E-05/yr, the<br />

final CDF corresponds to 5.00E-07/yr.<br />

This will be analyzed in detail in the<br />

following sections; however, the<br />

disadvantage is that the SPSA results<br />

will be optimistic if seismic-induced<br />

failures are excluded on a single<br />

screening criterion because equipment<br />

that is critical to plant safety shutdown<br />

can be used in various seismic-induced<br />

initiating events and contributes to the<br />

CDF of each initi ating event.<br />

In the case of the United States,<br />

ASME/ANS RA-S–2008, which can be<br />

called a PRA standard, describes as<br />

follows [2]:<br />

p DEVELOP seismic fragilities <strong>for</strong> all<br />

those structures, systems, or components,<br />

or combination thereof,<br />

identified by the systems analysis.<br />

p If screening of high seismic capacity<br />

components is per<strong>for</strong>med,<br />

DESCRIBE fully the basis <strong>for</strong><br />

screening and supporting documents.<br />

For example, it is acceptable<br />

to apply the guidance given in<br />

EPRI NP-6041-SL, Rev. 1, and<br />

NUREG/CR-4334 to screen out<br />

components with high seismic<br />

capacity. However, CHOOSE the<br />

screening level high enough that<br />

the contribution to core damage<br />

frequency and large early release<br />

frequency from the screened-out<br />

components is not significant.<br />

This provides a vague criterion that<br />

fragility analysis should be per<strong>for</strong>med<br />

to reflect all SSCs in the SPSA model<br />

and screening can be applied, and this<br />

case should not have a serious impact<br />

on both CDF and large early release<br />

frequency (LERF). Next, guidance<br />

suggesting the US SPSA methodology<br />

recommends determining the detailed<br />

fragility analysis SSCs through the<br />

steps below [1].<br />

p Step 1: Screen Inherently Rugged<br />

Structures, Systems, and Components.<br />

p Step 2: Assign Initial, Representative<br />

Fragility Values <strong>for</strong> Structures,<br />

Systems, and Components.<br />


Environment and Safety<br />

Equipment Selection Methodology of Seismic Probability Safety Assessment <strong>for</strong> <strong>Nuclear</strong> <strong>Power</strong> Plant ı Junghyun Ryu and Moosung Jae

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


p Step 3: Create and Quantify an<br />

Initial Seismic Probability Risk<br />

Assessment Model.<br />

p Step 4: Rank the Systems,<br />

Structures, and Components by<br />

Importance Measure.<br />

p Step 5: Per<strong>for</strong>m Detailed Fragility<br />

Calculations <strong>for</strong> the Risk- Important<br />

Structures, Systems, and Components.<br />

p Step 6: Document the Seismic<br />

Equipment List Screening Process.<br />

All equipment except “inherently<br />

rugged structures, systems, and components”<br />

is considered in the SPSA<br />

model, so the evaluation result is quite<br />

conservative and the SPSA model<br />

must be established in advance to<br />

per<strong>for</strong>m this analysis. In addition,<br />

since there are no specific screening<br />

criteria <strong>for</strong> detailed fragility analysis,<br />

there is a weakness in that the results<br />

may differ depending on the judgment<br />

criteria of the per<strong>for</strong>ming expert. As<br />

mentioned above, it is recommended<br />

that all SSCs should be included in the<br />

SEL and qualitatively selected and<br />

removed within the range that does<br />

not affect the entire CDF or LERF;<br />

however, no methodo logy can find a<br />

systematic method <strong>for</strong> this.<br />

3 Sensitivity analysis <strong>for</strong><br />

single HCLPF screening<br />

criteria<br />

The <strong>Nuclear</strong> Energy Institute (NEI)<br />

presents the screening criteria <strong>for</strong><br />

per<strong>for</strong>ming fragility calculation [3],<br />

which is similar to the HCLPF<br />

screening criteria applied in Korea. To<br />

confirm the validity of the screening<br />

criteria presented in this report, we<br />

per<strong>for</strong>med a sensitivity analysis using<br />

the top 50 cutsets of SPSA based on<br />

the data contained in this report. The<br />

sensitivity analysis consisted of SPSA<br />

models assuming 4th seismic acceleration<br />

intervals <strong>for</strong> convenience. As a<br />

quantification tool, software using<br />

minimal cutsets upper bound (MCUB)<br />

methodology was used, but overestimated<br />

CDF was derived from high<br />

seismic acceleration interval and<br />

quantified using FTeMC [4] software<br />

with Monte Carlo simulation. The top<br />

50 cutsets presented in the report<br />

consist of 22 seismic-induced failures,<br />

ten random failure events, and four<br />

human errors events, as shown in<br />

Tables 1 and 2. These cutsets are also<br />

considered in the success sequence.<br />

In this paper, sensitivity analysis<br />

is carried out on two cases. First,<br />

we analyzed the changes in CDF<br />

according to the application of the<br />

screening criteria using the results of<br />

the convolution with the single piece<br />

of equipment; and second, we analyzed<br />

the changes in CDF when each<br />

piece of equipment with similar<br />

HCLPF size is not considered in the<br />

model. To utilize the results of the<br />

first, single piece of equipment convolute<br />

with seismic hazard curve. The<br />

convolution value was derived using<br />

PRASSE [5], which is used in the<br />

quantification analysis of seismicinduced<br />

initiating events with binary<br />

decision diagram method and Monte<br />

Carlo simulation. The results are<br />

shown in Table 3.<br />

As Table 2 shows, the higher the<br />

HCLPF value, the smaller the convolution<br />

result, but it can be seen that<br />

there is a slight difference, depending<br />

on β R and β U . Table 4 shows the<br />

sensitivity analysis according to two<br />

screening criteria.<br />

When screening criteria 5.00E-5/<br />

yr is applied, four pieces of equipment<br />

are excluded from the model, and the<br />

CDF reduction is 0.5 % based on the<br />

base model, so there is no significant<br />

effect. With screening criteria 1.00E-<br />

4/yr, the change in the overall CDF is<br />

around 4 %, indicating that applying<br />

higher screening criteria can have a<br />

significant impact on the overall results.<br />

There<strong>for</strong>e, the appropriate<br />

screening criteria was found to be effective<br />

because they might not<br />

have a significant impact on the<br />

entire CDF.<br />

Second, we examine the amount of<br />

change in CDF <strong>for</strong> equipment with<br />

similar HCLPF. Even with similar<br />

HCLPF value, the Am value may show<br />

large difference according to the β<br />

value, and the effect may be very<br />

different depending on the seismic<br />

acceleration intervals determined. In<br />

Component Description A m β U β R HCLPF<br />

SEIS-BS-CRANE Seismic failure of polar crane damages reactor vessel causing core damage 1.34 0.3 0.3 0.50<br />

SEIS-RC-RPV Seismic failure of Reactor Vessel 1.34 0.3 0.3 0.50<br />

SEIS-RC-RPVINT Seismic failure of Vessel Internals causes ATWS 1.34 0.3 0.3 0.50<br />

SEIS-SW-FO-TNK Seismic failure of ESW diesel fuel oil tank 1-SW-TK-1 1.25 0.24 0.37 0.45<br />

SEIS-RH-PUMPS Seismic failure of MD RHR pumps 1.03 0.17 0.31 0.45<br />

SEIS-EE-FO-TKS Seismic failure of underground EDG Fuel Oil tanks 1-EE-TK-2A/B 1.07 0.3 0.3 0.40<br />

SEIS-EP-EDG Seismic Failure of the Emergency Diesel Generators (EDG-1, 2, 3) 1.07 0.3 0.3 0.40<br />

SEIS-EP-LCC Seismic failure of Load Control Centers 1-EP-LCC-1H/J,1H/J-1 0.97 0.24 0.31 0.39<br />

SEIS-EP-BATT Seismic failure of station batteries 1(2)-EPD-B-1A/B 1.02 0.19 0.42 0.35<br />

SEIS-EE-DAY-TK Seismic failure of EDG fuel oil day tanks 1.08 0.45 0.24 0.33<br />

SEIS-EP-XFRMR Seismic failure of 4kv-480v trans<strong>for</strong>mers 1(2)-EP-TRAN-1H/J and 1H/J1 0.83 0.24 0.25 0.37<br />


Seismic failure of Bearing Cooling HXs results in flood of the Emergency<br />

Switchgear Room<br />

0.77 0.25 0.22 0.35<br />

SEIS-CC-CCHX Seismic failure of CCW HX results in flood of the Emergency Switchgear Room 0.68 0.32 0.18 0.29<br />

SEIS-EP-MCC-CV Seismic failure of Motor Control Centers in Cable Vault and Tunnel 0.68 0.32 0.22 0.28<br />

SEIS-EP-MCC-SB Seismic failure of Motor Control Centers in Service Building 0.69 0.23 0.36 0.26<br />

SEIS-EE-FO-PMP Seismic failure of the EDG Fuel Oil Transfer Pumps 0.68 0.3 0.3 0.25<br />

SEIS-FW-TDAFW Seismic failure of Turbine-driven Aux feed water pump 0.68 0.3 0.3 0.25<br />

SEIS-CS-RWST Seismic failure of Refueling Water Storage Tank 1-CS-TK-1 0.55 0.3 0.2 0.24<br />


Seismic failure of the Turbine building causes loss of canal(SW) and leads to<br />

core damage<br />

0.53 0.42 0.23 0.17<br />

SEIS-EP-LOOP Seismic failure of offsite power 0.4 0.22 0.22 0.19<br />

SEIS-CN-ECST Seismic failure of Emergency Condensate Storage Tank 1-CN-TK-1 0.32 0.2 0.07 0.20<br />

SEIS-EP-AAC-DG Seismic failure of Alternate AC Diesel (batteries) 0.13 0.32 0.24 0.05<br />

| Tab. 1.<br />

Equipment fragility data <strong>for</strong> sensitivity case [3].<br />

Environment and Safety<br />

Equipment Selection Methodology of Seismic Probability Safety Assessment <strong>for</strong> <strong>Nuclear</strong> <strong>Power</strong> Plant ı Junghyun Ryu and Moosung Jae

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Component Description Mean Error Factor Distribution Type<br />

1EEEDG-FR-1EEEG1 EDG-1 Fails to run 1.90E-02 1.26 Log normal<br />

1EEEDG-TM-1EEEG1 EDG-1 in Test or Maintenance 1.90E-02 1.21 Log normal<br />

1SACMP-FS-1SAC2C Service Air Compressor 1-SA-C-2 fails to start 4.22E-03 1.12 Log normal<br />

2EEEDG-FR-2EEEG1 EDG-2 Fails to run 1.90E-02 1.26 Log normal<br />

2EEEDG-TM-2EEEG1 EDG-2 in Test or Maintenance 1.90E-02 1.21 Log normal<br />

3EEEDG-FR-3EEEG1 EDG-3 Fails to run 1.90E-02 1.26 Log normal<br />

3EEEDG-TM-3EEEG1 EDG-3 in Test or Maintenance 1.90E-02 1.21 Log normal<br />

AACEDG-TM-DG0M Alternate AC diesel in Test or Maintenance 3.07E-02 1.49 Log normal<br />

HEP-C-1AFW-SLO<br />



PROB-RCCA-2<br />

Operator fails to align alternate AFW source<br />

during seismic event – low g<br />

Operator fails to start Emergency SW pump<br />

during seismic event – low g<br />

Operator fails to start Instrument Air compressor<br />

during seismic event – low g<br />

Fraction of Control Rods that fail to insert that exceeds<br />

emergency boration success criteria<br />

| Tab. 2.<br />

Random failure and human error probability <strong>for</strong> sensitivity case [3].<br />

2.64E-02 3 Log normal<br />

2.54E-03 12.8 Log normal<br />

3.60E-03 3 Log normal<br />

5.50E-01 1 Uni<strong>for</strong>m<br />

PROB-RCPSL-182 Probability of 182 gpm RCP seal leak 1.98E-01 1 Uni<strong>for</strong>m<br />

REC-RHRHX-SLO Operator fails to align RHR during seismic event – low g 3.48E-01 1 Uni<strong>for</strong>m<br />


Component Case Convolution probability HCLPF<br />


SEIS-RC-RPV Case 1<br />

3.11E-05 0.50<br />

SEIS-RC-RPVINT Under 5.00E-05<br />

3.11E-05 0.50<br />

| Tab. 3.<br />

Convolution result <strong>for</strong> sensitivity case.<br />

| Tab. 4.<br />

Sensitivity analysis result <strong>for</strong> screening criteria.<br />

addition, the accident sequence varies<br />

depending on the plant characteristics,<br />

so the target equipment contributes<br />

to plant safety which can have<br />

3.11E-05 0.50<br />

SEIS-SW-FO-TNK 4.20E-05 0.45<br />


5.89E-05 0.45<br />

SEIS-EE-FO-TKS 6.60E-05 0.40<br />

SEIS-EP-EDG Case 2<br />

6.60E-05 0.40<br />

SEIS-EP-LCC Under 1.00E-04<br />

8.07E-05 0.39<br />

SEIS-EP-BATT 8.61E-05 0.35<br />

SEIS-EE-DAY-TK 8.62E-05 0.33<br />


1.14E-04 0.37<br />

SEIS-BC-BCHX 1.38E-04 0.35<br />

SEIS-CC-CCHX 2.08E-04 0.29<br />

SEIS-EP-MCC-CV 2.16E-04 0.28<br />

SEIS-EP-MCC-SB 2.26E-04 0.26<br />

SEIS-EE-FO-PMP 2.33E-04 0.25<br />

–<br />

SEIS-FW-TDAFW 2.33E-04 0.25<br />

SEIS-CS-RWST 3.46E-04 0.24<br />

SEIS-BS-TBLDG 4.69E-04 0.17<br />

SEIS-EP-LOOP 6.56E-04 0.19<br />

SEIS-CN-ECST 9.39E-04 0.20<br />

SEIS-EP-AAC-DG 5.21E-03 0.05<br />

Case<br />

Screening<br />

Criteria<br />

# of screening out<br />

component<br />

CDF<br />

Base model - - 6.38E-04/yr -<br />

-ΔCDF<br />

(%)<br />

Case 1 Under 5.00E-05 4 6.35E-04/yr 0.50 %<br />

Case 2 Under 1.00E-04 10 6.14E-04/yr 3.81 %<br />

a very different effect on the entire<br />

CDF. There<strong>for</strong>e, screening based on<br />

single HCLPF criteria will not be able<br />

to understand all the effects without<br />

per<strong>for</strong>ming various sensitivity analyses<br />

and may have an optimistic<br />

effect on the overall results. Table 5<br />

shows the analysis of the effect on<br />

CDF that is reduced when four pieces<br />

of equipment with similar HCLPF<br />

(battery; emergency diesel generator<br />

fuel oil day tank; 4kV-480V trans<strong>for</strong>mer;<br />

and bearing cooling heat<br />

exchanger) are excluded from the<br />

model.<br />

As a result, it can be seen that in the<br />

case of the battery, the reduction rate<br />

of CDF is very large, cor responding to<br />

2.25 %, but in the case of the trans<strong>for</strong>mer,<br />

it is very small, corresponding<br />

to 0.2 %. There<strong>for</strong>e, it can be seen that<br />

if a single HCLPF criterion is applied<br />

and screened-out, the effect on the<br />

entire CDF is different <strong>for</strong> each piece of<br />

equipment, and thus further analysis<br />

is required.<br />

4 Equipment selection<br />

methodology<br />

As discussed in the previous section,<br />

equipment selection <strong>for</strong> SPSA is not<br />

being carried out using a rational<br />

approach, such as by single screening<br />

criteria or conservatively considering<br />

all equipment. In this paper, we propose<br />

a methodology <strong>for</strong> an equipment<br />

selection methodology based on all<br />

three analysis steps of SPSA.<br />

4.1 Determination of site<br />

seismic hazard region<br />

using PSHA<br />

PSHA is the step of evaluating the<br />

site-specific seismic probability that is<br />

the input to the SPSA. Basically, the<br />

probability of an earthquake is the<br />

most important step because it directly<br />

Environment and Safety<br />

Equipment Selection Methodology of Seismic Probability Safety Assessment <strong>for</strong> <strong>Nuclear</strong> <strong>Power</strong> Plant ı Junghyun Ryu and Moosung Jae

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Component Description Am βu βr HCLPF ΔCDF(%)<br />


SEIS-EP-BATT Seismic failure of station batteries 1.02 0.19 0.42 0.35 -2.25%<br />

SEIS-EE-DAY-TK Seismic failure of EDG fuel oil day tanks 1.08 0.45 0.24 0.33 -2.22%<br />

SEIS-EP-XFRMR Seismic failure of 4kv-480v trans<strong>for</strong>mers 0.83 0.24 0.25 0.37 -0.20%<br />

SEIS-BC-BCHX Seismic failure of Bearing Cooling HXs 0.77 0.25 0.22 0.35 -0.96%<br />

| Tab. 5.<br />

Sensitivity analysis result <strong>for</strong> various equipment exclusion.<br />

Site name Pr (a>SSE) Pr (a>1.0g) Seismic CDF Seismic CDF Proportion (Over 1.0g)<br />

Seabrook 4.62E-04 3.80E-06 2.99E-05 13 %<br />

Krsko 4.00E-04 1.40E-06 5.96E-05 2 %<br />

Limerick 2.29E-04 1.79E-06 5.72E-06 31 %<br />

IP2 2.12E-04 1.75E-06 8.40E-06 21 %<br />

Millstone 3 1.82E-04 1.47E-06 8.85E-06 17 %<br />

Surry 1.32E-04 6.00E-07 2.37E-05 3 %<br />

Average 2.61E-04 1.79E-06 1.99E-05 14 %<br />

| Tab. 6.<br />

Comparison of PSHA and Seismic CDF <strong>for</strong> reference plant [6, 7].<br />

affects the probability of a seismicinduced<br />

initiating event. In this paper,<br />

we reviewed the correlation between<br />

various SPSA cases and PSHA of each<br />

site. Table 6 shows the specific values<br />

of the site-specific seismic hazard<br />

curve and CDF values <strong>for</strong> each nuclear<br />

power plant [6, 7].<br />

Although an attempt was made<br />

to review more cases of PSHA<br />

and seismic core damage frequency<br />

(SCDF), the published data are<br />

limited and insufficient to infer an<br />

accurate correlation; however, the<br />

trend can be confirmed using published<br />

data. The values that represent<br />

the correlation between site SPHA<br />

and SCDF are the probability of<br />

exceeding 1.0g and the probability of<br />

exceeding the safety shutdown earthquake<br />

(SSE). First, the probability of<br />

seismic acceleration exceeding 1.0 g<br />

corresponds to a very large earthquake<br />

magnitude, so most power<br />

plants assume that direct core<br />

damage occurs. However, as a result<br />

of reviewing the case of power plants<br />

based on this value, it is difficult to<br />

represent the characteristics of<br />

seismic hazards based on Pr(a > 1.0 g),<br />

since the core damage frequency <strong>for</strong><br />

earthquakes exceeding 1.0g accounts<br />

<strong>for</strong> a range of approximately 2 % –<br />

14 %.<br />

Next, we reviewed the probability<br />

of exceeding SSE. Since SSE is the reference<br />

value <strong>for</strong> seismic design of<br />

safety-related SSC, it can be assumed<br />

that in the event of an earthquake<br />

below SSE, most equipment can<br />

maintain its function, which has a<br />

significant impact on plant safety<br />

be<strong>for</strong>e and after the seismic acceleration<br />

corresponding to this value.<br />

When plotting the correlation<br />

between the probability of occurrence<br />

| Fig. 1.<br />

Relationship of Seismic Core Damage Frequency (SCDF) and probability <strong>for</strong> SSE.<br />

of SSE and CDF, it can be seen that the<br />

correlation between the two variables<br />

is high, as shown in Figure 1.<br />

There<strong>for</strong>e, if this value is broadly<br />

divided into four types, as shown in<br />

Table 7, it is possible to classify the<br />

possibility of earthquake occurrence<br />

on the site by reoccurrence period.<br />

In the methodology proposed in<br />

this paper, the probability of SSE<br />

occurrence is basically divided into<br />

four major sections, and we intend to<br />

use this as the basis <strong>for</strong> determining<br />

the equipment groups that need to be<br />

subjected to fragility analysis. In the<br />

case of Region A, since earthquakes<br />

are very likely to occur, it is suggested<br />

that the number of equipment groups<br />

that need to be analyzed <strong>for</strong> fragility is<br />

most; and in the case of Region D,<br />

the equipment target <strong>for</strong> fragility<br />

ana lysis is least because of the<br />

lowest probability of earthquakes. To<br />

examine the adequacy of the four<br />

distinct values, the data on the<br />

seismic hazard values of power plants<br />

operating in the US [8] are examined<br />

as shown in Figure 2.<br />

Based on these values, it can be<br />

seen that the 61 power plants comprise<br />

7 % in Region A, 21 % in Region<br />

B, and 26 % in Region C. Lastly,<br />

Region D occupies 46 %. It can be seen<br />

that the criteria are valid <strong>for</strong> classifying<br />

the overall seismic hazard<br />

trend of a specific site.<br />

4.2 Determination of<br />

equipment group <strong>for</strong><br />

fragility analysis<br />

The second step of the SPSA is seismic<br />

fragility analysis. Seismic fragility<br />

analysis is per<strong>for</strong>med by a structural<br />

expert, and it is a step to evaluate the<br />

seismic resistance <strong>for</strong> the weakest<br />

failure mode of each piece of<br />

Environment and Safety<br />

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<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Category Region name Pr (a>SSE) Recurrence period (yr)<br />

Region A Very high risk region Over 4.00E-04/yr Under 2,500 year<br />

Region B High risk region 2.00E-04/yr ∼ 4.00E-04/yr 2,500 year ∼ 5,000 year<br />

Region C Medium risk region 1.00E-04/yr ∼ 2.00E-04/yr 5,000 year ∼ 10,000 year<br />

Region D Low risk region Under 1.00E-04/yr Over 10,000 year<br />

| Tab. 7.<br />

Categorized site region <strong>for</strong> PSHA.<br />


| Fig. 2.<br />

Hazard curve <strong>for</strong> US nuclear power plant [8].<br />

equipment and to determine the median<br />

acceleration capacity (A m ); with β R<br />

and β U representing the variability.<br />

Seismic fragility analysis requires the<br />

greatest consumption of time and<br />

budget among the three steps, as<br />

structural experts with the particular<br />

expertise need to per<strong>for</strong>m the analysis,<br />

as well as review and conduct on-site<br />

verification of many data such as<br />

design, construction, and installation<br />

works. There<strong>for</strong>e, the methodology<br />

proposed in this paper is a method that<br />

suggests a way to minimize such analysis.<br />

First, through a survey of<br />

sufficient data on the fragility of<br />

general equipment, we examine the<br />

characteristics of the fragility of equipment<br />

generally installed in power<br />

plants. A great deal of research has<br />

been conducted on the fragility of<br />

general equipment. Among this<br />

research, according to the SPID report<br />

[9], the fragility of SSCs can be largely<br />

classified into three, and the first<br />

inherently rugged SSCs generally have<br />

very high seismic resistance, so there is<br />

no need to include inherently rugged<br />

SSCs in the SPSA model. Second, SSCs<br />

with somewhat high seismic resistance<br />

are reflected in the SPSA model if the<br />

impact on SPSA is significant after<br />

reviewing the magnitude. In the last<br />

case, SSCs must be considered in the<br />

SPSA model. In this paper, equipment<br />

groups are classified into four types<br />

based on generic fragility data, and<br />

the results are shown in Table 8.<br />

Table 8 refers to three reports [1,<br />

10, 11], and the equipment is categorized<br />

conservatively based on the lowest<br />

HCLPF value. The equipment<br />

group according to HCLPF value has<br />

the following characteristics:<br />

p Group I (HCLPF: 0.2g or less):<br />

Generally, the seismic acceleration<br />

corresponding to plant SSE is taken<br />

to 0.2 g, so when an earthquake<br />

occurs, SSCs will fail with a<br />

probability of around 5 %. This<br />

equipment group includes offsite<br />

power sources, which are generally<br />

considered the most vulnerable of<br />

plant installations, fixed electrical<br />

panels without anchor bolts,<br />

non-safety class buildings, and<br />

yard tanks. In this case, the equipment<br />

corresponding to the most<br />

vulnerable group should always<br />

be considered regardless of the<br />

magnitude of the site seismic<br />

hazard curve. Typical yard<br />

tanks include condensate storage<br />

tank and refueling water storage<br />

tank.<br />

p Group II (HCLPF: 0.2g to 0.4g):<br />

Typically, this equipment group<br />

comprises active components<br />

which are operated by power<br />

source, including fixed electrical<br />

panels using anchor bolts, relays,<br />

batteries, small tanks, and<br />

non-safety grade diesel generators.<br />

p Group III (HCLPF: 0.4g to 0.6g):<br />

Generally, this equipment group<br />

corresponds to the safety-related<br />

class component; mainly the<br />

equipment and power sources<br />

considered <strong>for</strong> accident mitigation.<br />

p Group IV (HCLPF: 0.6g or higher):<br />

Generally, this equipment group<br />

includes various valves, pipes that<br />

make up the inherently rugged<br />

SSCs, including pressurizer, steam<br />

generators, and equipment installed<br />

in the main system, such as<br />

safety injection tanks. The main<br />

component of this group is equipment<br />

that directly causes core<br />

damage.<br />

After applying these groups to<br />

individual power plants, there may be<br />

cases different from the general<br />

fragility due to a plant’s unique<br />

design characteristics or constructability.<br />

There<strong>for</strong>e, equipment different<br />

from the general fragility should be<br />

included in the analysis through a<br />

walkdown that is essential <strong>for</strong> SPSA.<br />

4.3 Seismic plant response<br />

analysis (SPRA)<br />

The final step is the plant seismic<br />

response analysis, which is divided<br />

into an analysis of equipment that can<br />

Environment and Safety<br />

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<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Group Category Equipment HCLPF<br />


