Nuclear Plant Journal - Digital Versions
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<strong>Nuclear</strong><br />
<strong>Plant</strong><br />
<strong>Journal</strong><br />
<strong>Plant</strong> Maintenance &<br />
Advanced Reactors Issue<br />
September-October 2008<br />
Volume 26 No. 5<br />
ISSN: 0892-2055<br />
Vermont Yankee, USA
KEY QUESTION FOR THE FUTURE<br />
How can I improve<br />
plant performance<br />
Look to AREVA NP for the global expertise to deliver<br />
a full spectrum of innovative, integrated solutions.<br />
For your peace of mind, we have the right resources to deliver the best value and quality engineering<br />
solutions. With U.S. market leadership and global resources, AREVA NP provides unmatched expertise<br />
for project execution and equipment reliability. With the opening of our BWR Center of Excellence<br />
in San Jose, we offer the most comprehensive engineering services in the industry to improve plant<br />
performance. Expect certainty. Count on AREVA NP. www.us.areva.com<br />
© Copyright 2008 AREVA NP Inc.
your trusted partner in mission<br />
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the power generation industry<br />
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More Experiences, More Resources and the Most Advanced Technology to Support <strong>Plant</strong> Inspections.<br />
Zetec, founded in 1968, is the leading supplier of nondestructive evaluation (NDE) inspection solutions<br />
based on integrated multi-method technologies - eddy current, ultrasonic, remote field, and magnetic<br />
flux leakage.<br />
Zetec, is your complete NDE testing solution: systems, instrumentation, software products, supplies,<br />
calibration, repair, training, and inspection services, all offered worldwide. In addition, our customers<br />
bring Zetec hundreds of new nondestructive testing challenges. Our accomplished team of industry<br />
and technical experts—application engineers, probe designers, machinists, and assemblers—are ready<br />
to meet those challenges.<br />
For more information on Zetec products or services, go to www.ZETEC.com
©2008 EDF Group<br />
AREVA EPR now under construction in France.<br />
Your Partner for <strong>Nuclear</strong> Power<br />
UniStar is charting a new course to America’s energy future with a<br />
fleet of AREVA’s advanced design U.S. EPR nuclear power plants.<br />
UniStar’s business model of flexible ownership and operations<br />
provides certainty of energy when and where you need it.<br />
To find out more about UniStar, call 410.470.4400 or visit<br />
www.unistarnuclear.com.<br />
For information on AREVA’s U.S. EPR, visit<br />
www.us.areva-np.com.<br />
For monthly photo updates of construction<br />
progress, send your e-mail address to<br />
info@unistarnuclear.com.
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong><br />
September-October 2008, Volume 26 No 5<br />
<strong>Plant</strong> Maintenance &<br />
Advanced Reactor Issue<br />
26th Year of Publication<br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong> is published by<br />
EQES, Inc.six times a year in February,<br />
April, June, August, October and December<br />
(Directory).<br />
The subscription rate for non-qualified<br />
readers in the United States is $150.00<br />
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is 0892-2055/02/$3.00+$.80.<br />
© Copyright 2008 by EQES, Inc.<br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong> is a registered<br />
trademark of EQES, Inc.<br />
Printed in the USA.<br />
Staff<br />
Senior Publisher and Editor<br />
Newal K. Agnihotri<br />
Publisher and Sales Manager<br />
Anu Agnihotri<br />
Editorial & Marketing Assistant<br />
Michelle Yong<br />
Administrative Assistant<br />
QingQing Zhu<br />
Articles & Reports<br />
Technologies of National Importance 16<br />
By Tsutomu Ohkubo, Japan Atomic Energy Agency, Japan<br />
Modeling & Simulation Advances Brighten Future <strong>Nuclear</strong> Power 18<br />
By Hussein Khalil, Argonne National Laboratory<br />
Energy & Desalination Projects 22<br />
By Ratan Kumar Sinha, Bhabha Atomic Research Centre, India<br />
A <strong>Plant</strong> with Simplified Design 24<br />
By John Higgins, GE Hitachi <strong>Nuclear</strong> Energy<br />
A Forward Thinking Design 27<br />
By Ray Ganthner, AREVA<br />
A Passively Safe Design 32<br />
By Ed Cummins, Westinghouse Electric Company<br />
A Market-Ready Design 34<br />
By Ken Petrunik, Atomic Energy of Canada Limited, Canada<br />
Generation IV Advanced <strong>Nuclear</strong> Energy Systems 42<br />
By Jacques Bouchard, French Commissariat a l'Energie Atomique, France<br />
and Ralph Bennett, Idaho National Laboratory<br />
Innovative Reactor Designs 46<br />
A Report by IAEA, Vienna, Austria<br />
Guidance For New Vendors 52<br />
By John Nakoski, U.S. <strong>Nuclear</strong> Regulatory Commission<br />
Road Map for Future Energy 54<br />
By John Cleveland, International Atomic Energy Agency, Vienna, Austria<br />
Vermont's Largest Source of Electricity 61<br />
By Tyler Lamberts, Entergy <strong>Nuclear</strong> Operations, Inc.<br />
Industry Innovations<br />
Intelligent Monitoring Technology 59<br />
By Chris Demars, Exelon <strong>Nuclear</strong><br />
Departments<br />
New Energy News 8<br />
Utility, Industry & Corporation 10<br />
New Products, Services & Contracts 12<br />
New Documents 14<br />
Meeting & Training Calendar 15<br />
<strong>Journal</strong> Services<br />
List of Advertisers 6<br />
Advertiser Web Directory 14<br />
On The Cover<br />
Vermont Yankee is a nuclear site located<br />
in Vermont. The plant is currently owned<br />
by Entergy <strong>Nuclear</strong> Vermont Yankee, LLC,<br />
and operated by Entergy’s nuclear business<br />
function. The unit is a boiling water<br />
reactor designed by General Electric Co.,<br />
and has a net generating capacity of 587<br />
dependable megawatts. See page 61 for<br />
a profi le.<br />
Mailing Identification Statement<br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong> (ISSN 0892-2055) is published bimonthly in February,<br />
April, June, August, October and December by EQES, Inc., 799 Roosevelt Road,<br />
Building 6, Suite 208, Glen Ellyn, IL 60137-5925. The <strong>Journal</strong> is available costfree<br />
to qualified readers worldwide. The subscription rate for non-qualified readers<br />
is $150.00 per year. The cost for non-qualified, non-U.S. readers is $180.00. Periodicals (permit<br />
number 000-739) postage paid at the Glen Ellyn, IL 60137 and additional mailing offices. POSTMAS-<br />
TER: Send address changes to <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong> (EQES, Inc.), 799 Roosevelt Road, Building 6,<br />
Suite 208, Glen Ellyn, IL 60137-5925.<br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 5
List of Advertisers & NPJ Rapid Response<br />
Page Advertiser Contact Fax/Email<br />
19 Atomic Energy of Canada Limited Heather Smith (905) 403-7565<br />
2 AREVA NP, Inc. Donna Gaddy-Bowen (434) 832-3840<br />
31 Babcock & Wilcox Canada Ltd Yvette Amor (519) 621-9681<br />
21 Bechtel Power www.bechtel.com<br />
45 Bigge Power Constructors Andrew Wierda (510) 639-4053<br />
37 Black & Veatch Keith Gusich (913) 458-2491<br />
15 Ceradyne Patti Bass (714) 675-6565<br />
41 Climax Portable Machine Tools, Inc. Debra Horn dhorn@cpmt.com<br />
47 Data Systems & Solutions Romain Desgeorge 33 (0) 4 76 61 17 07<br />
49 Day & Zimmermann NPS David Bronczyk (215) 299-8395<br />
29 Enertech Tom Schell tschell@curtisswright.com<br />
7 GE Hitachi <strong>Nuclear</strong> Energy Mark Marano (910) 362-5017<br />
25 HSB Global Standards Louise Hamburger louise_hamburger@hsbct.com<br />
38 Meggitt Safety Systems Jennifer Cetta (805) 584-9157<br />
43 National Enrichment Facility Dana Starr (575) 394-0175<br />
11 NPTS, Inc. Rebecca Broman (716) 876-8004<br />
55 <strong>Nuclear</strong> Logistics Inc. Craig Irish (978) 250-0245<br />
43 Power House Tool, Inc. Laura Patterson (815) 727-4835<br />
8 Proto-Power Corporation Christopher D’Angelo (860) 446-8292<br />
39 The Shaw Group Inc. Holly Nava (856) 482-3155<br />
51 Thermo Fisher Scientific Tony Chapman (315) 451-9421<br />
64 Trentec, Inc. Arlene Corkhill (714) 528-0128<br />
13 Underwater Construction Charles Vallance (321) 779-4462<br />
4 UniStar <strong>Nuclear</strong> Energy Mary Klett (410)470-5606<br />
9 UniTech Services Group Steve Hofstatter (413) 543-2975<br />
35 Urenco Enrichment Company Ltd Please e-mail enquiries@urenco.com<br />
26 Westerman Companies Jim Christian (740) 569-4111<br />
63 Westinghouse Electric Company LLC Karen Fischetti (412) 374-3244<br />
17 WM Symposia, Inc. Mary E. Young mary@wmarizona.org<br />
3 Zetec, Inc. Katina Baarslag (425) 974-2678<br />
Information may be directly obtained from advertisers by faxing this page to the individual advertiser after completing<br />
the bottom part of the Rapid Response Fax Form. Advertisers’ web sites are listed in the Web Directory Listings<br />
on page 14.<br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong> Rapid Response Fax Form<br />
From the September-October 2008<br />
issue of <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong><br />
To: _________________________ Company: __________________ Fax: ___________________<br />
From: _______________________ Company: __________________ Fax: ___________________<br />
Address:_____________________ City: _______________________ State: _____ Zip: _________<br />
Phone: ______________________ E-mail: _____________________<br />
I am interested in obtaining information on: __________________________________________________<br />
Comments: _____________________________________________________________________________<br />
6 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
GE Hitachi<br />
<strong>Nuclear</strong> Energy<br />
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plans on track to succeed Choosing a<br />
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decades, GE Hitachi <strong>Nuclear</strong> Energy (GEH) has<br />
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GEH’s commitment and leadership can make<br />
your plans a reality. To learn more visit<br />
ge.com/nuclear
New Energy News<br />
COLA<br />
AmerenUE, a Missouri-based utility<br />
subsidiary of Ameren Corporation submitted<br />
a combined Construction and Operating<br />
License Application (COLA) to<br />
the U.S. <strong>Nuclear</strong> Regulatory Commission<br />
(NRC) for a potential new nuclear power<br />
plant in Callaway County, Missouri.<br />
The 8,000-page application seeks<br />
regulatory approvals to potentially build<br />
a new 1,600-megawatt pressurized water<br />
reactor adjacent to AmerenUE’s singleunit,<br />
1,190-megawatt Callaway electric<br />
generating plant which accounts for 19<br />
percent of the company’s total generation.<br />
Since the Callaway <strong>Plant</strong> came on line<br />
in December 1984, it has achieved the<br />
fourth highest generation output among<br />
the nation’s 104 nuclear power units.<br />
Contact: Mike Cleary, telephone: (573)<br />
681-7137, email: mcleary@ameren.com.<br />
Strategy Report<br />
The U.S. Department of Energy<br />
(DOE) and the U.S. <strong>Nuclear</strong> Regulatory<br />
Commission (NRC) delivered to<br />
Congress the Next Generation <strong>Nuclear</strong><br />
<strong>Plant</strong> (NGNP) Licensing Strategy Report<br />
which describes the licensing approach,<br />
the analytical tools, the research and<br />
development activities and the estimated<br />
resources required to license an advanced<br />
reactor design by 2017 and begin operation<br />
by 2021. The NGNP represents a new<br />
concept for nuclear energy utilization,<br />
in which a gas-cooled reactor provides<br />
process heat for any number of industrial<br />
applications including electricity<br />
production, hydrogen production, coalto-liquids,<br />
shale oil recovery, fertilizer<br />
production, and other applications that<br />
meet significant industrial needs.<br />
Visit <strong>Nuclear</strong>.gov to read the joint<br />
Licensing Strategy Report and to learn<br />
more about DOE’s Office of <strong>Nuclear</strong><br />
Energy.<br />
Contact: Angela Hill, telephone:<br />
(202) 586-4940.<br />
Loan Guarantee<br />
Dominion Virginia Power submitted<br />
to the U.S. Department of Energy<br />
the first part of an application for a loan<br />
guarantee as it considers a third nuclear<br />
reactor at the North Anna Power Station<br />
in Central Virginia.<br />
“Today’s filing is another important<br />
step in the process began more than seven<br />
years ago to position ourselves to be<br />
among the first to get a license for a new<br />
nuclear unit,” said Mark F. McGettrick,<br />
president and chief executive officer of<br />
Dominion Generation.<br />
Contact: Richard Zuercher, telephone: (804)<br />
273-3825, email: Richard.Zuercher@Dom.com.<br />
Proto-Power has been serving the nuclear community for decades<br />
so we understand the critical role this resource plays in our future.<br />
Our mission is to<br />
provide the highest<br />
quality engineering, design and project management in the<br />
industry. With experienced account managers focusing on a single<br />
client, in-depth resources throughout our organization, and the<br />
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operations and offer guidance for tomorrow’s challenges.<br />
SOLUTIONS<br />
for Our Energy Future<br />
Proto-Power. Vision for the nuclear future.<br />
PROTO-POWER CORPORATION<br />
a Zachry Group Company<br />
Groton, CT • 860.446.9725<br />
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DELIVERING ENGINEERING SOLUTIONS TO THE NUCLEAR POWER INDUSTRY<br />
Circle 107 on Reader Service Form<br />
8 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
Construction Agreement<br />
After the signing of a framework<br />
agreement, November 26, 2007 in Beijing,<br />
in the presence of both head of States of<br />
France and China, EDF and the Chinese<br />
electricity producer China Guangdong<br />
<strong>Nuclear</strong> Power Holding Company signed<br />
the final agreements in Beijing for the<br />
creation of a joint venture company to<br />
be called Guangdong Taishan <strong>Nuclear</strong><br />
Power Joint Venture Company Limited<br />
(TNPC). The aim of the joint venture is<br />
to construct and operate two nuclear EPR<br />
power stations at Taishan in the province<br />
of Guangdong, modeled on the existing<br />
EPR reactor built by EDF at Flamanville<br />
in Normandy, France.<br />
Preliminary work at the Taishan Unit 1<br />
site started in late 2007 and the first concrete<br />
pouring is scheduled for autumn<br />
2009, less than two years after the one<br />
at Flamanville 3. Some contracts have<br />
already been signed with Areva and Alstom<br />
for the supply of the nuclear and the<br />
turbine equipment respectively. The first<br />
unit should be commissioned at the end<br />
of 2013 and the second in 2015. At the<br />
height of construction work, over sixty<br />
EDF experts will be on-site at Taishan.<br />
Contact: Carole Trivi, telephone: 33<br />
1 40 42 44 19.<br />
123 Agreement<br />
Statement by the Prime Minister of India<br />
"We welcome the decision earlier<br />
today of the <strong>Nuclear</strong> Suppliers Group to<br />
adjust its guidelines to enable full civil<br />
nuclear cooperation with India. This is a<br />
forward-looking and momentous decision.<br />
It marks the end of India's decades long<br />
isolation from the nuclear mainstream<br />
and of the technology denial regime. It is<br />
a recognition of India's impeccable nonproliferation<br />
credentials and its status as<br />
a state with advanced nuclear technology.<br />
It will give an impetus to India's pursuit<br />
of environmentally sustainable economic<br />
growth."<br />
Contact: telephone: 43 1 2600-0,<br />
fax: 43 1 2600-7.<br />
Application Submitted<br />
Progress Energy Florida, a subsidiary<br />
of Progress Energy submitted a combined<br />
license (COL) application with the<br />
<strong>Nuclear</strong> Regulatory Commission (NRC)<br />
to construct a new nuclear power plant in<br />
Levy County, Florida.<br />
The application, submitted to the<br />
NRC on July 30, 2008, included the request<br />
to build two Westinghouse AP1000<br />
nuclear reactors at the site. <strong>Nuclear</strong> power<br />
is a key component of Progress Energy<br />
Florida’s balanced solution strategy to<br />
meet Florida’s long-term energy needs.<br />
<strong>Nuclear</strong> power, along with additional renewable<br />
energy resources and expanded<br />
energy-efficiency programs, is Progress<br />
Energy Florida’s strategy to address climate<br />
change and the need for greater fuel<br />
diversity.<br />
Contact: telephone: (919) 546-6189.<br />
Your<br />
Map To<br />
Offsite<br />
Metal<br />
Decon<br />
CO2, Plastic Bead, Ultrasonic, High Pressure Water, Steam<br />
Excavation Started<br />
The Shandong <strong>Nuclear</strong> Power Company<br />
with Westinghouse Electric Company<br />
LLC and its consortium partner<br />
The Shaw Group Inc. broke ground one<br />
month earlier than scheduled on the Haiyang<br />
<strong>Nuclear</strong> Power Facility in Shandong<br />
Province.<br />
The Haiyang facility will house two<br />
nuclear plants, each deploying Westinghouse’s<br />
AP1000 technology. Excavation<br />
for the first of the two plants will take approximately<br />
three months to create a hole<br />
12 meters deep (39 feet) that will house<br />
the nuclear reactor and turbine buildings.<br />
The volume of the excavation is approximately<br />
48,916 cubic meters or about 19.5<br />
Olympic-size swimming pools. When<br />
completed, a base for the plant nearly<br />
175 meters wide (570 feet) by 250 meters<br />
long (840 feet) will exist.<br />
Contact: Vaughn Gilbert, telephone:<br />
(412) 374-3896, email:<br />
gilberthv@westinghouse.com. <br />
Outage management of customer equipment; long term storage<br />
HEPA ventilation; tools and scaffolding; HP instruments<br />
Transport<br />
Services<br />
11 Licensed<br />
Facilities<br />
www.<br />
NPJOnline.<br />
com<br />
(800) 344-3824<br />
www.unitech.ws<br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 9
Utility, Industry & Corporation<br />
Utility<br />
Utility Achievement<br />
Award<br />
Constellation Energy announced<br />
that its Calvert Cliffs <strong>Nuclear</strong> Power<br />
<strong>Plant</strong> (CCNPP) has been awarded the<br />
American <strong>Nuclear</strong> Society’s 2008<br />
Utility Achievement Award for sustained<br />
outstanding performance. Jim Spina, vice<br />
president at CCNPP, accepted the award on<br />
behalf of the approximately 780 employees<br />
at Calvert Cliffs at a conference hosted by<br />
the American <strong>Nuclear</strong> Society in Amelia<br />
Island, Florida.<br />
Calvert Cliffs was recognized for<br />
demonstrating a prolonged dedication to<br />
safe nuclear generation as evidenced by a<br />
record high capacity factor and the highest<br />
site generation in four of the last five<br />
years.<br />
Contact: Dave Fitz, telephone: (888)<br />
232-1919<br />
Reader Service Card & Cost-free<br />
Subscription Cards<br />
1. The reader service inquiries are now available<br />
online by logging on to requestinfo.NPJOnline.com<br />
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the web site subscribe.NPJOnline.com to request or<br />
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version without any charge. Additional subscription<br />
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international readers other than US & Canada .<br />
Contact NPJ@goinfo.com for details.<br />
License Renewal<br />
Dominion, owner of the Kewaunee<br />
Power Station, filed an application to renew<br />
the facility’s operating license with the U.S.<br />
<strong>Nuclear</strong> Regulatory Commission (NRC).<br />
The 568-megawatt nuclear unit is<br />
licensed to operate through December 21,<br />
2013. With a renewed license, the station<br />
would be able to provide Wisconsin with<br />
safe, clean and reliable electricity through<br />
December 21, 2033.<br />
Contact: Mark Kanz, telephone: (920)<br />
388-8198.<br />
Joint Effort<br />
Entergy <strong>Nuclear</strong> and the Taiwan<br />
Power Company announced a joint effort<br />
tapping Entergy’s experience in license<br />
renewal efforts to allow for long term<br />
operations at Taiwan’s Kuosheng <strong>Nuclear</strong><br />
Power <strong>Plant</strong>.<br />
The Institute of <strong>Nuclear</strong> Energy Research,<br />
Taiwan, which advances nuclear<br />
technology and assures national nuclear<br />
<strong>Nuclear</strong><br />
<strong>Plant</strong><br />
<strong>Journal</strong><br />
An International Publication<br />
A <strong>Digital</strong> (electronic) Version of NPJ<br />
is Now Available!<br />
safety, has been contracted by the Taiwan<br />
Power Company to initiate a project<br />
to allow for extended operation of TPC’s<br />
Kuosheng plant, a dual unit site with boiling<br />
water reactors constructed in the early<br />
1980s.<br />
Contact: Mike Bowling, telephone:<br />
(601) 368-5655, email: mbowling@entergy.com.<br />
New Website<br />
Exelon <strong>Nuclear</strong> announced the launch<br />
of a new Texas-based Web site intended<br />
to keep the public updated and informed<br />
about the company’s proposed Victoria<br />
County nuclear plant. The Web site address<br />
is www.Exelon<strong>Nuclear</strong>Texas.com.<br />
Contact: Bill Harris, telephone: (361)<br />
578-2705.<br />
Industry<br />
ITER<br />
Commissariat français à l’énergie<br />
atomique, Cadarache, France, is the host<br />
for the ITER project constructing an<br />
experimental nuclear fusion reactor using<br />
hydrogen isotopes.<br />
The site was selected in 2005 by the<br />
international partners of the ITER project<br />
(India, China, South Korea, Japan, Russia,<br />
the United States and the European Union).<br />
It will take 10 years to build the project and<br />
a further 20 years of scientific experiments<br />
to prove that fusion can become a new<br />
reliable source of energy.<br />
Contact: Benoit Gausseron, telephone:<br />
(212) 757-9340, email:<br />
benoit.gausseron@investinfrance.org.<br />
Award for Technology<br />
The U.S. Department of Energy<br />
(DOE) awarded up to $15 million to<br />
34 research organizations as part of the<br />
Department’s Advanced Fuel Cycle<br />
Initiative (AFCI).<br />
For a list of recipients please go to,<br />
http://nuclear.gov/newsroom/2008PRs/<br />
AwardedProjects.pdf.<br />
Contact: Angela Hill, telephone: (202)<br />
586-4940.<br />
10 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
Corporation<br />
Purchase<br />
The Babcock & Wilcox Company<br />
(B&W), a subsidiary of McDermott<br />
International, Inc. announced that an<br />
affiliate of B&W has entered into a<br />
definitive agreement to acquire <strong>Nuclear</strong><br />
Fuel Services, Inc. (NFS) of Erwin, Tenn.,<br />
a provider of specialty nuclear fuels and<br />
related services. The acquisition supports<br />
B&W’s strategic goal of being a leading<br />
provider of nuclear manufacturing and<br />
service businesses for government and<br />
commercial markets.<br />
Contact: Steve Stultz, telephone: (330)<br />
860-6124, email: sstultz@babcock.com.<br />
New Facility<br />
Day & Zimmermann, announced<br />
its Maintenance and Modification unit has<br />
completed the acquisition of a fabrication<br />
and machining facility in Moss Point,<br />
Mississippi, from Industrial Maintenance<br />
and Machine, Inc. The facility will be<br />
operated by DZ Atlantic, a wholly owned<br />
subsidiary of Day & Zimmermann.<br />
“Having a fabrication facility will allow<br />
us to meet the needs of our existing<br />
customers and further develop customer<br />
relationships in other targeted industries,”<br />
said Mike McMahon, President of Day &<br />
Zimmermann’s Maintenance and Modification<br />
operation.<br />
The facility consists of a 180,000-sq.-<br />
ft. shop situated on 20 acres of land, and<br />
will give DZ Atlantic a significant range of<br />
capabilities including machining, mobile<br />
machining, structural steel production,<br />
piping, skids, and specialty welding.<br />
Contact: Maureen Omrod, telephone:<br />
(215) 299-2234, email:<br />
Maureen.Omrod@DayZim.com.<br />
Acquisition<br />
ENERCON, a 700-employee firm<br />
serving energy and environmental clients<br />
nationwide, has acquired EPIC Consulting,<br />
Inc. of Marietta, Georgia, an environmental<br />
and geotechnical firm specializing in highly<br />
customized solutions primarily in energyand<br />
environmentally-related businesses.<br />
ENERCON Vice President John Corn<br />
said, “EPIC is an excellent fit for ENER-<br />
CON and will complement our environmental<br />
and technical services. EPIC and<br />
ENERCON’s environmental division provide<br />
similar services but their geotechnical<br />
expertise is a great addition to our<br />
services.<br />
Contact: Peggy Striegel, telephone:<br />
(918) 740-5584, email: peggy@striegela.com.<br />
Platform for Future<br />
Growth<br />
Numet Engineering Ltd., a supplier<br />
of specialized, high-reliability precision<br />
engineered systems and equipment for<br />
the nuclear energy & hazardous waste<br />
management sectors has been acquired<br />
by the ODIM Group. The company will<br />
continue to operate as Numet Engineering<br />
Ltd. And will continue to exclusively<br />
focus towards the nuclear power industry.<br />
For the ODIM Group, the Numet<br />
acquisition brings a strong and well respected<br />
presence to the Canadian nuclear<br />
power sector.<br />
Contact: Bill Potter, telephone: (705)<br />
743-2708, email: bill.potter@numet.com.<br />
Software<br />
Exelon Corporation, recently selected<br />
Scientech’s award-winning PMAX<br />
software as its tool for on-line thermal<br />
performance monitoring at its 17 nuclear<br />
generating units. PMAX is renowned for<br />
its ability as a software tool to assist engineers<br />
and operators to identify megawatt<br />
losses and reveal plant thermal performance<br />
inefficiencies – in a real-time environment.<br />
PMAX has been adopted worldwide<br />
by over 300 thermal power plants (nuclear,<br />
fossil, and combined cycle), and is now<br />
the on-line thermal performance tool of<br />
choice at 59 of the nation’s 104 nuclear<br />
power plant units.<br />
Contact: Ed Hollis, telephone: (301)<br />
371-7485, email: ehollis@curtisswright.com.<br />
Module, Construction<br />
Westinghouse Electric Company<br />
and The Shaw Group Inc. signed a letter<br />
of intent (LOI) to form a joint venture<br />
to fabricate and assemble structural and<br />
equipment modules for AP1000 nuclear<br />
power plants to be built in the United<br />
States and selected global markets in<br />
which in-country supply is not available.<br />
Under terms of the LOI, Westinghouse<br />
and Shaw will each hold ownership shares<br />
in the joint venture.<br />
The new company, Global Modular<br />
Solutions LLC, will construct a 600,000<br />
sq. ft. facility in Lake Charles, Louisiana<br />
that is scheduled to begin operation in<br />
the late summer of 2009. When fully<br />
operational, the facility is expected to<br />
employ as many as 1,400 workers.<br />
Contact: Vaughn Gilbert, telephone: (412)<br />
374-3896, email: gilberhv@westinghouse.com.<br />
Strategy & Research<br />
Dr. Kathryn Jackson has been<br />
appointed to the position of vice president,<br />
Strategy, Research and Technology at<br />
Westinghouse Electric Company. Dr.<br />
Jackson was previously the executive vice<br />
president of River System Operations and<br />
Environment at Tennessee Valley Authority<br />
(TVA), where she has served since 1998.<br />
She holds a master’s in Industrial<br />
Engineering Management from the<br />
University of Pittsburgh (1983) and master’s<br />
and doctorate degrees in Engineering<br />
and Public Policy from Carnegie Mellon<br />
University (1987 and 1990).<br />
Contact: Vaughn Gilbert, telephone:<br />
(412) 374-3896, email:<br />
gilberhv@westinghouse.com.<br />
<br />
NPTS, Inc.<br />
an Engineering, Design, and<br />
Construction Management firm has<br />
current and anticipated openings for the<br />
following positions:<br />
Licensing, USAR & Regulatory<br />
•<br />
Engineers<br />
Engineering Design (All Disciplines)<br />
•<br />
Sr. Project Managers (All<br />
•<br />
Disciplines)<br />
Sr. Project Planners (All Disciplines)<br />
•<br />
Power Upgrade Project Engineers<br />
•<br />
Construction Management, Planners,<br />
•<br />
Schedulers, Estimators<br />
• Resident Engineers (All Disciplines)<br />
• Operations Support Engineers<br />
• Operations Training Instructors<br />
• Procurement Specialists &<br />
Expeditors<br />
• Start-up & Commissioning<br />
Engineers<br />
For Power Uprates, New Builds, Life<br />
Extension, Upgrades, Modification<br />
and Maintenance Projects<br />
Please forward Resumes to:<br />
NPTS, Inc.<br />
2060 Sheridan Drive<br />
Buffalo, New York 14221<br />
Phone: 716.876.8066<br />
Fax: 716.876.8004<br />
E-mail: rbroman@npts.net<br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 11
New Products, Services & Contracts<br />
New Products<br />
Ultrasonic Flaw<br />
Detectors<br />
GE Sensing & Inspection Technologies<br />
introduces a new family of ultrasonic<br />
flaw detectors, providing inspectors with<br />
a flexible platform, as inspection needs<br />
change. The Phasor family incorporates<br />
conventional and phased array ultrasound<br />
technology in three upgradeable models:<br />
Phasor CV, Phasor 16/16 Weld and Phasor<br />
XS. The tiered platform offers inspectors<br />
the opportunity to select the model<br />
that best suits their specific application in<br />
oil & gas, power generation, aerospace or<br />
transportation.<br />
Contact: Amanda Fontaine,<br />
email: Amanda.fontaine4@ge.com.<br />
Walking Robot<br />
Zetec, Inc., the total solution<br />
nondestructive testing (NDT) provider<br />
for the Power Generation industry,<br />
announced it will launch the industry’s<br />
most flexible and functional tube sheet<br />
walking robot at the 27th Steam Generator<br />
NDE Workshop.<br />
Small in size and weighing less than<br />
35 lbs, the ZR-100 provides ultimate<br />
flexibility in reaching all of the tubes<br />
within the tube sheet without complex<br />
repositioning motions. This provides<br />
quick and efficient motion in positioning<br />
the ZR-100 to a target zone or specific<br />
tube. All of this is accomplished while<br />
providing industry leading speed.<br />
The ZR-100 can transverse across the<br />
tube sheet at speeds of up to 5 feet per<br />
minute for large moves and can achieve<br />
tube-to-tube speeds during test or repair<br />
operations of up to 4 inches/second. The<br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>’s<br />
Product & Service Directory 2009<br />
2009 Directory<br />
All nuclear power industry suppliers who are not listed<br />
in the 2008 Directory may register for the 2009 Directory<br />
by sending an email to npj@goinfo.com with complete<br />
contact information.<br />
Suppliers listed in <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>'s 2008<br />
Directory will receive the 2009 Directory mailing<br />
with a list of their products and services as they<br />
appeared in the 2008 Directory.<br />
Deadlines:<br />
Input Form- November 12, 2008<br />
Ad Committment- November 12, 2008<br />
Contact:<br />
Email: npj@goinfo.com<br />
Telephone: 630-858-6161, ext. 103<br />
FAx: 630-858-8787<br />
<strong>Nuclear</strong><br />
<strong>Plant</strong><br />
<strong>Journal</strong><br />
An International Publication<br />
Product & Service Directory 2009<br />
ZR-100 utilizes built-in Machine Vision<br />
for secondary tube verification for all<br />
attached tooling.<br />
Contact: Katina Baarslag, telephone:<br />
(425) 974-2678, email: KBaarslag@zetec.com.<br />
Services<br />
Inspection Time<br />
Reduced<br />
Toshiba GE Turbine Components<br />
(TGTC) has reduced the time required to<br />
inspect and measure steam turbine blades<br />
from 280 minutes to 45 minutes by using<br />
the MAXOS non-contact measurement<br />
system from Steintek GmbH (Greding,<br />
Germany). The coordinate measuring<br />
machine (CMM) used in the past to<br />
inspect the blades was not only slow but<br />
was unable to access hard-to-reach areas<br />
such as dovetail hooks and fillets. The<br />
MAXOS uses five axes to reach every<br />
point on the blades and also generates<br />
specific and accurate measurements of<br />
critical areas. Resulting measurements<br />
are reported instantly and the need<br />
for additional manual inspection is<br />
eliminated.<br />
“The MAXOS optical scanner provides<br />
the best possible accuracy, eliminates<br />
the need for matt coating, and<br />
integrates easily with our engineering<br />
and production processes,” said Tomio<br />
Kubota, President of TGTC. “Our trials<br />
also demonstrated that the MAXOS is<br />
significantly faster than the other systems<br />
that we considered. The expertise and<br />
professionalism that were evident during<br />
this trial gave us the confidence to adopt<br />
this new technology.”<br />
Contact: NVision Inc (Southlake,<br />
TX and Wixom, MI), telephone: (248)<br />
268-2525, email: sales@nvision3d.com.<br />
ASME Renewal<br />
Certificates<br />
TechPrecision Corporation, a<br />
manufacturer of large-scale, highprecision<br />
machined metal fabrications for<br />
the alternative energy, medical, nuclear,<br />
12 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
defense, aerospace and other commercial<br />
industries, announced that its wholly<br />
owned subsidiary, Ranor, Inc. received<br />
its renewal Certificates of Authorization<br />
from the American Society of Mechanical<br />
Engineers (“ASME”). The Certificates<br />
of Authorization cover the Company’s<br />
facilities in Westminster, Massachusetts<br />
and are an integral part of Ranor’s<br />
ongoing business plan to be a supplier to<br />
the emerging nuclear renaissance.<br />
Contact: Amanda Lleshdedaj,<br />
telephone: (310) 477-9800, email:<br />
Amanda.lleshdedaj@ccgir.com.<br />
Contracts<br />
Turbine Island<br />
Alstom signed a contract worth over<br />
200 million euros with China Guang Dong<br />
<strong>Nuclear</strong> Power Company (CGNPC) for<br />
the engineering and procurement of the<br />
complete turbine island for the nuclear<br />
power plant to be built in Taishan (southwestern<br />
province of Guangdong). Taishan<br />
will be China’s first EPR power plant.<br />
This contract follows the $300<br />
million order (including around $100<br />
million for Alstom) booked in February<br />
2008 and won in partnership with the<br />
Chinese industrial group and Alstom’s<br />
long-standing partner, Dongfang Electric<br />
Company. This first order is for the supply<br />
of two 1,750 MW Arabelle turbinegenerator<br />
packages for the Taishan<br />
nuclear plant.<br />
Contact: Philippe Kasse, telephone:<br />
33 1 41 49 29 82/33 08, email:<br />
philippe.kasse@chq.alstom.com.<br />
<strong>Nuclear</strong> Fuel Assemblies<br />
AREVA has signed a contract with<br />
Taiwan Power Company (Taipower) to<br />
supply boiling water reactor fuel assemblies<br />
for units 1 and 2 of the Chinshan<br />
and Kuosheng nuclear power plants. The<br />
award, worth more than $200 million,<br />
is the conclusion of an invitation to bid<br />
launched in June 2007.<br />
The scope of work includes five<br />
firm reload batches and three optional<br />
reload batches for each unit. AREVA<br />
will provide core monitoring system<br />
assistance in addition to the fabrication<br />
service, reload fuel design, licensing<br />
analysis and operation support.<br />
Contact: Laurence Pernot, telephone:<br />
(301) 841-1694, email: Laurence.<br />
pernot@areva.com.<br />
Project Contract<br />
SNC-Lavalin <strong>Nuclear</strong> and Murray<br />
& Roberts announced that Pebble Bed<br />
Modular Reactor (Pty) Ltd has awarded<br />
their joint venture company, Murray &<br />
Roberts SNC-Lavalin <strong>Nuclear</strong> (Pty) Ltd.<br />
(MRSLN), a contract to provide engineering,<br />
procurement, project and construction<br />
management services for Phase<br />
II of the Pebble Bed Modular Reactor<br />
(PBMR) Demonstration Power <strong>Plant</strong> at<br />
Koeberg, South Africa.<br />
Phase II of the project entails<br />
construction of a commercial scale power<br />
plant at Koeberg near Cape Town, which<br />
is subject to obtaining a nuclear licence<br />
from the National <strong>Nuclear</strong> Regulator<br />
and a positive Record of Decision on the<br />
Environmental Impact Assessment.<br />
Contact: Gillian MacCormack,<br />
telephone: (514) 393-8000 ext. 7354.<br />
Steam Generator<br />
Treatment<br />
Studsvik has received an order for<br />
the treatment and metal recycling of<br />
three steam generators. The customer<br />
is Vattenfall Ringhals in Scandinavia,<br />
and the order is received under the<br />
existing Memorandum of Understanding<br />
concerning treatment of large components<br />
signed in 2006. The steam generators are<br />
planned to be delivered to Studsvik during<br />
fall 2008 and the treatment is planned to<br />
start during the first quarter 2009. The<br />
contract value is SEK 34 million.<br />
Contact: Magnus Groth, telephone:<br />