I Electrical Equipment Offsite power 0.09<br />

Structures and Tanks Large flat-bottomed tank 0.15<br />

Electrical Equipment 480 VAC Motor Control Center (MCC) / Switchgear / Bus (unanchored) 0.17<br />

Electrical Equipment 5 kV+ AC Bus / Switchgear (Unanchored) 0.17<br />

Electrical Equipment DC Motor Control Center (MCC) / Switchgear / Bus (Unanchored) 0.17<br />

Electrical Equipment Panel (Electric or Instrument) (Unanchored) 0.17<br />

Pump & Compressors Diesel-driven pump (non-safety) 0.17<br />

Structures and Tanks Turbine building 0.17<br />

Structures and Tanks Other non-safety buildings 0.17<br />

Piping Buried piping (non-safety pipe) 0.18<br />

Piping Distributed piping system (welded steel pipe) 0.19<br />

II Electrical Equipment Cable tray 0.21<br />

Pump & Compressors Motor-driven pump 0.25<br />

Pump & Compressors Turbine-driven pump 0.25<br />

Electrical Equipment 480 VAC Motor Control Center (MCC) / Switchgear / Bus (Anchored) 0.26<br />

Piping Distributed piping system (cast iron pipe) 0.26<br />

Structures and Tanks Small tank 0.30<br />

HX & HVAC Chiller 0.31<br />

Piping Small LOCA (SLOCA) 0.31<br />

HX & HVAC Air handling unit (AHU) / Room Cooler / Fan 0.33<br />

Electrical Equipment Panel (Electric or Instrument) (Anchored) 0.34<br />

Electrical Equipment Relay Chatter Failure (During Seismic Event) 0.34<br />

Pump & Compressors Large vertical, centrifugal pump (motor-driven, non-safety) 0.34<br />

Electrical Equipment Batteries 0.35<br />

Structures and Tanks Reactor internals and core assembly 0.35<br />

Piping Buried piping (safety pipe) 0.36<br />

Electrical Equipment Trans<strong>for</strong>mer (Indoor) 0.37<br />

Structures and Tanks Containment building / drywell 0.38<br />

Electrical Equipment Emergency Diesel Generator (EDG) 0.40<br />

III Pump & Compressors Diesel-driven pump (safety) 0.45<br />

HX & HVAC HVAC duct 0.48<br />

Pump & Compressors Reactor coolant pump / Recirculation pump 0.50<br />

Pump & Compressors Compressor 0.51<br />

Structures and Tanks Reactor building / auxiliary building / control building / fuel building 0.51<br />

Structures and Tanks Safety diesel generator building 0.51<br />

Structures and Tanks Other safety buildings 0.51<br />

Electrical Equipment 4 kV+ AC Bus / Switchgear (Anchored) 0.52<br />

Piping Medium LOCA (MLOCA) 0.53<br />

Electrical Equipment Battery Charger 0.55<br />

Electrical Equipment Inverter 0.55<br />

Structures and Tanks Reactor pressure vessel (RPV) 0.58<br />

Valves Large (>10”D) Motor-Operated Valve (MOV) / Damper 0.60<br />

Valves Small (

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

cause an initial event and equipment<br />

that should be used to mitigate safety<br />

shutdown such as engineering safety<br />

feature when a seismic-induced initial<br />

event occurs. Equipment that may<br />

cause an initial event should be<br />

included conservatively without a<br />

screening process, as all of the<br />

initial events may cause direct core<br />

damage. However, this equipment<br />

may be excluded if the frequency of<br />

occurrence is calculated to be below<br />

1.00E-07/yr, which corresponds to<br />

the typical initial event screening<br />

criteria [12]. Table 9 shows equipment<br />

that can initiate seismic event<br />

in general light-water reactor nuclear<br />

power plant.<br />

In general, seismic events can<br />

occur in seismic-induced loss of<br />

coolant accident, loss of power, loss of<br />

control, loss of ultimate heat sink,<br />

main steam line break, and anticipated<br />

transient without scram.<br />

Next, when the seismic-induced<br />

initiating event occurs, an analysis<br />

of the equipment considered to<br />

mitigate the event is per<strong>for</strong>med. Like<br />

all external event analysis, SPSA<br />

basically uses the event tree and fault<br />

Seismic initiating event<br />

Loss of<br />

Coolant Accident<br />

(LOCA)<br />

Loss of <strong>Power</strong><br />

(LOP)<br />

Loss of Control<br />

(LOC)<br />

Loss of Heat Sink<br />

(LOHS)<br />

Large<br />

Medium<br />

Small<br />

Small-Small<br />

Loss of offsite power<br />

Station black out<br />

Loss of KV<br />

Loss of DC power<br />

Loss of AC power<br />

Main control room<br />

Component cooling<br />

water<br />

Ultimate heat sink<br />

Main steam line break (MSLB)<br />

Anticipate transient without scram (ATWS)<br />

Building failure<br />

| Tab. 9.<br />

Equipment <strong>for</strong> seismic induced initiating event.<br />

CASE<br />

| Tab. 10.<br />

Sensitivity analysis result <strong>for</strong> importance analysis of internal events PSA.<br />

tree used in the internal event PSA.<br />

Here, non-safety and non-seismic<br />

equipment that cannot be used conservatively<br />

is excluded from the<br />

model. However, although seismicinduced<br />

initiating events are different<br />

from internal events, the primary<br />

heat removal, which is essential <strong>for</strong><br />

mitigation of the accident, must be<br />

per<strong>for</strong>med by the same equipment<br />

and procedure. There<strong>for</strong>e, in the<br />

ASME Standard, which is considered<br />

the standard of the PSA, equipment<br />

corresponding to the standard can be<br />

listed based on FV (Fussell-Vesely)<br />

importance 0.005 higher and RAW<br />

(Risk Achievement Worth) value of<br />

two or higher, which are the criteria<br />

that can be used to classify significant<br />

basic events. To confirm the application<br />

of importance value, a sensitivity<br />

analysis was per<strong>for</strong>med on the<br />

reference nuclear power plant. In<br />

Table 10, we show the difference between<br />

the results of the existing SPSA<br />

and CDF when only the equipment<br />

that was evaluated as important in<br />

internal events was modeled.<br />

As a result, it is confirmed that the<br />

results of the sensitivity analysis<br />

Equipment<br />

Pressurizer, Steam generator, RCP, RCS<br />

RCP seal<br />

Impulse lines<br />

EDG<br />

DC panel, Inverter, Battery, Charger<br />

AC Panel<br />

Main control board, MCR HVAC<br />

Pump, Heat exchanger, Piping<br />

Pump, Heat exchanger,<br />

Intake structure<br />

Control rods<br />

Containment building<br />

Service building<br />

Turbine building<br />

# of basic event<br />

<strong>for</strong> SPSA model<br />

% of<br />

baseline CDF<br />

Base model 1,798 100.0 %<br />

Only FV important basic event 49 97.0 %<br />

Only RAW important basic event 217 95.9 %<br />

Intersection FV and RAW<br />

important basic event<br />

24 95.5 %<br />

Union FV or RAW important basic event 242 97.6 %<br />

showed no significant difference<br />

between the case of modeling all<br />

equipment and the case of modeling<br />

only equipment that was evaluated as<br />

important in internal events. When<br />

242 items of equipment, 13 % of the<br />

total of 1798, were modeled in the<br />

SPSA model, the CDF was 97.6 % of<br />

the baseline CDF, and even when<br />

49 pieces of equipment classified as<br />

important in basic events based on FV<br />

are modeled, a result equivalent<br />

to 97 % of the baseline CDF was<br />

obtained. In addition, even when only<br />

24 pieces of equipment were considered,<br />

a value corresponding to<br />

95.5 % of the CDF value of the base<br />

model was derived, indicating that<br />

the importance of other equipment<br />

other than this was much lower<br />

than expected. There<strong>for</strong>e, it is<br />

judged that the equipment selection<br />

methodology based on the values of<br />

the internal event FV and RAW is<br />

much more rational than the<br />

existing HCLPF-based method, and is<br />

an efficient method that can<br />

reflect the characteristics of the<br />

SPSA model.<br />

4.4 Summary of the proposed<br />

equipment selection<br />

methodology<br />

The procedure is shown in Figure 3, a<br />

flow chart of the method of selecting<br />

equipment <strong>for</strong> seismic events over the<br />

three steps described above.<br />

First, based on the site seismic<br />

hazard curve obtained as a result of<br />

PSHA, the SSE value is checked,<br />

which is the design standard of the<br />

power plant, and probability value<br />

exceeding SSE value is checked.<br />

After that, the region of ​the site is<br />

determined by checking the SSE<br />

reoccurrence period according to<br />

the SSE excess probability value.<br />

Through this value, the group of<br />

equipment that should per<strong>for</strong>m<br />

fragility analysis is determined. Next,<br />

based on the results of the PSA<br />

importance of internal events, basic<br />

events with an FV value of 0.005 or<br />

more or RAW value of two or more,<br />

are listed and described in terms of<br />

the function of the system. However,<br />

basic events related to non-seismic<br />

equipment and human error are<br />

excluded among the results of the<br />

PSA importance of internal events.<br />

Finally, equipment is derived<br />

through cross-examination between<br />

the ‐equipment required <strong>for</strong> the<br />

function of the mitigating equipment<br />

derived from an internal event and the<br />

equipment group subject to fragility<br />

analysis.<br />


Environment and Safety<br />

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<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


| Fig. 3.<br />

Flowchart <strong>for</strong> the equipment selection methodology.<br />

5 Case study <strong>for</strong> the<br />

equipment selection<br />

methodology<br />

In this paper, we compare and<br />

analyze the seismic PSA results of the<br />

reference nuclear power plant using<br />

the proposed methodology. Existing<br />

PSA results could not be considered,<br />

as detailed data on the walkdown<br />

could not be collected. However, even<br />

if the walkdown is excluded, the<br />

results are sufficient to demonstrate<br />

the effectiveness of the methodology<br />

proposed in this paper. The first step<br />

is the analysis of site-specific PSHA<br />

results. The review of the corresponding<br />

hazard curves indicates<br />

that the probability of earthquake<br />

occurrence is relatively low, and the<br />

probability of exceeding 0.2g on the<br />

SSE basis corresponds to 1.06E-04/yr,<br />

corresponding to the Region C<br />

medium risk region proposed in this<br />

paper. In terms of the likelihood of<br />

earthquake occurrence, it is a value<br />

that is approximately 54 % of the total<br />

power plants in terms of seismic<br />

hazard considered by Figure 2. Based<br />

on this, the plant needs to review<br />

equipment groups I and II subject to<br />

fragility analysis. Fragility analysis<br />

groups I and II include general<br />

vulnerable equipment including<br />

offsite power sources and yard tanks,<br />

as well as general active equipment.<br />

The next analysis is the plant seismic<br />

response analysis using internal<br />

events PSA results; 74 basic events<br />

were selected with an FV value of<br />

0.005 or more and 253 basic events<br />

with a RAW value of two or more.<br />

Among them, 30 pieces of equipment<br />

that satisfy both FV and RAW are<br />

considered, along with 297 basic<br />

events. Excluding 20 human error<br />

basic events and 26 non-seismic basic<br />

events, 251 basic events are considered.<br />

In addition, except <strong>for</strong> valves,<br />

flow elements, flow transmitters,<br />

radiation transmitters, dampers and<br />

filters, which are inherently rugged<br />

SSCs, the total 136 basic events are<br />

derived.<br />

The 136 basic events derived can<br />

be divided into related systems and<br />

equipment types to achieve the<br />

following 14 critical system functions.<br />

Here, if we review the 14 important<br />

functions, we can confirm easily the<br />

unique operating characteristics of<br />

the reference plant.<br />

p Auxiliary feedwater (AF) supply<br />

p AF pump room cooling<br />

p Emergency power supply<br />

p Component cooling water supply<br />

p Diesel generator fuel supply<br />

p Diesel generator room cooling<br />

p Essential chilled water supply<br />

p High-pressure injection<br />

p Low-pressure injection<br />

p Plant control<br />

p Ultimate heat sink<br />

p Reactor containment cooling<br />

p Safety actuation signal<br />

p Ultimate heat sink pump room<br />

cooling<br />

In the end, cross-examining 14 critical<br />

systems and functions with fragility<br />

equipment groups I and II, the results<br />

shown in Table 11 are obtained.<br />

The number of equipment items derived<br />

through cross-examination is a<br />

total of 23, which is a very small result<br />

considering the overall equipment in<br />

nuclear power plants. How ever, when<br />

reviewing the pre viously analyzed<br />

SPSA results, it can be confirmed that<br />

all devices are considered important in<br />

the existing SPSA model except <strong>for</strong> the<br />

three pieces of equipment that initiate<br />

seismic-induced initiating events, so<br />

the methodology proposed in this<br />

paper is very efficient. It can be<br />

confirmed that it is reasonable.<br />

6 Conclusion<br />

In this study, a methodology of the<br />

equipment selection <strong>for</strong> SPSA is proposed.<br />

The single HCLPF screening<br />

criterion which has been applied <strong>for</strong><br />

SPSA reflects some of the site-specific<br />

PSHA results but does not reflect the<br />

plant design characteristics, so if the<br />

single HCLPF screening criterion is<br />

applied to the model based on this, an<br />

optimistic evaluation can be made.<br />

The methodology proposed in this<br />

paper has the following advantages:<br />

p Safety aspects: Single HCLPF<br />

screening criteria is not applied,<br />

and all equipment is not reflected<br />

in the model using general fragility<br />

data, so realistic and reasonable<br />

SPSA results are expected.<br />

Environment and Safety<br />

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<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Group Category Equipment Mitigation Function Selected equipment<br />

I Electrical Equipment Offsite power – –<br />

Structures and Tanks Large flat-bottomed tank Auxiliary feed water supply Condensate storage tank<br />

Electrical Equipment 480 VAC Motor Control Center (MCC) /<br />

Switchgear / Bus (unanchored)<br />

High pressure injection<br />

Low pressure injection<br />

Diesel generator fuel supply<br />

Plant control<br />

Refueling water storage tank<br />

Refueling water storage tank<br />

EDG storage tank<br />

Load center<br />

Motor control center<br />

Electrical Equipment 5 kV+ AC Bus / Switchgear (Unanchored) Plant control Switchgear<br />

Electrical Equipment<br />

DC Motor Control Center (MCC) / Switchgear<br />

/ Bus (Unanchored)<br />

Plant control<br />

DC Panel<br />

Electrical Equipment Panel (Electric or Instrument) (Unanchored) Plant control Electrical panel<br />

Pump & Compressors Diesel-driven pump (non-safety) – –<br />

Structures and Tanks Turbine building – –<br />

Structures and Tanks Other non-safety buildings – –<br />

Piping Buried piping (non-safety pipe) – –<br />

Piping Distributed piping system (welded steel pipe) – –<br />

II Electrical Equipment Cable tray – –<br />

Pump & Compressors Motor-driven pump Auxiliary feed water supply AF motor driven pump<br />

Component cooling water supply<br />

Essential chilled water supply<br />

Diesel generator fuel supply<br />

High pressure injection<br />

Low pressure injection<br />

Ultimate heat sink<br />

Component cooling water pump<br />

Essential chilled water pump<br />

EDG fuel pump<br />

HPSI pump<br />

LPSI pump<br />

NSCW pump<br />

Pump & Compressors Turbine-driven pump Auxiliary feed water supply AF turbine driven pump<br />

Electrical Equipment 480 VAC Motor Control Center (MCC) /<br />

Switchgear / Bus (Anchored)<br />

Plant control<br />

Piping Distributed piping system (cast iron pipe) – –<br />

Load center<br />

Motor control center<br />

Structures and Tanks Small tank Component cooling water supply Component cooling surge tank<br />

HX & HVAC Chiller Essential chilled water supply Chiller<br />

Piping Small LOCA (SLOCA) – –<br />

HX & HVAC Air handling unit (AHU) / Room Cooler / Fan AF pump room cooling AF room AHU<br />

Diesel generator room cooling<br />

Reactor containment cooling<br />

Ultimate heat sink pump room<br />

cooling<br />

EDG room AHU<br />

RCFC fan<br />

NSCW room AHU<br />

Electrical Equipment Panel (Electric or Instrument) (Anchored) Plant control Electrical panel<br />

Electrical Equipment Relay Chatter Failure (During Seismic Event) Plant control Electrical panel<br />

Pump & Compressors<br />

Large vertical, centrifugal pump (motor-driven,<br />

non-safety)<br />

– –<br />

Electrical Equipment Batteries Plant control Electrical panel<br />

Structures and Tanks Reactor internals and core assembly – –<br />

Piping Buried piping (safety pipe) – –<br />

Electrical Equipment Trans<strong>for</strong>mer (Indoor) – –<br />

Structures and Tanks Containment building / drywell – –<br />

Electrical Equipment Emergency Diesel Generator (EDG) Emergency power supply Emergency diesel generator<br />


| Tab. 11.<br />

Selected equipment <strong>for</strong> fragility analysis.<br />

p Economics aspects: As the largest<br />

portion of the required manpower<br />

<strong>for</strong> SPSA is the detailed fragility<br />

analysis work, the methodology<br />

proposed in this paper is economically<br />

beneficial as the number of<br />

pieces of equipment subject to<br />

SPSA decreases.<br />

The equipment selection methodology<br />

proposed in this paper requires<br />

analysis of all three parts of<br />

SPSA, unlike the previously proposed<br />

methodology. This means that selecting<br />

equipment through consideration<br />

of only one part, such as PSHA, may<br />

lead to incorrect results. SPSA has its<br />

own uncertainty, so if one factor<br />

affects several steps, the uncertainty<br />

becomes very large. There<strong>for</strong>e, the<br />

analysis of equipment not essential <strong>for</strong><br />

SPSA can increase this uncertainty, so<br />

it can be said that it is necessary <strong>for</strong> a<br />

much more realistic analysis that is<br />

not considered in advance.<br />

In recent years, SPSA has evolved<br />

from a single unit criterion evaluation<br />

to an evaluation of multiple units<br />

operating at one site. In this case, the<br />

number of equipment items handled<br />

by the SPSA increases, resulting in<br />

Environment and Safety<br />

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<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


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unnecessary model enlargement. There<strong>for</strong>e, even in such<br />

case, it is necessary to select equipment that has a significant<br />

influence on seismic events, and applying the equipment<br />

selection methodology proposed in this paper can<br />

contribute to the simplification of the SPSA model and the<br />

accuracy of the analysis of the quantitative results, and<br />

minimize unnecessary analysis. There<strong>for</strong>e, it is expected<br />

that uncertainty errors will be minimized.<br />

References<br />

[1] Seismic Probabilistic Risk Assessment Implementation Guide. USA: Electric <strong>Power</strong> Research Institute;<br />

2013, 3002000709.<br />

[2] Standard <strong>for</strong> Level 1/Large Early Release Frequency Probabilistic Risk Assessment <strong>for</strong> <strong>Nuclear</strong><br />

<strong>Power</strong> Plant Applications. USA: The American Society of Mechanical Engineers; 2009, ASME/ANS<br />

Ra-SA-2009.<br />

[3] White Paper: Criterion <strong>for</strong> Capacity-based Selection of SSCs <strong>for</strong> Per<strong>for</strong>ming Fragility Analysis in a<br />

Seismic Risk-based Evaluation. USA: <strong>Nuclear</strong> Energy Institute; 2012.<br />

[4] FTeMC Quick Guide Fault Tree Top Event Probability Evaluation Using Monte Carlo Simulation.<br />

Korea: Korea Atomic Energy Research Institute; 2017, KAERI-ISA- memo-FTeMC-01, Rev. 1.<br />

[5] Development and Validation of the Seismic Probabilistic Safety Assessment Software PRASSE.<br />

Korea: Korea Atomic Energy Research Institute; 2012, KAERI/TR-4649.<br />

[6] Probabilistic safety assessment <strong>for</strong> seismic events. IAEA; 1993, TECDOC-724.<br />

[7] M Vermaut, Ph Monette P Shah R.D Campbell, Methodology and results of the seismic probabilistic<br />

safety assessment of Krško <strong>Nuclear</strong> <strong>Power</strong> Plant, <strong>Nuclear</strong> Engineering and Design,<br />

2 May 1998, Volume 182, Issue 1, Pages 59-72<br />

[8] Risk Assessment of Operational Events Handbook Volume 2 – External Events. USA: <strong>Nuclear</strong><br />

Regulatory Commission; 2008, Revision 1.01<br />

[9] Seismic Evaluation Guidance Screening, Prioritization and Implementation Details (SPID) <strong>for</strong> the<br />

Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic. USA: Electric <strong>Power</strong><br />

Research Institute; 2013, 1025287<br />

[10] A Methodology <strong>for</strong> Analyzing Precursors to Earthquake- Initiated and Fire-Initiated Accident Sequences.<br />

USA: <strong>Nuclear</strong> Regulatory Commission; 1998, NUREG/CR-6544.<br />

[11] Surry Seismic Probabilistic Risk Assessment Pilot Plant Review. USA: Electric <strong>Power</strong> Research Institute;<br />

2010, 1020756.<br />

[12] Identification of External Hazards <strong>for</strong> Analysis in Probabilistic Risk Assessment. USA: Electric<br />

<strong>Power</strong> Research Institute; 2011, 1022997.<br />

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Junghyun Ryu<br />

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Department of <strong>Nuclear</strong> Engineering,<br />

Hanyang University, Seoul, Republic of Korea<br />

yarya@hanyang.ac.kr<br />

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ISSN 1431-5254<br />

Junghyun Ryu is a doctoral student in the Department of <strong>Nuclear</strong> Engineering at<br />

Hanyang University. He works in FNC Technology in South Korea.<br />

Prof Dr Moosung Jae<br />

Professor at Hanyang University, South Korea<br />

jae@hanyang.ac.kr<br />

Moosung Jae is Professor of Department of <strong>Nuclear</strong> Engineering, Hanyang<br />

University, Seoul, Korea. He is now Director of MUR Risk Assessment Group<br />

( MURRG Center) at HYU.<br />

Prior to joining Hanyang University, Seoul, Korea, he had worked <strong>for</strong> as a senior<br />

researcher <strong>for</strong> Korea Atomic Energy Research Institute (KAERI).<br />

He received B.S. and M.S. in <strong>Nuclear</strong> Engineering fromSeoul National University and<br />

Ph.D. in <strong>Nuclear</strong> Engineering from University of Cali<strong>for</strong>nia, Los Angeles, USA (1992).<br />

Now he is a member of INSAG (<strong>International</strong> <strong>Nuclear</strong> Safety Group), IAEA as well as<br />

a member of the National Academy of Engineering of KOREA.<br />

Environment and Safety<br />

Equipment Selection Methodology of Seismic Probability Safety Assessment <strong>for</strong> <strong>Nuclear</strong> <strong>Power</strong> Plant ı Junghyun Ryu and Moosung Jae

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Experimental Study of Convective<br />

Heat Transfer Through Fuel Pins of a<br />

<strong>Nuclear</strong> <strong>Power</strong> Plant<br />

Atif Mehmood, Ajmal Shah, Mazhar Iqbal, Ali Riaz, Muhammad Ahsan Kaleem and Abdul Quddus<br />

Cylindrical heat sources are used to model the fuel pins in the reactor core of a nuclear power plant and also in the<br />

spent fuel storage. This research was aimed at studying convection through slender cylinders arranged in a square<br />

array, placed inside a ventilated enclosure. Four cylindrical heat sources were manufactured, each having L/D ratio of<br />