46 155 22 10 86. <br />
www.<br />
radiation<br />
training.com<br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 13
New Documents<br />
EPRI<br />
1. BWR Vessel and Internals Project,<br />
Evaluation of RAMA Fluence<br />
Methodology Calculational<br />
Uncertainty, Product ID: 1016938,<br />
Published July 2008.<br />
This report documents the overall<br />
calculational uncertainty associated with<br />
the application of the Radiation Application<br />
Modeling Application (RAMA) Fluence<br />
Methodology to BWR reactor pressure<br />
vessel fluence evaluations.<br />
2. Feasibility of Direct Disposal of<br />
Dual-Purpose Canisters in a High-<br />
Level Waste Repository, Product ID:<br />
1018051, Published August 2008.<br />
A deep geologic repository at Yucca<br />
Mountain, Nevada, has been proposed for<br />
the disposal of commercial spent nuclear<br />
fuel (CSNF) and other nuclear fuel and<br />
high level radioactive waste (HLW) from<br />
defense and nuclear weapons programs.<br />
Atomic Energy of<br />
Canada Limited<br />
www.aecl.ca<br />
AREVA NP, Inc.<br />
www.us.areva.com<br />
Babcock & Wilcox<br />
Canada Ltd.<br />
www.babcock.com/bwc<br />
Bechtel Power<br />
www.bechtel.com<br />
Bigge Power Constructors<br />
www.bigge.com<br />
Black & Veatch<br />
www.bv.com<br />
Ceradyne<br />
www.ceradyne.com<br />
Climax Portable<br />
Machine Tools, Inc.<br />
www.cpmt.com<br />
Data Systems & Solutions<br />
www.ds-s.com<br />
The U.S. Department of Energy<br />
(DOE) has proposed a standardized<br />
transportation, aging and disposal (TAD)<br />
canister for emplacement of CSNF at<br />
Yucca Mountain.<br />
3. Study to Identify Potential<br />
Improvements of Operation<br />
Tools and Support Systems–Non-<br />
Proprietary, Product ID: 1016730,<br />
Published August 2008.<br />
This project analyzed safety<br />
significant events (SSEs) in several<br />
nuclear power plants to identify where<br />
improvements in instrumentation and<br />
control (I&C) and information technology<br />
(IT) could prevent or mitigate some<br />
of these events. This report identifies<br />
potential improvement paths that could<br />
enhance reliability and availability for<br />
implementation consideration by utilities<br />
where appropriate at their own plants.<br />
4. Program on Technology Innovation:<br />
Using Information Technology<br />
to Increase <strong>Nuclear</strong> Power <strong>Plant</strong><br />
Performance, Product ID: 1016962,<br />
Published August 2008.<br />
As current nuclear power plants<br />
(NPPs) continue to operate for the next<br />
20–30 years, certain issues are driving the<br />
plants to come up with new ways of doing<br />
work. Solutions to these issues may<br />
be possible using modern information<br />
technology (IT). This can include the use<br />
of both software and hardware and can<br />
encompass traditional corporate IT systems<br />
as well as plant instrumentation and<br />
control (I&C) systems.<br />
The above document may be obtained<br />
from EPRI Order and Conference Center,<br />
1300 West WT Harris Blvd., Charlotte,<br />
NC 28262; telephone: (800) 313-3774,<br />
email: orders@epri.com.<br />
<br />
NPJ Advertiser Web Directory<br />
Day & Zimmermann NPS Thermo Fisher Scientific<br />
www.dznps.com<br />
www.thermo.com/cidtec<br />
Enertech<br />
www.enertechnuclear.com<br />
GE Hitachi <strong>Nuclear</strong> Energy<br />
www.ge.com/nuclear<br />
HSB Global Standards<br />
www.hsbgsnuclear.com<br />
Meggitt Safety Systems<br />
www.meggittsafety.com<br />
National Enrichment Facility<br />
www.nefnm.com<br />
NPTS, Inc.<br />
www.npts.net<br />
<strong>Nuclear</strong> Logistics Inc.<br />
www.nuclearlogistics.com<br />
Power House Tool, Inc.<br />
www.powerhousetool.com<br />
Proto-Power Corporation<br />
www.protopower.com<br />
The Shaw Group Inc.<br />
www.shawgrp.com<br />
Trentec, Inc.<br />
www.trentec.com<br />
Underwater Construction<br />
www.uccdive.com<br />
UniStar <strong>Nuclear</strong> Energy<br />
www.unistarnuclear.com<br />
UniTech Services Group<br />
www.unitech.ws<br />
Urenco Enrichment Company<br />
Ltd.<br />
www.urenco.com<br />
Westerman Companies<br />
www.westermancompanies.com<br />
Westinghouse Electric<br />
Company LLC<br />
www.westinghousenuclear.com<br />
WM Symposia, Inc.<br />
www.wmsym.org<br />
Zetec, Inc.<br />
www.zetec.com<br />
14 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
Meeting & Training Calendar<br />
1. NEI International Uranium Fuel<br />
Seminar, October 19-22, 2008, The<br />
Westin Tarbor Center, Denver Colorado.<br />
Contact: <strong>Nuclear</strong> Energy Institute,<br />
Janet Schluester, telephone:<br />
(202) 739-8098, email: jrs@nei.org.<br />
2. Technical Meeting on the International<br />
Decommissioning Network,<br />
October 20-24, 2008, Vienna, Austria.<br />
Contact: International Atomic<br />
Energy Agency, P. Dinner, email:<br />
P.Dinner@iaea.org.<br />
th<br />
3. EPRI 7 International Decommissioning<br />
& Radioactive Waste Workshop,<br />
October 28-October 30, 2008,<br />
Hotel Hilton Lyon, Lyon, France.<br />
Contact: Electric Power Research<br />
Institute, Sean Bushart, telephone:<br />
(650) 855-2978, email: Sbushart@epri.com.<br />
4. 23rd Canadian <strong>Nuclear</strong> Society<br />
<strong>Nuclear</strong> Simulation Symposium,<br />
November 2-4, 2008, Ottawa,<br />
Ontario, Canada. Contact: Denise<br />
Rouben, CNS, telephone: (416) 977-<br />
7620, email: cns-snc@on.aibn.com.<br />
5. Future Power, November 4-5, 2008,<br />
London. Contact: <strong>Nuclear</strong> Engineering<br />
International, telephone:<br />
44 0 208 2697 812, website: www.<br />
neimagazine.com/futurepower.<br />
6. Winter Meeting and <strong>Nuclear</strong> Technology<br />
Expo, November 9-13, 2008,<br />
Reno, Nevada. Contact: American<br />
<strong>Nuclear</strong> Society, telephone: (708)<br />
579-8316).<br />
7. Technical Meeting to Maintain and<br />
Update the <strong>Nuclear</strong> Fuel Cycle Information<br />
System, November 12-14,<br />
2008, Vienna, Austria. Contact: International<br />
Atomic Energy Agency,<br />
M. Ceyhan, email: M.Ceyhan@iaea.org.<br />
th<br />
8. 8 International Conference on<br />
CANDU Maintenance, November 16-<br />
18, 2008, Metro Toronto Convention<br />
Centre and InterContinental Toronto<br />
Centre Hotel, Toronto, Ontario.<br />
Contact: Denise Rouben, CNS,<br />
telephone: (416-977-7620, email:<br />
cns-snc@on.aibn.com.<br />
9. November 17-20, 2008, Las Vegas,<br />
Nevada. Contact: Argonne National<br />
Laboratory, Lawrence Boing,<br />
telephone: (630) 252-6729, email:<br />
lboing@anl.gov.<br />
th<br />
10. 46 Semiannual <strong>Nuclear</strong> Fuel Management<br />
Seminar, November 17-20,<br />
2008, Atlanta, Georgia. Contact:<br />
Christina DeLance, NAC International,<br />
telephone: (678) 328-1281,<br />
email: cdelance@nacintl.com.<br />
11. Boiler and Reactor Feedpump Turbine<br />
Workshop, November 18-20,<br />
2008, Nashville Marriot at Vanderbilt<br />
University, Nashville, Tennessee.<br />
Contact: Electric Power Research<br />
Institute, Linda Parrish, telephone:<br />
(704) 5952-2000.<br />
12. The <strong>Nuclear</strong> Power Congress 2008,<br />
December 9-10, 2008, The Ritz-<br />
Carlton Golf Resort, Naples Florida.<br />
Contact: Kristy Perkins, American<br />
Conference Institute, email:<br />
k.perkins@americanconference.<br />
com.<br />
13. WM 2008 Phoenix, Waste Management<br />
for the <strong>Nuclear</strong> Renaissance,<br />
March 1-5, 2009, Phoenix, Arizona.<br />
Contact: WMS Administration,<br />
telephone: (520) 696-0399, email:<br />
papers@wmarizona.org.<br />
14. World <strong>Nuclear</strong> Fuel Cycle 2009,<br />
April 22-24, 2009, Sydney, Australia.<br />
Contact: Stuart Cloke, World<br />
<strong>Nuclear</strong> telephone: 44 207 451 1520,<br />
email: cloke@world-nuclear.org.<br />
15. Annual Meeting on <strong>Nuclear</strong> Technology,<br />
May 12-14, 2009, Congress<br />
Center Dresden, Germany. Contact:<br />
dbcm GmbH, telephone: 49 02241<br />
93897 0, email: info@dbcm.de. <br />
Neutron Absorber<br />
Materials<br />
BORAL ® Composite<br />
BORTEC ® MMC<br />
Borated Aluminum<br />
Enriched Boron<br />
Natural Boron Carbide<br />
418-693-0227 nuclear@ceradyne.com www.ceradyne.com<br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 15
Technologies of National<br />
Importance<br />
By Tsutomu Ohkubo, Japan Atomic<br />
Energy Agency.<br />
1. Please provide a brief description of<br />
RMWR 300MW(e)/X.<br />
The reduced-moderation water<br />
reactor (RMWR) is a BWR-type reactor<br />
being developed to ensure the sustainable<br />
energy supply in the future through<br />
multiple recycling of plutonium based on<br />
the well-developed and experienced LWR<br />
technologies. The RMWR core consists<br />
of hexagonal fuel assemblies with MOX<br />
fuel rods arranged in the triangular tightlattice<br />
configuration. Therefore, it can<br />
attain a fissile plutonium conversion ratio<br />
or the breeding ratio over 1.0 under the<br />
relatively hard or fast neutron spectrum.<br />
The conceptual design of RMWR<br />
300MWe with the passive safety<br />
features has been accomplished in main<br />
cooperation with Hitachi Ltd. aiming at<br />
the electric power generation using the<br />
small 330MWe/955MWt RMWR core<br />
with the discharge burn-up of 65GWd/t<br />
and the operation cycle of 25 months under<br />
the multiple recycling situation. The core<br />
consists of 282 hexagonal fuel bundles,<br />
each of which has 217 fuel rods with<br />
the outer diameter of 13.0 mm arranged<br />
in the triangular lattice with 1.3 mm gap<br />
width between rods. The MOX part is<br />
shortened around 0.2 m high and two<br />
MOX parts are piled up with an internal<br />
blanket region, forming the double-flatcore<br />
configuration to attain the negative<br />
void reactivity coefficients, as shown in<br />
the figure. Adding the upper and lower<br />
blanket regions, the total axial length is<br />
1.32m. The control rods are Y-shaped<br />
ones with the follower structure above the<br />
neutron absorber material region.<br />
The core is cooled by the natural<br />
circulation of the water coolant under<br />
the same operating conditions as BWRs,<br />
i.e. 7.2MPa and 561K. A breeding ratio<br />
Responses to questions by Newal<br />
Agnihotri, Editor of <strong>Nuclear</strong> <strong>Plant</strong><br />
<strong>Journal</strong>.<br />
of 1.01 and the negative void reactivity<br />
coefficients are simultaneously realized<br />
in the design. The fuel cycle concept is<br />
a closed one and the simplified PUREX<br />
method, in which purification processes<br />
for Pu and U are eliminated, is considered<br />
for the reprocessing process. Minor<br />
actinides (MAs) could be recycled in<br />
MOX with the enhanced proliferation<br />
resistance, when MA recovery and MA-<br />
MOX fuel fabrication processes are<br />
established.<br />
In order to overcome what is called<br />
the scale demerit for small reactors,<br />
the plant systems is simplified and the<br />
passive safety features are introduced in<br />
the present plant system design. One of<br />
the major passive safety features is the<br />
natural circulation core cooling system,<br />
and other passive safety concepts, such as<br />
the gravity steam-water separation in the<br />
upper plenum, the accumulator injection<br />
system, the isolation condenser system<br />
and the passive containment cooling<br />
system, are also intended to be utilized<br />
to improve the economy and to enhance<br />
the reliability and the safety. In the<br />
present safety system, a hybrid one with<br />
the combination of the passive and the<br />
active components is proposed and has<br />
been evaluated to reduce the cost for the<br />
reactor components.<br />
Although no prototype for this<br />
reactor concept has been established, a<br />
Tsutomu Ohkubo<br />
Tsutomu Okubo joined Japan Atomic<br />
Energy Research Institute (JAERI, it<br />
is now Japan Atomic Energy Agency<br />
(JAEA) since October 2005) in 1978<br />
and worked for advanced water reactors<br />
design research, reactor thermalhydraulics<br />
and safety engineering. He<br />
is currently working on the development<br />
of the reduced-moderation type water<br />
reactor named FLWR as the Senior<br />
Principal Researcher. He is a member of<br />
the Atomic Energy Society of Japan.<br />
large scale experimental program for the<br />
critical heat flux in the tight-lattice rod<br />
bundle was already conducted under the<br />
reactor operating conditions and the core<br />
cooling capability was demonstrated.<br />
Some irradiation tests are necessary for<br />
the highly enriched MOX fuel rods up<br />
to at high burn-up. Since this reactor<br />
concept is based on the well-developed<br />
and experienced LWR technologies up to<br />
now, it is expected to be realized without<br />
serious difficulties. It would be ready for<br />
commercialization in 2020s. This reactor<br />
concept was also nominated as the High<br />
Conversion BWR (HC-BWR) of the<br />
advanced BWRs in the International<br />
Near-Term Deployment (INTD).<br />
2. Does your reactor include a<br />
containment building If yes, please<br />
describe the characteristics of your<br />
containment building.<br />
It has a steel containment system<br />
to facilitate heat transfer from inside to<br />
outside as a part of the passive containment<br />
cooling system.<br />
3. What has JAEA done in using nuclear<br />
energy in applications other than<br />
power production, including District<br />
Heating, Seawater Desalination and<br />
Transportation<br />
JAEA has been developing the<br />
hydrogen production technologies using<br />
16 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
Very High Temperature Gas-cooled<br />
Reactor (VHTR) and Sodium-cooled Fast<br />
Breeder Reactor (FBR) with different<br />
demonstration FBR will be operated<br />
around 2025 and a commercialized FBR<br />
will be developed before 2050.<br />
JAEA also participates in all four GIF<br />
VHTR Projects of hydrogen production,<br />
fuel, material and code development.<br />
JAEA is a world front runner of the VHTR<br />
and hydrogen production technologies<br />
and is willing to cooperate with foreign<br />
organization developing the VHTRhydrogen.<br />
Toshiba, MHI, Fuji electric and<br />
<strong>Nuclear</strong> Fuel Industries take part in the<br />
Next Generation <strong>Nuclear</strong> <strong>Plant</strong> (NGNP)<br />
program. What they will achieve depends<br />
on the budget they will acquire.<br />
Bird’s-eye view of core and cross sectional view of fuel assembly<br />
methods. VHTR can be also used for<br />
desalination and district heating.<br />
4. Who are JAEA’s partners in<br />
producing hydrogen utilizing nuclear<br />
energy Has a prototype already been<br />
tested Please include a schedule for<br />
application of hydrogen technology for<br />
transportation in Japan.<br />
Japanese industries such as Toshiba,<br />
MHI etc. are working with JAEA<br />
to develop the hydrogen production<br />
technology using VHTR.<br />
An experimental facility to produce<br />
hydrogen of 30 liter /h using Iodine and<br />
Sulfur (IS) method was constructed for<br />
VHTR. The successful 1 week operation<br />
was completed to confirm its chemical<br />
process and establish the control<br />
technology.<br />
Though the prototype has not<br />
been constructed, the research and<br />
developments for hydrogen production<br />
with the IS process and achievement<br />
of higher efficiency than the previous<br />
method is being planned.<br />
Japanese Atomic Energy Commission<br />
recently stated that VHTR-hydrogen plays<br />
a key role to reduce CO2 emission and it<br />
will be commercialized during 2020-2030<br />
for application including transportation.<br />
JAEA contributes to the fast reactor<br />
system development out of Generation IV<br />
reactors, especially a lot in the Sodiumcooled<br />
Fast Reactor (SFR) program as<br />
leaders.<br />
Japan has a national development<br />
plan as a “key technology of national<br />
importance” among the government,<br />
utilities, industries and JAEA that a<br />
Contact: Tsutomu Ohkubo, Japan<br />
Atomic Energy Agency, 4002 Narita-<br />
Cho, Oarai-Machi, Ibraki-Ken 311-<br />
1393, Japan; telephone: 81-29-267-1919<br />
ext 6480, fax: 81-29-266-3675, email:<br />
ohkubo.tsutomu@jaea.go.jp. <br />
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5. What is JAEA’s contribution to<br />
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tangible results<br />
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<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 17
Modeling & Simulation Advances<br />
Brighten Future <strong>Nuclear</strong> Power<br />
By Hussein Khalil, Argonne National<br />
Laboratory.<br />
Bob Hill and Jim Cahalan from the<br />
<strong>Nuclear</strong> Engineering Division and<br />
Andrew Siegel from the Mathematics<br />
& Computer Science Division also<br />
contributed.<br />
1. What applications have you currently<br />
undertaken for design, operation, or<br />
construction of nuclear power plants<br />
Applications of leadership class<br />
computers for nuclear energy R&D at<br />
Argonne have so far focused mainly on<br />
development and design of advanced<br />
sodium cooled fast reactors (SFR), which<br />
target sustainable energy generation, waste<br />
minimization, assured passive safety,<br />
and competitive economics. To enable<br />
these applications we are developing<br />
a modern computational framework<br />
that uses advanced software tools and<br />
computational methods for simulation<br />
of multi-physics (neutronic, thermalhydraulic,<br />
mechanical, etc.) phenomena<br />
in complex reactor geometries. This<br />
framework, named SHARP (Simulation<br />
for High-efficiency Advanced Reactor<br />
Prototyping), enables high-fidelity<br />
simulation of reactor behavior taking<br />
advantage of the enormous computing<br />
power afforded by leadership class<br />
computers. Its design provides flexibility<br />
to employ less detailed (faster running)<br />
models and to couple the different<br />
physics modules tightly or loosely<br />
depending on problem characteristics<br />
and accuracy requirements. A key<br />
goal of our development is to integrate<br />
improved methods for characterizing the<br />
uncertainty in predicted quantities within<br />
the analysis framework.<br />
To demonstrate the benefit of<br />
leadership class computers for SFR<br />
analysis, two computationally intensive<br />
applications of computational fluid<br />
dynamics (CFD) techniques are being<br />
Responses to questions by Newal<br />
Agnihotri, Editor of <strong>Nuclear</strong> <strong>Plant</strong><br />
<strong>Journal</strong>.<br />
Hussein Khalil<br />
Hussein S. Khalil is director of<br />
Argonne’s <strong>Nuclear</strong> Engineering Division<br />
and is responsible for the Lab’s research<br />
carried out using the IBM Blue Gene/P<br />
supercomputer at Argonne’s Leadership<br />
Class Computing Facility (see http://<br />
www.alcf.anl.gov):<br />
• Detailed characterization of turbulent<br />
coolant flow and heat transfer in SFR<br />
wire-wrapped fuel pin bundles.<br />
• Investigation of transient flow<br />
fluctuations (thermal striping) in<br />
the SFR upper plenum region where<br />
the coolant discharged from fuel<br />
assemblies mixes.<br />
Additionally, we are performing<br />
high-order, multigroup neutron transport<br />
calculations for a highly detailed model<br />
of a SFR core.<br />
Blue Gene/P was just officially<br />
clocked as the fastest computer in the<br />
world dedicated to open science and is<br />
the third fastest computer in the world<br />
overall. It uses 163,840 parallel compute<br />
nodes to execute at a clock rate of nearly<br />
0.56 petaflops (1 petaflop = 10 15 floating<br />
point operations per second) with a total<br />
RAM of 80 terabytes. A simulation that<br />
would take two years on a standard PC<br />
can now be done in ten minutes. Access<br />
to BG/P is granted using a competitive<br />
peer reviewed process.<br />
2. Will your system analysis cut down<br />
the capital cost of NPP <strong>Nuclear</strong> Steam<br />
Supply System by making the fuel and<br />
the thermal hydraulics more effi cient<br />
on nuclear reactor technology and<br />
nuclear non-proliferation. He has<br />
worked at Argonne since 1983 and<br />
became a Senior Scientist in 2001. He<br />
has a Ph.D. from MIT (1983) and an<br />
MBA from the University of Chicago<br />
(1996).<br />
Dr. Khalil is an internationally<br />
recognized expert in nuclear reactor<br />
physics and engineering. His research<br />
has centered on the advancement of<br />
reactor analysis methods and their<br />
application for fast reactor design<br />
optimization.<br />
The computational capabilities<br />
under development will enable more<br />
precise representation (modeling) of the<br />
reactor and power plant configuration<br />
and more accurate solution of the<br />
equations describing reactor neutronic,<br />
irradiation, thermal, fluid flow, and<br />
structural/mechanical behavior. This<br />
degree of modeling fidelity, combined<br />
with enhanced capability for uncertainty<br />
characterization, will make it possible<br />
to design and operate reactors closer<br />
to the true physical capabilities of the<br />
fuel, materials of construction and<br />
components. When these high fidelity<br />
modeling capabilities are employed in the<br />
design process, unnecessary conservatism<br />
in reactor design and operation can be<br />
reduced without compromising safety<br />
assurance.<br />
3. Will your computational tools also<br />
facilitate optimizing the usage of fuel by<br />
providing assistance in designing and in<br />
operation<br />
The high fidelity models being<br />
integrated in SHARP allow greatly<br />
improved characterization of fuel<br />
operating conditions over its lifetime.<br />
Results of these advanced models can be<br />
employed in models of fuel behavior (a<br />
key component of the overall code system)<br />
to support the optimization of fuel design<br />
(Continued on page 20)<br />
18 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
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Modeling &...<br />
Continued from page 18<br />
and to provide the necessary assurance<br />
of fuel integrity over its operating life<br />
considering both normal (operational)<br />
and abnormal conditions.<br />
Fuel behavior models currently<br />
available have limited predictive capacity.<br />
They rely extensively on the results of<br />
fuel property and irradiation tests and<br />
post-irradiation examinations. A large<br />
number of in-pile tests are typically<br />
needed to encompass the fuel operating<br />
conditions of interest, and the duration of<br />
these tests may be several years to reach<br />
the targeted discharge burnup.<br />
The high cost and protracted nature<br />
of these tests create a strong incentive to<br />
develop computational models of fuel<br />
behavior that have greater predictive<br />
capability and are less dependent on<br />
empirical testing. Advancement of such<br />
capabilities is pursued in parallel with<br />
the (reactor) modeling and simulation<br />
efforts described here, with the aim of<br />
appropriately integrating or coordinating<br />
their application in the future.<br />
4. Who are your global partners in this<br />
effort<br />
Argonne is leading a team of U.S.<br />
national laboratories (Idaho, Oak Ridge<br />
and Lawrence Livermore National<br />
Laboratories) and several universities<br />
in the advancement of reactor modeling<br />
and simulation capabilities centered<br />
on the SHARP code and the effective<br />
use of leadership class computers,<br />
including the IBM Blue Gene/P. This<br />
national effort is sponsored by the U.S.<br />
Department of Energy and is carried out<br />
in cooperation with the French Atomic<br />
Energy Commission (CEA) and the<br />
Japanese Atomic Energy Agency (JAEA).<br />
Cooperative activities currently underway<br />
include (a) joint definition of benchmark<br />
problems that can be used to test the<br />
existing and developmental code systems<br />
in each country, (b) joint comparison and<br />
assessment of benchmark results, and<br />
(c) joint assessment and improvement<br />
of enabling software tools, e.g., tools for<br />
geometry description, mesh generation,<br />
data management, solution decomposition<br />
and parallelization and visualization of<br />
results.<br />
5. Please describe your plans with<br />
your current technology for assisting<br />
research, design, and operation of<br />
nuclear power plants in the next fi ve<br />
years<br />
Our current plans are focused on<br />
continued development, testing and<br />
integration of the SHARP code. The<br />
development effort will be guided and<br />
focused by applications supporting the<br />
development of conceptual designs for<br />
advanced reactor systems and confirmation<br />
of their safety. Their main initial use<br />
will be to complement experimental<br />
measurements in the qualification of the<br />
existing analysis tools and to investigate<br />
design options and operating conditions<br />
that cannot be explored reliably with<br />
existing tools.<br />
Although separate- and integraleffects<br />
measurements will continue to<br />
SFR Bundle<br />
be needed for validation of the models<br />
used in reactor design, the advanced<br />
capabilities under development will make<br />
it possible to optimize the experimental<br />
campaigns and to support greater use of<br />
“numerical prototyping” in the design of<br />
reactor components and systems.<br />
6. Please share any other details, which<br />
you may like to bring to the attention<br />
of our readership in the nuclear power<br />
industry.<br />
The code systems in use today for<br />
reactor development and design were<br />
initiated more than thirty years ago and<br />
were designed to accommodate the<br />
computing resources, tools and methods<br />
that were available at the time. We are<br />
targeting a vastly improved capability<br />
that exploits advances in computers and<br />
software tools to facilitate reactor design<br />
optimization, provide increased assurance<br />
of performance and safety characteristics,<br />
and reduce the need for large scale integral<br />
experiments to characterize or validate<br />
performance.<br />
In addition to the improved<br />
ability to predict reactor behavior, we<br />
envision a vastly superior process for<br />
development, design and licensing of<br />
future reactors. This process would<br />
integrate all significant aspects of the<br />
design to influence optimized design<br />
choices at the conceptual stage of the<br />
design. It would also support evolution<br />
from the conceptual stage to the detailed<br />
design of realizable components. Finally,<br />
it would provide for automated transfer<br />
of design specifications to instructions<br />
for manufacture and assembly, enabling<br />
the manufacture of parts and components<br />
to close tolerances and assured fit at the<br />
time of assembly.<br />
7. Do have enough funding to realize<br />
your plans in the next fi ve years<br />
We are grateful for the sponsorship<br />
the U.S. Department of Energy provides<br />
for our effort to advance modeling and<br />
simulation of nuclear reactors, as well as<br />
for the its past and continuing investment<br />
in high-performance computers and<br />
the software needed to make effective<br />
use of these computers. Our progress<br />
on development and application of the<br />
reactor simulation tools, centered on<br />
the SHARP code, obviously depends<br />
on the funding support we receive over<br />
the next five years – not only for code<br />
development but also for application and<br />
validation studies and quality assurance.<br />
We are optimistic about the prospect<br />
for this funding, because the benefit of<br />
this research for advancing the use of<br />
nuclear energy is increasingly recognized<br />
by the technical community and<br />
policymakers. At the same time, we are<br />
extremely interested in partnerships with<br />
commercial organizations that can provide<br />
additional resources for accelerating<br />
our development and validation efforts<br />
and bringing their products to bear on<br />
the commercial design, licensing and<br />
operation of nuclear power plants.<br />
Contact: Hussein S. Khalil, Argonne<br />
National Laboratory, 9700 S. Cass<br />
Avenue, Bldg 208, Argonne, IL 60439;<br />
telephone: (630) 252-1456, fax: (630)<br />
252-4780, email: hkhalil@anl.gov. <br />
20 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
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Energy & Desalination Projects<br />
By Ratan Kumar Sinha, Bhabha Atomic<br />
Research Centre, India.<br />
1. Is your project part of Generation<br />
IV International Forum, International<br />
Project on Innovative <strong>Nuclear</strong><br />
Reactors and Fuel Cycles (INPRO)<br />
If so, please provide details of your<br />
project’s involvement with the above<br />
organizations.<br />
India is not a member of Generation-<br />
IV International Forum. The design and<br />
development of Advanced Heavy Water<br />
Reactor (AHWR) has been carried out at<br />
Bhabha Atomic Research Centre (BARC)<br />
without any external collaboration.<br />
The IAEA’s International Project on<br />
Innovative <strong>Nuclear</strong> reactors and fuel<br />
cycles (INPRO) has stipulated a set of<br />
requirements and criteria that should be<br />
fulfilled by the innovative nuclear reactors<br />
and fuel cycles of the future. AHWR<br />
served as a case study for validating these<br />
requirements and criteria.<br />
2. Please provide a brief description of<br />
AHWR.<br />
AHWR is a 300 MWe, vertical,<br />
pressure tube type, boiling light water<br />
cooled, and heavy water moderated<br />
reactor. The reactor incorporates a<br />
number of passive safety features and is<br />
associated with a fuel cycle having reduced<br />
environmental impact. At the same time,<br />
the reactor possesses several features,<br />
which are likely to reduce its capital and<br />
operating costs. In the Indian context,<br />
AHWR will serve as a platform for the<br />
timely development and demonstration<br />
of the reactor and fuel cycle technologies<br />
required to be in place before large scale<br />
thorium utilisation in the future. The<br />
AHWR fuel cycle has, however, enough<br />
flexibility to accommodate a large variety<br />
of fuelling options.<br />
The reactor uses thorium based oxide<br />
fuel with in-situ generated Uranium-233<br />
and Plutonium, recovered from the spent<br />
Responses to questions by Newal<br />
Agnihotri, Editor of <strong>Nuclear</strong> <strong>Plant</strong><br />
<strong>Journal</strong>.<br />
Ratan Kumar Sinha<br />
Mr. Ratan Kumar Sinha graduated<br />
in Mechanical Engineering in 1972<br />
and received training in nuclear<br />
engineering, at postgraduate level, in<br />
the training school of Bhabha Atomic<br />
Research Centre (BARC) Mumbai,<br />
India. He has thirty-fi ve years of<br />
experience in the area of development<br />
of reactor engineering technologies for<br />
components and systems of pressure<br />
tube type research and power reactors.<br />
At present he is serving as Director,<br />
Reactor Design & Development<br />
Group and, Director Design,<br />
fuel of water cooled reactors, serving<br />
as fissile materials under equilibrium<br />
conditions. It addresses the requirement<br />
of sustainability of nuclear fuel resource<br />
through the use of a closed fuel cycle<br />
along with thorium.<br />
Incidentally, on account of the use of<br />
thorium based fuel, with no production<br />
of additional Plutonium and the presence<br />
of high energy gamma emitting daughter<br />
products of Uranium-232, the reactor is<br />
considered to have inherent proliferation<br />
resistant features. The production of<br />
minor actinides, in this reactor, is reduced<br />
by nearly one order of magnitude, in<br />
Manufacturing & Automation Group,<br />
BARC. His current responsibilities<br />
include directing programmes<br />
for new advanced reactors under<br />
design and development at BARC<br />
to utilise thorium. These include,<br />
the Advanced Heavy Water Reactor<br />
(AHWR), which produces most of its<br />
power from thorium, and has several<br />
innovative passive safety features.<br />
He is also responsible for the design<br />
and development of a Compact High<br />
Temperature Reactor (CHTR), which<br />
is a technology demonstrator for future<br />
Indian High Temperature Reactors<br />
intended for hydrogen generation.<br />
Mr. Sinha is a nationally and<br />
internationally recognized expert in<br />
the area of nuclear reactor technology.<br />
For the past four years he has been the<br />
Chairman of the Steering Committee<br />
of INPRO, the IAEA’s International<br />
Project on Innovative <strong>Nuclear</strong> Reactors<br />
and Fuel Cycles.<br />
Mr. Sinha has received several awards<br />
and honours. He was elected a Fellow<br />
of the Indian National Academy of<br />
Engineering in the year 1998. He<br />
has been an elected member of the<br />
Executive Committee of the Indian<br />
<strong>Nuclear</strong> Society for the last eight years.<br />
comparison with conventional reactors,<br />
thus substantially reducing the burden<br />
of managing the inventory of long-lived<br />
radioactive waste.<br />
In AHWR, light water at 259 ° C<br />
enters the core through 452 feeders, each<br />
connected to a single vertical pressure tube,<br />
in which heat is transferred from nuclear<br />
fuel leading to boiling of the coolant. The<br />
steam water mixture produced in these<br />
pressure tubes rises through tail pipes<br />
leading to four steam drums in which<br />
steam, at nominal conditions of 70 bar<br />
pressure and 270 ° C, is separated and taken<br />
to the turbine cycle. The plant is designed<br />
22 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
to produce 300 MWe electricity along<br />
with 500 m 3 /day of desalinated water.<br />
The inherent and passive safety features<br />
of the reactor include negative void<br />
coefficient of reactivity, full power core<br />
heat removal using natural circulation,<br />
shut down decay heat removal backed<br />
up by natural circulation, a passive shut<br />
down device to address a postulated<br />
insider threat of disablement of the two<br />
main shut down systems, passive cooling<br />
of concrete structures surrounding the<br />
main heat transport system piping, and<br />
passive isolation of containment as<br />
well as passive cooling of containment<br />
environment following a postulated loss<br />
of coolant accident.<br />
The reactor is provided with a<br />
double containment. A 6000 m 3 capacity<br />
water tank located inside the primary<br />
containment, near its top, serves as a heat<br />
sink for a range of postulated scenarios in<br />
which the main coolant supply to the core<br />
and/or the cooling water to the condenser<br />
is not available. With the help of this<br />
heat sink and other passive features, the<br />
reactor is designed for providing a grace<br />
period of at least three days following<br />
any postulated scenario affecting the<br />
plant. Thus, even without any external<br />
source of power, coolant and operator<br />
actions, safety of the reactor is assured<br />
for practically an indefinite period.<br />
The new design features of the<br />
reactor have been validated with the help<br />
of several large experimental facilities. A<br />
large Critical Facility designed to validate<br />
the reactor physics design of AHWR has<br />
recently been commissioned at BARC.<br />
The safety related features of AHWR have<br />
been subjected to a pre-licensing design<br />
appraisal by the Indian Atomic Energy<br />
Regulatory Board. The design of the<br />
nuclear systems is nearly complete and is<br />
available for initiating the construction of<br />
the plant in the near future.<br />
3. What has Bhabha Atomic Research<br />
Centre done in using nuclear energy in<br />
application other than power production,<br />
including District Heating, Sea Water<br />
Desalination and Transportation<br />
The Bhabha Atomic Research Centre<br />
has helped the domestic development of<br />
all required technologies, materials and<br />
hardware necessary for the Pressurised<br />
Heavy Water Reactor programme. It<br />
has also been engaged in providing the<br />
inspection and maintenance support, as<br />
needed in some critical areas for these<br />
reactors. District heating is not a major<br />
requirement in most of India with tropical<br />
climate conditions. However, the Indian<br />
programme includes 220 MWe (750<br />
MWth) PHWRs that may be effectively<br />
deployed for a variety of applications<br />
requiring small/medium power reactors.<br />
BARC has got a very active programme<br />
in sea water desalination and its work<br />
covers a number of technologies,<br />
including membrane based technologies<br />
and evaporation based technologies<br />
for desalination and potable water<br />