6.1. These heat sources were placed inside an enclosure which had one inlet and three outlets. At first, optimum<br />

configuration of outlets <strong>for</strong> heat transfer was found by keeping the inlet speed of air and heat flux constant. This<br />

optimum outlet configuration was then used to study the effect of changing heat flux on heat transfer. Speed of air was<br />

kept constant at 0.7 m/s throughout this study. The heat flux was changed, ranging from 79.83 W/m 2 to 513.59 W/m 2<br />

and Rayleigh number changed accordingly from 2.093x10 9 to 8.575x10 9 . It was observed that by increasing Rayleigh<br />

number, Nusselt number decreased along with the heat transfer.<br />

1 Introduction<br />

Convective heat transfer is one of the<br />

important areas of research in heat<br />

transfer. Efficient heat transfer from<br />

heat sources has been an important<br />

topic of research over the past few<br />

years. Cylindrical heat sources are<br />

widely used in many industries<br />

such as building, solar and nuclear<br />

industry. In nuclear industry, fuel pins<br />

inside reactor core and spent fuel<br />

storage can be modeled as cylindrical<br />

heat sources with convective fluid<br />

flowing around them. Due to such<br />

vast applications and importance,<br />

researchers around the world have<br />

studied heat transfer through cylindrical<br />

heat sources both experimentally<br />

and computationally to solve issues<br />

related to temperature control.<br />

There are many configurations<br />

in which convective heat transfer<br />

through cylindrical heat sources can<br />

be studied. The medium through<br />

which the heat is transferred, could be<br />

finite or infinite, the heat sources<br />

could be single or multiple, and could<br />

be ranged in a horizontal, vertical or<br />

inclined orientation, the convective<br />

fluid could be air, water or any other<br />

suitable fluid, the type of convection<br />

could be free, <strong>for</strong>ced or mixed convection<br />

and the enclosure could be<br />

ventilated or non-ventilated. All these<br />

different types of configurations<br />

have been studied previously. Popiel<br />

[1] reviewed heat transfer through<br />

slender cylinders in which he used the<br />

data put <strong>for</strong>th by Cebeci [2] and<br />

differentiated between heat transfer<br />

through thick and slender cylinders.<br />

This is important since thick cylinders<br />

can be approximated as flat plates and<br />

thin cylinders cannot. Griffiths and<br />

Davis [3] studied convection in 1922,<br />

where they studied vertical cylinders<br />

in isothermal condition. They did<br />

experiments on various cylinders,<br />

keeping their diameters constant and<br />

varying the length of the cylinders<br />

and determined the average Nusselt<br />

numbers. Later on, Morgan [4] used<br />

that data to find two correlations,<br />

which are valid over their respective<br />

ranges of dimensionless number. Fujii<br />

et al. [5] did experiments on isothermal<br />

vertical cylinders, using three<br />

different fluids i.e. water (Pr = 5),<br />

mobiltherm oil (Pr = 100) and spindle<br />

oil (Pr = 100). Separate correlations<br />

were then developed <strong>for</strong> the local<br />

Nusselt numbers <strong>for</strong> each of the fluids<br />

used. Jarral and Campo [6] carried<br />

out an experimental study, using<br />

isoflux boundary condition and determined<br />

local Nusselt numbers in terms<br />

of Rayleigh number <strong>for</strong> air. They did<br />

their experimentation using cylinders<br />

with three different slenderness<br />

ratios. Ali Riaz et al. [7] per<strong>for</strong>med<br />

experiments on vertical cylinders and<br />

horizontal cylinders to find the heat<br />

transfer coefficients. The Nusselt<br />

number was observed to decrease<br />

from the bottom of the cylinder,<br />

towards the top, up to a certain point<br />

after which it started to increase. The<br />

reason behind this is that thermal<br />

boundary layer thickness increases<br />

from bottom to top of the cylinder. In<br />

the horizontal configuration, however,<br />

the local Nusselt number was<br />

least at the outlet and maximum at the<br />

inlet. Arshad et al. [8] per<strong>for</strong>med<br />

experiments with natural convection<br />

at high Rayleigh numbers using nine<br />

cylinders in a 3x3 array placed<br />

vertically and enclosed inside an<br />

enclosure. It was observed that<br />

surface temperatures increased up to<br />

a specific point and then decreased,<br />

this is attributed to mixing, which<br />

results in increase in heat transfer.<br />

K. Hata et al. [9] studied heat transfer<br />

through natural convection in laminar<br />

region, using vertical rods placed in a<br />

7x7 array placed in liquid sodium.<br />

The effect of pitch-to-diameter (P/D)<br />

ratio, array size, bundle geometry and<br />

Rayleigh number on heat transfer,<br />

was observed by calculating Nusselt<br />

number under different conditions.<br />

Yuji Isahai and Naozo Hattori [10]<br />

worked on a numerical study of heat<br />

transfer through natural convection in<br />

a heated rod bundle, that was placed<br />

vertically in an equilateral triangle<br />

configuration inside an enclosure.<br />

They used a total of 19 cylinders in<br />

their study in a hexagonal <strong>for</strong>mation,<br />

with the center cylinder dedicated <strong>for</strong><br />

instrumentation. Five different P/D<br />

ratios were studied which varied<br />

between minimum of 1.1 to maximum<br />

of 7.0. Abdul Jabbar Khalifa and Zaid<br />

Ali [11] studied heat transfer through<br />

natural convection in single and<br />

in multiple cylinders. For multiple<br />

cylinder configuration, they used nine<br />

cylinders in a 3 x 3 array, out of which<br />

only three cylinders were heated. A<br />

square array configuration was used<br />

<strong>for</strong> cylinders having a P/D ratio of 2.<br />

The fluid used <strong>for</strong> heat transfer was<br />

water. Heat flux was varied and its<br />

effect was studied. K. Tehseen et al.<br />

[12] per<strong>for</strong>med a numerical study,<br />

using ANSYS, to find the effects of<br />

different orientations of core on heat<br />

transfer along bare circular tubes and<br />

tube bundles. They found that the<br />

heat transfer was directly proportional<br />

to the heated length and<br />

inversely proportional to the inside<br />

diameter of the tube.<br />

37<br />


Research and Innovation<br />

Experimental Study of Convective Heat Transfer Through Fuel Pins of a <strong>Nuclear</strong> <strong>Power</strong> Plant ı Atif Mehmood, Ajmal Shah, Mazhar Iqbal, Ali Riaz, Muhammad Ahsan Kaleem and Abdul Quddus

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


Convection heat transfer is affected<br />

by geometry parameters of enclosure<br />

such as position and area of vents.<br />

There has been limited research<br />

considering such parameters and,<br />

there<strong>for</strong>e, this found the motivation<br />

behind the present experimental<br />

study. In this present work, cylindrical<br />

heat sources were used, which can be<br />

used to model fuel pins inside a<br />

nuclear reactor or spent fuel inside<br />

spent fuel storage. Heat transfer<br />

through a 2x2 array of heat sources in<br />

square configurations was studied.<br />

Heat transfer between heat sources<br />

and air was tabulated and compared<br />

in different arrangements of the<br />

outlets present on the enclosure. The<br />

results were then plotted showing the<br />

relation between Nusselt number and<br />

modified Rayleigh number.<br />

2 Mathematical Modeling<br />

The equations used <strong>for</strong> the present<br />

system were:<br />

Q in = P(2.1)<br />

Q out = Q in = V × I(2.2)<br />

Due to small thickness of aluminum<br />

tube, all the heat from the heat source<br />

is conducted to the outer surface.<br />

Q out = Q conduction =Q convection + Q radiation<br />

(2.3)<br />

Q convection = h avg A s (T s,avg – T b )(2.4)<br />

Where, h avg is the average heat<br />

transfer coefficient.<br />

And<br />

For calculating the local heat transfer<br />

coefficients, the following <strong>for</strong>mula<br />

was used.<br />

Q convection = h x A s (T x – T b )(2.5)<br />

Heat losses through radiation can be<br />

calculated by:<br />

(2.6)<br />

ε is 0.04 <strong>for</strong> Aluminum and σ is<br />

5.67x10 -8 W/m 2 K 4 . Due to small<br />

emissivity of Aluminum and a very low<br />

value of σ, radiation can be neglected<br />

and it can be assumed that all the heat<br />

transfer from the outer surface of heat<br />

sources is via con vection.<br />

The <strong>for</strong>mulae <strong>for</strong> Nusselt number<br />

and modified Rayleigh number are:<br />

(2.7)<br />

<br />

(2.8)<br />

<br />

(a)<br />

| Fig. 1.<br />

Thermocouple locations (a) front view of a heat source (b) Top view of the enclosure.<br />

3 Experimental Setup<br />

Four identical sources each having<br />

a diameter of 50.8 mm and a<br />

slenderness ratio of 6.1 were manufactured.<br />

Each of these heat sources<br />

had a rated power of 1000 W. Outer<br />

surface of heat sources was made of<br />

Aluminum. Each of the heat sources<br />

was equipped with four K-type<br />

thermocouples, the positions shown<br />

in Figure 1. In Figure 1(b), the stars<br />

show the thermocouple facing<br />

the flow channel and the triangles<br />

represent the thermocouples on the<br />

opposite side of the flow channel.<br />

The heat sources were placed<br />

inside a wooden enclosure of dimensions<br />

20”x20”x20”. There were three<br />

outlets of same size (4”x2”) in the<br />

enclosure, one at the top and one<br />

each on right top and left top of the<br />

enclosure. For the measurement of<br />

the ambient temperature, a thermocouple<br />

was installed near inner<br />

boundary of the wooden enclosure.<br />

To convert the data provided by the<br />

thermocouples into temperatures,<br />

PANGU data acquisition system was<br />

| Fig. 2.<br />

Experimental setup.<br />

(b)<br />

used. All these components of the<br />

experimental setup are shown in<br />

Figure 2.<br />

In this research work, data<br />

acquisition system was used which<br />

had a calibration uncertainty of 1 °C<br />

and data scatter was observed to be<br />

5.1 °C. Total uncertainty in measurement<br />

of temperature was calculated to<br />

be 4.8 %. Uncertainty in voltage<br />

was 1 % with a full scale reading of<br />

600 V and in current, the uncertainty<br />

was 1.5 % with a full scale reading of<br />

10 A. Anemometer that was used<br />

<strong>for</strong> the measurement of velocity of air<br />

had an uncertainty of 0.03 m/s.<br />

Also, different dimensions that<br />

were measured e.g., diameter of the<br />

aluminum pipes, length of aluminum<br />

pipes and size of inlet and outlet vents<br />

had an uncertainty of about 2 mm.<br />

Each experiment was repeated thrice<br />

so that the results are reported with<br />

precision and accuracy. Mean values<br />

were shown in the graphs and error<br />

bars were displayed, representing<br />

standard deviation in the results.<br />

4 Results and Discussions<br />

In the experimentation phase, the<br />

outlet configuration was first<br />

optimized to give best heat transfer<br />

and then the further experimentation<br />

was carried out using different heat<br />

fluxes.<br />

4.1 Optimization of Outlets<br />

Configuration<br />

Five cases were studied, each case<br />

differing from the other on the basis of<br />

locations of the outlets’ opening.<br />

In the natural convection case, top<br />

outlet was opened and inlet at the<br />

bottom was opened, but the fan was<br />

remained switched off. Air flowed<br />

freely and naturally from the bottom<br />

and exited through the top outlet. In<br />

other four cases, bottom inlet was<br />

opened with the fan operating at a<br />

Research and Innovation<br />

Experimental Study of Convective Heat Transfer Through Fuel Pins of a <strong>Nuclear</strong> <strong>Power</strong> Plant ı Atif Mehmood, Ajmal Shah, Mazhar Iqbal, Ali Riaz, Muhammad Ahsan Kaleem and Abdul Quddus

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

(a)<br />

(c)<br />

| Fig. 3.<br />

Variation of Nu with outlet configuration <strong>for</strong> (a) cylinder 1, (b) cylinder 2, (c) cylinder 3, (d) cylinder 4.<br />

constant speed of 0.7 m/s. The<br />

corresponding outlets were opened<br />

<strong>for</strong> each case and temperatures were<br />

measured which were then used to<br />

find the values of heat transfer<br />

coefficients <strong>for</strong> each case.<br />

In Figure 3, the near outlet refers<br />

to the outlet which is closest to that<br />

particular cylinder and far outlet is the<br />

outlet which is at a distance from that<br />

cylinder. For example, <strong>for</strong> cylinders<br />

1 and 2, near outlet is the left outlet<br />

and far outlet is the right outlet as<br />

shown in Figure 1.<br />

It is evident from the plots above,<br />

that <strong>for</strong> any given cylinder, the heat<br />

transfer is the least in case of natural<br />

convection and improved <strong>for</strong> <strong>for</strong>ced<br />

convection as more outlets were then<br />

opened. Heat transfer is maximum<br />

when all the three outlets were<br />

opened. This gave an optimum configuration<br />

of outlets with respect to<br />

heat transfer.<br />

(b)<br />

(d)<br />

value of 8.575x10 9 at heat flux of<br />

513.59 W/m 2 . The results <strong>for</strong> these<br />

five cases are shown in Figure 4.<br />

It is clear <strong>for</strong>m the plotted points<br />

that the heat transfer dropped as the<br />

(a)<br />

flux and modified Rayliegh number<br />

was increased. Nusselt number did<br />

not change much as the modified<br />

Rayleigh number increased from<br />

2.09x10 9 to 5.79x10 9 . Beyond that<br />

point, with increase in the modified<br />

Rayleigh number, there is a certain<br />

decrease in the values of Nusselt<br />

number. The reason behind this<br />

behaviour is attributed to the fact that<br />

the outlets were not large enough to<br />

allow heat transfer away from the<br />

cylinders and the temperatures were<br />

seen to go higher as the heat flux and<br />

modified Rayleigh number of the<br />

heaters was increased.<br />

5 Conclusions<br />

Heat sources were manufactured<br />

and arranged in a square array<br />

inside a ventilated enclosure with<br />

one inlet and three outlets at<br />

three different walls of the enclosure.<br />

Experiments were carried out <strong>for</strong><br />

five different outlet configurations at<br />

constant heat flux and constant<br />

speed of air <strong>for</strong>m the inlet to<br />

determine the optimum configuration<br />

of the outlets <strong>for</strong> heat transfer.<br />

After that, experiments were carried<br />

out <strong>for</strong> five different heat fluxes<br />

ranging from 79.83 W/m 2 to<br />

513.59 W/m 2 and the results were<br />

plotted in the <strong>for</strong>m of dimensionless<br />

numbers.<br />

(b)<br />


4.2 Effect of Heat Flux<br />

on Heat Transfer<br />

A total of five different heat fluxes<br />

were used to study their effect on heat<br />

transfer. Heat fluxes were changed<br />

from 79.83 W/m 2 to 513.59 W/m 2 .<br />

The data was also translated in the<br />

<strong>for</strong>m of dimensionless numbers. Due<br />

to change in heat fluxes, modified<br />

Rayleigh number, Ra* changed from a<br />

minimum value of 2.093x10 9 at heat<br />

flux of 79.83 W/m 2 to a maximum<br />

(c)<br />

(d)<br />

| Fig. 4.<br />

Variation of Nu with modified Ra <strong>for</strong> (a) cylinder 1, (b) cylinder 2, (c) cylinder 3, (d) cylinder 4.<br />

Research and Innovation<br />

Experimental Study of Convective Heat Transfer Through Fuel Pins of a <strong>Nuclear</strong> <strong>Power</strong> Plant ı Atif Mehmood, Ajmal Shah, Mazhar Iqbal, Ali Riaz, Muhammad Ahsan Kaleem and Abdul Quddus

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


The results of the experiments can<br />

be summarized as following:<br />

p Heat transfer improved by almost<br />

25 % in the cases with <strong>for</strong>ced<br />

convection i.e., when the fan was<br />

switched on and all three outlets<br />

were opened as compared to the<br />

natural convection case when the<br />

fan was off.<br />

p Heat transfer from the cylinders<br />

seemed to remain almost constant<br />

with increase in heat flux and the<br />

modified Rayleigh number up to a<br />

certain range and then decreased<br />

as the modified Rayleigh number<br />

was increased further.<br />

p The reason could be attributed to<br />

the fact that the size of the outlets<br />

was small (4” x 2”) as compared<br />

to the size of the enclosure<br />

(20”x20”x20”).<br />

Nomenclature<br />

h<br />

Q<br />

A<br />

Heat transfer coefficient<br />

Heat Supplied<br />

Area<br />

q” Heat Flux<br />

σ<br />

β<br />

k<br />

Ra*<br />

V<br />

I<br />

P<br />

ε<br />

α<br />

ν<br />

Nu<br />

Stefan-Boltzmann constant<br />

Volumetric thermal expansion coefficient<br />

Thermal conductivity<br />

Modified Rayleigh Number<br />

Voltage supplied<br />

Current<br />

<strong>Power</strong> supplied<br />

Emissivity<br />

Thermal diffusivity<br />

Momentum diffusivity<br />

Nusselt Number<br />

Subscripts<br />

x Local value<br />

avg<br />

s<br />

b<br />

Average value<br />

Value at surface<br />

Bulk<br />

References<br />

[1] C. O. Popiel, “Free Convection Heat Transfer from Vertical<br />

Slender Cylinders: A Review,” Heat Transfer Engineering,<br />

vol. 29:6, pp. 521-536, 2008.<br />

[2] T. Cebeci, “ Laminar-Free-Convective-Heat Transfer from<br />

the Outer Surface of a Vertical Slender Circular Cylinder,”<br />

Proc. 5th <strong>International</strong> Heat Transfer Conference, vol. 3,<br />

pp. 15-19, 1974.<br />

[3] E. Griffiths and A. Davis, “The transmission of heat by<br />

radiation and convection,” Food Investigation Board,<br />

Special Report, vol. 9, 1922.<br />

[4] V. T. Morgan, “The Overall Convective Heat Transfer from<br />

Smooth Circular Cylinders,” Advances in Heat Transfer,<br />

vol. 11, pp. 199-264, 1975.<br />

[5] T. Fujii, M. Takeuchi, M. Fujii, K. Suzaki and H. Uehara,<br />

“ Experiments on natural-convection heat transfer from<br />

the outer surface of a vertical cylinder to liquids,”<br />

<strong>International</strong> <strong>Journal</strong> of Heat and Mass Transfer, vol. 13,<br />

no. 5, pp. 753-770, 1970.<br />

[6] S. Jarall and A. Campo, “Experimental Study of Natural<br />

Convection from Electrically Heated Vertical Cylinders<br />

Immersed in Air,” Experimental Heat Transfer, vol. 18,<br />

no. 3, pp. 127-134, 2005.<br />

[7] A. Riaz, A. Shah, A. Basit and M. Iqbal, “Experimental<br />

Study of Laminar Natural Convection Heat Transfer from<br />

Slender Circular Cylinder in Air Quiescent Medium,” in<br />

Proceedings of 2019 16th <strong>International</strong> Bhurban<br />

Conference on Applied Sciences & Technology (IBCAST),<br />

Islamabad, 2019.<br />

[8] M. Arshad, M. Inayat and I. Chughtai, “Experimental<br />

study of natural convection heat transfer from an<br />

enclosed assembly of thin vertical cylinders,” Applied<br />

Thermal Engineering, vol. 31, no. 1, pp. 20-27, 2011.<br />

[9] K. Hata, K. Fukuda and T. Mizuuchi, “Laminar natural<br />

convection heat transfer from vertical 7x7 rod bundles<br />

in liquid sodium,” <strong>Journal</strong> of <strong>Nuclear</strong> Engineering and<br />

Radiation Science, 2018.<br />

[10] N. Hattori and Y. Isahai, “A Numerical Study of Natural<br />

Convection in a Vertical Cylinder Bundle,” Heat Transfer-<br />

Asian Research, vol. 32, no. 4, 2003.<br />

[11] A. J. N. Khalifa and Z. A. Hussien, “Natural convection<br />

heat transfer from a single and multiple heated thin<br />

cylinders in water,” Heat Mass Transfer, vol. 51,<br />

pp. 1579-1586, 2015.<br />

[12] K. Tehseen, K. Qureshi, M. Basit, R. Nawaz, W. Siddique<br />

and R. Khan, “Computational Heat Transfer Analysis of<br />

Tubes and Tube Bundles with Supercritical Water as<br />


POWER, vol. 65, no. 11-12, pp. 588-594, 2020.<br />

Authors<br />

Atif Mehmood<br />

Lecturer, Department of<br />

Mechanical Engineering,<br />

Pakistan Institute of<br />

Engineering and Applied<br />

Sciences (PIEAS), Nilore,<br />

Islamabad<br />

atifmehmood@<br />

pieas.edu.pk<br />

Lead author of this article, Mr. Atif Mehmood, is a<br />

lecturer at Department of Mechanical Engineering,<br />

Pakistan Institute of Engineering and Applied Sciences<br />

(PIEAS), Islamabad. He completed his bachelors in<br />

mechanical engineering from National University of<br />

Sciences and Technology (NUST), Islamabad be<strong>for</strong>e<br />

joining PIEAS <strong>for</strong> post-graduate studies. He has an<br />

interest in thermal hydraulics as well as design<br />

engineering.<br />

Ajmal Shah<br />

Professor, Department of Mechanical Engineering,<br />

Pakistan Institute of Engineering and Applied Sciences<br />

(PIEAS), Nilore, Islamabad<br />

ajmal@pieas.edu.pk<br />

Mazhar Iqbal<br />

Lecturer, Department of Mechanical Engineering,<br />

Pakistan Institute of Engineering and Applied Sciences<br />

(PIEAS), Nilore, Islamabad<br />

mazhariqbal@<br />

pieas.edu.pk<br />

Ali Riaz<br />

Post-graduate fellow, Department of Mechanical<br />

Engineering, Pakistan Institute of Engineering and<br />

Applied Sciences (PIEAS), Nilore, Islamabad<br />

aliriaz3.60@gmail.com<br />

Muhammad Ahsan Kaleem<br />

Lecturer, Department of Mechanical Engineering,<br />

Pakistan Institute of Engineering and Applied Sciences<br />

(PIEAS), Nilore, Islamabad<br />

ahsankaleem@pieas.edu.pk<br />

Abdul Quddus<br />

PhD scholar, Department of Mechanical Engineering,<br />

Pakistan Institute of Engineering and Applied Sciences<br />

(PIEAS), Nilore, Islamabad<br />

engquddus613@yahoo.com<br />

Research and Innovation<br />

Experimental Study of Convective Heat Transfer Through Fuel Pins of a <strong>Nuclear</strong> <strong>Power</strong> Plant ı Atif Mehmood, Ajmal Shah, Mazhar Iqbal, Ali Riaz, Muhammad Ahsan Kaleem and Abdul Quddus

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Preliminary CFD Analysis of Innovative<br />

Decay Heat Removal System <strong>for</strong> the<br />

European Sodium Fast Reactor Concept<br />

Aleksander Grah, Haileyesus Tsige-Tamirat, Joel Guidez, Antoine Gerschenfeld, Konstantin Mikityuk,<br />

Janos Bodi and Enrico Girardi<br />

A major safety design objective <strong>for</strong> sodium cooled fast reactors is the practical elimination of the prolonged failure<br />

of the decay heat removal function. In order to achieve this objective, innovative decay heat removal system concepts<br />

are being developed. This paper discusses the preliminary CFD analysis of the design of an Innovative decay heat<br />

removal system implemented in the reactor pit which replaces the safety vessel of the European Sodium Fast Reactor.<br />

The efficiency of the oil cooling system is evaluated. A representative part of the reactor pit structure is modelled. The<br />

heat path is from the reactor vessel wall through a gap, an insulation layer into a concrete structure. The heat sinks are<br />

the oil cooling system and the ambient. A simplified case is used <strong>for</strong> verification with an analytical solution. It is shown<br />

that the oil cooling system is capable to remove the heat from the structure and keep the temperature there below<br />

70 degrees Celsius. At nominal conditions about 3 MW have to be removed and at a top oil cooling temperature of<br />

approximately 200 degrees.<br />


1 Introduction<br />

Sodium-cooled Fast Reactors (SFRs)<br />

are among the most advanced concept<br />

of the Generation-IV reactor systems<br />

[1] with extensive design, construction,<br />

operation and decommissioning<br />

experience. Several experimental, prototype<br />

and power reactors have been<br />

built in Europe and elsewhere [2].<br />

Currently, there are two com mercial<br />

operating SFRs in Russia as well as<br />

experimental reactors ope rating in<br />

China and India. SFRs have several<br />

attractive features which allow them to<br />

meet the Generation-IV <strong>International</strong><br />

Forum (GIF) objectives <strong>for</strong> future<br />

nuclear energy systems in terms of<br />

safety, sustainability, economics and<br />

proliferation re sistance. The fast neutron<br />

spectrum in connection with low<br />

neutron absorption and low moderation<br />

characteristic leads to enhanced<br />

neutron economy, which allows in<br />

addition to efficient fission power<br />

generation, the breeding of nuclear<br />

fuel and burning of trans uranic and<br />

minor actinides. In addition, the outstanding<br />

heat transfer property of<br />

sodium enables the design of a compact<br />

reactor core and high- per<strong>for</strong>mance,<br />

low-pressure heat transfer<br />

system. The SFR technology is in<br />

particular suitable to meet the twin<br />

missions of sustainability and improved<br />

economics both by efficient<br />

resource utilization and the effective<br />

management of plutonium and other<br />

minor actinides in a closed fuel cycle<br />

and by efficient power production<br />

using innovative power conversion<br />

systems.<br />

The current SFR technology R&D<br />

focuses on innovations to improve<br />

plant economics by enhancing reliability<br />

and reduction of capital cost as<br />

well as to enhance safety up to the<br />

level of the requirements <strong>for</strong> the<br />

Generation-IV plants. The safety goals<br />

<strong>for</strong> Generation IV reactors require<br />

excelling in safety and reliability; the<br />

reduction of the likelihood and degree<br />

of reactor core damage; and the<br />

elimination of the need <strong>for</strong> offsite<br />

emergency response.<br />

In order to meet these goals innovative<br />

safety concepts are needed in<br />

all areas of safety design of the<br />

fundamental safety functions. Beside<br />

reliable and robust reactivity control<br />

mechanism, highly reliable heat removal<br />

systems <strong>for</strong> assuring adequate<br />

cooling of safety relevant components<br />

and structures and effective options<br />

<strong>for</strong> dealing with severe accidents need<br />

to be implemented.<br />

Within the FP7 Euratom project<br />

CP-ESFR [3], the concept of the<br />

European Sodium-cooled Fast Reactor<br />

(ESFR) has been proposed and is<br />

currently being further improved<br />

within the Euratom Horizon-2020<br />

project ESFR-SMART [4] where new<br />

safety provisions have been proposed<br />

considering safety objectives envisaged<br />

<strong>for</strong> Generation-IV reactors and the<br />

update of European and international<br />

safety frameworks [5], taking into<br />

account the Fukushima accident. The<br />

proposed measures aim at enhancing<br />

safety and improving the robustness<br />

of the safety demonstration. The<br />

ESFR- SMART project aims to improve<br />

the reactor safety to the level of the<br />

requirements <strong>for</strong> the Generation-IV<br />

reactors, and make a proposal <strong>for</strong><br />

new safety options, based on both<br />

present and previous projects experiences.<br />

In line with these objectives,<br />

several innovative approaches to the<br />

safety design of the ESFR concepts<br />

are being investigated. One of the<br />

safety innovations concerns the<br />

re- placement of the safety vessel with<br />

reactor pit which is able to provide<br />

improved operational and safety<br />

functions of both the decay hear<br />

removal system and the containment<br />

function. With this innovation, the<br />

aim is to discard the safety vessel and<br />

to keep its safety functions, i.e., containment<br />

of the primary sodium in<br />

case of the reactor vessel leak without<br />

reduction of the primary sodium free<br />

level below the intermediate heat<br />

exchanger windows, by modifying the<br />

reactor pit geometry and by using the<br />

metallic liner on the reactor pit<br />

surface. In addition, this approach<br />

provides options to implement a new<br />

decay heat removal option within the<br />

reactor pit. With the implementation<br />

of this decay heat removal system, an<br />

additional safety provision is provided<br />

which could allow to practically<br />

eliminate a major safety design<br />

concern <strong>for</strong> sodium cooled fast<br />

reactors that of the prolonged failure<br />

of the decay heat removal function.<br />

This paper discusses the preliminary<br />

CFD analysis of the design of<br />

the innovative decay heat removal<br />

system implemented in the reactor pit<br />

which consists of oil and water based<br />

loops that provides capability to<br />

remove the entire decay heat of the<br />

reactor. Furthermore, the paper<br />

explains the overall design of the<br />

reactor pit including the preliminary<br />

Research and Innovation<br />

Preliminary CFD Analysis of Innovative Decay Heat Removal System <strong>for</strong> the European Sodium Fast Reactor Concept ı<br />