production in a cost-effective as well<br />
as energy-efficient manner. The Indian<br />
Madras Atomic Power Station (MAPS),<br />
for example, is being coupled with a large<br />
desalination plant.<br />
4. Who are Bhabha Atomic Research<br />
Centre’s partners in producing hydrogen<br />
utilizing nuclear energy Has a prototype<br />
already been tested Please include a<br />
schedule for application of hydrogen<br />
technology for transportation in India.<br />
Bhabha Atomic Research Centre<br />
has been working on the development<br />
of technologies for producing hydrogen<br />
using water splitting reactions. Its current<br />
activities in this area include conventional<br />
electrolysis, high temperature electrolysis,<br />
and chemico-thermal processes for<br />
hydrogen generation. Bhabha Atomic<br />
Research Centre is one of the several<br />
research organizations, academic<br />
institutions and industrial partners that<br />
have contributed towards preparation of a<br />
national hydrogen energy road map.<br />
5. What is Bhabha Atomic Research<br />
Centre’s contribution to Generation IV<br />
reactors and what is the schedule for the<br />
industry to see some tangible results<br />
India is not a member of Generation-<br />
IV. However, the Advanced Heavy Water<br />
Reactor mentioned above fulfils/exceeds<br />
all the requirements stipulated by INPRO,<br />
for the next generation nuclear reactors.<br />
The design of this demonstration reactor<br />
has reached an adequately advanced level,<br />
and the construction of the reactor is<br />
planned to be initiated in the near future.<br />
6. Who is the manufacturer of forgings<br />
for reactor pressure vessels in India<br />
The domestic Indian nuclear<br />
power programme is currently based<br />
on Pressurised Heavy Water Reactors<br />
(PHWRs) and pool type Fast Breeder<br />
Reactors (FBRs). These reactors do not<br />
require reactor pressure vessels. Major<br />
components for the Indian nuclear reactor<br />
programme have been manufactured by<br />
several industries in the governmental<br />
(public sector) as well as private sector in<br />
India.<br />
Contact: Ratan Kumar Sinha, Bhabha<br />
Atomic Research Centre, BARC, Mumbai,<br />
400085; email: redamin@barc.gov.in. <br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 23
A <strong>Plant</strong> with Simplified Design<br />
By John Higgins, GE Hitachi <strong>Nuclear</strong><br />
Energy.<br />
1. How does the ESBWR minimize<br />
damage to the fuel in case of a loss of<br />
coolant accident (LOCA)<br />
With the ESBWR, the fuel remains<br />
covered and well-cooled through all<br />
operational events including the unlikely<br />
event of a LOCA. This ensures that the<br />
fuel temperature remains at or below the<br />
fuel’s normal operating temperature. The<br />
ESBWR builds on the outstanding safety<br />
record of the world’s established BWR<br />
fleet.<br />
2. How has the ESBWR improved the<br />
reactor water chemistry to minimize<br />
affect on the fuel and on reactor<br />
internals during normal operation and<br />
during accident conditions<br />
The ESBWR operates well within<br />
the industry-established BWR water<br />
chemistry guidelines (specifically the<br />
limits on feedwater iron levels), which<br />
effectively precludes the buildup of iron<br />
oxide deposits on fuel elements and<br />
reactor internals.<br />
One of the goals of maintaining<br />
good BWR water chemistry during plant<br />
operation is to minimize the potential for<br />
developing intergranular stress corrosion<br />
cracking (IGSCC) on reactor internals.<br />
For the ESBWR, the potential for IGSCC<br />
resistance is addressed through the use<br />
of significantly improved materials, such<br />
as Type 316 <strong>Nuclear</strong> Grade stainless<br />
steel and stabilized nickel-base niobium<br />
modified Alloy 600 and Alloy 82. The<br />
ESBWR design has significantly reduced<br />
the number of welds needed, and along<br />
with the use of improved materials, the<br />
potential for cracking is substantially<br />
reduced.<br />
3. What fuel and fuel cladding material<br />
design enhancements have been made<br />
in ESBWR to ensure minimum damage<br />
Responses to questions by Newal<br />
Agnihotri, Editor of <strong>Nuclear</strong> <strong>Plant</strong><br />
<strong>Journal</strong>.<br />
John Higgins<br />
John Higgins serves as Vice President,<br />
ESBWR Projects, for GE Hitachi<br />
<strong>Nuclear</strong> Energy. Higgins joined the<br />
company in 2005 as the Project Manager<br />
responsible for a joint Department of<br />
Energy initiative to advance the design<br />
of the next-generation boiling water<br />
reactor technology, the ESBWR. In<br />
of the fuel during normal operation, and<br />
during accident scenarios<br />
The ESBWR takes advantage of years<br />
of operating experience with BWRs and<br />
offers improvements to address typical<br />
fuel cladding problems. Global <strong>Nuclear</strong><br />
Fuel - a joint venture of GE, Hitachi and<br />
Toshiba formed to produce BWR fuel -<br />
will supply ESBWR fuel incorporating<br />
the following features :<br />
• Debris filtration devices to trap<br />
debris material before it reaches the<br />
fuel rods in order to prevent debris<br />
fretting<br />
• Pellet cladding interaction (PCI)<br />
resistant fuel rod technology to<br />
prevent PCI failures<br />
• Corrosion resistant cladding to<br />
prevent fuel failures due to build-up<br />
or chemical intrusion events<br />
4. What innovative fuel cycles have<br />
been used in ESBWR to maximize fuel<br />
effi ciency<br />
Building on years of operating<br />
experience and advanced fuel design, the<br />
ESBWR core design provides numerous<br />
2006, Higgins assumed additional<br />
responsibilities for overall deployment<br />
planning for the ESBWR, and in<br />
2007, he was promoted to his current<br />
position. In 2008, Higgins assumed<br />
overall management responsibility<br />
for the global ESBWR business,<br />
responsible for completing the NRC<br />
certifi cation process, fi nalizing the<br />
detailed design, establishing the<br />
advanced modularization requirements<br />
and construction methods, and<br />
commercialization of the technology.<br />
Higgins is a degreed engineer with<br />
30 years of professional experience<br />
supporting both nuclear and fossil<br />
projects. During his career, Higgins has<br />
accumulated a broad base of experience<br />
that includes licensed merchant marine<br />
offi cer, nuclear start-up engineer, and<br />
business unit manager.<br />
options for our customers that will help<br />
to minimize outage lengths, support<br />
high discharge exposure and reduce<br />
enrichment requirements for fuel cycles.<br />
ESBWR cores can support a wide range<br />
of refueling cycle intervals ranging from<br />
12 to 24 months.<br />
5. How has ESBWR ensured a longer<br />
cable life to ensure a longer plant life<br />
To ensure a longer cable life, the<br />
ESBWR utilizes a comprehensive quality<br />
assurance (QA) program, a disciplined<br />
electrical design regimen, the most<br />
stringent nuclear industry standards<br />
and rigorous qualification testing. GEH<br />
participates in the development of<br />
consensus nuclear power industry cable<br />
standards, which are based on research<br />
and testing specifically for nuclear<br />
power plant applications. The ESBWR<br />
electrical engineering team continues<br />
to utilize disciplined practices and<br />
state-of-the art design tools to build on<br />
GEH’s nuclear legacy. GEH’s QA and<br />
equipment qualification programs include<br />
evaluation of life-limiting mechanisms,<br />
24 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
special material selection and rigorous<br />
proof testing of cable performance.<br />
The qualification proof testing includes<br />
condition monitoring, flammability,<br />
radiation exposure, simulation of the<br />
60-year life span, mechanical stress and<br />
LOCA testing.<br />
6. What is the plant life of ESBWR<br />
The design life for the ESBWR plant<br />
and all its major components is 60 years.<br />
7. What enhancements have been made<br />
in the control station design to ensure<br />
improved human-system interface<br />
By utilizing experienced plant<br />
operators and human factors engineering<br />
concepts, the ESBWR control room was<br />
developed, designed and evaluated using<br />
an integrated top-down design process<br />
that uses state-of-the-art methods and<br />
technology. The control room was<br />
developed to meet the review criteria<br />
detailed in industry standards, including<br />
Standard Review Plan Chapter 18 from<br />
NUREG-0800 and also NUREG 0711.<br />
Significant ESBWR control room<br />
enhancements include:<br />
• A wide display panel<br />
• Alarm filtering and prioritization<br />
• Computerized procedures<br />
• <strong>Plant</strong> and system automation<br />
• Video workstations<br />
• Advanced trending<br />
• Comprehensive human factors<br />
engineering<br />
8. How has the current instrumentation<br />
and control system in the ESBWR been<br />
upgraded from the previous GE Hitachi<br />
<strong>Nuclear</strong> Energy designs to ensure a<br />
reliable plant operation with longer<br />
plant life<br />
The ESBWR Distributed Control<br />
and Information System (DCIS) has<br />
four divisions and is designed with no<br />
single failure points that could affect<br />
the performance of support systems.<br />
In addition, DCIS contains sufficient<br />
redundancy so that even in the unlikely<br />
event of a failure while maintenance is<br />
being performed, there would be no impact<br />
on safety system functions. ESBWR<br />
automation systems ensure consistent<br />
and conservative plant operation, either<br />
remotely or control room dispatched, for<br />
plant functions such as pulling control<br />
rods to critical, heat-up/pressurization,<br />
turbine roll and synchronization, and<br />
power operation. Key control systems<br />
are triply redundant to improve the safety<br />
and reliability of the plant.<br />
9. How has information technology<br />
been used to survey and self-diagnose<br />
problems in the systems, structure and<br />
components of the ESBWR<br />
Major ESBWR plant components<br />
are fully instrumented to support on-line<br />
monitoring for equipment degradation and<br />
maintenance. Examples of parameters<br />
included in the on-line condition<br />
monitoring include:<br />
• Flow rate, suction pressure, discharge<br />
pressure and speed for pumps<br />
• Current, voltage, power and running<br />
hours for motors<br />
• Flow rates, differential pressure and<br />
inlet/outlet temperatures for heat<br />
exchangers.<br />
Similarly, large rotating machines<br />
(or small inaccessible machines) like<br />
feedwater pumps and drywell cooling<br />
(Continued on page 26)<br />
The world is once again turning to nuclear<br />
power to meet its future energy needs.You<br />
can rely on the leadership and experience<br />
of HSB Global Standards for all RCC-M and<br />
ASME code inspection and certification<br />
requirements.<br />
• The world leader in nuclear plant &<br />
equipment inspections and certifications<br />
• More than 400 engineers, inspectors and<br />
auditors worldwide<br />
• Our accreditation to perform reviews in<br />
multiple countries simplifies the process<br />
of exporting pressure equipment<br />
• We provide complete certification &<br />
training in ASME and RCC-M code<br />
compliance<br />
Go to www.hsbgsnuclear.com for more<br />
information, local contacts or to request a<br />
nuclear code training program.<br />
NUCLEAR CERTIFICATION<br />
North America Toll-free: 800-417-3437 x25434<br />
Worldwide: +1 860-722-5434<br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 25
A <strong>Plant</strong>...<br />
Continued from page 25<br />
fans are equipped with instrumentation to<br />
support high-speed vibration monitoring<br />
and other condition evaluation techniques.<br />
The alarms from the advanced condition<br />
monitoring are integrated into the plant<br />
displays and alarm system.<br />
The DC Power Supply utilizes the<br />
latest proven technology to monitor battery<br />
voltage and provides for a “battery<br />
maintenance” feature that maintains<br />
batteries at full charge. Uninterruptible<br />
Power Supply (UPS) systems have input<br />
voltage electronic switching that protect<br />
from grid-induced spikes that could trip<br />
the safety-related DC power from their<br />
inverters.<br />
10. How does ESBWR handle unstable<br />
and disruptive phenomena, such as<br />
water hammer<br />
The passive safety design of the ES-<br />
BWR has an enhanced design capability<br />
to mitigate disruptive phenomena. In the<br />
case of water hammer, an improved system<br />
layout and enhanced features mitigate<br />
the probability of water hammer and<br />
potential consequences. Those improved<br />
features include various system design<br />
and layouts, such as surge tanks, automatic<br />
air release/vacuum valves installed<br />
at high points in system piping and at the<br />
pump discharge, proper valve actuation<br />
times to minimize water hammer, procedural<br />
requirements ensuring proper line<br />
filling prior to system operation and after<br />
maintenance operations, and the use of<br />
a check valve at each pump discharge to<br />
prevent backflow into the pump.<br />
11. Is GE Hitachi <strong>Nuclear</strong> Energy<br />
exploring options to manufacture reactor<br />
pressure vessels given the fact that there<br />
are very few manufacturers in the world<br />
to meet the required demand<br />
Because of GEH’s continued<br />
involvement in building and uprating<br />
nuclear plants around the world, we have<br />
maintained a robust manufacturing and<br />
supply chain, which serves us well as<br />
we engage in the nuclear renaissance.<br />
However we recognize that the demand<br />
is increasing, and as we have done in the<br />
past, we continually explore additional<br />
options for manufacturing reactor pressure<br />
vessels and other large components.<br />
12. How does the economy of the<br />
ESBWR compare with its previous<br />
designs<br />
As GEH’s next evolution of advanced<br />
BWR technology, the ESBWR offers a<br />
simplified design providing improved<br />
safety, excellent economics, better plant<br />
security, a broad seismic design envelope<br />
and operational flexibility that increases<br />
plant availability.<br />
Contact: Ned Glascock, GE Hitachi<br />
<strong>Nuclear</strong> Energy, 3901 Castle Hayne<br />
Road, Wilmington, NC 28402; email:<br />
Edward.glascock@ge.com. <br />
26 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
A Forward Thinking Design<br />
By Ray Ganthner, AREVA.<br />
1. What were AREVA’s objectives<br />
in introducing the EPR to the global<br />
market<br />
World-wide energy demand is<br />
increasing at an accelerating pace. At<br />
the same time, there are environmental<br />
challenges to consider. The world needs<br />
more CO2-free nuclear energy to provide<br />
certainty of energy supply to the economy.<br />
AREVA’s objective was to develop and<br />
offer a design that would most effectively<br />
meet the demand for a reliable source<br />
of power generation. The EPR is an<br />
evolutionary design based on mature,<br />
yet greatly enhanced, technology that<br />
improves safety and performance. That’s<br />
why we call it the evolutionary power<br />
reactor. The EPR has many innovative<br />
design features; but they are all based<br />
on proven technologies to provide the<br />
confidence and certainty of design the<br />
public and plant operators demand.<br />
2: What innovative fuel cycles have<br />
been used in the EPR to maximize fuel<br />
effi ciency<br />
The EPR design provides enhanced<br />
and flexible fuel utilization. The EPR has<br />
increased thermal margin compared with<br />
existing plants. The linear heat rate has<br />
been reduced and the coolant flow per<br />
assembly has been increased compared<br />
with a typical Pressurized Water Reactor<br />
plant. Therefore, the EPR has improved<br />
flexibility in designing fuel cycles from<br />
12 months to 24 months. Using our<br />
proven gadolinium burnable absorber and<br />
axial blankets, coupled with a new heavy<br />
neutron reflector around the core, we are<br />
able to design extremely efficient cores<br />
that minimize uranium requirements.<br />
This is important with the price of<br />
Uranium being much higher than it was<br />
only a few years ago.<br />
Responses to questions by Newal<br />
Agnihotri, Editor of <strong>Nuclear</strong> <strong>Plant</strong><br />
<strong>Journal</strong>.<br />
Ray Ganthner<br />
Ray Ganthner is AREVA NP Inc.<br />
senior vice president, New <strong>Plant</strong>s<br />
Deployment. He joined the company<br />
in 1980 and is currently responsible<br />
for certifi cation of advanced reactor<br />
designs for deployment in North<br />
America, including light water reactors<br />
and advanced high temperature gas<br />
reactors.<br />
3. What fuel and fuel cladding material<br />
design enhancements have been made<br />
in EPR to ensure minimum damage of<br />
the fuel during normal operation, and<br />
during accident scenarios<br />
AREVA utilizes our latest, most<br />
advanced cladding material, M5 TM , for<br />
EPR fuel. This cladding is already in use<br />
in many operating reactors world wide.<br />
The testing and operating history has<br />
verified that this cladding has superior<br />
mechanical properties and significantly<br />
reduces cladding oxidation as compared<br />
with standard zirconium or zircoloy-4<br />
material. The experience of this advanced<br />
material developed by AREVA increases<br />
the certainty of optimum fuel performance<br />
in the most challenging operating<br />
environments. Additionally, the plant is<br />
designed to operate in the range where<br />
Ganthner became Manager of Group<br />
and New Projects in 1991 and<br />
among his notable achievements was<br />
completion of the Bellefonte and WNP-1<br />
nuclear power plants. In 1994, Ganthner<br />
was named vice president of engineering<br />
and project services, where he was<br />
responsible for commercial nuclear<br />
power plant products and services.<br />
Ganthner’s executive responsibilities<br />
were expanded to include business<br />
development in 1996.<br />
In 1997, Ganthner was called upon to<br />
sponsor an international team to develop<br />
advanced nuclear fuel designs for<br />
introduction in the U.S. and Europe, and<br />
in 2000, he returned to the commercial<br />
nuclear power plant business where he<br />
was responsible for plant systems and<br />
analysis, and engineering programs<br />
with offi ces in Virginia, North Carolina<br />
and Massachusetts. Ganthner holds a<br />
Bachelor of Science in Naval Science<br />
from the U.S. Naval Academy and a<br />
Master of Business Administration from<br />
Lynchburg College.<br />
there are significant margins and thus the<br />
fuel starts out with a better margin in the<br />
case of any operational transients or in the<br />
unlikely event of an accident transient.<br />
4. How has EPR improved the reactor<br />
water chemistry to minimize affect on<br />
the fuel and on reactor internals during<br />
normal operation and during accident<br />
conditions<br />
The EPR is designed to be compatible<br />
with the latest water chemistry limits<br />
specified by the Electric Power Research<br />
Institute. All materials in contact with<br />
the reactor coolant are selected to be<br />
compatible with these requirements.<br />
Furthermore, pH control of the reactor<br />
coolant is made easier by the use of<br />
enriched B10 for soluble reactivity<br />
control, which decreases the amount of<br />
(Continued on page 28)<br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 27
A Forward...<br />
Continued from page 27<br />
boric acid required as compared with<br />
most previous plant designs.<br />
5. How do the EPR’s active and<br />
passive safety systems, including onsite<br />
and offsite emergency power sources,<br />
minimize damage to the fuel in case of a<br />
loss of coolant accident (LOCA)<br />
The EPR concept first and foremost<br />
is to design in prevention of fuel damage,<br />
then mitigation. Each of the EPR’s<br />
four independent safety trains has the<br />
designed capacity to provide the full<br />
safety function. Each of the four systems<br />
has its own dedicated emergency power<br />
source supplied by a separate diesel<br />
generator. The EPR’s safety margins are<br />
approximately a factor of 100 better than<br />
the regulatory requirements. These four<br />
safety systems are activated by automatic<br />
digital protection systems and can also be<br />
controlled by reactor operators. During an<br />
anomalous event you don’t want to rely on<br />
the laws of physics and the engineering<br />
alone, but you also want to have control<br />
of what is happening inside your plant. In<br />
addition, the lower power density of EPR<br />
fuel compared to other designs provides<br />
greater safety margins.<br />
As an example of the EPR’s forwardthinking<br />
design philosophy when it<br />
comes to safety, the EPR design led the<br />
industry by providing the extra margin of<br />
safety against airplane crash now being<br />
proposed in the recent NRC rulemaking.<br />
Its “double walled” containment and four<br />
physically separated safety trains provide<br />
the certainty of protection against a<br />
potentially severe threat to containment<br />
integrity.<br />
6. What is the plant life of EPR<br />
The EPR is designed for a plant<br />
life of 60 years. But even longer plant<br />
life is possible largely due to the use of<br />
more advanced materials and welding<br />
techniques. The metallurgical properties<br />
of Inconel 690 greatly improve the life<br />
of steam generator tubes. Reactor vessel<br />
materials, weld materials, and even the<br />
location of welds all work together to<br />
optimize and probably eventually achieve<br />
actual plant lifetimes in the unprecedented<br />
range of 60 to100 years.<br />
7. What enhancements have been made<br />
in the steam generator to ensure a longer<br />
plant life How long are the EPR steam<br />
generators expected to last<br />
The EPR steam generators are<br />
designed to last the entire plant design<br />
life of 60 years. This is due to the<br />
significant enhancements in materials<br />
and fabrication techniques incorporated<br />
into all of AREVA’s steam generators<br />
over the last 20 years. Thermallytreated<br />
alloy-690 tubing with full depth<br />
hydraulic expansion in the tube sheets<br />
virtually eliminates the potential for<br />
stress-corrosion cracking observed in<br />
many of the current generation plants<br />
that used mill-annealed alloy-600 tubing.<br />
Tube support plates are fabricated using<br />
410 SS, which has been proven to reduce<br />
fouling. Anti-vibration bars made of 405<br />
stainless steel are meticulously installed<br />
in such a way that unwanted tube wear is<br />
virtually eliminated.<br />
8. How does the EPR ensure a longer<br />
cable life to facilitate a longer plant life<br />
Cable technology has improved since<br />
existing plants were built, and longer<br />
life cables are available from various<br />
manufacturers. We are working with these<br />
manufacturers to develop even longer life<br />
cables that are compatible with the design<br />
life of the EPR.<br />
9. How has the current instrumentation<br />
and control system in the EPR been<br />
upgraded from the previous AREVA<br />
designs to ensure a reliable plant<br />
operation with longer plant life<br />
As in previous AREVA designs, the<br />
EPR I&C system design pays specific<br />
attention to safety and ensuring a high<br />
level of operational flexibility in order<br />
to meet the needs of reliable electric<br />
generation. The notable upgrade in the<br />
EPR I&C system is the TELEPERM XS<br />
digital control equipment. <strong>Digital</strong> I&C<br />
systems offer improved reliability over<br />
analog systems, and do not suffer the same<br />
types of degradation problems that occur<br />
with analog systems to support longer<br />
life. In addition, I&C systems implement<br />
advanced functionality, such as partial<br />
trips, to respond to a plant disturbance<br />
while maintaining operation. The overall<br />
design of I&C systems and associated<br />
equipment complies with requirements<br />
imposed by the process, nuclear safety<br />
and operating conditions.<br />
10. What enhancements have been made<br />
in the control station design to ensure<br />
improved human-system interface<br />
A great deal of consideration was<br />
given at the design stage by human-factor<br />
engineers for enhancing the reliability<br />
of operators’ actions during operation,<br />
testing and maintenance phases. The Main<br />
Control Room (MCR) is the centralized<br />
location used by the operators to supervise<br />
and control plant processes. The MCR<br />
is ergonomically designed using stateof-the-art<br />
human factors principles. It<br />
will be equipped with information rich<br />
screen-based indication and controls for<br />
both safety-related and non-safety related<br />
functions, computer-based procedures,<br />
and alarm display screens.<br />
The MCR provides the operator<br />
with a clear understanding of the plant<br />
status including severe accident. The<br />
enhanced human system interface (HSI)<br />
elements will provide significantly<br />
more information to the operator in a<br />
more efficient way versus conventional<br />
displays. These upgrades are expected<br />
to increase situation awareness, without<br />
creating information overload. The<br />
increased automation will help to<br />
minimize operator error and assists in<br />
(Continued on page 30)<br />
28 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
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A Forward...<br />
Continued from page 28<br />
error detection and recovery capability.<br />
With a well-designed system<br />
overview, the decision making process is<br />
made easier because the “data collection”<br />
mode required when using conventional<br />
panels is minimized. Alarm displays and<br />
computerized procedures will have ties<br />
to the indications and controls that the<br />
operator requires to make procedure step<br />
decisions.<br />
The MCR is equipped with:<br />
• Two screen-based workstations for<br />
the operators<br />
• A screen-based workstation for<br />
presenting information to the shift<br />
supervisor and the safety engineer<br />
• An additional workstation for a<br />
third operator to monitor auxiliary<br />
systems<br />
• An auxiliary panel to bring the plant<br />
to cold shutdown using safety grade<br />
displays and control<br />
• Large plant overview panels that give<br />
information on the status and main<br />
parameters of the plant<br />
11. How has information technology<br />
been used to survey and self-diagnose<br />
problems in the systems, structure, and<br />
components in EPR<br />
I&C systems, along with specialized<br />
diagnostic systems, provide advanced<br />
capabilities for the collection and storage<br />
of information regarding plant equipment.<br />
This information can be transferred<br />
to business management systems for<br />
analysis to support a wide variety of<br />
operational and maintenance objectives.<br />
12. How does the EPR handle unstable<br />
and disruptive phenomena, such as<br />
water hammer<br />
Unstable or undesired disruptive<br />
phenomena are handled at the engineering<br />
and design stage by specific design rules<br />
set to eliminate the problem. For example,<br />
geometries that could lead to rapid steam<br />
collapse or rapid valve movements are<br />
avoided, limiting the potential for water<br />
hammer. Flow assisted corrosion is<br />
limited by employing design limits on<br />
liquid velocity and water chemistry or by<br />
specification of more robust materials,<br />
for example, stainless steel or chromiummolybdenum<br />
pipe.<br />
13. What enhancements have been made<br />
in the designs and construction of EPR<br />
to control fi re and smoke in the plant<br />
affecting safety critical systems<br />
The U.S. EPR is a robust design<br />
with increased safety margin with respect<br />
to fire safe shutdown capability. The<br />
physical separation of safety system<br />
trains and the redundancy of safety<br />
systems minimize the possible effect of<br />
smoke and fire on critical safety systems.<br />
We designed redundant safety systems to<br />
exceed regulatory requirements. The four<br />
train safety concept means you can have<br />
one train in maintenance, one train may<br />
be affected by a fire, and the remaining<br />
train or trains required for safe shutdown<br />
are still available. Since each safety<br />
train is independent and located within a<br />
physically separate building, propagation<br />
of fire between divisions is eliminated.<br />
14. How does the economy of the EPR<br />
compare with its previous designs<br />
The EPR original design objective<br />
was to make the plant at least 10 percent<br />
more economic to operate than existing<br />
plants. We think we have achieved that by<br />
increasing the power level, reducing the<br />
numbers of components, and eliminating<br />
unnecessary maintenance activities. With<br />
the four independent operating trains,<br />
online maintenance has been made<br />
possible. As a result, the plant’s output in<br />
megawatt hours is higher, so fixed costs<br />
are spread over more megawatts. The<br />
EPR has a higher thermal efficiency and<br />
projected lifetime availability between 92<br />
and 95 percent.<br />
15. Is AREVA exploring options to<br />
manufacture reactor pressure vessels<br />
given the fact that there are very few<br />
manufactures in the world to meet the<br />
required demand<br />
The demand for all these new reactors<br />
around the world is a challenge for heavy<br />
component manufacturers, and AREVA<br />
is involved with the global supply of<br />
these components. In fact, we’ve been<br />
consistently investing in manufacturing to<br />
make sure we are ready when the expected<br />
demand for more nuclear energy is finally<br />
realized. We’ve been in the process of<br />
upgrading and expanding all of our heavy<br />
component shops. We’ve completed two<br />
large expansions of our manufacturing<br />
plant at Chalon, France and recently<br />
acquired a large steel forging plant in<br />
France. We’re also looking at building<br />
a large component manufacturing plant<br />
in the USA. A part of our strategy is<br />
to continuously evaluate the global<br />
marketplace and the forging business to<br />
determine whether we need to develop<br />
in-house ultra-heavy forging capability<br />
and if so, when this would make the<br />
most business sense. We’re completely<br />
committed to the expansion of clean<br />
nuclear energy, so we look at everything<br />
involved. AREVA’s tremendous domestic<br />
and global resources and our EPR design<br />
currently under construction in Finland<br />
and France, together provide a significant<br />
cost and schedule certainty for more<br />
nuclear energy to become a reality.<br />
Contact: Susan M. Hess, AREVA NP<br />
Inc., 3315 Old Forest Road, Lynchburg,<br />
VA 24501; telephone: (434) 832-2379,<br />
fax: (434) 382-2379, email:<br />
Susan.Hess@areva.com.<br />
<br />
www.<br />
NPJOnline.<br />
com<br />
30 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
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A Passively Safe Design<br />
By Ed Cummins, Westinghouse Electric<br />
Company.<br />
1. How does AP1000 minimize damage<br />
to the fuel in case of a loss of coolant<br />
accident (LOCA)<br />
The AP1000 Passive Core Cooling<br />
Systems together with other safety<br />
features is designed to protect the fuel in<br />
case of a LOCA.<br />
Regarding Accident scenarios,<br />
the AP1000 meets the U. S. NRC<br />
deterministic-safety and probabilistic-risk<br />
criteria with large margins. The safety<br />
analysis is documented in the AP1000<br />
Design Control Document (DCD) and<br />
Probabilistic Risk Assessment (PRA).<br />
Results of the PRA show a very low core<br />
damage frequency (CDF) that is 1/100 of<br />
the CDF of currently operating plants.<br />
The Advisory Council on Reactor<br />
Safeguards (ACRS) and the U.S. NRC<br />
have scrutinized the AP1000 Passive<br />
Safety Systems and ruled that they meet<br />
the U.S. NRC core cooling criteria, and<br />
other safety criteria such as Three Mile<br />
Island lessons learned.<br />
2. How has AP1000 improved the<br />
reactor water chemistry to minimize<br />
affect on the fuel and on reactor<br />
internals during normal operation and<br />
during accident conditions<br />
Zinc addition; a soluble zinc<br />
compound is added to the coolant as a<br />
means to reduce radiation fields within<br />
the primary system and to reduce the<br />
potential for crud-induced power shift<br />
(CIPS). The zinc used may be either<br />
natural zinc or zinc depleted of 64Zn.<br />
3. What fuel and fuel cladding material<br />
design enhancements have been made<br />
in AP1000 to ensure minimum damage<br />
of the fuel during normal operation, and<br />
during accident scenarios<br />
The use of ZIRLO cladding material;<br />
ZIRLO cladding material combines<br />
neutron economy (low absorption crosssection);<br />
high corrosion resistance to<br />
Responses to questions by Newal<br />
Agnihotri, Editor of <strong>Nuclear</strong> <strong>Plant</strong><br />
<strong>Journal</strong>.<br />
Ed Cummins<br />
Ed Cummins has spent his 32-year<br />
Westinghouse career in a variety of<br />
assignments in project management,<br />
coolant, fuel, and fission products; and<br />
high strength and ductility at operating<br />
temperatures. ZIRLO is an advanced<br />
zirconium based alloy that has the same<br />
or similar properties and advantages as<br />
Zircaloy-4 and was developed to support<br />
extended fuel burn up.<br />
Regarding accident scenarios,<br />
the AP1000 meets the U. S. NRC<br />
deterministic-safety and probabilistic-risk<br />
criteria with large margins. The safety<br />
analysis is documented in the AP1000<br />
Design Control Document (DCD) and<br />
Probabilistic Risk Assessment (PRA).<br />
Results of the PRA show a very low core<br />
damage frequency (CDF) that is 1/100 of<br />
the CDF of currently operating plants.<br />
4. What innovative fuel cycles have<br />
been used in AP1000 to maximize fuel<br />
effi ciency<br />
The AP1000 is designed to use an 18<br />
month or 16/20 month alternating cycle<br />
for optimum economics.<br />
5. How has AP1000 ensured a longer<br />
cable life to ensure a longer plant life<br />
The AP1000 instrumentation and<br />
control systems are designed in accordance<br />
with guidance provided in applicable<br />
portions of the following and<br />
other related standards: IEEE 383-1974,<br />
engineering management and new plant<br />
design.<br />
In March of 2000, Westinghouse initiated<br />
development of the AP1000 plant<br />
designed to be competitive with natural<br />
gas fi red combined cycle plants. He<br />
is currently Vice President, <strong>Nuclear</strong><br />
Power <strong>Plant</strong> Regulatory Affairs and<br />
Standardization, responsible for the<br />
licensing and commercialization of the<br />
AP1000.<br />
Mr. Cummins holds a Bachelor of<br />
Science Degree from the U.S. Naval<br />
Academy, a Master of Science Degree<br />
in Engineering Applied Science from<br />
the University of California, Davis,<br />
Livermore and a Master of Business<br />
Administration from Duquesne<br />
University.<br />
“IEEE Standard for Type Test of Class IE<br />
Electric Cables, Field Splices, and Connections<br />
for <strong>Nuclear</strong> Power Generating<br />
Stations.”<br />
6. What is the plant life of AP1000<br />
The AP1000 has a 60 year design<br />
life.<br />
7. What enhancements have been made<br />
in the control station design to ensure<br />
improved human-system interface<br />
Use of digital Instrumentation and<br />
Control systems.<br />
8. How has the current instrumentation<br />
and control system in AP1000<br />
been upgraded from the previous<br />
Westinghouse designs to ensure a<br />
reliable plant operation with longer<br />
plant life<br />
Use of digital Instrumentation and<br />
Control systems with rigorous adherence<br />
to NRC developed Human Factors<br />
Engineering guidance.<br />
9. How has information technology<br />
been used to survey and self-diagnose<br />
problems in the systems, structure, and<br />
components in AP1000<br />
Design Reliability Assurance Program<br />
(D-RAP); the AP1000 D-RAP is<br />
32 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
implemented as an integral part of the<br />
AP1000 design process to provide confidence<br />
that reliability is designed into the<br />
plant and that the important reliability<br />
assumptions made as part of the AP1000<br />
probabilistic risk assessment (PRA) will<br />
remain valid throughout plant life. The<br />
PRA quantifies plant response to a spectrum<br />