Aleksander Grah, Haileyesus Tsige-Tamirat, Joel Guidez, Antoine Gerschenfeld, Konstantin Mikityuk, Janos Bodi and Enrico Girardi

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


| Fig. 1.<br />

ESFR concept plant layout.<br />

parametric calculations of temperature<br />

distributions <strong>for</strong> various design<br />

scenarios. The efficiency of the oil<br />

cooling system is evaluated. A representative<br />

part of the reactor pit<br />

structure is modelled. The heat path is<br />

from the reactor vessel wall through a<br />

gap, an insulation layer into a concrete<br />

structure. The heat sinks are the<br />

oil cooling system and the ambient.<br />

The oil cooling system is situated at<br />

the metal liner which is mounted<br />

along the insulation layer. In the gap.<br />

A simplified case is used <strong>for</strong> verification<br />

with an analytical solution.<br />

Furthermore, it is shown that the oil<br />

cooling system is capable to remove<br />

the heat from the structure and keep<br />

the temperature there below 70 °C. At<br />

nominal conditions about 3 MW have<br />

to be removed and at a top cooling<br />

temperature of approximately 200 °C.<br />

2 ESFR system layout<br />

2.1 ESFR design approach<br />

The ESFR concept has been developed<br />

within the FP7 CP-ESFR project (2009<br />

to 2013), considering past European<br />

experience in SFR technology in<br />

particular in the French SPX2 project<br />

and in European Fast Reactor project<br />

which involved a wider European<br />

cooperation. For both projects, the<br />

operational experience of the Superphenix<br />

reactor SPX [6][7] provided<br />

valuable inputs to improve safety,<br />

reliability, in service inspection and<br />

repair and economics.<br />

The ESFR concept is a large pool<br />

type industrial Sodium Fast Reactor of<br />

1,500 MWe/3,600 MW th . The design<br />

objectives <strong>for</strong> ESFR include simplification<br />

of structures, improved In-service<br />

inspection and repair capabilities,<br />

reduction of risks related to sodium<br />

fires and to the water/sodium<br />

reaction, improved fuel maintenance,<br />

core catcher with the capability <strong>for</strong> a<br />

whole core discharge and improved<br />

robustness against external hazards<br />

[5]. The ESFR core is composed of<br />

two zones of inner and outer fuel<br />

assemblies, and 3 rows of reflectors.<br />

There are two independent control<br />

rod assembly systems with additional<br />

passive reactivity insertion mechanisms.<br />

The core design aims at a fuel<br />

management scheme with a flexible<br />

breeding and minor actinide burning<br />

strategy.<br />

The ESFR concept plant layout is<br />

sketched in Figure 1. It is based on<br />

options already considered in previous<br />

and existing pool sodium<br />

fast reactors, with several potential<br />

improvements regarding safety, inspection<br />

and manufacturing. A<br />

particular attention is also given to<br />

compactness. The reactor vessel is<br />

cooled with sodium (submerged weir)<br />

and is surrounded by a reactor pit<br />

which replaces the safety vessel<br />

including all its function <strong>for</strong> normal<br />

operation and accident conditions.<br />

The reactor vault can be inspected <strong>for</strong><br />

maintenance.<br />

The secondary system comprises<br />

six 600 MWth parallel and independent<br />

sodium loops, each connected<br />

to an intermediate heat<br />

exchanger (IHX) located in the reactor<br />

vessel. Each loop (see Figure 1)<br />

includes one Mechanical Secondary<br />

Pump (MSP), modular Steam Generators<br />

(SG) and one Sodium Dump<br />

Vessel (SDV). In addition to the<br />

normal power removal, the secondary<br />

system provides the DHR function in<br />

case of primary pumps trip.<br />

The main design objective <strong>for</strong> the<br />

Decay Heat Removal (DHR) system is<br />

to practically eliminate the prolonged<br />

loss of the decay heat removal<br />

function. In order to achieve this<br />

objective, three independent DHR<br />

systems have been implemented (see<br />

Figure 1). The first DHR system,<br />

DHRS-1, is provided by sodium/air<br />

heat exchangers connected to each<br />

IHX. These loops replace the Direct<br />

Reactor Cooling (DRC) System in the<br />

previous design and have various<br />

advantages such as avoiding additional<br />

roof penetrations and maintaining<br />

cold column in the IHXs [5]. The<br />

second DHR system, DHRS-2, is<br />

provided by cooling the steam generator<br />

modules by air in natural or <strong>for</strong>ced<br />

convection through hatch openings,<br />

as is done in the Phenix reactor, providing<br />

the heat sink <strong>for</strong> the secondary<br />

loop. The third DHR system, DHRS-3,<br />

is implemented in the reactor pit with<br />

two independent cooling circuits, one<br />

with oil heat exchanger brazed on the<br />

liner and one with water inside the<br />

concrete. The water loop is capable to<br />

maintain the whole pit at temperatures<br />

below 70 °C. Due to the removal<br />

of the safety vessel, the DHRS-3 which<br />

is directly attached to the liner is<br />

expected to be more efficient and to<br />

assure a large part of the Decay Heat<br />

Removal close to 100 %.<br />

The safety analyses being currently<br />

per<strong>for</strong>med indicate that with the<br />

current DHR concept the demonstration<br />

of practical elimination of a<br />

prolonged loss of the DHR function is<br />

feasible.<br />

2.2 Description of the reactor<br />

pit design<br />

One of the main innovative approaches<br />

to the safety design of the ESFR<br />

concept concerns the replacement of<br />

the safety vessel with reactor pit that<br />

aims to improve the operational and<br />

safety functions of both the decay heat<br />

removal systems and the containment.<br />

All existing SFRs have a safety<br />

vessel (see an example of Super phenix<br />

safety vessel Figure 2) around the<br />

main vessel. The function of this safety<br />

vessel is to confine the primary sodium<br />

in case of the main vessel leakage, so<br />

as to avoid lowering of the primary<br />

sodium free level below the inlet<br />

windows of the intermediate heat<br />

exchangers and thus to provide an<br />

efficient natural convection through<br />

the core. In case of the main vessel<br />

leakage, the reactor is not recoverable<br />

and the core must be unloaded. Due to<br />

the need to wait <strong>for</strong> reduction of the<br />

residual power of the assemblies, this<br />

handling could take a significant<br />

duration (i.e. higher than one year)<br />

especially in the ESFR-SMART design<br />

without external sodium storage. The<br />

safety vessel must there<strong>for</strong>e remain<br />

filled with sodium <strong>for</strong> a long time. The<br />

Research and Innovation<br />

Preliminary CFD Analysis of Innovative Decay Heat Removal System <strong>for</strong> the European Sodium Fast Reactor Concept ı<br />

Aleksander Grah, Haileyesus Tsige-Tamirat, Joel Guidez, Antoine Gerschenfeld, Konstantin Mikityuk, Janos Bodi and Enrico Girardi

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

| Fig. 2.<br />

Arrival of the safety vessel inside the reactor<br />

pit of Superphenix.<br />

potential danger in these conditions is<br />

that the reactor pit is not designed to<br />

withstand a sodium leak from this<br />

safety vessel. Moreover, this safety<br />

vessel leak would also lead to interruption<br />

of the core cooling by natural<br />

convection, leading to a very difficult<br />

overall situation.<br />

A number of measures have been<br />

taken to prevent leakage of the safety<br />

vessel: slight overpressure between<br />

the two vessels to detect a possible<br />

leak and choice of different materials<br />

to avoid a common failure mode on<br />

corrosion. It is recalled that it is a<br />

problem of corrosion on welds that<br />

led to the leakage of the Superphenix<br />

storage drum vessel, and that this leak<br />

was taken up by the safety vessel of<br />

this storage drum [3].<br />

The scenarios of vessel leakage are<br />

diverse, from corrosion leakage to<br />

leakage on a severe accident with<br />

mechanical energy release. This leads<br />

to high uncertainties in the temperatures<br />

and leakage rates, which<br />

make it difficult to demonstrate the<br />

safety vessel mechanical strength<br />

against the corresponding thermal<br />

shocks. Moreover, the French licensing<br />

authority requires considering the<br />

double leakages in order to verify that<br />

the situation does not lead to cliff-edge<br />

effect in term of radiological releases<br />

in the environment. This demonstration<br />

was required after the SPX<br />

external sodium storage leak. It is<br />

particularly required if the core<br />

unloading is high.<br />

The safety vessel is a proven option,<br />

especially demonstrated during the<br />

incident at the Superphénix storage<br />

drum [6], and is adopted in all existing<br />

fast reactors. However, the evolution<br />

of safety standards leads us to look at<br />

other options where its functions<br />

could be directly taken over by a<br />

reactor pit capable of withstanding a<br />

sodium leak, and thus a long-term<br />

mitigation situation. It was an option<br />

that had already been looked at in the<br />

EFR project with a vessel anchored in<br />

the pit, which option was later abandoned<br />

<strong>for</strong> reasons of feasibility and<br />

design difficulties.<br />

The proposed design of the reactor<br />

pit is composed of the following<br />

domains (see Figure 3, Figure 4 and<br />

Figure 5):<br />

p A mixed concrete/metal structure<br />

with a water cooling system inside<br />

the concrete supports the thick<br />

metal slab to which the reactor<br />

vessel is attached. Together with<br />

the reactor roof, it provides a<br />

sealed containment which must<br />

keep its integrity in all the cases of<br />

normal or accidental operations<br />

[1].<br />

p Inside the concrete/metal structure,<br />

blocks of insulating materials<br />

(non-reactive with sodium) are<br />

installed. Alumina is selected as<br />

reference material <strong>for</strong> the insulation<br />

blocks. A conventional insulation<br />

layer could be considered in<br />

future to increase insulation effects<br />

(outside the scope of the paper).<br />

A metallic liner is placed on the<br />

surface of the insulation blocks. The<br />

gap between the reactor vessel and<br />

the liner must be small enough<br />

(350 mm was chosen) to avoid<br />

decrease of the primary sodium free<br />

level below the intermediate heat<br />

exchanger (IHX) windows in case of<br />

sodium leakage from the reactor<br />

vessel. During normal operation, the<br />

primary sodium free level is 1,350 mm<br />

below the roof. In case of primary<br />

sodium leak about 300 m 3 of sodium<br />

will leave the reactor vessel to fill the<br />

gap and the new equilibrium free level<br />

of the primary sodium will be about<br />

3,070 mm below the reactor roof.<br />

| Fig. 4.<br />

Drawing of the reactor pit design.<br />

With this level of sodium inside the<br />

primary circuit, there is still a<br />

1,090 mm sodium level above the<br />

upper IHX openings, which allows a<br />

sodium inflow into the IHX and a good<br />

natural convection and core cooling. If<br />

the sodium temperature decrease to<br />

180 °C , the volume of remaining<br />

sodium will decrease of about 205 m 3<br />

due to the change in the density, It<br />

would cause the sodium level to go<br />

down to around 50 mm above the IHX<br />

entry bottom. And the natural convection<br />

remains possible.<br />

p The oil cooling system is installed<br />

next to or even inside the liner<br />

(Figure 3).<br />

| Fig. 3.<br />

Detail of the reactor pit design (1 reactor vessel, 2 liner, 3 insulation,<br />

4 concrete/metal structure, 5 oil cooling system (DHRS-3.1), 6 water<br />

cooling system (DHRS-3.2, 7 gap)).<br />


Research and Innovation<br />

Preliminary CFD Analysis of Innovative Decay Heat Removal System <strong>for</strong> the European Sodium Fast Reactor Concept ı<br />

Aleksander Grah, Haileyesus Tsige-Tamirat, Joel Guidez, Antoine Gerschenfeld, Konstantin Mikityuk, Janos Bodi and Enrico Girardi

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


(a)<br />

| Fig. 5.<br />

Drawings of the reactor pit cooling systems: (a) oil and (b) water.<br />

p Finally, a special concrete with<br />

alumina (aluminous concrete)<br />

which could withstand, without<br />

significant chemical reaction with<br />

sodium, a leakage of the liner<br />

could be used between the liner<br />

and the insulation (blocks of<br />

alumina).<br />

Two independent active cooling<br />

systems are proposed in the reactor pit<br />

(we use the acronyms DHRS-3 <strong>for</strong> the<br />

combination of these two systems):<br />

p The oil cooling system (DHRS-3.1)<br />

close to the liner (Figure 5a). The<br />

oil under <strong>for</strong>ced convection can<br />

remove the heat transferred by<br />

radiation from the reactor vessel at<br />

high temperature. Conversely to<br />

water, the adopted synthetic oil is<br />

resistant to high temperatures<br />

above 300 °C and reacts with<br />

sodium without producing hydrogen.<br />

As an example the commercial<br />

oil called ”Therminol SP“ [8] can<br />

be used in normal operation at<br />

temperatures up to 315 °C.<br />

p The water cooling system<br />

(DHRS-3.2) <strong>for</strong> the concrete<br />

cooling is installed in the concrete<br />

(Figure 5b) and aims at maintaining<br />

the concrete temperature<br />

under 70 °C in all possible situations,<br />

even if the oil system is lost.<br />

Both oil and water circuits work<br />

during normal operation and have to<br />

maintain the concrete temperature<br />

below 70 °C. This margin is intended<br />

(b)<br />

to ensure the concrete integrity and to<br />

protect it from thermal degradation.<br />

During the reactor shutdown the oil<br />

system alone has to be able after few<br />

days to remove all the decay heat<br />

generated by the fuel. In case of the<br />

reactor vessel leak and loss of the oil<br />

system, the water system should be<br />

able to remove the decay heat<br />

generated by the core and to maintain<br />

the concrete below 70 °C.<br />

2.3 Main design-basis<br />

scenarios<br />

The reactor pit must be designed <strong>for</strong><br />

the following three main scenarios:<br />

p Scenario 1: Normal operation:<br />

The main vessel temperature is at<br />

about 400 °C. The operation of the<br />

oil cooling system is sufficient to<br />

maintain the correct thermal<br />

conditions in the pit (i.e. less than<br />

70 °C <strong>for</strong> the concrete of the mixed<br />

structures).<br />

p Scenario 2: Operation in exceptional<br />

decay heat removal<br />

regime: The safety studies should<br />

take into account exceptional<br />

situations of successive losses of<br />

decay heat removal systems. In this<br />

case, in exceptional situations of<br />

Categories 3 and 4, the reactor<br />

vessel is allowed to reach temperature<br />

of 650 °C. The two cooling<br />

systems (oil and water) must make<br />

it possible to maintain the concrete<br />

temperature below 70 °C while<br />

playing an important role in the<br />

decay heat removal (in this study<br />

only the oil cooling system is taken<br />

into account; further publication is<br />

following).<br />

p Scenario 3: Operation in accident<br />

situation of sodium leakage:<br />

In a situation of little leak,<br />

vigorous sodium cooling is possible<br />

with the redundant and available<br />

DHRS, to bring the sodium to a<br />

temperature corresponding to the<br />

handling temperature (180 °C).<br />

There<strong>for</strong>e, the maximum temperature<br />

of the sodium in the gap<br />

should not exceed 200 °C. The<br />

demonstration of the oil cooling<br />

system availability in case of<br />

reactor vessel leakage is difficult<br />

and we assume as hypothesis that<br />

the oil cooling system is no longer<br />

available. The operation of the<br />

water cooling system alone must<br />

be sufficient to maintain the<br />

concrete temperature below 70 °C<br />

(further publication).<br />

NB: But other studies need later to<br />

be per<strong>for</strong>med, taking in account a<br />

leakage (or combined) due to a high<br />

thermal transient (e.g., due to failure<br />

of DHR system) and if the failure is<br />

due to the severe accident. In this case<br />

the sodium temperature will be higher<br />

3 Computational fluid<br />

dynamics analysis<br />

3.1 CFD model<br />

The objective is the development of a<br />

simple Computational Fluid Dynamics<br />

(CFD) model to compute the steadystate<br />

heat transfer from the reactor<br />

vessel through the reactor pit. The<br />

computations are per<strong>for</strong>med with<br />

ANSYS CFX which is a parallelized<br />

high-per<strong>for</strong>mance CFD software tool<br />

[9]. The steady-state model is based<br />

on finite volume technique applied to<br />

solve the Navier-Stokes equations. To<br />

achieve results with low computation<br />

time, the reactor pit is divided into<br />

several sections and the CFD analysis<br />

is per<strong>for</strong>med <strong>for</strong> every section. The<br />

drawing of the geometry <strong>for</strong> one<br />

section, the “elementary cell” of the<br />

reactor pit structure, is shown in<br />

Figure 6. For the calculation example,<br />

the oil cooling system is installed<br />

inside the liner of the special wavy<br />

shape (see Figure 7).<br />

The following simplifications and<br />

assumptions are applied at this stage<br />

of computations:<br />

p The gap between the reactor vessel<br />

and the liner is considered as<br />

vacuum <strong>for</strong> first CFD computations<br />

to minimize the CPU time.<br />

Research and Innovation<br />

Preliminary CFD Analysis of Innovative Decay Heat Removal System <strong>for</strong> the European Sodium Fast Reactor Concept ı<br />

Aleksander Grah, Haileyesus Tsige-Tamirat, Joel Guidez, Antoine Gerschenfeld, Konstantin Mikityuk, Janos Bodi and Enrico Girardi

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

| Fig. 6.<br />

Drawing of an “elementary cell” of the reactor pit model.<br />

| Fig. 7.<br />

The CFD model of an “elementary cell” of the reactor pit model.<br />

p The material of the insulation layer<br />

is assumed to be glass wool <strong>for</strong> the<br />

current study.<br />

p Only steady state computations are<br />

per<strong>for</strong>med.<br />

Fast solution is important to be able to<br />

per<strong>for</strong>m the large amounts of<br />

parametric studies. The resulting<br />

CFD model geometry is shown in<br />

Figure 6. The domains <strong>for</strong> the model<br />

are as follows (l is the thermal<br />

conductivity):<br />

p The outer surface of the reactor<br />

vessel wall is shown on the lefthand<br />

side (red).<br />

p The gas gap (white); l =<br />

0.026 W m -1 K -1 . However, <strong>for</strong> the<br />

computations nearly-vacuum is<br />

assumed here as worst case and the<br />

heat transport parameters (density<br />

and specific heat) are set to very<br />

low values and the solution of the<br />

flow is switched off.<br />

p The stainless steel liner of the wavy<br />

shape is proposed (blue); l =<br />

60.5 W m -1 K -1 with the pipes of the<br />

oil cooling system inside (dark<br />

blue).<br />

p The insulation layer (yellow); l =<br />

0.04 W m -1 K -1 .<br />

p The concrete structure (grey); l =<br />

1.4 W m -1 K -1 .<br />

p The right-hand side surface interfaces<br />

the environment (black).<br />

The boundary conditions <strong>for</strong> the CFD<br />

model in Figure 6 are as follows:<br />

p Fixed temperature of the reactor<br />

vessel wall (red surface at the left)<br />

<strong>for</strong> different axial levels (T vw =<br />

400 °C, 500 °C, 600 °C and 700 °C).<br />

This is intended to represent the<br />

different levels of the heat source<br />

inside the vessel.<br />

p The heat sink is the environment<br />

outside of the concrete structure<br />

(black surface at the right). As a<br />

first approximation the following<br />

data is taken <strong>for</strong> the heat transfer:<br />

T a = 50 °C, k = 6 W m -2 K -1 , where<br />

k is the heat exchange coefficient.<br />

For later calculations, the heat<br />

sinks will include the water cooling<br />

system.<br />

p The other outer surfaces are<br />

defined as “symmetric”, i.e.<br />

adiabatic.<br />

Calculations were made first<br />

without taking into account the oil<br />

cooling system in order to see if<br />

one could reach, with the only<br />

heat removal to the environment,<br />

required temperature level in the<br />

concrete domain.<br />

3.1 Analytical solution<br />

<strong>for</strong> verification<br />

The current CFD model without<br />

local heat sinks (the cooling systems)<br />

is convenient <strong>for</strong> verification by<br />

com parison to an analytical solution.<br />

This requires one-dimensional<br />

model of the heat transfer by radiation<br />

in the gap and by heat conduction<br />

in the solids. The heat<br />

path is from the hot reactor vessel<br />

wall towards the outer concrete<br />

surface facing the environment.<br />

The heat flow due to radiation in the<br />

gap between a hot surface (index 1)<br />

and a cold surface (index 2) can be<br />

written as:<br />

(1)<br />

Hereby T is the (absolute) temperature<br />

and σ = 5.67⋅10 -8 W m -2 K -4 is the<br />

Stefan-Boltzmann constant. Assuming<br />

emissivity of stainless steel <strong>for</strong> both<br />

surfaces (ε1 = ε2 = ε = 0.4, a typical<br />

value <strong>for</strong> stainless steel) and full<br />

visibility (F 12 = 1) and <strong>for</strong> A 1 ≈ A 2 = A<br />

the temperature of the cold surface<br />

yields:<br />

(2)<br />

Heat conduction through a cylindrical<br />

solid wall in direction of the radius r<br />

can be written as<br />

<br />

(2)<br />

Hereby l is the thermal conductivity<br />

of the solid and A(r) = 2πrh the<br />

cylindrical surface at the radius r and<br />

<strong>for</strong> a segment of the height h. The<br />

integration between the radii r 1 and r 2<br />

with the corresponding temperatures<br />

T 1 and T 2 yields the temperature at the<br />

outer radius<br />

<br />

(1)<br />

3.3 CFD results without<br />

the cooling system<br />

For this computation the oil cooling<br />

channels in the liner in Figure 6<br />

are not taken into account. The liner<br />

is considered to be flat (equal<br />

thickness) and without pipes. The<br />

temperature along the heat path<br />

in the centre of the domains is<br />

shown in Figure 8. In Figure 8a,<br />

the constant-temperature boundary<br />

condition is applied at the primary<br />

vessel wall (T vw = 400 °C, 500 °C,<br />

600 °C and 700 °C, red surface in<br />

Figure 7). In the nearly-vacuum<br />

domain of the gap the heat transfer<br />

takes place by means of radiation and<br />

the temperature is almost identical to<br />

the boundary condition. The temperature<br />

of the metal liner is also very<br />

close to the temperature of the vessel<br />

(since the cooling system is not<br />

modelled in this case). As expected,<br />

the main temperature drop takes<br />

place in the insulation layer. However,<br />

the temperature in the concrete is<br />

mostly above the required limit of<br />

70 °C as shown in Figure 8b. Furthermore,<br />

in Figure 8a and b the CFD<br />

results are compared with the analytical<br />

solutions of Eq. (2) and Eq. (4).<br />

The maximum deviation of the<br />

temperature is less than 5 °C.<br />


Research and Innovation<br />

Preliminary CFD Analysis of Innovative Decay Heat Removal System <strong>for</strong> the European Sodium Fast Reactor Concept ı<br />

Aleksander Grah, Haileyesus Tsige-Tamirat, Joel Guidez, Antoine Gerschenfeld, Konstantin Mikityuk, Janos Bodi and Enrico Girardi

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


(a)<br />

| Fig. 8.<br />

The temperature along the heat path (in x direction); computation (CFD) and analytical solution (Ana); a) the whole length; b) the concrete domain.<br />