of initiating events to demonstrate<br />
the low probability of core damage and<br />
resultant risk to the public. PRA input<br />
includes specific values for the reliability<br />
of the various structures, systems, and<br />
components in the plant that are used to<br />
respond to postulated initiating events.<br />
10. How does AP1000 handle unstable<br />
and disruptive phenomena, such as<br />
water hammer<br />
The AP1000 is designed to minimize<br />
phenomena such as water hammer by<br />
incorporating industry lessons learned.<br />
The layout of the startup feed water<br />
piping and the main feed water line<br />
include features to minimize the potential<br />
for water hammer.<br />
The potential for water hammer,<br />
stratification, and striping is additionally<br />
reduced by the use of separate startup<br />
feed water piping and nozzles for each<br />
steam generator. The startup feed water<br />
nozzle is located at an elevation that is the<br />
same as the main feed water nozzle and<br />
is rotated circumferentially away from<br />
the main feed water nozzle. A startup<br />
feed water spray system independent<br />
of the main feed water feed ring is used<br />
to introduce startup feed water into the<br />
steam generator.<br />
11. How does the economy of AP1000<br />
compare with its previous designs<br />
The AP1000 is designed to be<br />
simpler, with less systems and equipment,<br />
and thus more economic.<br />
12. What enhancements have been<br />
made in the designs and construction of<br />
AP1000 to control fi re and smoke in the<br />
plant affecting safety critical systems<br />
Separation and fire areas. As<br />
presented in the AP1000 Design Control<br />
Document (DCD); fire areas are three<br />
dimensional spaces designed to contain<br />
a fire that may exist within them. They<br />
are separated by fire barriers, fire barrier<br />
penetration protection, and other devices,<br />
such as those within the heating and air<br />
conditioning ducts that isolate a fire to<br />
within the fire area.<br />
13. What enhancements have been made<br />
in the steam generator to ensure a longer<br />
plant life<br />
Use of Alloy 690 tubes; Nickelchromium-iron<br />
alloy in various forms is<br />
used for parts where high velocities could<br />
otherwise lead to erosion/corrosion to<br />
help increase component life.<br />
14. How long are the AP1000 steam<br />
generators expected to last<br />
The AP1000 plant is being designed<br />
to meet the ALWR utility requirements<br />
specified in Volume III of the ALWR<br />
Utility Requirements Document (URD).<br />
The URD states that for <strong>Plant</strong> Design<br />
Life, “The plant shall be designed to<br />
operate for 60 years without necessity<br />
for an extended refurbishment outage.<br />
The plant shall be designed to permit<br />
expeditious component replacement for<br />
obsolescence and failure over a lifetime<br />
of 60 years.”<br />
15. Does AP1000, having a passive<br />
safety system, still need an onsite and<br />
offsite emergency power<br />
No, not for safety. To minimize the<br />
challenges to the passive safety systems,<br />
the AP1000 design does include nonsafety<br />
connections to the site power grid<br />
and 2 non-safety diesel generators.<br />
Contact: Scott Shaw, Westinghouse<br />
<strong>Nuclear</strong>, 4350 Northern Pike, Monroeville,<br />
PA 15146; telephone: (412) 374-6737,<br />
email: shawsa@westinghouse.com. <br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>’s<br />
Product & Service Directory 2009<br />
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are not listed in the 2008 Directory may register<br />
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<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 33
A Market-Ready Design<br />
By Ken Petrunik, Atomic Energy of<br />
Canada Limited.<br />
Background<br />
Atomic Energy of Canada Limited’s<br />
(AECL’s) newest CANDU ® (CANada<br />
Deuterium Uranium) reactor, the ACR-<br />
1000 ® (Advanced CANDU Reactor ® ),<br />
is a 1200 MWe-class Generation III+<br />
nuclear power plant with a 60-year design<br />
life, including a mid-life pressure-tube<br />
replacement. It is a light-water-cooled,<br />
heavy-water-moderated pressure-tube<br />
reactor, with low-enriched uranium fuel<br />
(LEU), which has evolved from the<br />
well-established CANDU line. It retains<br />
proven CANDU design features while<br />
incorporating innovations and state-ofthe-art<br />
technologies to enhance safety,<br />
operation, maintenance, performance and<br />
economics.<br />
A key strategy in designing the ACR-<br />
1000 was to expand the Instrumentation<br />
and Control (I&C) and Information<br />
Technology (IT) systems by designingin<br />
and integrating operations and<br />
maintenance (O&M) functions. SMART<br />
CANDU modules allow on-line health<br />
monitoring of systems and components.<br />
Maximum use of modularization and<br />
‘open-top’, parallel construction—which<br />
have already been demonstrated at the<br />
Qinshan Phase III CANDU units, both<br />
delivered under budget and ahead of<br />
schedule—are key to AECL’s ACR-1000<br />
new-build project model.<br />
AECL is currently focusing on nearterm<br />
opportunities to build ACR-1000<br />
plants in Canada. CANDU reactors, now<br />
operating successfully on four continents,<br />
have already demonstrated that the<br />
technology can be easily localized in<br />
other countries—due to a core comprised<br />
of a large number of small, identical fuel<br />
channel components. Recent offshore<br />
new-build projects have also proven that<br />
nuclear power plants can be built on time<br />
and on budget.<br />
Responses to questions by Newal<br />
Agnihotri, Editor of <strong>Nuclear</strong> <strong>Plant</strong><br />
<strong>Journal</strong>.<br />
Ken Petrunik<br />
Ken Petrunik, PhD is President,<br />
CANDU Reactor Division, Atomic<br />
Energy of Canada Limited (AECL).<br />
He also holds the AECL corporate<br />
position of Executive Vice-President<br />
and Chief Operating Offi cer. Dr<br />
Petrunik has spent more than 30<br />
1. How do the economics of ACR-1000<br />
compare with those of other Generation<br />
III+ reactors<br />
The ACR-1000 has evolved from<br />
the CANDU 6 design, and has attractive<br />
economics. It is designed to achieve lower<br />
specific capital cost, shorter construction<br />
schedule, higher plant capacity factor,<br />
lower operating cost, increased operating<br />
life and enhanced ease of operation.<br />
The ACR-1000’s economics are fully<br />
competitive with numbers published in<br />
the literature for other Generation III+<br />
designs.<br />
2. What is the status of ACR-1000<br />
licensing<br />
ACR technology had extensive<br />
pre-project review from the United<br />
States <strong>Nuclear</strong> Regulatory Commission<br />
(USNRC, 2001-02), the Canadian <strong>Nuclear</strong><br />
Safety Commission (CNSC, 2003-06)<br />
and, more recently, by the UK regulator<br />
(2007-08). Findings were positive. On<br />
April 1, 2008, AECL and CNSC signed<br />
a Memorandum of Understanding for<br />
performing a pre-project design review<br />
on ACR-1000, which will be conducted<br />
in two phases through 2009.<br />
The key submission for this preproject<br />
design review is the 3,000-page,<br />
years with AECL, leading the company<br />
through design, licensing, construction<br />
and commissioning of CANDU power<br />
stations around the world. The teams he<br />
assembled were instrumental in bringing<br />
in all of our recent new-build reactor<br />
projects into service on time or ahead<br />
of schedule, and on budget. Dr Petrunik<br />
introduced open top construction and<br />
modularization technology to CANDU<br />
power plants, and also led the fi rst use<br />
in Canada of authorized electronic<br />
documentation for AECL projects, the<br />
model for future projects. More recently,<br />
in his role of Chief Operating Offi cer,<br />
he has further deepened his already<br />
excellent relationships with customers<br />
and governments, working to develop<br />
markets for AECL’s market-ready ACR-<br />
1000 and world leading CANDU 6.<br />
20-chapter, Generic Safety Case Report<br />
(GSCR), submitted on June 30, 2008.<br />
This report provides an integral picture<br />
of the ACR-1000 safety design and<br />
bounding safety analysis. Being in the<br />
format of a Preliminary Safety Analysis<br />
Report (PSAR), it is comprehensive and<br />
self-contained.<br />
3. Is AECL exploring options to<br />
manufacture reactor pressure vessels<br />
given the fact that there are very few<br />
manufactures in the world to meet the<br />
required demand<br />
The ACR-1000 and all CANDU<br />
reactors are pressure-tube reactors. Thus,<br />
they do not have high-pressure reactor<br />
vessels typical of light water reactors,<br />
or the associated supply difficulty with<br />
heavy forgings. The only large forgings<br />
for ACR-1000 are related to the steam<br />
generators, for which there are alternate<br />
suppliers. There is a robust supply<br />
chain for pressure tubes with alternative<br />
suppliers in North America and overseas,<br />
with recent supply availability clearly<br />
demonstrated in refurbishment projects.<br />
4. What fuel and fuel cladding material<br />
design enhancements have been made in<br />
(Continued on page 36)<br />
34 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
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Enriching the future
A Market...<br />
Continued from page 34<br />
ACR-1000 to ensure minimum damage<br />
of the fuel during normal operation, and<br />
during accident scenarios<br />
Reference fuel for the ACR-1000<br />
is the 43-element CANFLEX-ACR<br />
(CANDU FLEXible) bundle, which<br />
incorporates 42 elements with 11.5 mm<br />
OD, 2.4% enriched LEU and one 20-mmdiameter<br />
central element with burnable<br />
neutron absorbers (BNA). Sheath material<br />
is Zircaloy-4.<br />
ACR-1000 fuel acceptance criteria<br />
for normal operation were used to<br />
systematically evaluate any potential<br />
damage mechanisms that could affect<br />
fuel robustness. This ensures that fuel<br />
cannot be damaged in fulfilling design<br />
requirements for normal operation.<br />
Design changes, listed below, help to<br />
minimize fuel damage during normal<br />
operation and accidents:<br />
• More highly subdivided 43-element<br />
CANFLEX-ACR fuel bundle,<br />
lowering fuel element ratings and<br />
reducing the power-related damage<br />
mechanisms<br />
• Fuel pellet geometry optimized to<br />
minimize sheath strains and fission<br />
gas pressure<br />
• CANLUB interlayer thickness<br />
increased to improve resistance to<br />
damage due to power ramp failures<br />
• Fuel sheath thickness defined to<br />
maintain its intrinsic collapsibility<br />
• Fuel bundle endplate geometry<br />
modified to improve irradiated fuel<br />
bundle strength during refuelling<br />
operations<br />
• Use of CANFLEX-ACR fuel<br />
bundle with AECL’s patented flowenhancing<br />
sheath appendages,<br />
providing increased margin to dryout<br />
in postulated accident conditions<br />
• Central fuel element containing<br />
BNAs to control the coolant void<br />
reactivity, thus minimizing potential<br />
for fuel damage in the case of a<br />
postulated large-break loss-ofcoolant<br />
accident (LOCA)<br />
5. How does ACR-1000 minimize<br />
damage to the fuel in case of a loss-ofcoolant<br />
accident<br />
The ACR-1000 design has<br />
incorporated some new features to<br />
minimize fuel damage that might occur<br />
during a postulated large-break Loss-of-<br />
Coolant Accident:<br />
• Reduced core lattice pitch (distance<br />
between the fuel channels), reducing<br />
the coolant void reactivity (CVR)<br />
during a postulated large-break<br />
LOCA<br />
• Increased calandria-tube diameter,<br />
resulting in reduced moderatorto-fuel<br />
ratio, which reduces the<br />
moderator volume and, hence,<br />
reduces the CVR<br />
• Enhanced fuel design, with the centre<br />
element containing zirconia with<br />
BNAs, further reducing the CVR<br />
All of the above features combine<br />
to give a small negative CVR value for<br />
nominal end-of-life conditions, such that<br />
the power transient during a large-break<br />
LOCA is benign.<br />
Changes to the fuel design make the<br />
fuel less susceptible to failure during a<br />
LOCA. As above (Question 4), the more<br />
subdivided CANFLEX-ACR fuel bundle<br />
lowers fuel element ratings and reduces<br />
the power-related damage mechanisms<br />
while fuel pellet geometry minimizes<br />
sheath strains and fission-gas pressure,<br />
ACR-1000 Four Unit Layout<br />
reducing the likelihood of fuel failures<br />
during power transients.<br />
Finally, the ACR-1000 design has<br />
retained the two independent fast-acting<br />
reactor shutdown systems, which are the<br />
well-established means of limiting the<br />
reactivity transient during a postulated<br />
large-break LOCA in traditional CANDU<br />
reactors. As a result of all of these<br />
enhancements, calculations show that<br />
during a postulated large-break LOCA,<br />
there will be no fuel failures in the ACR-<br />
1000 reactor design.<br />
6. What innovative fuel cycles have<br />
been used in ACR-1000 to maximize fuel<br />
effi ciency and to minimize concerns of<br />
proliferation<br />
The reference fuel for the ACR-<br />
1000 has a uniform 2.4% enrichment.<br />
The ACR-1000 uses the advanced<br />
CANFLEX ® fuel bundle, developed<br />
as the optimal carrier for CANDU<br />
advanced fuel cycles. Development is<br />
underway to increase enrichment and<br />
burnup, to further improve economics.<br />
In addition, Recovered Uranium (RU)<br />
from conventional reprocessing can be<br />
burned efficiently in the ACR-1000, with<br />
the addition of fissile LEU or plutonium<br />
(Pu). The reactor can operate with a<br />
full core of 2.4% LEU, or with RU plus<br />
fissile to 2.4% Heavy Element (HE). The<br />
on-power refuelling capability permits<br />
switching back and forth between the two<br />
fuel types, without any hardware changes<br />
to the safety/control systems.<br />
Additionally, spent ACR-1000 fuel<br />
with a residual fissile content of about<br />
1%, opens the possibility of its re-use<br />
in existing CANDU reactors. The ACR-<br />
1000 is also amenable to thorium fuel<br />
cycles. The simplest case, feasible in the<br />
short term, is the Once-Through Cycle<br />
(OTT). This is easy to implement, with<br />
no reprocessing required, to achieve a<br />
burnup of about 21,000 MWd/TeHE.<br />
This cycle also creates a “reservoir” of<br />
Uranium-233 (233U) for future use. In<br />
the longer term, a closed-cycle option<br />
offers burnups to 40,000 MWd/TeHE.<br />
Spent fuel is reprocessed to recycle 233U,<br />
and burnup can be tailored by adding Pu<br />
to fresh bundles.<br />
Proliferation-resistance results from<br />
a combination of technical design features,<br />
operational modalities, institutional<br />
arrangements and safeguards measures.<br />
In CANDU technology, these features are<br />
strongly linked and self-enforced, with<br />
the result that their combination is greater<br />
than the sum of the parts. CANDU technology<br />
has always incorporated intrinsic<br />
proliferation-resistance features—derived<br />
from the fundamental physics of naturaluranium<br />
or LEU-fuelled reactors.<br />
While these inherent barriers<br />
minimize the attractiveness of CANDU<br />
technology as a target for proliferation,<br />
external measures provide verification<br />
(Continued on page 38)<br />
36 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
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Continued from page 36<br />
and deterrence through timely detection.<br />
International Atomic Energy Agency<br />
(IAEA) safeguards have been successfully<br />
incorporated in CANDU reactors for<br />
decades, and have evolved over time.<br />
7. What enhancements have been made<br />
in the control station design to ensure<br />
improved human-system interface<br />
Improvements in computer technology—particularly<br />
digital communications<br />
and distributed systems—provided<br />
a significant opportunity to improve the<br />
human system interface in the ACR-1000<br />
Main Control Room (MCR), as follows:<br />
• Enlarged main operator console,<br />
with more computer display stations,<br />
allowing for control and monitoring<br />
at the console instead of at the standto-operate<br />
panels of the past designs;<br />
automating standard manual control<br />
sequences reduces the chance of<br />
human error.<br />
• Improved main operator console and<br />
shift interrogation console, providing<br />
work stations for monitoring safety<br />
and production functions, and for<br />
administrative functions; large<br />
work areas for paperwork and<br />
documentation with easy-access<br />
document storage<br />
• Large-screen displays and a small<br />
section of hardwired backup panels<br />
providing plant overview information<br />
for situation awareness.<br />
• Automated safety system testing,<br />
which can be initiated from the<br />
main operator console and reduces<br />
operator workload<br />
• Highly-effective CANDU Alarm<br />
Message List System (CAMLS),<br />
filtering the alarm message stream to<br />
ensure only pertinent alarms appear<br />
• Seismically-qualified MCR, allowing<br />
operator to remain there following<br />
a seismic event and handle it using<br />
familiar interfaces<br />
8. How has information technology<br />
been used to survey and self-diagnose<br />
problems in the systems, structure, and<br />
components in ACR-1000 How does<br />
this ensure reliable operation and longer<br />
plant life<br />
From smart sensors to increased<br />
process and large equipment diagnostic<br />
monitoring and assessments, the new<br />
digital technologies will enhance<br />
ACR-1000 diagnostic and prognostic<br />
or condition-monitoring capabilities,<br />
including smart sensors and control<br />
elements, vibration-monitoring and<br />
neutronics analysis from the CANDU<br />
6 reference design. ACR 1000 will be<br />
incorporating significantly different<br />
designs and levels of integration than the<br />
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areas as well, will greatly impact the<br />
functionality performance of the plant<br />
industrial network systems:<br />
Network Design Topologies<br />
• The new “distribution system”<br />
• Operator Support<br />
• Operator rounds Logs<br />
• Video surveillance<br />
(Continued on page 40)<br />