The dotted red line denotes the temperature limit of 70°C.<br />

| Fig. 9.<br />

Temperature field, T vw = 700 °C, T cc = 200 °C.<br />

3.4 CFD results with<br />

the cooling system<br />

For this computation a constant<br />

average temperature T cc at the oil<br />

cooling channel walls (Figures 5 and<br />

6) is set as boundary condition. The<br />

aim is to understand the interaction<br />

between the reactor vessel wall<br />

temperature, the oil cooling system<br />

temperature, and the maximum<br />

concrete temperature. The final goal is<br />

to estimate the power removed by the<br />

oil cooling system.<br />

In Figure 9 the temperature field<br />

<strong>for</strong> the computed elementary cell is<br />

depicted. The hot reactor vessel with<br />

the constant temperature of the vessel<br />

wall equal to T vw = 700 °C is on the<br />

left-hand side. The temperature of the<br />

oil cooling channel walls is set to T cc =<br />

200 °C. Most of the heat transfer takes<br />

place between the vessel wall and the<br />

cooling channel. The maximum<br />

concrete temperature is slightly below<br />

70 °C.<br />

In Figure 10 the temperature is<br />

shown along the main heat path (from<br />

the hot main vessel wall at the left<br />

towards the ambient on the right). In<br />

Figure 10a the temperature is shown<br />

along the x axis (symmetry centre line<br />

of the elementary cell) and along a<br />

(b)<br />

line cutting the centre of one of<br />

the cooling channels. The case<br />

corresponds to the temperature field<br />

in Figure 9. As expected, the hottest<br />

location of the liner is at the centre<br />

line (maximum distance to the cooling<br />

channels). The temperature minimum<br />

is defined by the cooling channel temperature<br />

(T cc = 200 °C). In Figure 10b<br />

and 10c the temperature is given <strong>for</strong><br />

four different vessel wall temperatures.<br />

For all cases, the concrete<br />

temperature remains below the limit<br />

(70 °C) denoted by the red dotted<br />

line. The case with T vw = 700 °C corresponds<br />

to Figure 10a and to the<br />

temperature field shown in Figure 9.<br />

In Figure 11a the maximum concrete<br />

temperature T c is shown versus<br />

the cooling channel temperature <strong>for</strong><br />

different vessel wall temperatures.<br />

The case discussed in Figure 9 and in<br />

10a corresponds to the point T vw =<br />

700 °C and T cc = 200 °C which is<br />

slightly below the dotted red line.<br />

Consequently, according to this model<br />

the temperature of the cooling channel<br />

wall has to be maintained below<br />

200 °C to keep the concrete temperatures<br />

below the limit. For lower vessel<br />

wall temperatures the cooling channel<br />

wall temperature can be higher.<br />

(a) (b) (c)<br />

| Fig. 10.<br />

Temperature profile along the x-axis <strong>for</strong> a cooling channel temperature of T cc = 200°C.<br />

Research and Innovation<br />

Preliminary CFD Analysis of Innovative Decay Heat Removal System <strong>for</strong> the European Sodium Fast Reactor Concept ı<br />

Aleksander Grah, Haileyesus Tsige-Tamirat, Joel Guidez, Antoine Gerschenfeld, Konstantin Mikityuk, Janos Bodi and Enrico Girardi

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

(a)<br />

| Fig. 11.<br />

a) The maximum concrete temperature and b) <strong>Power</strong> removed by the oil cooling system at differ-ent temperatures of the reactor vessel and the oil cooling<br />

channel wall.<br />

Parametric calculations of the<br />

power removed by the oil cooling<br />

system at different vessel temperatures<br />

are shown in Figure 11b. They<br />

are evaluated with grey body factors<br />

of vessel and liner equal to 0.4 that<br />

is a typical value <strong>for</strong> stainless steel.<br />

If necessary this value could be<br />

increased by liner surface processing<br />

to increase its emissivity and the<br />

associated decay heat removal capacity.<br />

Calculations were per<strong>for</strong>med<br />

<strong>for</strong> various oil temperatures (Figure<br />

11b). An average temperature of<br />

about 200 °C is proposed <strong>for</strong> operation<br />

which is far below the operating range<br />

of the “Therminol SP” oil (315 °C).<br />

The power removed by radiation<br />

from the reactor vessel at 400 °C<br />

(~nominal core inlet temperature) is<br />

about 3 kW/m 2 .The surface of the<br />

reactor vessel, radiating towards<br />

the oil cooling system, is about<br />

1,050 m 2 . So at nominal operation,<br />

approximately 3 MW will have to be<br />

removed in order to maintain the<br />

oil at an average temperature of<br />

approximately 200 °C.<br />

In exceptional regime of decay<br />

heat removal (situations in category 3<br />

and 4) the system can then remove<br />

(the main vessel being at 650 °C),<br />

a power of about 15 MW. This<br />

value doesn’t take into account the<br />

exchanges by gas convection between<br />

vessel and liner, and could be also<br />

increased by special processing of the<br />

liner surface to increase its emissivity<br />

coefficient. The value of 15 MW corresponds<br />

to the decay heat power<br />

level after about three days.<br />

For Scenario 2 (See Section “Main<br />

design-basis scenarios”) we assume<br />

that the average oil temperature (and<br />

there<strong>for</strong>e the liner) is at ~200 °C. For<br />

Scenario 3 we assume that the other<br />

DHRS guarantee that the primary<br />

sodium is at about 200 °C and the oil<br />

system is out of operation, there<strong>for</strong>e<br />

the liner will be also at ~200 °C. Thus,<br />

the conducted preliminary analysis<br />

based on the use of the oil cooling<br />

system alone is potentially applicable<br />

to all three scenarios. Nevertheless,<br />

the necessity of the water cooling<br />

system will be analysed at the next<br />

phase of the (more detailed) analysis.<br />

NB: thermal exchange have been<br />

calculated only with radiative and<br />

conduction effects. The exchanges by<br />

convection have not been taken in<br />

account and should be estimated later.<br />

4 Conclusion<br />

Within the Euratom H-2020 ESFR-<br />

SMART project, various design<br />

options have been proposed to<br />

improve the safety of the ESFR<br />

concept to the level of the requirements<br />

<strong>for</strong> the Generation-IV reactors<br />

based on both present and previous<br />

projects experiences. One of the main<br />

innovative approaches to the safety<br />

design of the ESFR concepts concerns<br />

the re-placement of the safety vessel<br />

with reactor pit in order to provide<br />

improved operational and safety<br />

functions of both the decay hear<br />

removal system and the containment.<br />

The focus of the present paper is the<br />

preliminary CFD analyses of the<br />

innovative decay heat removal design<br />

implemented in the reactor pit.<br />

Two active (<strong>for</strong>ced-convection)<br />

cooling systems are proposed: an oil<br />

cooling system close to the metallic<br />

liner and a water cooling system in<br />

the reactor pit concrete structure.<br />

The main goals of the cooling systems<br />

are:<br />

1) to keep the reactor vessel at<br />

acceptable temperature level and<br />

there<strong>for</strong>e provide its integrity and<br />

2) to maintain the concrete temperature<br />

below 70 °C in the three<br />

scenarios of operation envisaged<br />

(b)<br />

(normal operation, decay heat<br />

removal without and with primary<br />

sodium leak).<br />

The conducted CFD analysis in the<br />

present paper is concentrated on the<br />

evaluation of the oil cooling system. It<br />

considers vessel wall temperatures up<br />

to 700 °C. The main heat transfer<br />

mechanism from the wall towards a<br />

metal liner at the other side of the gap<br />

is assumed to be thermal radiation.<br />

It can be shown, in the scenarios<br />

studied, with the oil cooling system<br />

designed to keep the liner temperature<br />

at around 200 °C, that this system<br />

alone can fulfill the goals to safely<br />

remove the residual heat of about<br />

3 MW in nominal conditions and to<br />

maintain the temperature of the<br />

concrete of the reactor pit below 70 °C<br />

in the two accidental situations.<br />

The water cooling system seems only<br />

a supplementary safety protection<br />

against any pit structure concrete<br />

damage and will be subject of further<br />

publication.<br />

Acknowledgement<br />

The research leading to these results<br />

has received funding from the<br />

Euratom research and training<br />

programme 2014-2018 under grant<br />

agreement No 754501.<br />

References<br />

[1] A Technology Roadmap <strong>for</strong> Generation IV <strong>Nuclear</strong> Energy<br />

Systems, U.S. Department of Energy <strong>Nuclear</strong> Energy Research<br />

Advisory Committee and the Generation IV <strong>International</strong><br />

Forum (Dec. 2002).<br />

[2] Aoto, K., et al., A summary of sodium-cooled fast reactor<br />

development, Prog. Nucl. Ener. 77, 247- 265, (2014).<br />

[3] G.L. Fiorini and A. Vasile, European Commission – 7 th<br />

Framework Programme The Collaborative Project on European<br />

Sodium Fast Reactor (CP-ESFR), Nucl. Eng. and Des.241,<br />

3461– 3469, (2011).<br />

[4] K. Mikityuk, et al., ESFR-SMART: new Horizon-2020 project on<br />

SFR safety, IAEA-CN245-450, Proceedings of <strong>International</strong><br />

Conference on Fast Reactors and Related Fuel Cycles: Next<br />

Generation <strong>Nuclear</strong> Systems <strong>for</strong> Sustainable, Development<br />

FR17, Yekaterinburg, Russia, 26-29 June, (2017).<br />


Research and Innovation<br />

Preliminary CFD Analysis of Innovative Decay Heat Removal System <strong>for</strong> the European Sodium Fast Reactor Concept ı<br />

Aleksander Grah, Haileyesus Tsige-Tamirat, Joel Guidez, Antoine Gerschenfeld, Konstantin Mikityuk, Janos Bodi and Enrico Girardi

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


[5] J. Guidez, A. Rineiski, G. Prêle, E. Girardi, J. Bodi, K. Mikityuk,<br />

Proposal of new safety measures <strong>for</strong> European Sodium Fast<br />

Reactor to be evaluated in framework of Horizon-2020<br />

ESFR-SMART project, Proceedings of the <strong>International</strong><br />

Congress on Advances in <strong>Nuclear</strong> <strong>Power</strong> Plants (ICAPP 2018),<br />

Charlotte, North Carolina, USA, 8-11 April ,(2018).<br />

[6] J. Guidez and G. Prêle, “Superphenix. Technical and scientific<br />

achievements”, Edition Springer, ISBN 978-94-6239-245-8,<br />

(2017).<br />

[7] Guidez, J., Gerschenfeld, A., Grah, A., Tsige-Tamirat, H.,<br />

Mikityuk, K., Bodi, J., Girardi, E., “European Sodium Fast<br />

Reactor: innovative design of reactor pit aiming at suppression<br />

of safety vessel”, ICAPP 2019, May 12 – 15, Juan-les-pins,<br />

France.<br />

[8] Therminol® SP Heat Transfer Fluid:<br />

https://www.therminol.com/products/Therminol-SP<br />

[9] ANSYS CFX: https://www.ansys.com/products/fluids/ansys-cfx<br />

Authors<br />

Aleksander Grah<br />

JRC, Petten, Netherlands<br />

aleksander.grah<br />

@ec.europa.eu<br />

Dr. Aleksander Grah is Scientific Officer at the<br />

European Commission Joint Research Centre, Petten,<br />

The Netherlands. Currently, he is working in the<br />

field of Safety of advanced reactors applying Computational<br />

Fluid Mechanics (CFD).<br />

Haileyesus<br />

Tsige-Tamirat<br />

JRC, Petten, Netherlands<br />

haileyesus.tsige-tamirat<br />

@ec.europa.eu<br />

Antoine Gerschenfeld<br />

CEA, Paris-Saclay, France<br />

antoine.gerschenfeld<br />

@cea.fr<br />

Antoine Gerschenfeld is a research engineer at the<br />

Fluid Mechanics Section of the French Commissariat<br />

à l’Energie Atomique, specializing in the thermalhydraulics<br />

of Generation 4 nuclear reactors. He is also<br />

an assistant professor at Ecole Polytechnique and a<br />

member of the Framatome scientific advisory board.<br />

Konstantin Mikityuk<br />

PSI, Villigen AG,<br />

Switzerland<br />

konstantin.mikityuk<br />

@psi.ch<br />

In charge of Advanced <strong>Nuclear</strong> Systems group at<br />

Paul Scherrer Institute, Switzerland. Main interest:<br />

research related to design and safety assessment of<br />

Generation-IV fast-spectrum nuclear reactors<br />

Janos Bodi<br />

PSI, Villigen AG,<br />

Switzerland<br />

janos.bodi@psi.ch<br />

Dr. H. Tsige-Tamirat is Scientific Officer at the<br />

European Commission Joint Research Centre, Petten,<br />

The Netherlands. Currently, he is working in the field<br />

of Safety of advanced reactors.<br />

Joel Guidez<br />

CEA, Paris-Saclay, France<br />

joel.guidez@cea.fr<br />

Janos Bodi is a nuclear engineering professional.<br />

His main professional activities are related to<br />

neutronic and thermal-hydraulic analysis of<br />

Sodium-cooled Fast Reactors. Furthermore, he<br />

contributes to the reactor design development<br />

of Generation IV reactor concepts.<br />

Enrico Girardi<br />

EDF, Paris-Saclay, France<br />

enrico.girardi@edf.fr<br />

Joel Guidez retired in March 2020 while remaining<br />

scientific advisor to the CEA and working in ESFR<br />

SMART European project. Be<strong>for</strong>e this he was member<br />

of the operational committee of the office of the High<br />

Commissioner, honorary President of the ST7 SFEN<br />

section, member of the RGN editorial board,<br />

representative of France at the RSWG of GIF, president<br />

of the GCFS – French safety advisory group – tripartite<br />

CEA/EDF/AREVA, and scientific manager of the GEN<br />

IV segment at CEA.<br />

Dr. Enrico Girardi is currently an Expert on Reactor<br />

Physics, Modelling & Simulation at Eléctricité de<br />

France R&D Safety and Fuel Cycle group. His main<br />

research fields are on both PWR, SMR and SFR reactor<br />

physics, and in particular on neutronic modelling. In<br />

the last 15 years he has contributed to several reactor<br />

calculation chain development at EDF and has been<br />

leading the EDF R&D “Generation IV reactor” project<br />

from 2015 to 2018.<br />

Research and Innovation<br />

Preliminary CFD Analysis of Innovative Decay Heat Removal System <strong>for</strong> the European Sodium Fast Reactor Concept ı<br />

Aleksander Grah, Haileyesus Tsige-Tamirat, Joel Guidez, Antoine Gerschenfeld, Konstantin Mikityuk, Janos Bodi and Enrico Girardi

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Study on Verification of SPACE Code<br />

Based on an MSGTR Experiment at the<br />

ATLAS-PAFS Facility<br />

Kyungho Nam<br />

A Multiple Steam Generator Tube Rupture (MSGTR) accident is defined in Korea as an accident in which more than<br />

five U-tubes of a steam generator break down. To understand the thermal hydraulic phenomena of a MSGTR accident,<br />

an experimental study was conducted by KAERI. The experiment was conducted to simulate the transient phenomena<br />

caused by the rupture of five U-tubes, and to validate the heat removal capacity of the Passive Auxiliary Feedwater<br />

System (PAFS) during the transient.<br />

In this paper, an MSGTR experiment<br />

at the ATLAS-PAFS test facility was<br />

simulated using the SPACE code to<br />

verify the prediction capability of this<br />

code <strong>for</strong> multiple failure accident,<br />

which is involved in the design extension<br />

condition. The overall system<br />

transient behavior obtained using<br />

SPACE code showed similar trends<br />

with the experimental results in terms<br />

of factors such as the system pressure,<br />

mass flow rate, and collapsed water<br />

level on the component. Additionally,<br />

a sensitivity analysis was conducted<br />

using the experimental correlation<br />

<strong>for</strong> PAFS which is included in the<br />

wall condensation model as an<br />

option in SPACE code. As a sensitivity<br />

calculation results, it is also recommended<br />

that the PAFS model in<br />

SPACE code be applied to obtain more<br />

accurate prediction results about the<br />

PAFS operation by per<strong>for</strong>ming a<br />

safety analysis of the APR+ nuclear<br />

power plant.<br />

1 Introduction<br />

1.1 Background<br />

Following the Fukushima nuclear<br />

disaster in 2011 and based on the<br />

lessons learned from the accident,<br />

there have been many changes to the<br />

relevant safety design criteria and/or<br />

regulations around the world. The<br />

<strong>Nuclear</strong> Safety and Security Commission<br />

(NSSC) in Korea has required<br />

plant-specific accident management<br />

plans, which extend beyond design<br />

basis accidents to include severe<br />

accidents. The revised regulation in<br />

Korea has determined a list of multiple<br />

failure accidents that must be considered<br />

<strong>for</strong> any accident management<br />

plan [1]. Multiple failure accidents<br />

should be considered <strong>for</strong> Design<br />

Extension Conditions, which is<br />

defined by the <strong>International</strong> Atomic<br />

Energy Agency (IAEA) Specific Safety<br />

Requirement [2, 3].<br />

The Multiple Steam Generator<br />

Tube Rupture (MSGTR) accident is<br />

selected as one of the multiple failure<br />

accidents by the Korean NSSC, and it<br />

is defined as an accident in which<br />

more than five U-tubes of a steam<br />

generator rupture. In a MSGTR<br />

accident, the pressurized primary<br />

coolant leaks into the secondary<br />

system and thus exposes radioactive<br />

material. There<strong>for</strong>e, it is important<br />

that the extent of the leak is limited<br />

and that the pressure drop across the<br />

break be kept as low as possible to<br />

reduce the radioactive release.<br />

Compared to single tube rupture,<br />

MSGTR causes quicker depressurization<br />

of the reactor coolant system<br />

(RCS) and places greater demand on<br />

the inventory makeup process.<br />

To elucidate the thermal hydraulic<br />

process of the MSGTR accident,<br />

an experimental study using the<br />

Advanced Test Loop <strong>for</strong> Accident<br />

Simulation (ATLAS) facility was<br />

conducted by the Korea Atomic<br />

Energy Research Institute (KAERI)<br />

[4]. The experiment simulated the<br />

rupture of five steam generator tubes,<br />

and the results showed that the<br />

Passive Auxiliary Feedwater System<br />

(PAFS) adopted in the Advanced<br />

<strong>Power</strong> Reactor Plus (APR+) had<br />

sufficient cooling capacity to mitigate<br />

the accident. The PAFS is one of the<br />

advanced safety features of a passive<br />

cooling system that allow it to<br />

replace a conventional active Auxiliary<br />

Feed-water System (AFWS) [5]. In a<br />

typical <strong>Nuclear</strong> <strong>Power</strong> Plant (NPP),<br />

a motor-driven or turbine-driven<br />

auxiliary feedwater is supplied after<br />

the wide-range water level of a steam<br />

generator is decreased below the low<br />

steam generator level. However, to<br />

confirm the cooling capability of<br />

PAFS compared to that of AFWS, an<br />

experimental scenario was conducted<br />

in which the PAFS was supplied to<br />

an intact SG instead of auxiliary<br />

feedwater. The PAFS is a passive<br />

system capable of condensing the<br />

steam generated in a steam generator<br />

and feeding the condensed water to<br />

the steam generator using gravity.<br />

For current safety analyses of<br />

Korean nuclear power plants, thermalhydraulic<br />

safety analysis codes<br />

supplied by <strong>for</strong>eign vendors such<br />

as Westinghouse and Combustion<br />

Engineering have been used. The<br />

Ministry Of Trade, Industry and<br />

Energy (MOTIE) in Korea launched<br />

the ‘Nu-Tech 2012’ project to improve<br />

the competitiveness of the Korean<br />

nuclear industry in 2006, and the<br />

Korean nuclear industry developed<br />

the SPACE (Safety and Per<strong>for</strong>mance<br />

Analysis CodE <strong>for</strong> nuclear power<br />

plants) code [6]. This code was<br />

approved by the Korean NSSC in<br />

2017, and it will replace outdated<br />

vendor-supplied codes and will be<br />

used <strong>for</strong> safety analyses of operating<br />

nuclear power plants in Korea as well<br />

as the design of advanced reactors.<br />

While the SPACE code will mainly be<br />

used in safety analyses, the code with<br />

best-estimate capabilities will be able<br />

to cover per<strong>for</strong>mance analysis as well.<br />

The programming language <strong>for</strong> the<br />

SPACE code is C++, and this<br />

code adopts the advanced physical<br />

modeling of two-phase flows, namely<br />

two-fluid three-field models which<br />

comprise gas, continuous liquid, and<br />

droplet fields.<br />

According to the General Safety<br />

Requirement (GSR) of IAEA, any<br />

calculation method and computer<br />

codes used in safety analysis must<br />

undergo verification and validation<br />

[7]. Further, verification calculations<br />

of system tests or integral tests are<br />

used to validate the general consistency<br />

of the revision [8]. There<strong>for</strong>e,<br />

a verification calculation based on<br />

integral effect experiments is needed<br />

to improve the reliability of the prediction<br />

results of the SPACE code.<br />


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In particular, the multiple failure<br />

accident condition is the new safety<br />

design criteria, so the verification<br />

process should be per<strong>for</strong>med. To this<br />

end, the first objective of this study is<br />

to verify a multiple failure accident<br />

calculation capability of the SPACE<br />

code, while the second objective is to<br />

confirm the transient phenomena of<br />

MSGTR and the cooling effect of PAFS<br />

during MSGTR with respect to the<br />

first objective, as mentioned above.<br />

The heat loss phenomenon is a<br />

measure of the total heat transfer of<br />

heat from either conduction, convection,<br />

radiation, or any combination<br />

of these. Newton’s law of cooling<br />

states that the rate of heat loss of an<br />

object is directly proportional to the<br />

difference in the temperature between<br />

the object and its surroundings. Under<br />

the experimental conditions of high<br />

temperature and high pressure, the<br />

heat loss is particularly likely to<br />

increase because of the temperature<br />

difference between the experiment<br />

component and the surrounding<br />

atmosphere. This physical phenomenon<br />

can affect the heat transfer<br />

experiment, and it plays an important<br />

role in the per<strong>for</strong>mance of the system.<br />

The heat loss is a function of area in<br />

accordance with the convective heat<br />

transfer equation. According to the<br />

design document, the ATLAS facility<br />

has a relatively large surface area to<br />

volume ratio which is in accordance<br />

with the design characteristic [9]. For<br />

this reason, additional work to<br />

confirm the heat loss effect on the<br />

ATLAS-PAFS facility was also conducted<br />

to confirm the heat loss effect<br />

on the system behavior of the integral<br />

test facility.<br />

1.2 A brief description<br />

of ATLAS-PAFS facility<br />

As mentioned previously, KAERI has<br />

been operating an integral effect test<br />

facility, ‘ATLAS’, <strong>for</strong> the transient and<br />

accident simulation of a Pressurized<br />

Water Reactor (PWR) [10]. The<br />

reference plant of ATLAS is the<br />

APR1400, which has been developed<br />

by the Korean nuclear industry.<br />

ATLAS has the same two-loop features<br />

as the APR1400, and it is designed<br />

using the scaling method to simulate<br />

various test scenarios as realistically<br />

as possible [9, 10]. This test facility<br />

also includes design features of the<br />

OPR1000, which is Korean standard<br />

NPP, such as a cold-leg injection mode<br />

<strong>for</strong> safety injection and a low pressure<br />

safety injection mode.<br />

As mentioned above, the PAFS is<br />

one of the advanced passive safety<br />

systems adopted in the APR+, and an<br />

experimental program is currently<br />

underway at KAERI to validate the<br />

cooling and operational per<strong>for</strong>mance<br />

of the PAFS [11]. The main objective<br />

of the ATLAS-PAFS integral effect test<br />

is to investigate the thermal hydraulic<br />

behavior in the primary and<br />

secondary systems of the ATLAS<br />

during a transient at which PAFS is<br />

actuated. The PAFS facility is<br />

described in further detail below.<br />

2 Description of<br />

ATLAS-PAFS model<br />

<strong>for</strong> MSGTR scenario<br />

using SPACE code<br />

2.1 Experiment scenario<br />

To validate the prediction capability<br />

of the SPACE code, the experimental<br />

in<strong>for</strong>mation provided by KAERI was<br />

utilized [4]. The target scenario <strong>for</strong><br />

the experiment is a MSGTR with a<br />

PAFS operation occurrence and<br />

asymmetric cooling. To initiate the<br />

MSGTR transient, first, the break<br />

valve at the SGTR simulation pipe line<br />

is opened. By opening the break valve,<br />

the primary system inventory was<br />

discharged from the hot side of the<br />

lower plenum to the upper location of<br />

the steam generator secondary hot<br />

side. Next, the primary system began<br />

to be depressurized and the secondary<br />

side water level of the steam generator<br />

increased. At the same time as the<br />

HSGL signal occurrence, reactor trip<br />

occurred, and the core power started<br />

to decrease following the decay curve.<br />

For the transient calculation, the<br />

decay power curve was considered<br />

along with the tables <strong>for</strong> time versus<br />

power. The main feedwater isolation<br />

valves (MFIVs) and the main steam<br />

isolation valves (MSIVs) <strong>for</strong> two steam<br />

generators were closed after delay<br />

times. The main steam safety valves<br />

(MSSVs) on the steam line opened<br />

due to the pressure increase of SG-1,<br />

and these valves were kept in cyclic<br />

operation of opening and closing to<br />

protect the primary and secondary<br />

systems from over-pressurization. The<br />

accident caused the depressurization<br />

of RCS and resulted in a Low PZR<br />

Pressure (LPP) signal. Further, the<br />

Safety Injection Pumps (SIPs) began<br />

after delay times. It was assumed that<br />

only one safety injection pump per<br />

train was operated <strong>for</strong> the experiment<br />

scenario. In accordance with this<br />

assumption, SIP-1 and SIP-3 were<br />

available. The injection flow rate was<br />

applied using pressure – mass flow<br />

curve based on experiment data. To<br />

simulate an accident management<br />

measure based on the cooling per<strong>for</strong>mance<br />

of PAFS during a MSGTR,<br />

the PAFS was supplied to an intact<br />

SG-2 instead of auxiliary feedwater<br />

after the water level of the SG-2 fell<br />

below the PAFS operation set point<br />

due to the decay power. It was also<br />

assumed that the auxiliary feed<br />

water system would not work <strong>for</strong> the<br />

assessment of PAFS cooling capability.<br />

After the initiation of the PAFS, the<br />

decay heat was removed from the RCS<br />

by the natural convection of the PAFS.<br />

Finally, the whole system was cooled<br />

down in a stable manner with the<br />

successful operation of SIPs, MSSVs,<br />

and PAFS.<br />

2.2 Brief overview of model<br />

in<strong>for</strong>mation<br />

For the experimental simulation,<br />

the ATLAS-PAFS test facility was<br />

modeled using SPACE code as shown<br />

in Figure 1. In the calculation, the<br />

decay power was imposed in<br />

accordance with the experiment, and<br />

the operation logics of the safety<br />

systems, such as safety injection and<br />

PAFS, were reflected. The geometrical<br />

and material in<strong>for</strong>mation of the<br />

components and pipe lines in the<br />

ATLAS-PAFS test facility was also<br />

reflected [9, 12]. The reactor vessel<br />

was separated to simulate the core,<br />

bypass flow, reactor vessel lower<br />

plenum, and the reactor vessel upper<br />

head. The core model included the top<br />

and bottom inactive core regions,<br />

average channel, and hot channel.<br />

The safety injection system had four<br />

independent trains and a direct vessel<br />

injection (DVI) mode. Each train of<br />

the safety injection system consisted<br />

of safety injection pumps (SIPs). Injection<br />

lines could be aligned to either<br />

reactor vessel down-comer <strong>for</strong> DVI<br />

injection. The pressurizer was modeled<br />

as a single component with<br />

10 vertical nodes. The lower part was<br />

connected to hot leg through a surge<br />

line separated into five nodes. The hot<br />

legs and cold legs were modeled with<br />

four cells each, while the intermediate<br />

legs were modeled with five nodes.<br />

The steam generator included five<br />

nodes <strong>for</strong> the evaporator and two<br />

nodes <strong>for</strong> the economizer. The main<br />

steam safety valves were modeled into<br />

three separate groups; each group<br />

was operated on a different set point<br />

of pressure in the steam generator<br />

dome.<br />

2.3 SGTR simulation facility<br />

The geometry of the ATLAS facility,<br />

which was composed of a SGTR<br />

simulation pipe and connected with<br />

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| Fig. 1.<br />

Modeling diagram of ATLAS and SGTR simulation pipe using SPACE code.<br />

a PAFS facility, was modeled as<br />

shown in Figure 1. In the ATLAS<br />

facility with the SGTR simulation<br />

pipe, the primary system inventory<br />

was discharged from the hot side<br />

of the lower plenum to the upper<br />

location of the SG-1 secondary<br />

hot-side to simulate a MSGTR<br />

accident. It is composed of a break<br />

simulation valve, an orifice flow<br />

meter, an orifice, and break nozzles.<br />

The ‘PIPE’ component option of SPACE<br />

code was used to model the upstream<br />

pipe, which was part of the SGTR<br />

simulation pipe; this pipe section was<br />

geometrically divided into 10 nodes.<br />

The break nozzle was installed to<br />

simulate a five-tube rupture with a<br />

non-choking orifice. This tube section<br />

is modeled using the ‘CELL’ component<br />

option of SPACE code. The<br />

input of the inner diameter is<br />

1.756 mm and the total flow area is<br />

the summation of the five-tube area.<br />

An orifice with a 1.68 mm hole was<br />

installed at the end of the break<br />

nozzles to simulate the choking flow<br />

condition at tube rupture, and the<br />

break nozzles were designed to<br />

maintain the equivalent pressure drop<br />

in the case of the non-choking flow<br />

condition. The mass flow rate through<br />

the SGTR simulation pipe was<br />

measured using an orifice flow meter.<br />

These orifice and orifice flow meter<br />

sections were modeled using the<br />

‘FACE’ component option of SPACE<br />

code. The experiment began by<br />

opening an initiation valve to simulate<br />

a MSGTR on the SG-1. This valve was<br />

modeled as the ‘TRIP VALV’ component<br />

option of SPACE code, and it<br />

was opened at the start of the transient<br />

calculation. The MSGTR occurs<br />

on the tubes of SG-1, which are<br />

connected to the SGTR simulation<br />

pipe; five break nozzles are opened in<br />

total [4].<br />

2.4 Passive Auxiliary<br />

Feedwater System (PAFS)<br />

The steam supply and return water<br />

line connected the PCHX to the SG-2<br />

of the ATLAS [12]. There<strong>for</strong>e, as<br />

shown in Figure 2, the PAFS was<br />

modeled by adding junctions at the<br />

main steam line and at the economizer<br />

nozzle as the inlet and outlet of<br />

the PAFS, respectively. The steam<br />

supply line and the return water line<br />

were divided into 24 nodes and<br />

31 nodes, respectively. The diameters<br />

of the steam supply line and the return<br />

water line were about 0.04 m and<br />

0.03 m, respectively. The PAFS<br />

operation valve was connected to the<br />

return water line and the feed water<br />

line, and it was modeled as a trip<br />

valve. This valve open signal was<br />

synchronized to the instant that the<br />

collapsed water level in the steam<br />

generator reached the set point of the<br />

low steam generator level. Once the<br />

valve was open, the latched option<br />

made it impossible <strong>for</strong> the valve to<br />

close again. The end of the return line<br />

was connected to the bottom nozzle<br />

of the steam generator economizer<br />

volume. The PCHX was the most<br />

important component of PAFS, and it<br />

was filled with condensate water and<br />

the return water line on steady state<br />

condition. The condensation tube of<br />

PCHX was modeled with 24 nodes as<br />

shown in Figure 3. The length of the<br />

horizontal nodes was about 0.23 m<br />

and the horizontal part of the<br />

PCHX was modeled as 1.806 m. An<br />

inclination of 3° was applied to the<br />

horizontal tube region while an<br />

inclination of 41.2° was applied to one<br />

inlet node and one outlet node to<br />

simulate a U-shaped bend. These<br />

design values were determined to<br />

prevent the condensation-induced<br />

water hammer inside the tube of<br />

PCHX [13]. The area of the PCHX pipe<br />

component was about 22.35 cm 2 ,<br />

which equals the summation of the<br />

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| Fig. 2.<br />

Connection nodding diagram of PAFS to ATLAS steam generator.<br />

| Fig. 3.<br />

Nodding diagram of PCHX.<br />

three-tube area. The connected<br />

heat structures were modeled as a<br />

cylindrical shape. The inner and outer<br />

coordinates were the inner and outer<br />

radii of the tube, respectively. The<br />

number of tubes was used as an input<br />

<strong>for</strong> equivalent heat transferring area.<br />

The top and bottom headers of PCHX,<br />

which both play roles in preventing<br />

the vibration of the PCHX tube in<br />

the PCCT, were modeled as cell<br />

components. The Passive Condensate<br />

Cooling Tank (PCCT) of PAFS was<br />

designed as a rectangular pool. When<br />

the PAFS was actuated, the heat<br />

transfer from the PCHX caused the<br />

pool water in the PCCT to evaporate,<br />

after which the steam flowed through<br />

the upper pipe on the top. This upper<br />

pipe was connected to the upper cells<br />

of the PCCT. Finally, the ‘TFBC’<br />

component, which was connected to<br />

the upper pipe, played a role in maintaining<br />

the atmospheric pressure. The<br />

water pool of the PCCT served as a<br />

heat sink. The core decay heat was<br />

transferred through the condensation<br />

of steam inside the tubes. There<strong>for</strong>e,<br />

the extracted heat increased the<br />

temperature of the pool water. It is<br />

expected that rigorous boiling<br />

occurred at the tube outside the<br />

surface, and a strong buoyancy flow<br />

was also expected. In order to simulate<br />

natural convection by buoyancy flow<br />

in the PCCT, the 3D option of SPACE<br />

code was applied to model the PCCT<br />

facility as illustrated in Figure 4. This<br />

PCCT model was divided into<br />

182 cells, where the y direction and<br />

the z direction were respectively<br />

divided into 13 and 14 cells. The<br />

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| Fig. 4.<br />

Nodding diagram of PCCT using ‘3D’ option of SPACE code.<br />

cooling water was filled up to nine<br />

nodes of the z direction, and the upper<br />

nodes were in atmosphere condition.<br />

The initial cooling water level in PCCT<br />

was about 3.8 m while the water<br />

temperature was 28.8 °C.<br />

3 Results and discussions<br />

3.1 Steady-state results<br />

Be<strong>for</strong>e using the SPACE model <strong>for</strong><br />

transient analyses, a consistent set of<br />

parameters must be obtained by the<br />

steady-state initialization process.<br />

Table 1 lists the initial conditions of<br />

the experiment, calculated results of<br />

the SPACE code, and difference. The<br />

initial core power generated by the<br />

heater rods was 1.627 MW, and the<br />

heat loss of the primary piping into<br />

the atmosphere was estimated to<br />

be about 97.1 kW based on the<br />

in<strong>for</strong>mation of the initial experiment<br />

condition. During the initial conditions,<br />

the major thermal hydraulic<br />

parameters, including the system<br />

pressure, fluid temperature, and mass<br />

flow rate, were reasonably consistent<br />

with the experiment condition.<br />

The calculation time started from<br />

-1000.0 s to 0.0 s and all design<br />

parameters converged after -500.0 s.<br />

Parameter Experiment SPACE code Difference (%)<br />

Primary system<br />

Core power (MW) 1.627 1.627 0.00<br />

Heat loss (kW) 97.1 97.2 0.40<br />

Pressurizer (PZR) pressure (MPa) 15.52 15.50 0.13<br />

PZR level (m) 3.71 3.71 0.00<br />

Cold leg flow rate (kg/s) 1.9728 1.941 1.61<br />

Core inlet temp. (°C) 292.0 289.4 0.89<br />

Core outlet temp. (°C) 327.5 325.2 0.70<br />

Secondary system<br />

Steam flow rate (kg/s) 0.4019 0.4153 3.33<br />

Feed water flow rate (kg/s) 0.4209 0.4194 0.36<br />

Feed water temp. (°C) 233.3 233.3 0.00<br />

Steam dome pressure (MPa) 7.83 7.81 0.20<br />

Steam temp. (°C) 295.7 294.2 0.51<br />

SG collapsed water level (m) 4.97 4.97 0.00<br />

Passive Auxiliary Feedwater System (PAFS)<br />

Initial PCCT level (m) 3.8 3.8 0.00<br />

Initial PCCT temp. (°C) 28.8 28.8 0.00<br />

| Tab. 1.<br />

Comparison results between initial condition of experiment and steady state results of SPACE code.<br />

After checking the steady state condition,<br />

the transient calculation<br />

started at 0.0 s of the steady state<br />

condition.<br />

3.2 Transient analysis results<br />

According to the agreement of the<br />

ATLAS Domestic Standard Problem-05,<br />

the experimental data should be<br />

confidential. There<strong>for</strong>e, all of the<br />

experiment results in this paper<br />

including the time frame (t*) were<br />

divided by an arbitrary value and<br />

plotted on the non-dimensional axis.<br />

3.2.1 Sequence of transient<br />

calculation result<br />

Table 2 summarizes the measured<br />

and calculated sequences of a<br />

MSGTR-PAFS test. The transient was<br />

initiated with the MSGTR in normal<br />

condition. To initiate the MSGTR<br />

transient, the break valve was opened<br />

at 0.0 s with the pressurizer heater off<br />

according to the experiment scenario.<br />

The break flow contributed to the<br />

steam generator overfill, and the<br />

collapsed water level of the SG-1<br />

reached a set point of HSGL signal,<br />

leading to the generation of the HSGL<br />

signal. This signal was generated at a<br />

later time in the calculation case than<br />

it was in the experiment case. The<br />

reactor trip occurred simultaneously<br />

due to the HSGL signal, and the MSIVs<br />

and MFIVs were also closed after their<br />

respective delay times. After the<br />

MSIVs were closed, the pressure in<br />

steam generator was controlled by the<br />

cyclic operation of MSSVs. According<br />

to the experimental assumptions, the<br />

heater power which simulates the<br />

decay power had the ANS-73 decay<br />

curve applied <strong>for</strong> the experimental<br />

condition, and the heater power<br />

started after the reactor trip. For the<br />

calculation, time verse decay power<br />

data was applied in accordance with<br />

the experiment in<strong>for</strong>mation [4]. The<br />

time at which the MSSVs first opened<br />

in the experiment and the calculation<br />

results were the same after the HSGL<br />

signal was triggered. During this<br />

steam generator overfill period, the<br />

pressure in the primary system<br />

decreased substantially. As a result,<br />

when the primary system pressure<br />

decreased to the set point of the LPP<br />

trip, the LPP signal was triggered, and<br />

the SIP injection was initiated after<br />

the delay time. The collapsed water<br />

level of SG-2 decreased due to the<br />

decay power, then reached a set point<br />

of PAFS operation. Finally, the PAFS<br />

operation valve automatically opened<br />

and passive cooling using PAFS<br />

was initiated.<br />


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<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Event Experiment (t = t*) SPACE code (t = t*) Remarks<br />


MSGTR initiation 0.0000 0.0000 PZR heater off at same time as experiment scenario<br />

HSGL signal 0.0011 0.0024 SG collapsed water level > set-point<br />

Reactor trip 0.0011 0.0024<br />

MSIV close 0.0015 0.0028<br />

MFIV close 0.0018 0.0031<br />

3.2.2 Primary system behaviors<br />

The decay power was one of the important<br />

factors in this analysis, and<br />

the calculation results were consistent<br />

with the experimental result, as<br />

shown in Figure 5. Figure 6 shows<br />

the measured and calculated break<br />

flow, which were normalized by the<br />

maximum value of the experiment.<br />

The break flow rate largely depended<br />

on the pressure difference between<br />

the primary and the secondary<br />

systems. There<strong>for</strong>e, as shown in<br />

Figure 6, the peak flow rate occurred<br />

at the beginning of transient. After the<br />

occurrence of peak flow, the break<br />

flow in the experiment case oscillated<br />

and gradually decreased. On the other<br />

hand, the calculation result shows<br />

that the break flow was maintained<br />

constantly. The break flow was deeply<br />

related to the difference between<br />

the pressures of the primary and<br />

secondary systems. After the pressure<br />

Coincident with HSGL<br />

Decay power start 0.0023 0.0036 Delay times after reactor trip<br />

MSSV first opening 0.0026 0.0033 SG pressure > set-point<br />

LPP signal 0.0248 0.0232 PZR pressure < set-point<br />

SIP injection start 0.0276 0.0260 Delay times after LPP signal<br />

PAFS operation start 0.7204 0.7173 SG collapsed water level < set-point<br />

| Tab. 2.<br />

Sequence of transient analysis result.<br />

of the primary system was reduced by<br />

the break, the pressure of the PZR<br />

was predicted to be lower than the<br />

experiment results (t* < 0.2 in<br />

Figure 7). The pressure of the primary<br />

system was relatively lower than that<br />

obtained in the experiment results,<br />

thus indicating a lower break flow<br />

rate. Due to the lower break flow rate,<br />

the pressure of PZR was maintained<br />

somewhat higher than the experiment<br />

result, which resulted in the break<br />

| Fig. 5.<br />

Comparison result of core power.<br />

| Fig. 6.<br />

Comparison result of MSGTR break flow rate.<br />

| Fig. 7.<br />

Comparison result of system pressure.<br />

| Fig. 8.<br />

Comparison result of SIP injection rate.<br />

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Study on Verification of SPACE Code Based on an MSGTR Experiment at the ATLAS-PAFS Facility ı Kyungho Nam

<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

| Fig. 9.<br />

Comparison result of SG collapsed water level.<br />

| Fig. 10.<br />

Comparison result of PAFS line mass flow rate.<br />


| Fig. 11.<br />

Comparison result of PCCT water level.<br />

| Fig. 12.<br />

Comparison result of fluid temperature at PAFS line after PAFS operation.<br />

flow rate being maintained (t* > 0.2<br />

in Figure 7). At the moment the<br />

break occurred, the pressure in the<br />

pressurizer began to decrease rapidly<br />

due to the loss of coolant through the<br />

break point. The depressurization<br />

rates of the pressurizer obtained from<br />

the experiment and the calculation<br />

results were almost the same, as<br />

shown in Figure 7. The pressure in the<br />

pressurizer continually decreased and<br />

reached the set point of the LPP signal.<br />

As the LPP signal was generated,<br />

the SIP injection was initiated, as<br />

shown in Figure 8. During the<br />

beginning phase of SIP injection, the<br />

calculation result of the SIP flow<br />

rate was slightly higher than the<br />

experimental result. This was<br />

attributed to that fact that the PZR<br />

pressure according to the calculation<br />

result remained lower than that<br />

obtained in the experiment result.<br />

As the accident progressed, the SIP<br />

injection rate obtained in the calculation<br />

result was consistent with the<br />

experiment result. When the RCS<br />

pressure reached the saturation<br />

pressure due to the decay power,<br />

plateau pressure was observed. In<br />

addition, the water injected by the SIP<br />

contributed to the primary system<br />

pressure. Thus, as shown in Figure 7,<br />

the primary system pressure was<br />

maintained.<br />

3.2.3 Secondary system<br />

behaviors<br />

After the MSIVs were closed by the<br />

HSGL signal, the pressure in the steam<br />

generator was maintained within the<br />

range of the opening and closing set<br />

points of the MSSVs. The transient<br />

behavior of pressure in the steam<br />

generator shows that the calculation<br />

result is consistent with the experimental<br />

results. Following the PAFS<br />

operation, the pressure in an intact<br />

SG-2 drastically decreased due to the<br />

cooling by PAFS; the SG-2 depressurization<br />

trend is similar to the core<br />

cooling rate trend.<br />

Figure 9 shows a comparison of<br />

the SG collapsed water levels in the<br />

experiment and in the SPACE calculation.<br />

The water level of broken SG-1<br />

increased and reached the full water<br />

level. By contrast, that of an intact<br />

SG-2 decreased rapidly and reached a<br />

set point of the PAFS operation. As the<br />

PAFS signal was triggered, the PAFS<br />

operation valve automatically opened,<br />

and the main steam from the SG-2<br />

flowed into the steam supply line. As<br />

shown in Figure 10, the mass flow<br />

peaked, and then natural circulation<br />

flow was <strong>for</strong>med. The mass flow rate<br />

in the case of SPACE calculation was<br />

consistent with the experimental<br />

result. The main steam from the steam<br />

supply line flowed into the condensation<br />

tubes, and the condensate<br />

circulated through the return water<br />

line to the economizer of the<br />

steam generator. After the PAFS was<br />

actuated, the water level also<br />

increased due to the thermal<br />

expansion, as shown in Figure 11.<br />

3.2.4 Wall condensation heat<br />

transfer model <strong>for</strong> PCHX<br />

The wall condensation heat transfer<br />

rate in PCHX is one of the dominant<br />

factors <strong>for</strong> determining the PAFS<br />

cooling capability. There<strong>for</strong>e, the<br />

many precedent studies related<br />

to condensation in PCHX were<br />

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<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />


per<strong>for</strong>med. The wall condensation<br />

models were incorporated into the<br />

SPACE code heat transfer package. For<br />

pure steam condensation, like the<br />

problem addressed in this paper, the<br />

same model used in RELAP5/MOD3.3<br />

was selected as a default model. The<br />

maximum value among Nusselt’s<br />

[14], Shah’s [15], and Chato’s [16]<br />

correlations is used to consider the<br />

geometric and turbulent effects. The<br />

experimental correlation <strong>for</strong> PAFS is<br />

also included in the wall condensation<br />

model as an option in SPACE code<br />

[17]. Based on the calculation results<br />

mentioned in chapter 3.2.1, this<br />

model was applied. In this section,<br />

the default model and experiment<br />

correlation model <strong>for</strong> PAFS were<br />

compared to confirm the prediction<br />

ability of SPACE code <strong>for</strong> PCHX<br />

cooling per<strong>for</strong>mance. Chen’s model,<br />

which is the default option in SPACE<br />

code, is applied to the outside of PCHX<br />

[18].<br />

The fluid temperature after PAFS<br />

operation is shown in Figure 12. The<br />

calculation result shows that the fluid<br />

temperature on the return water line<br />

is higher than the experiment, and<br />

that the default option case is highest.<br />

That means that the calculation<br />

results which were obtained using the<br />

default option and the PAFS model<br />

underestimated cooling per<strong>for</strong>mance.<br />

However, the calculation using the<br />

PAFS model was more accurate than<br />

the default option.<br />

4 Conclusions<br />

In this study, a MSGTR experiment<br />

with the PAFS operation per<strong>for</strong>med by<br />

KAERI was simulated using the SPACE<br />

code. This study focused on verifying<br />

the prediction capability of the SPACE<br />

code <strong>for</strong> MSGTR accidents, which is<br />

one of the multiple failure accidents,<br />

and to evaluate the cooling capacity<br />

of the PAFS during such an accident.<br />

The calculation results showed that<br />

the overall system transient results<br />

using the SPACE code showed<br />

comparatively similar trends with the<br />

experimental results in terms of the<br />

system pressure, mass flow rate, and<br />

collapsed water level in components.<br />

There<strong>for</strong>e, it was concluded that the<br />

SPACE code reasonably predicted the<br />

experiment. And, the default model,<br />

which uses the maximum value<br />

among Nusselt’s, Shah’s, and Chato’s<br />

and the experiment correlation<br />

model <strong>for</strong> PAFS in SPACE code were<br />

compared to confirm the prediction<br />

ability of SPACE code <strong>for</strong> PCHX<br />

cooling per<strong>for</strong>mance. The calculation<br />

result using the PAFS model was more<br />

accurately estimated than the default<br />

option.<br />

Based on the present calculation<br />

results, the following conclusions<br />

were obtained in this study. First, the<br />

experiment and calculation results<br />

showed that the cooling capability of<br />

PAFS is sufficient to replace the active<br />

auxiliary feedwater system during<br />

MSGTR transient. Additionally, it is<br />

recommended that the PAFS model in<br />

SPACE code be applied <strong>for</strong> more<br />

accurate prediction results <strong>for</strong> PAFS<br />

operation by per<strong>for</strong>ming a safety<br />

analysis of an APR+ nuclear power<br />

plant.<br />

Acknowledgements<br />

This work was per<strong>for</strong>med within the<br />

program of the fifth ATLAS Domestic<br />

Standard Problem (DSP-05), which<br />

was organized by the Korea Atomic<br />

Energy Research Institute (KAERI) in<br />

collaboration with the Korea Institute<br />

of <strong>Nuclear</strong> Safety (KINS) under the<br />

national nuclear R&D program funded<br />

by the Ministry of Education<br />

(MOE) of the Korean government.<br />

The authors are also grateful to the<br />

fifth ATLAS DSP-05 program participants:<br />

KAERI <strong>for</strong> the experimental data<br />

and to the council of the fifth DSP-<br />

05 program <strong>for</strong> providing the opportunity<br />

to publish the results.<br />

Acronyms and Abbreviations<br />

NPP<br />

HSGL<br />

LPP<br />

MSIV<br />

MFIV<br />

<strong>Nuclear</strong> <strong>Power</strong> Plant<br />

High Steam Generator Level<br />

Low Pressurizer Pressure<br />

Main Steam Isolation Valve<br />

Main Feed-water Isolation Valve<br />

MSSV Main Steam Safety Valve<br />

SIP<br />

PAFS<br />

PCCT<br />

Safety Injection Pump<br />

Passive Auxiliary Feedwater System<br />

Passive Condensate Cooling Tank<br />

PCHX Passive Condensation Heat Exchanger<br />

HL<br />

CL<br />

IL<br />

SG<br />

Hot Leg<br />

Cold Leg<br />

Intermediate Leg<br />

Steam Generator<br />

References<br />

[1] Korean NSSC, Regulations on the Scope of Accident<br />

Management and the detailed criteria of Accident<br />

Management Capability Evaluation, Notification No. 2017-<br />

34 of the <strong>Nuclear</strong> Safety and Security Commission in Korea,<br />

Rev.1 (2017).<br />

[2] IAEA, Safety of <strong>Nuclear</strong> <strong>Power</strong> Plants: Design, IAEA Specific<br />

Safety Requirements No. SSR-2/1, Rev.1 (2016).<br />

[3] J.H. Koh et al., Current status of design extension conditions<br />

technology development <strong>for</strong> prevention of severe accident<br />

in nuclear power plants, 13th <strong>International</strong> Conference on<br />

probabilistic Safety Assessment and Management (PSAM<br />

13) (2016).<br />

[4] KAERI, Experimental Study on the Multiple Steam Generator<br />

Tube Rupture with Passive Auxiliary Feedwater System<br />

Operation, KAERI/TR-8010/2020 (2020).<br />

[5] J. Cheon et al., The Development of a Passive Auxiliary<br />

Feedwater System in APR+, ICAPP2010, San Diego, USA<br />

(2010).<br />

[6] S.J. Ha et al., Development of the SPACE code <strong>for</strong> nuclear<br />

power plants, <strong>Nuclear</strong> Engineering and Technology 43,<br />

(2011) 45-62.<br />

[7] IAEA, Safety Assessment <strong>for</strong> Facilities and Activities, IAEA<br />

Safety Standards Series No. GSR Part 4, Rev.1 (2016).<br />

[8] Petruzzi, A., D’Auria, F., Thermal-Hydraulic System Codes in<br />

<strong>Nuclear</strong> Reactor Safety and Qualification Procedures,<br />

Science and Technology of <strong>Nuclear</strong> Installations, Vol. 2008,<br />

Article ID 460795, (2008) p.16.<br />

[9] KAERI, Description report of ATLAS facility and<br />

instrumentation, KAERI/TR-7218/2018 (2018).<br />

[10] Y.S. Kim et al., Commissioning of the ATLAS thermalhydraulic<br />

integral test facility, Annals of <strong>Nuclear</strong> Energy,<br />

Vol. 35, (2008) pp. 1791-1799.<br />

[11] K.H. Kang et al., Separate and integral effect tests <strong>for</strong><br />

validation of cooling and operational per<strong>for</strong>mance of the<br />

APR+ Passive Auxiliary Feedwater System, <strong>Nuclear</strong><br />

Engineering and Technology 44, No.6, (2012).<br />

[12] KAERI, Description report of ATLAS-PAFS Facility and<br />

Instrumentation, KAERI/TR-5545/2014 (2014).<br />

[13] B.U. Bae et al., Design of condensation heat exchanger <strong>for</strong><br />

the PAFS(Passive Auxiliary Feedwater System) of APR+<br />

( Advanced <strong>Power</strong> Reactor Plus)”, Annals of <strong>Nuclear</strong> Energy,<br />

Vol.46, (2012) 134-143.<br />

[14] Nusselt, W.A., The surface Condensation of Water Vapor,<br />

Zieschrift Ver. Deut. Ing., 60, (1916) p.541.<br />

[15] Shah, M.M., A general correlation <strong>for</strong> heat transfer during<br />

film condensation inside pipes, Int. J. Heat Mass Transfer,<br />

22, (1979) p.547.<br />

[16] Chato, J.C., Laminar Condensation inside Horizontal and<br />

Inclined Tubes, American society of heating, refrigeration<br />

and air conditioning engineering journal, 4, (1962) p.52.<br />

[17] B.U. Bae et al., Evaluation of mechanistic wall condensation<br />

models <strong>for</strong> horizontal heat exchanger in PAFS (Passive<br />

Auxiliary Feedwater System), Annals of <strong>Nuclear</strong> Energy,<br />

Vol.107, (2017) 53-61.<br />

[18] K.Y. Choi et al., Development of a wall-to-fluid heat transfer<br />

package <strong>for</strong> the SPACE code, <strong>Nuclear</strong> Engineering and<br />

Technology, Vol.41, No.9 (2009).<br />

Author<br />

Kyungho Nam<br />

Associate<br />

research engineer<br />

Korea Hydro &<br />

<strong>Nuclear</strong> <strong>Power</strong> Co., LTD.<br />

Central Research Institute<br />

Safety Analysis Group<br />

Deajeon, Republic of<br />

Korea<br />

Kyungho Nam. Associate research engineer at Korea<br />

Hydro & <strong>Nuclear</strong> <strong>Power</strong> Central Research Institute<br />

(KHNP CRI). Majored in nuclear engineering and have<br />

a master’s degree 4 years of experience in nuclear<br />

safety analysis, and the current work is also safety<br />

analysis using safety analysis code and containment<br />

P/T analysis.<br />

Research and Innovation<br />

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<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Top<br />

IAEA Presents New Plat<strong>for</strong>m<br />

on Small Modular Reactors<br />

and Their Applications<br />

(iaea) The IAEA presented its newly<br />

established Plat<strong>for</strong>m on Small Modular<br />

Reactors (SMRs) and their Applications,<br />

aimed at supporting countries<br />

worldwide in the development and<br />

deployment of this emerging nuclear<br />

power technology, during an event on<br />

the margins of the of the 65 th IAEA<br />

General Conference.<br />

With more than 70 SMR designs<br />

under development in 17 countries<br />

and the first SMR units already in<br />

operation in Russia, SMRs and their<br />

smaller cousins, microreactors (MRs),<br />

are <strong>for</strong>ecast to play an increasingly<br />

important role in helping the global<br />

energy transition to net zero. Still, the<br />

technology, its safety and economic<br />

competitiveness must be de monstrated<br />

be<strong>for</strong>e SMRs can be more widely<br />

deployed, panellists agreed.<br />

The IAEA’s new Plat<strong>for</strong>m on SMRs<br />

and their Applications will assist<br />

countries in addressing these and<br />

related challenges. Using as a reference<br />

the <strong>Nuclear</strong> Energy Series publication,<br />

Technology Roadmap <strong>for</strong> Small<br />

Modular Reactor Deployment, the<br />

Plat<strong>for</strong>m provides experts with a<br />

one-stop shop to access the IAEA’s full<br />

array of support and expertise on<br />

SMRs, from technology development<br />

and deployment (including nonelectric<br />

applications) to nuclear safety,<br />

security and safeguards.<br />

“High standards of nuclear safety,<br />

security and non-proliferation must be<br />

ensured <strong>for</strong> SMR deployment,” Mikhail<br />

Chudakov, IAEA Deputy Director<br />

General and Head of the Department<br />

of <strong>Nuclear</strong> Energy said in his opening<br />

remarks. “But beyond this, it is<br />

generally recognized that if SMRs are<br />

going to be successful, they will need<br />

to be economically com petitive with<br />

respect to other clean energy alternatives.<br />

Achieving that will require<br />

accelerating their technological development<br />

and readiness level.”<br />

The IAEA has in place several<br />

activities related to SMRs to support<br />

its Member States through cooperation<br />

in SMR design, development and<br />

deployment and to serve as a hub <strong>for</strong><br />

sharing SMR regulatory knowledge<br />

and experience. Although the IAEA<br />

safety standards can generally be<br />

applied to SMRs, global experts from<br />

the SMR Regulators’ Forum are<br />

working on a tailor-made solution to<br />

help national authorities regulate this<br />

new class of nuclear power reactors,<br />

which are expected to generate up to<br />

300 megawatts (electrical) (MW(e))<br />

of power depending on their design.<br />

Participants spoke about challenges<br />

facing SMR development and<br />

deployment and how the new Plat<strong>for</strong>m<br />

could be used to help countries<br />

address them.<br />

“Potential topics <strong>for</strong> medium to<br />

long term activities include supply<br />

chain development, the development<br />

of industrial codes and standards, and<br />

suitable deployment strategies,” said<br />

Marco Ricotti, a professor of nuclear<br />

engineering at Italy’s Politecnico<br />

di Milano, who chairs the IAEA’s<br />

Technical Working Group <strong>for</strong> Small<br />

and Medium-sized or Modular<br />

Reactors (TWG-SMR).<br />

“Based on discussions within the<br />

(SMR Regulators’) Forum and also<br />

within our own regulatory <strong>for</strong>ums at<br />

work, we feel strongly that it is not<br />

realistic in the near term to develop<br />

detailed guidance <strong>for</strong> every technology,”<br />

said Marcel de Vos, who works<br />

on new reactor licensing at the<br />

Canadian <strong>Nuclear</strong> Safety Commission.<br />

“The pragmatic approach in our view<br />

is that we need to work with what we<br />

have and make calculated improvements<br />

as experiences are gathered<br />

and gained.”<br />

“There is a lot of work ahead of us,<br />

but we are working efficiently and in<br />

the right direction with Member<br />

States,” said Lydie Evrad, Deputy<br />

Director General and Head of the<br />

Department of <strong>Nuclear</strong> Safety and<br />

<strong>Nuclear</strong> Security.<br />

“The Plat<strong>for</strong>m is a very powerful<br />

interdepartmental mechanism,<br />

bringing together expertise from<br />

across the organization on SMRs,”<br />

said Stefano Monti, Chair of the<br />

Plat<strong>for</strong>m Implementation Team and<br />

Head of the IAEA’s <strong>Nuclear</strong> <strong>Power</strong><br />

Technology Development Section.<br />

| www.iaea.org<br />

Foratom: Limited Attention<br />

to <strong>Nuclear</strong> in Commission’s<br />

Energy Price Communication<br />

(<strong>for</strong>atom) FORATOM would have<br />

liked to see today’s communication<br />

from the Commission pay closer<br />

attention to the role which low-carbon<br />

and dispatchable nuclear can play in<br />

mitigating the current energy crisis.<br />

By including European nuclear in its<br />

toolkit of measures to tackle energy<br />

prices, it would have a unique opportunity<br />

of limiting its dependence on<br />

carbon intensive natural gas imports,<br />

thereby reducing its exposure to<br />

wholesale price fluctuations and its<br />

carbon footprint.<br />

“As highlighted in the communication,<br />

the current price increases are<br />

being driven by higher natural gas<br />

prices on the global market”, states<br />

Yves Desbazeille, FORATOM Director<br />

General. “There<strong>for</strong>e, as the EU moves<br />

to increase its share of variable renewables,<br />

it is essential that EU policy<br />

supports other low-carbon European<br />

sources to ensure reduced dependency<br />

on imports.”<br />

The Communication also highlights<br />

the effects which lower availability<br />

of renewables has had on the<br />

market, leading to supply constraints.<br />

Because nuclear can provide both<br />

baseload and dispatchable electricity,<br />

it acts as a perfect counterbalance at<br />

times when renewables are unavailable.<br />

As noted in the Communication,<br />

nuclear currently accounts <strong>for</strong><br />

around 25% of the electricity mix in<br />

the EU.<br />

With industrial activity ramping up<br />

post COVID, this has led to an increase<br />

in demand <strong>for</strong> energy. “It would be a<br />

mistake to treat this as a short-term<br />

issue. It is clear that demand <strong>for</strong><br />

electricity is expected to increase<br />

dramatically in the push to decarbonise<br />

Europe’s economy” adds<br />

Mr. Desbazeille. “There<strong>for</strong>e, the EU<br />

needs to already be putting solutions<br />

in place today to ensure that it is able<br />

to generate enough low-carbon<br />

electricity in Europe to meet growing<br />

demand. This means supporting the<br />

development of nuclear energy”.<br />

The Communication also makes<br />

reference to the sustainable finance<br />

taxonomy, reiterating the point that a<br />

complementary Delegated Act (CDA)<br />

‘will cover nuclear energy subject to<br />

and consistent with the results of the<br />

specific review process underway in<br />

accordance with the EU Taxonomy<br />

Regulation’. As this review is now<br />

complete, and the experts have overall<br />

concluded that nuclear is taxonomy<br />

compliant, we urge the Commission to<br />

urgently publish the CDA to avoid<br />

nuclear being unfairly penalised.<br />

| www.<strong>for</strong>atom.org<br />

NEI’s New Ad Campaign<br />

“See the Light” Embraces a<br />

Carbon-Free Future<br />

(nei) What’s New? The <strong>Nuclear</strong> Energy<br />

Institute launched its “See the Light”<br />

ad campaign, which speaks to the<br />

critical role nuclear energy plays in our<br />

carbon-free future. With eye-catching<br />

bursts of light illumi nating a dark<br />

room, the ad emphasizes that a<br />

brighter future must be powered by a<br />

diverse energy system championed by<br />

wind, solar and nuclear.<br />

57<br />

NEWS<br />


<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

58<br />

NEWS<br />

Fast Facts:<br />

p The new ad addresses broader<br />

issues around the need to increase<br />

investments in carbon-free technologies,<br />

while also committing to<br />

more nuclear energy alongside<br />

wind and solar.<br />

p The “See the Light” campaign will<br />

appear primarily inside the<br />

Beltway on digital and social media<br />

plat<strong>for</strong>ms. The campaign’s paid<br />

media strategy aims to share the<br />

ad’s compelling visuals and<br />

connect how we power our daily<br />

lives and address the climate crisis.<br />

p NEI’s creative advertising and<br />

media buying are consolidated<br />

under Bully Pulpit Interactive<br />

(BPI).<br />

What Maria Korsnick, president and<br />

chief executive officer of NEI, has to<br />

say: “Our new ‘See the Light’ campaign<br />

demonstrates that we are at a pivotal<br />

moment – thinking differently about<br />

the best way to reach a carbon-free<br />

future. To make this future a reality, it<br />

will require policymakers and the<br />

private sector to increase investments<br />

and support policies that position<br />

nuclear as the backbone of our energy<br />

future. <strong>Nuclear</strong> energy, paired with<br />

wind and solar, can be the source<br />

that powers us to a brighter future –<br />

helping us meet the challenges of<br />

electricity production, job creation<br />

and decarbonizing our economy.”<br />

Big Picture: The ad campaign<br />

mirrors increasing attention <strong>for</strong><br />

solutions to address the climate crisis<br />

along with the funding necessary<br />

to meet climate goals. Policymakers<br />

and the private sector are taking<br />

steps to invest in that future – making<br />

nuclear energy a key element of these<br />

ef<strong>for</strong>ts.<br />

This year, the Biden administration<br />

proposed a record-high budget<br />

proposal of $1.9 billion <strong>for</strong> nuclear<br />

programs and has pledged to bring<br />

carbon emissions from electricity<br />

generation close to zero by 2035.<br />

Congress has also introduced several<br />

proposals to support nuclear plants,<br />

demonstrating the strong bipartisan<br />

support <strong>for</strong> nuclear as a source of<br />

reliable carbon-free power that can<br />

power the grid while also generating<br />

clean hydrogen and providing<br />

well-paying, long term jobs. And in<br />

the private sector, utilities are making<br />

bold commitments to reach net zero<br />

carbon emissions by 2050 or sooner –<br />

viewing nuclear energy as a key<br />

element to reaching their commitments.<br />

The ad can be found on NEI’s<br />

website and YouTube, as well as our<br />

social media channels under the<br />

hashtag #SeeTheLight.<br />

| www.nei.org<br />

“More Uranium Development<br />

Needed to Meet Demands of<br />

Growing <strong>Nuclear</strong> Fleet” –<br />

World <strong>Nuclear</strong> Association<br />

launches the <strong>Nuclear</strong> Fuel<br />

Report 2021<br />

(wna) World <strong>Nuclear</strong> Association<br />

launched the 2021 edition of The<br />

<strong>Nuclear</strong> Fuel Report, concluding that<br />

the positive trend in nuclear<br />

Operating Results July 2021<br />

Plant name Country Nominal<br />

capacity<br />

Type<br />

gross<br />

[MW]<br />

net<br />

[MW]<br />

Operating<br />

time<br />

generator<br />

[h]<br />

Energy generated, gross<br />

[MWh]<br />

Month Year Since<br />

commissioning<br />

Time availability<br />

[%]<br />

Energy availability<br />

[%] *) Energy utilisation<br />

[%] *)<br />

Month Year Month Year Month Year<br />

OL1 Olkiluoto BWR FI 920 890 744 675 009 4 274 075 281 308 792 100.00 92.43 100.00 91.22 98.62 91.33<br />

OL2 Olkiluoto BWR FI 920 890 744 667 969 3 908 343 270 811 065 100.00 83.91 99.61 83.43 97.59 83.51<br />

KCB Borssele 1) PWR NL 512 484 735 348 976 2 020 948 174 089 745 94.00 78.85 94.03 90.41 91.35 77.57<br />

KKB 1 Beznau 7) PWR CH 380 365 744 277 593 1 808 997 135 020 383 100.00 93.89 100.00 93.47 98.14 93.51<br />

KKB 2 Beznau 7) PWR CH 380 365 744 276 265 1 922 699 142 299 000 100.00 100.00 100.00 100.00 97.68 99.48<br />

KKG Gösgen 7) PWR CH 1060 1010 744 774 841 4 448 202 335 334 791 100.00 94.67 99.91 82.38 98.25 82.49<br />

CNT-I Trillo PWR ES 1066 1003 744 779 046 4 063 296 268 087 144 100.00 76.69 100.00 75.98 97.32 74.41<br />

Dukovany B1 PWR CZ 500 473 744 359 263 1 998 118 121 642 557 100.00 80.36 100.00 79.62 96.58 78.56<br />

Dukovany B2 PWR CZ 500 473 744 355 040 2 495 518 117 107 433 100.00 100.00 100.00 99.87 95.44 98.11<br />

Dukovany B3 PWR CZ 500 473 458 196 070 1 928 043 115 288 600 61.56 78.57 54.54 77.32 52.71 75.80<br />

Dukovany B4 PWR CZ 500 473 744 361 452 1 851 720 116 417 621 100.00 74.31 99.15 73.60 97.16 72.80<br />

Temelin B1 PWR CZ 1082 1032 744 803 159 4 000 384 133 571 674 100.00 73.34 99.96 72.53 99.77 72.68<br />

Temelin B2 PWR CZ 1086 1036 0 0 4 559 781 130 148 685 0 82.07 0 81.74 0 82.54<br />

Doel 1 PWR BE 467 445 744 342 458 2 042 436 142 056 277 100.00 85.94 99.92 85.28 98.60 86.14<br />

Doel 2 PWR BE 467 445 681 304 498 1 956 341 140 566 406 91.55 83.12 88.79 82.20 87.34 82.19<br />

Doel 3 PWR BE 1056 1006 744 763 114 5 335 922 276 548 232 100.00 99.45 100.00 99.09 96.29 98.72<br />

Doel 4 PWR BE 1086 1038 744 799 060 5 539 882 282 899 871 100.00 100.00 100.00 100.00 97.42 98.92<br />

Tihange 1 PWR BE 1009 962 744 736 020 5 101 732 312 968 707 100.00 100.00 99.77 99.62 98.35 99.62<br />

Tihange 2 PWR BE 1055 1008 744 762 548 4 118 827 269 822 643 100.00 81.81 99.34 77.32 97.98 77.32<br />

Tihange 3 PWR BE 1089 1038 744 790 300 5 460 436 292 114 095 100.00 100.00 99.88 99.96 98.02 99.09<br />

Plant name<br />

Type<br />

Nominal<br />

capacity<br />

gross<br />

[MW]<br />

net<br />

[MW]<br />

Operating<br />

time<br />

generator<br />

[h]<br />

Energy generated, gross<br />

[MWh]<br />

Time availability<br />

[%]<br />

Energy availability Energy utilisation<br />

[%] *) [%] *)<br />

Month Year Since Month Year Month Year Month Year<br />

commissioning<br />

GKN-II Neckarwestheim 1,2,4) DWR 1400 1310 539 727 270 6 063 720 357 415 264 72.38 86.84 72.38 86.82 69.93 85.33<br />

KBR Brokdorf DWR 1480 1410 744 1 040 747 7 032 265 378 295 594 100.00 100.00 99.64 99.92 94.21 93.19<br />

KKE Emsland DWR 1406 1335 744 1 007 106 6 474 751 375 485 452 100.00 92.15 100.00 92.02 96.20 90.56<br />

KKI-2 Isar DWR 1485 1410 744 1 058 590 7 357 319 384 786 362 100.00 100.00 99.97 99.99 95.39 97.06<br />

KRB C Gundremmingen SWR 1344 1288 744 983 282 6 807 442 357 285 208 100.00 100.00 100.00 99.65 97.52 98.85<br />

KWG Grohnde DWR 1430 1360 744 1 011 906 6 057 299 404 817 648 100.00 87.44 99.97 87.11 94.54 82.81<br />


<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

gene rating capacity projections that<br />

began in the previous (2019) report<br />

con tinues. In the period, two newcomer<br />

countries – Belarus and the<br />

United Arab Emirates – connected<br />

their first reactors to the grid; further<br />

reactors were commissioned in China,<br />

India, Pakistan and Russia; construction<br />

of new reactors was<br />

launched in China, Iran and Turkey;<br />

and many other countries are considering<br />

either expanding existing<br />

nuclear programmes (e.g. Argentina,<br />

Brazil, Bulgaria, Netherlands,<br />

Romania, South Africa) or building<br />

their first reactors (e.g. Egypt, Poland,<br />

Uzbekistan). The nuclear generation<br />

capacity is expected to grow by 2.6 %<br />

annually, reaching 615 GWe by 2040<br />

in the Reference Scenario.<br />

The report found that world<br />

uranium production dropped considerably<br />

from 63,207 tonnes of<br />

uranium (tU) in 2016 to 47,731 tU in<br />

2020. Unfavourable market conditions,<br />

compounded by the Covid-19<br />

pandemic, led to a sharp decrease in<br />

Operating Results August 2021<br />

investment in the development of new<br />

and existing mines. The currently<br />

depressed uranium market has caused<br />

not only a sharp decrease in uranium<br />

exploration activities (by 77 % from<br />

$2.12 billion in 2014 to nearly $483<br />

million in 2018) but also the curtailment<br />

of uranium production at<br />

existing mines, with more than<br />

20,500 tonnes of annual production<br />

being idled. Uranium production<br />

volumes at existing mines are projected<br />

to remain fairly stable until the<br />

late 2020s, then decreasing by more<br />

than half from 2030 to 2040<br />

Intense development of new projects<br />

will be needed in the current<br />

decade to avoid potential supply<br />

disruptions. A number of projects at<br />

very advanced stages of development<br />

are waiting <strong>for</strong> an improved supplydemand<br />

market situation in order to<br />

commence uranium production.<br />

According to the report, there will<br />

have to be a doubling in the development<br />

pipeline <strong>for</strong> new projects by<br />

2040. There are more than adequate<br />

project extensions, uranium resources<br />

and other projects in the pipeline to<br />

accomplish this need, but it is essential<br />

<strong>for</strong> the market to send the signals<br />

needed to launch the development of<br />

these projects.Beyond mining, the<br />

report found that:<br />

p In the conversion sector, near-term<br />

reactor requirements in UF6 will<br />

be largely covered by commercial<br />

inventories. By 2023, global conversion<br />

production is expected to<br />

meet requirements due to the<br />

ramp-up and restart of existing<br />

facilities. Nevertheless, in the longrun<br />

more conversion capacity will<br />

be needed.<br />

p In the enrichment sector, excess<br />

capacity is currently used <strong>for</strong><br />

underfeeding and tails re-enrichment,<br />

bringing in approximately<br />

6,000 – 8,000 tU in support of the<br />

undersupplied uranium market.<br />

This will largely be reduced over<br />

time, as enrichment requirements<br />

rise due to nuclear generating<br />

capacity growth.<br />

*) Net-based values<br />

(Czech and Swiss nuclear<br />

power plants<br />

gross-based)<br />

1) Refueling<br />

2) Inspection<br />

3) Repair<br />

4) Stretch-out-operation<br />

5) Stretch-in-operation<br />

6) Hereof traction supply<br />

7) Incl. steam supply<br />

BWR: Boiling<br />

Water Reactor<br />

PWR: Pressurised Water<br />

Reactor<br />

Source: VGB<br />

59<br />

NEWS<br />

Plant name Country Nominal<br />

capacity<br />

Type<br />

gross<br />

[MW]<br />

net<br />

[MW]<br />

Operating<br />

time<br />

generator<br />

[h]<br />

Energy generated, gross<br />

[MWh]<br />

Month Year Since<br />

commissioning<br />

Time availability<br />

[%]<br />

Energy availability<br />

[%] *) Energy utilisation<br />

[%] *)<br />

Month Year Month Year Month Year<br />

OL1 Olkiluoto BWR FI 920 890 744 678 638 4 952 713 281 987 430 100.00 93.40 99.94 92.33 99.15 92.32<br />

OL2 Olkiluoto BWR FI 920 890 744 676 351 4 584 694 271 487 416 100.00 85.96 100.00 85.54 98.81 85.46<br />

KCB Borssele PWR NL 512 484 744 371 004 2 391 952 174 460 749 100.00 81.55 100.00 90.37 97.31 80.09<br />

KKB 1 Beznau 7) PWR CH 380 365 744 278 008 2 087 005 135 298 391 100.00 94.67 100.00 94.30 98.25 94.11<br />

KKB 2 Beznau 1,2,7) PWR CH 380 365 131 48 079 1 970 778 142 347 079 17.61 89.49 17.37 89.46 16.58 88.90<br />

KKG Gösgen 7) PWR CH 1060 1010 744 775 908 5 224 110 336 110 699 100.00 95.35 99.82 84.61 98.39 84.52<br />

CNT-I Trillo PWR ES 1066 1003 744 779 119 4 842 415 268 866 263 100.00 79.67 100.00 79.05 97.35 77.33<br />

Dukovany B1 PWR CZ 500 473 744 361 297 2 359 415 122 003 854 100.00 82.87 100.00 82.22 97.12 80.93<br />

Dukovany B2 PWR CZ 500 473 744 356 903 2 852 421 117 464 336 100.00 100.00 100.00 99.89 95.94 97.84<br />

Dukovany B3 PWR CZ 500 473 744 361 147 2 289 190 115 649 747 100.00 81.31 100.00 80.22 97.08 78.52<br />

Dukovany B4 PWR CZ 500 473 744 364 221 2 215 940 116 781 842 100.00 77.59 100.00 76.97 97.91 76.01<br />

Temelin B1 PWR CZ 1086 1036 744 805 249 4 805 633 134 376 923 100.00 76.74 99.93 76.04 99.66 76.13<br />

Temelin B2 PWR CZ 1086 1036 102 91 707 4 651 488 130 240 392 13.71 73.35 11.35 72.76 11.35 73.45<br />

Doel 1 PWR BE 467 445 744 342 593 2 385 029 142 398 870 100.00 87.73 99.92 87.14 99.24 87.81<br />

Doel 2 PWR BE 467 445 744 342 520 2 298 860 140 908 925 100.00 85.27 99.98 84.47 98.34 84.25<br />

Doel 3 PWR BE 1056 1006 644 594 195 5 930 116 277 142 427 86.52 97.80 86.14 97.44 74.58 95.64<br />

Doel 4 PWR BE 1086 1038 744 796 782 6 336 664 283 696 653 100.00 100.00 100.00 100.00 97.12 98.69<br />

Tihange 1 PWR BE 1009 962 744 737 243 5 838 975 313 705 950 100.00 100.00 99.84 99.64 98.31 99.45<br />

Tihange 2 PWR BE 1055 1008 744 749 533 4 868 359 270 572 176 100.00 84.13 96.93 79.82 96.23 79.73<br />

Tihange 3 PWR BE 1089 1038 744 789 968 6 250 404 292 904 062 100.00 100.00 99.85 99.95 97.98 98.95<br />

Plant name<br />

Type<br />

Nominal<br />

capacity<br />

gross<br />

[MW]<br />

net<br />

[MW]<br />

Operating<br />

time<br />

generator<br />

[h]<br />

Energy generated, gross<br />

[MWh]<br />

Time availability<br />

[%]<br />

Energy availability Energy utilisation<br />

[%] *) [%] *)<br />

Month Year Since Month Year Month Year Month Year<br />

commissioning<br />

GKN-II Neckarwestheim DWR 1400 1310 736 1 010 380 7 074 100 358 425 644 100.00 88.52 100.00 88.50 97.21 86.84<br />

KBR Brokdorf DWR 1480 1410 744 1 038 035 8 070 300 379 333 629 100.00 100.00 98.65 99.76 94.05 93.30<br />

KKE Emsland DWR 1406 1335 744 1 008 876 7 483 627 376 494 328 100.00 93.15 100.00 93.04 96.39 91.30<br />

KKI-2 Isar DWR 1485 1410 744 1 072 168 8 429 487 385 858 530 100.00 100.00 100.00 99.99 96.67 97.01<br />

KRB C Gundremmingen SWR 1344 1288 744 985 716 7 793 158 358 270 924 100.00 100.00 100.00 99.70 97.78 98.72<br />

KWG Grohnde DWR 1430 1360 744 1 005 707 7 063 006 405 823 355 100.00 89.04 100.00 88.76 93.89 84.23<br />


<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

60<br />

NEWS<br />

p In the fuel fabrication sector, competition<br />

may become more intense<br />

from both the commercial and<br />

technological perspective, due to<br />

increased interest in developments<br />

of advanced fuels (e.g. <strong>for</strong> nonlight<br />

water reactors). <strong>Nuclear</strong> fuel<br />

demand increasing in Asia and<br />

decreasing in the West may cause<br />

fuel vendors to move from a<br />

regional to a more global market<br />

approach.<br />

Sama Bilbao y León, Director General<br />

of World <strong>Nuclear</strong> Association, said<br />

following the launch of the <strong>Nuclear</strong><br />

Fuel Report 2021:“Given its unique<br />

combination of attributes – reliability,<br />

af<strong>for</strong>dability, low-carbon and universal<br />

deployability – it is clear that<br />

nuclear energy will play an even larger<br />

role in the electricity and energy<br />

systems of tomorrow. The <strong>Nuclear</strong><br />

Fuel Report makes it clear that<br />

sufficient uranium resources exist to<br />

meet the expected growth, but<br />

uranium markets need to rebalance to<br />

incentivise investment in uranium<br />

mining to support the expansion of<br />

the global nuclear fleet.”<br />

| www.world-nuclear.org<br />

Company News<br />

Environmentalists Praise<br />

World’s Only Floating NPP<br />

<strong>for</strong> High Efficiency and<br />

Eco-friendliness<br />

(rosatom) ROSATOM’s floating<br />

nuclear power plant in the city of<br />

Pevek of Russia’s Chukotka Autonomous<br />

Okrug has been visited by a<br />

public expedition <strong>for</strong> the very first<br />

time. Led by Alexey Yekidin, a<br />

leading researcher at the Institute of<br />

Indus trial Ecology of the Ural Branch<br />

of the Russian Academy of Sciences,<br />

the expedition united ecologists,<br />

academics, and representatives of<br />

public associations.<br />

Participants were tasked with<br />

collecting and analysing data on the<br />

environmental and radiation safety of<br />

the floating nuclear power plant, as<br />

well as assessing the plant and its<br />

overall operation and communicating<br />

their findings to the public.<br />

The environmentalists carried out<br />

measure ments both at the station<br />

itself and in the surrounding area, as<br />

well as in the city of Pevek.<br />

Their findings showed that<br />

background radiation in both the<br />

vicinity of the floating nuclear power<br />

plant and in the city of Pevek arise<br />

exclusively from natural sources: i.e.,<br />

from natural radionuclides and<br />

cosmic radiation, and that the average<br />

value of said radiation does not exceed<br />

0.12 μSv/h in either location.<br />

“More than 20 measurements were<br />

taken in the FNPP’s industrial area, as<br />

well as in the surrounding area and in<br />

the city of Pevek, and no artificial<br />

radionuclides were found at the<br />

surveyed sites. It has, there<strong>for</strong>e, been<br />

concluded that the operation of the<br />

floating nuclear power plant does not<br />

negatively impact the region’s radioecological<br />

situation,” said Aleksey<br />

Yekidin.<br />

Alan Khasiev, chairman of the<br />

coordinating council of the interregional<br />

public environmental movement<br />

“Oka,” said: “Our programme<br />

has been in operation since 2010.<br />

During this time, we have carried out<br />

44 full-scale environmental monitoring<br />

expeditions to most of the<br />

Russian-designed NPPs operating in<br />

both Russia and abroad, as well as to<br />

those under construction. The ecological<br />

expedition to the floating<br />

nuclear power plant and around Pevek<br />

is the latest stage of this programme.”<br />

Green Party member and biologist<br />

Larisa Kosyuk said: “In the context of<br />

the carbon emissions regulation introduced<br />

by the EU, the FNPP project can<br />

serve as an example of green technologies<br />

in the energy sector. Such<br />

power plants will be especially useful<br />

in the regions of Russia’s Far North<br />

and the Far East, where there are no<br />

hydrological resources of energy, and<br />

the delivery of fuel – such as coal and<br />

oil products – is expensive. According<br />

to the <strong>International</strong> Energy Agency,<br />

nuclear energy is the second largest<br />

source of energy in the world,<br />

accounting <strong>for</strong> 10% of the world’s<br />

electricity generation.”<br />

Kirill Toropov, deputy director of<br />

the local branch of Rosenergoatom<br />

JSC, told the expedition participants<br />

about the changes that have taken<br />

place in the town since the floating<br />

nuclear power plant has been connected<br />

to the grid: “Not only did the<br />

arrival of the floating nuclear power<br />

plant introduce an additional source of<br />

energy, it opened a new chapter in the<br />

development of SMR technologies in<br />

the region’s energy sector and in that<br />

of the country as a whole. Since its<br />

commissioning, the FNPP has already<br />

established itself as a reliable and<br />

innovative source of thermal and<br />

electric energy. One cannot help bu<br />

notice its positive contribution to improving<br />

the environmental situation,<br />

both in Pevek itself (with a 30 %<br />

reduction in soot emissions from the<br />

Chaunskaya CHPP) and in the<br />

surrounding bodies of water (as<br />

evinced by the restoration of flora and<br />

fauna in the Chaunskaya Bay and the<br />

return of seals and other species of<br />

marine animals).”<br />

Ivan Leyushkin, head of the Pevek<br />

municipal district, noted the expedition’s<br />

importance <strong>for</strong> in<strong>for</strong>ming the<br />

region’s residents about the safety of<br />

nuclear generation. Leyushkin also<br />

spoke about the crucial role that the<br />

NPP is playing in the region’s development:<br />

“Since the floating nuclear<br />

power plant has started operation,<br />

107 million roubles’ worth of socially<br />

significant projects have been<br />

implemented in the Pevek district.<br />

Our cooperation with Rosatom will<br />

continue in the future. In September<br />

of this year, a cooperation agreement<br />

was signed between the governor of<br />

the Chukotka Autonomous Okrug<br />

Roman Kopin and the ROSATOM’s<br />

Director General Alexey Likhachev.”<br />

Facts<br />

Equipped with two KLT-40S reactors,<br />

the FNPP is a source of 70 MW’s worth<br />

of electrical energy and 50 Gcal/h’s<br />

worth of thermal energy. It comprises<br />

the floating power unit (FPU) “Akademik<br />

Lomonosov” and onshore infrastructure<br />

designed to supply heat and<br />

electricity from the FPU to consumers.<br />

Since its commissioning, the FNPP<br />

has carried out continuous industrial<br />

environmental monitoring, as well as<br />

monitoring of air, soil cover, sea water,<br />

bottom sediments, aquatic biological<br />

resources, as well as atmospheric<br />

emissions and the management of<br />

waste from production and consumption.<br />

The cost of the operation of the<br />

plant’s environmental protection<br />

system amounted to more than<br />

17 million roubles in 2020. The NPP’s<br />

operation also made it possible to<br />

prevent more than 300 thousand<br />

tonnes of atmospheric CO 2 equivalent<br />

emissions in 2019 and 2020.<br />

| www.rosatom.ru<br />

Market data<br />

(All in<strong>for</strong>mation is supplied without<br />

guarantee.)<br />

<strong>Nuclear</strong> Fuel Supply<br />

Market Data<br />

In<strong>for</strong>mation in current (nominal)<br />

U.S.-$. No inflation adjustment of<br />

prices on a base year. Separative work<br />

data <strong>for</strong> the <strong>for</strong>merly “secondary<br />

market”.<br />


<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

Uranium<br />

Prize range: Spot market [USD*/lb(US) U 3O 8]<br />

140.00<br />

120.00<br />

) 1<br />

Uranium prize range: Spot market [USD*/lb(US) U 3O 8]<br />

140.00<br />

) 1<br />

120.00<br />

61<br />

100.00<br />

100.00<br />

80.00<br />

80.00<br />

60.00<br />

40.00<br />

20.00<br />

Yearly average prices in real USD, base: US prices (1982 to1984) *<br />

60.00<br />

40.00<br />

20.00<br />

NEWS<br />

0.00<br />

1980<br />

1985<br />

1990<br />

1995<br />

2000<br />

2005<br />

2010<br />

2015<br />

2020<br />

2021<br />

Year<br />

* Actual nominal USD prices, not real prices referring to a base year. Sources: Energy Intelligence, Nukem; Bild/Figure: <strong>atw</strong> 2021<br />

* Actual nominal USD prices, not real prices referring to a base year. Year<br />

Sources: Energy Intelligence, Nukem; Bild/Figure: <strong>atw</strong> 2021<br />

| Uranium spot market prices from 1980 to 2021 and from 2009 to 2021. The price range is shown.<br />

In years with U.S. trade restrictions the unrestricted uranium spot market price is shown.<br />

Separative work: Spot market price range [USD*/kg UTA]<br />

Conversion: Spot conversion price range [USD*/kgU]<br />

180.00<br />

26.00<br />

) 1 ) 1<br />

160.00<br />

140.00<br />

0.00<br />

24.00<br />

22.00<br />

20.00<br />

Jan. 2009<br />

Jan. 2010<br />

Jan. 2011<br />

Jan. 2012<br />

Jan. 2013<br />

Jan. 2014<br />

Jan. 2015<br />

Jan. 2016<br />

Jan. 2017<br />

Jan. 2018<br />

Jan. 2019<br />

Jan. 2020<br />

Jan. 2021<br />

Jan. 2022<br />

120.00<br />

18.00<br />

16.00<br />

100.00<br />

14.00<br />

80.00<br />

12.00<br />

10.00<br />

60.00<br />

8.00<br />

40.00<br />

6.00<br />

20.00<br />

4.00<br />

2.00<br />

0.00<br />

0.00<br />

Jan. 2009<br />

Jan. 2010<br />

Jan. 2011<br />

Jan. 2012<br />

Jan. 2013<br />

Jan. 2014<br />

* Actual nominal USD prices, not real prices referring to a base year. Year<br />

Jan. 2015<br />

Jan. 2016<br />

Jan. 2017<br />

Jan. 2018<br />

Jan. 2019<br />

Jan. 2020<br />

Jan. 2021<br />

Jan. 2022<br />

Sources: Energy Intelligence, Nukem; Bild/Figure: <strong>atw</strong> 2021<br />

Jan. 2009<br />

Jan. 2010<br />

Jan. 2011<br />

Jan. 2012<br />

Jan. 2013<br />

Jan. 2014<br />

* Actual nominal USD prices, not real prices referring to a base year. Year<br />

Jan. 2015<br />

Jan. 2016<br />

Jan. 2017<br />

Jan. 2018<br />

Jan. 2019<br />

Jan. 2020<br />

Jan. 2021<br />

Jan. 2022<br />

Sources: Energy Intelligence, Nukem; Bild/Figure: <strong>atw</strong> 2021<br />

| Separative work and conversion market price ranges from 2009 to 2021. The price range is shown.<br />

)1<br />

In December 2009 Energy Intelligence changed the method of calculation <strong>for</strong> spot market prices. The change results in virtual price leaps.<br />

* Actual nominal USD prices, not real prices referring to a base year<br />

Sources: Energy Intelligence, Nukem; Bilder/Figures: <strong>atw</strong> 2021<br />

Uranium prices [US-$/lb U 3 O 8 ; 1 lb =<br />

453.53 g; 1 lb U 3 O 8 = 0.385 kg U].<br />

Conversion prices [US-$/kg U],<br />

Separative work [US-$/SWU (Separative<br />

work unit)].<br />

2017<br />

p Uranium: 19.25–26.50<br />

p Conversion: 4.50–6.75<br />

p Separative work: 39.00–50.00<br />

2018<br />

p Uranium: 21.75–29.20<br />

p Conversion: 6.00–14.50<br />

p Separative work: 34.00–42.00<br />

2019<br />

p Uranium: 23.90–29.10<br />

p Conversion: 13.50–23.00<br />

p Separative work: 41.00–52.00<br />

2020<br />

January to March 2020<br />

p Uranium: 24.10–27.40<br />

p Conversion: 21.50–23.50<br />

p Separative work: 45.00–53.00<br />

April 2020<br />

p Uranium: 27.50–34.00<br />

p Conversion: 21.50–23.50<br />

p Separative work: 45.00–52.00<br />

May 2020<br />

p Uranium: 33.50–34.50<br />

p Conversion: 21.50–23.50<br />

p Separative work: 48.00–52.00<br />

June 2020<br />

p Uranium: 33.00–33.50<br />

p Conversion: 21.50–23.50<br />

p Separative work: 49.00–52.00<br />

July 2020<br />

p Uranium: 32.50–33.20<br />

p Conversion: 21.50–23.50<br />

p Separative work: 50.50–53.50<br />

August 2020<br />

p Uranium: 30.50–32.25<br />

p Conversion: 21.50–23.50<br />

p Separative work: 51.00–54.00<br />

September 2020<br />

p Uranium: 29.90–30.75<br />

p Conversion: 21.00–22.00<br />

p Separative work: 51.00–54.00<br />

October 2020<br />

p Uranium: 28.90–30.20<br />

p Conversion: 21.00–22.00<br />

p Separative work: 51.00–53.00<br />

November 2020<br />

p Uranium: 28.75–30.25<br />

p Conversion: 19.00–22.00<br />

p Separative work: 51.00–53.00<br />

December 2020<br />

p Uranium: 29.50–30.40<br />

p Conversion: 19.00–22.00<br />

p Separative work: 51.00–53.00<br />

2021<br />

January 2021<br />

p Uranium: 29.50–30.50<br />

p Conversion: 19.00–22.00<br />

p Separative work: 51.00–53.00<br />

February 2021<br />

p Uranium: 28.75–29.10<br />

p Conversion: 20.00–22.00<br />

p Separative work: 52.00–54.00<br />

March 2021<br />

p Uranium: 27.25–31.00<br />

p Conversion: 20.00–22.00<br />

p Separative work: 52.00–55.00<br />

April 2021<br />

p Uranium: 28.40–31.00<br />

p Conversion: 19.00–21.00<br />

p Separative work: 51.00–54.00<br />

May 2021<br />

p Uranium: 29.15–31.35<br />

p Conversion: 19.50–21.50<br />

p Separative work: 52.00–54.00<br />

June 2021<br />

p Uranium:31.00–32.50<br />

p Conversion: 19.50–21.50<br />

p Separative work: 54.00–56.00<br />

July 2021<br />

p Uranium:32.20–32.50<br />

p Conversion: 19.00–21.00<br />

p Separative work: 54.00–56.00<br />

August 2021<br />

p Uranium:32.20–36.00<br />

p Conversion: 19.00–21.00<br />

p Separative work: 55.50–57.50<br />

| Source: Energy Intelligence<br />

www.energyintel.com<br />


<strong>atw</strong> Vol. 66 (2021) | Issue 6 ı November<br />

62<br />

For the Sake of <strong>Nuclear</strong> Safety and Security,<br />

it’s Time <strong>for</strong> Rogue Actors to Leave the Stage<br />


John Shepherd is<br />

editor-in-chief of the online<br />

publication<br />

New Energy 360 & World<br />

Battery News.<br />

Sources:<br />

Rafael Grossi on Iran:<br />

https://bit.ly/3E3BZzC<br />

University of Bristol<br />

report:<br />

https://bit.ly/3Ei04mv<br />

As I sat down to write this article, the UK’s newly installed <strong>for</strong>eign minister, Liz Truss, announced an upcoming<br />

meeting with her Iranian counterpart on the sidelines of the UN General Assembly in New York.<br />

Truss said she would tell Iran’s <strong>for</strong>eign minister, Amir<br />

Abdollahian, that his country should return “to the nuclear<br />

deal negotiating table be<strong>for</strong>e it is too late” – a phrase so<br />

often recycled by world leaders on the issue.<br />

Sadly, the words ‘too late’ have, in the case of Iran and<br />

its in-out-in-out adherence to international agreements,<br />

lost all meaning.<br />

To refresh our minds, Truss was referring to the Joint<br />

Comprehensive Plan of Action (JCPoA), an agreement on<br />

the Iranian nuclear programme supposedly reached in<br />

Vienna, SIX years ago, between Iran and the five permanent<br />

members of the UN Security Council – China, France,<br />

Russia, UK, US – plus Germany (the P5+1), together with<br />

the EU.<br />

Well, political leaders have since come and gone in<br />

many of the concerned nations and the Iran nuclear deal<br />

has ricocheted up and down the scale of geopolitical<br />

priorities. But one thing has not changed: Iran’s alternating<br />

acceptance and defiance of what it chooses to ‘like’ about<br />

the deal. This pattern of behaviour has gone on <strong>for</strong> so long<br />

that it defies belief.<br />

The revolving door of Iran’s diplomatic dance with the<br />

nuclear energy community is still turning. As this article<br />

was written, the director-general of the <strong>International</strong><br />

Atomic Energy Agency (IAEA), Rafael Grossi, was holding<br />

talks with Mohammad Eslami, Iran’s vice-president and<br />

head of the Atomic Energy Organization of the Islamic<br />

Republic of Iran (AEOI).<br />

Under normal circumstances, it would be uplifting to<br />

hear of such high-level talks, were it not <strong>for</strong> the fact that<br />

Mr Grossi is the third IAEA chief in my own humble career<br />

reporting on matters nuclear – and Iran was a recalcitrant<br />

player long be<strong>for</strong>e.<br />

I remember interviewing a <strong>for</strong>mer IAEA chief,<br />

Mohamed ElBaradei, at his office in Vienna about the<br />

Iran question some 20 years ago. I asked then what I<br />

ask again now… What have all these years of talking<br />

achieved?<br />

I suppose keeping a lid on potentially more terrifying<br />

nuclear behaviour by Iran is an achievement of sorts, but is<br />

that the best the world can hope <strong>for</strong>?<br />

And of course, Iran’s actions continue to embolden<br />

others, such as North Korea, who believe they have the<br />

right to benefit from nuclear energy without guaranteeing<br />

the peaceful nature of its use in their hands.<br />

Those of us who are of a ‘certain age’ remember the<br />

1995 deal between the US, Japan and South Korea, which<br />

was to establish the Korean Peninsula Energy Development<br />

Organization. That deal garnered financing to<br />

support the construction of two light water reactors in<br />

North Korea, in return <strong>for</strong> the state to freeze its illicit<br />

plutonium weapons programme. It was yet another<br />

ultimately doomed venture.<br />

Fast <strong>for</strong>ward to this year, and we once again have the<br />

IAEA “reiterating previous calls” <strong>for</strong> North Korea to comply<br />

with its obligations, including full and effective implementation<br />

of Non-Proliferation Treaty safeguards.<br />

The current head of the IAEA said North Korea’s<br />

secretive nuclear activities are (once again) “a cause <strong>for</strong><br />

serious concern… and deeply troubling”.<br />

It seems even the special relationship that <strong>for</strong>mer US<br />

President Donald Trump said he had established with the<br />

North Korean leader counted <strong>for</strong> nothing after all – not<br />

even after those “love letters” Kim Jong-un penned to the<br />

president. How fickle – and how unsurprising!<br />

So how do we redouble our ef<strong>for</strong>ts and work <strong>for</strong> a world<br />

in which all those who wish to use atomic power <strong>for</strong><br />

peaceful purposes adhere to the requirements to ensure<br />

real nuclear safety and security?<br />

I’m reminded of a quote attributed to <strong>for</strong>mer UK prime<br />

minister Sir Winston Churchill: “Success is not final, failure<br />

is not fatal: it is the courage to continue that counts.”<br />

Courage will indeed be needed and so will<br />

determination – and readers will know that our industry<br />

has determination, expertise and zeal in abundance.<br />

To demonstrate, I was interested to read of new<br />

cooperation on pioneering radiation mapping research<br />

carried out by researchers from the University of Bristol<br />

and Ukrainian researchers and engineers at the Chernobyl<br />

<strong>Nuclear</strong> <strong>Power</strong> Plant.<br />

The university team was given privileged access to the<br />

now infamous control room of the plant’s fourth reactor,<br />

where they deployed specially developed radiation<br />

mapping and scanning sensors. These were also deployed<br />

inside the New Safe Confinement, the protective structure<br />

erected to cover the remains of the destroyed reactor and<br />

the original ‘sarcophagus’, which was hastily constructed<br />

in the aftermath of the 1986 accident.<br />

The aim of that visit was to further explore the value of<br />

autonomous and semi-autonomous radiation mapping<br />

systems in high-radiation environments. The teams say<br />

that in deploying these systems in the exclusion zone and<br />

at the plant, researchers were able to better define the<br />

location and amount of residual radiological hazards.<br />

This work encapsulates the vision and drive that is the<br />

hallmark of research supported by the nuclear industry –<br />

harnessing advances in technology to address the legacy of<br />

Chernobyl (and Fukushima), thereby enhancing expertise<br />

and knowhow to strengthen technical capabilities <strong>for</strong> the<br />

years ahead.<br />

We must look to the future because I venture to suggest<br />

that is where nuclear’s best years lie, provided rogue actors<br />

are not allowed to continue with clandestine activities –<br />

which will cause immense harm to the reputation of the<br />

nuclear industry as a whole.<br />

Looking ahead, peaceful nuclear activities should also<br />

go beyond the generation of electricity, such as innovations<br />

in using heat from nuclear power plants <strong>for</strong> heating<br />

or cooling, or as an energy source towards the production<br />

of fresh water, hydrogen or other products.<br />

To return to the Churchill quotation, having the courage<br />

to continue with support and investments <strong>for</strong> nuclear<br />

technologies will indeed improve lives and combat the<br />

worse effects of climate change.<br />

However, world leaders and nuclear industry leaders<br />

need to exclude access to nuclear technologies – once and<br />

<strong>for</strong> all – to those who refuse to be demonstrably transparent<br />

about their activities. Those bad actors represent<br />

the greatest risk to nuclear safety and security and the<br />

biggest threat to the nuclear energy industry’s future.<br />

<strong>Nuclear</strong> Today<br />

For the Sake of <strong>Nuclear</strong> Safety and Security, it’s Time <strong>for</strong> Rogue Actors to Leave the Stage ı John Shepherd


Important notice <strong>for</strong> international subscribers!<br />

The nuclear phase out in Germany not only concerns the nuclear industry itself but literally<br />

all accompanying national players such as the specialized press.<br />

There<strong>for</strong>e, also <strong>atw</strong> – <strong>International</strong> <strong>Journal</strong> <strong>for</strong> <strong>Nuclear</strong> <strong>Power</strong> must shift its focus and will emphasize<br />

the national context much more than in previous years. One focus will be on the German nuclear<br />

decommissioning market and consistently many articles published will be primarily in German<br />

language.<br />

Despite the fact that much of the journal will be in German, starting January 2022, we plan<br />

to publish some English articles in every issue. Furthermore, the size of the magazine will be<br />

reduced according to the shrinking importance of the national German nuclear market.<br />

As a consequence of all the changes announced above, our international customers have<br />

a special right of immediate and easy termination of subscription which is effective on<br />

January 1 st , 2022.<br />

Nevertheless, to maintain our strong claim to foster nuclear competence a new online plat<strong>for</strong>m<br />

‘<strong>atw</strong> scientific’ will be introduced in November 2021. Completely in English language, ‘<strong>atw</strong> scientific’<br />

will be an open-source web plat<strong>for</strong>m <strong>for</strong> worldwide scientific nuclear content, where submissions<br />

will be pre-reviewed and DOIs will be assigned. More in<strong>for</strong>mation will be available soon!<br />

We would like to take this opportunity to warmly thank you – especially<br />

our gentle international readers – <strong>for</strong> staying with <strong>atw</strong> and we hope to<br />

meet you at the new international online plat<strong>for</strong>m ‘<strong>atw</strong> scientific’!

#52KT<br />

www.kerntechnik.com<br />

Media Partner<br />

52 nd KERNTECHNIK<br />

2022<br />

Get your ticket now!<br />

https://www.kerntechnik.com/kerntechnik-en/tickets/<br />

29 – 30 March 2022<br />

HYPERION Hotel, Leipzig<br />

Industry Meeting, Exhibition, Scientific Discourse,<br />

Young Scientists Workshop and Networking Plat<strong>for</strong>m<br />

What you can expect:<br />

p Plenary session with outstanding speakers<br />

discussing technical, economic and social challenges<br />

p Networking during Business Campus<br />

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research facilities and universitities<br />

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from the nuclear industry<br />

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Make sure you be there!

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