CABLES AND H2/O2 MONITORING SYSTEMS<br />
www.meggittsafety.com<br />
38 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
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Continued from page 38<br />
Health Physics Support<br />
• Personnel tracking Dosimetry Video<br />
surveillance<br />
• Portable monitoring<br />
Communications<br />
• Paging Telephone Cell phone<br />
• Radio<br />
Security<br />
• Access Personnel tracking<br />
• Video surveillance<br />
AECL has developed a suite of<br />
Operations and Maintenance (O&M)<br />
support applications, known as<br />
SMART CANDU, to assist the O&M<br />
organization. These track the health<br />
of key systems and components and<br />
provide diagnostic tools to identify and<br />
correct problems before they result in a<br />
loss of performance. SMART CANDU<br />
applications combine process, chemistry<br />
and inspection data to provide up-todate<br />
assessments of the current status<br />
of key plant systems and components.<br />
For example, ChemAND (Chemistry<br />
Analysis and Diagnostic) monitors water<br />
chemistry and ThermAND monitors heat<br />
transfer systems and components.<br />
Data are stored in a Life-of-<strong>Plant</strong><br />
Historian, where they can be easily<br />
retrieved and displayed to compare the<br />
current plant status with past behavior.<br />
The impact of plant operating conditions<br />
on the future performance of critical<br />
components in the system can be further<br />
assessed using one of the embedded<br />
analytical models that are interfaced<br />
with the plant data. This enables<br />
engineering staff to track, for example,<br />
thermal performance, fatigue usage, the<br />
performance of pump/motor sets and<br />
the results of inspection campaigns, and<br />
to predict the impact of plant operating<br />
conditions on steam generator fouling,<br />
activity transport and steam generator<br />
chemistry.<br />
Field tests at domestic CANDU<br />
utilities have demonstrated that these<br />
features greatly reduce the time required<br />
to diagnose problems and allow plant staff<br />
to operate in a more proactive mode.<br />
Thus, these new tools help to optimize<br />
ongoing operation and maintenance while<br />
allowing informed decision-making and<br />
planning for the future. This ensures<br />
reliable plant operation and leads to<br />
longer plant-life.<br />
9. How has the current instrumentation<br />
and control system in the ACR-1000<br />
been upgraded from the previous<br />
AECL designs to ensure reliable plant<br />
operation with longer plant life<br />
The ACR 1000 plant design uses<br />
a distributed control system (DCS) to<br />
perform plant monitoring and control<br />
functions previously implemented using<br />
centralized digital control computers,<br />
analog control devices and relay logic.<br />
The control strategies for the DCS control<br />
programs are based on previous CANDU<br />
designs but are implemented on a new<br />
hardware platform taking advantage of<br />
advances in computer technology and<br />
supplementing this process with new<br />
techniques and analyses. These new<br />
techniques allow system designers to<br />
take advantage of new features possible<br />
in a DCS application, and ensure the<br />
DCS achieves significant capital and<br />
operating cost reductions and improved<br />
safety through high operational and<br />
safety reliability, reduced I&C system<br />
complexity.<br />
In previous CANDUs, plant control<br />
was performed by centralized control<br />
computers (DCC), analog devices<br />
and relay logic. System control was<br />
performed by dual redundant computers<br />
that executed a set of control programs for<br />
monitoring, annunciation, and control of<br />
plant systems. In a second level, control<br />
devices such as analog controllers and<br />
programmable logic controllers (PLCs)<br />
handled lower-level control functions.<br />
The control and instrumentation<br />
design used in the ACR 1000 plant has<br />
separated the computer control system<br />
from the plant information system in<br />
recognition of the fact that the controls are<br />
less subject to change and more sensitive<br />
to the risk of change. The computer<br />
information systems and human-machine<br />
interaction systems, on the other hand,<br />
must be flexible, expandable and easy to<br />
upgrade to exploit evolving technology.<br />
The primary advantages of this<br />
evolutionary DCS design are as follows:<br />
• The significant elimination of C&I<br />
hardware components, wiring,<br />
cabling and wire terminations<br />
achieves significant capital and<br />
operating cost savings<br />
• <strong>Plant</strong> safety will be enhanced<br />
because the distributed architecture<br />
of the group control functions<br />
makes simultaneous loss of all these<br />
functions due to component failures<br />
incredible.<br />
• Improved software design tools,<br />
software reviewability and simplified<br />
operating environment will contribute<br />
to reduced software errors.<br />
• Elimination of intrusive hardware<br />
maintenance activities to modify<br />
functionality will also improve plant<br />
safety.<br />
The DCS design concept provides<br />
very high reliability and fault tolerance,<br />
minimizing the need to provide separate<br />
local control or hardwired backup. Faulttolerant<br />
features include channelization,<br />
redundancy and fail-safe outputs. Use of<br />
a single hardware platform for high- and<br />
low-level controls reduces maintenance<br />
errors by ensuring familiarity of the<br />
maintenance personnel with a single<br />
control system.<br />
<strong>Digital</strong> protection systems first<br />
formed part of the CANDU 6 product.<br />
The systems were called Programmable<br />
<strong>Digital</strong> Comparators (PDCs). The PDCs<br />
formed part of the shutdown systems in<br />
the reactors. They provided much of the<br />
process-related reactor trip coverage,<br />
increasing the potential for more complex<br />
trips. Using computer capabilities, it<br />
was possible to add self-checking and<br />
monitoring to the equipment. The<br />
actuation of the safety functions for the<br />
two shutdown systems in ACR-1000 will<br />
also be software-based, using proven<br />
methods from past and current reactor<br />
projects.<br />
For further information on the ACR-<br />
1000, see <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong><br />
Volume 26 No.1, January-February<br />
2008.<br />
*CANDU ® , Advanced CANDU<br />
Reactor ® , ACR-1000 ® and CANFLEX ®<br />
are registered trademarks of Atomic<br />
Energy of Canada Limited (AECL).<br />
SMART CANDU, CANFLEX-<br />
ACR and ChemAND are also<br />
AECL trademarks.<br />
Contact: Heather Smith, AECL,<br />
2251 Speakman Drive, Mississauga,<br />
Ontario. L5K 1B2 Canada; telephone:<br />
(905) 823-9060 ext 7541, fax: (905) 403-<br />
7565, email: smithh@aecl.ca. <br />
40 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
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Generation IV Advanced <strong>Nuclear</strong><br />
Energy Systems<br />
By Jacques Bouchard, French<br />
Commissariat à l’Energie Atomique,<br />
France and Ralph Bennett, Idaho<br />
National Laboratory.<br />
Generation IV first stepped on to the<br />
international scene in 2001 when nine<br />
countries joined together on a mission to<br />
develop and implement the next wave of<br />
safe and sustainable nuclear reactors, and<br />
created the Generation IV International<br />
Forum (GIF) to oversee it. Seven years of<br />
important changes in energy, environment<br />
and public acceptance have given the<br />
GIF a renewed sense of purpose. During<br />
those years an R&D program, with a<br />
framework covering technical and legal<br />
aspects, was created to meet the challenge<br />
of expanding nuclear energy throughout<br />
the 21st century.<br />
The GIF countries’ pledge<br />
to cooperate comes at a particularly<br />
urgent time. Worldwide greenhouse gas<br />
emissions grew 70 percent between 1970<br />
and 2004, and if current energy practices<br />
remain unchecked such emissions<br />
will have a devastating effect on the<br />
planet. The dramatic effect on climate<br />
of increased carbon emissions poses a<br />
problem that transcends national borders<br />
and politics. Safe, efficient nuclear<br />
energy must be a part of a serious effort<br />
to stabilize greenhouse gas levels.<br />
Making a significant difference<br />
in carbon emissions would require<br />
a large expansion of nuclear power.<br />
According to Princeton University’s<br />
Carbon Mitigation Initiative, increasing<br />
the number of nuclear power plants to<br />
1000 worldwide—more than double<br />
what it is today—could avoid one billion<br />
tons of carbon emissions per year by<br />
2055. Generation IV aims to develop<br />
reactors and their associated fuel cycles<br />
that assure their long term sustainability<br />
and allow them to address more than just<br />
electricity generation, thereby setting the<br />
stage for sustained expansion through the<br />
century.<br />
Jacques Bouchard<br />
Jacques Bouchard is Special Adviser<br />
to the Chairman of the French<br />
Commissariat à l’Energie Atomique.<br />
Mr. Bouchard has also served as<br />
chairman of the Generation IV<br />
International Forum since 2006.<br />
A Framework for R&D<br />
Collaboration<br />
The effort towards a new generation<br />
of nuclear energy systems started in July<br />
2001, when nine countries signed the GIF<br />
Charter. In doing so, France, Argentina,<br />
Brazil, Canada, Japan, the Republics of<br />
Korea and South Africa, the United States<br />
and United Kingdom signaled their mutual<br />
interest in new nuclear systems. Since<br />
then, Switzerland, Euratom (representing<br />
the nations of the Euratom Treaty), China<br />
and Russia have all become members of<br />
the GIF.<br />
To date, nine of the members have<br />
also acceded to a Framework Agreement, 1<br />
which allows its signatories to formally<br />
participate in the development of<br />
Generation IV nuclear systems. Under<br />
that agreement, System Arrangements<br />
provide the framework for collaboration<br />
on each type of reactor. These<br />
arrangements allow for cooperation<br />
with industry, academia and even other<br />
governments to accomplish the R&D.<br />
Each member finances its own research<br />
Ralph Bennett<br />
Ralph Bennett, PhD, is Director<br />
of International and Regional<br />
Partnerships, Idaho National<br />
Laboratory. In 1979, he earned a<br />
Ph.D. in nuclear engineering at MIT.<br />
He is also the Technical Director of the<br />
Generation IV International Forum.<br />
and development, chooses which systems<br />
it will work on, and shares and protects<br />
the intellectual property they develop<br />
collaboratively.<br />
<strong>Nuclear</strong> Power through<br />
the Generations<br />
The conventional paradigm for<br />
the history of nuclear reactors has been<br />
to separate different types of nuclear<br />
designs into “generations.” Generation I,<br />
dating from the 1950s and 60s, includes<br />
early prototypes in a number of countries.<br />
Generation II, the first commercial power<br />
plants, date from the 70s and 80s and<br />
include Pressurized Water Reactors<br />
and Boiling Water Reactors—designs<br />
generally utilizing water for coolant and<br />
slightly enriched uranium for fuel, almost<br />
all of which are still operating today.<br />
Most nuclear power plants being built<br />
now are categorized as Generation III—<br />
water-cooled reactors with more refined<br />
designs than their Generation II ancestors.<br />
This third generation has evolved to<br />
be both safer and more efficient, but<br />
42 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
is nevertheless focused on electricity<br />
generation and only a limited recycle of<br />
the plutonium generated during one cycle<br />
through the core.<br />
Worldwide projections of increased<br />
demand for electricity and new<br />
imperatives to reduce carbon emissions<br />
have lent special urgency to the promise<br />
of next generation systems. New<br />
markets, technical innovations and a<br />
rising acceptance of nuclear energy<br />
have produced the conditions needed<br />
for a revolution in nuclear technology.<br />
Economic competitiveness, improved<br />
safety, conservation of uranium resources<br />
and minimalization of waste, increased<br />
physical protection of the plants and<br />
added resistance to threats of nuclear<br />
proliferation are the new challenges<br />
posed to Generation IV reactors, which<br />
will ensure the sustainable development<br />
of nuclear energy.<br />
Goals for Generation IV<br />
Six different Generation IV nuclear<br />
reactor systems are currently being<br />
advanced. They were identified by an<br />
international group of over 100 experts<br />
who examined more than 130 proposals<br />
sent by specialists from around the<br />
world. The GIF took a top-down<br />
approach to choosing which designs were<br />
most promising versus the challenges<br />
of sustainability, safety, economics,<br />
proliferation resistance and physical<br />
protection. Further considerations<br />
included estimated R&D costs and time<br />
horizons. Though the final six systems<br />
selected have different strengths, each<br />
one was chosen for its unique potential<br />
to contribute to the new face of nuclear<br />
energy and advance toward the following<br />
eight goals:<br />
Sustainability–1: Generation IV<br />
nuclear energy systems will provide<br />
sustainable energy generation that meets<br />
clean air objectives and promotes longterm<br />
availability of systems and effective<br />
fuel utilization for worldwide energy<br />
production.<br />
Sustainability–2: Generation IV<br />
nuclear energy systems will minimize<br />
and manage their nuclear waste and<br />
notably reduce the long-term stewardship<br />
burden in the future, thereby improving<br />
protection for the public health and the<br />
environment.<br />
Economics–1: Generation IV nuclear<br />
energy systems will have a clear lifecycle<br />
cost advantage over other energy<br />
sources.<br />
Economics–2: Generation IV nuclear<br />
energy systems will have a level of<br />
financial risk comparable to other energy<br />
projects.<br />
Safety and Reliability–1: Generation<br />
IV nuclear energy systems operations<br />
will excel in safety and reliability.<br />
Safety and Reliability–2: Generation<br />
IV nuclear energy systems will have a<br />
very low likelihood and degree of reactor<br />
core damage.<br />
Safety and Reliability–3: Generation<br />
IV nuclear energy systems will eliminate<br />
the need for offsite emergency response.<br />
Proliferation Resistance and<br />
Physical Protection–1: Generation IV<br />
nuclear energy systems will increase the<br />
(Continued on page 44)<br />
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<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 43
Generation IV...<br />
Continued from page 43<br />
assurance that they are a very unattractive<br />
and the least desirable route for diversion<br />
or theft of weapons-usable materials, and<br />
provide increased physical protection<br />
against acts of terrorism.<br />
A short overview of each system<br />
follows:<br />
Sodium-Cooled Fast Reactor<br />
(SFR): The GIF is currently devoting<br />
much of its effort to this system. It<br />
uses liquid sodium for coolant, thereby<br />
gaining a high power density and lower<br />
coolant volume fraction. It features a<br />
closed fuel cycle, which is needed for fuel<br />
breeding and/or actinide management.<br />
The layout is flexible, with a pool layout<br />
(shown) or a compact loop layout. Either<br />
could be adjusted to produce a small-,<br />
medium- or large-sized reactor. The<br />
SFR can be economically competitive in<br />
electricity markets with innovations to<br />
reduce capital costs. The SFR is more<br />
efficient than thermal-spectrum reactors<br />
with open fuel cycles, with its potential<br />
to use both fissile and fertile isotopes<br />
of uranium. GIF has been taking a<br />
streamlined approach to developing SFR,<br />
by building upon technologies that are<br />
already being deployed throughout the<br />
world and advancing their performance.<br />
Progress in developing the SFR is well<br />
underway, with advances being made in<br />
fuel technology in France, compact heat<br />
exchangers in the United States, and<br />
design innovations underway in Japan 2 .<br />
Very High Temperature Reactor<br />
(VHTR): The GIF is also devoting much<br />
of its effort to this system. It is a heliumcooled<br />
thermal reactor that can achieve<br />
an outlet temperature approaching 900<br />
degrees Celsius. The ceramic fuel of<br />
the VHTR has a high degree of passive<br />
safety, and the high temperature gives it<br />
a high thermal efficiency approaching<br />
50%. The high temperature also allows<br />
the VHTR to be applied to hydrogen<br />
production and other high temperature<br />
process heat applications, as well as low<br />
temperature heat applications such as<br />
water desalination, thereby addressing<br />
non-electric energy needs. The primary<br />
areas of research involve fuels, high<br />
temperature materials, and hydrogen<br />
production processes, and virtually all<br />
of the GIF members are collaborating on<br />
this system.<br />
Gas-Cooled Fast Reactor (GFR):<br />
A fast-spectrum thermal reactor using<br />
helium coolant with an outlet temperature<br />
of 850 degrees Celsius. It is attractive<br />
because of its high efficiency and<br />
minimization of transuranic waste.<br />
Supercritical Water Reactor<br />
(SCWR): The SCWR design uses<br />
water above its critical point condition<br />
(374°C, 22.1 MPa) as the coolant. This<br />
avoids the need for steam generators<br />
and considerably reduces the size of the<br />
turbine generator. It is a flexible design,<br />
configurable as a fast or thermal reactor.<br />
Its thermal efficiency may exceed 45%,<br />
and its lower capital cost favors the<br />
economical production of electricity.<br />
Lead-Cooled Fast Reactor (LFR):<br />
This fast reactor uses molten lead or<br />
lead/bismuth as a coolant and has a<br />
high degree of safety since the coolant<br />
is less chemically reactive than sodium.<br />
It operates at a temperature higher than<br />
the SFR, which may allow its use for<br />
44 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
hydrogen production as well as electricity.<br />
It has a closed fuel cycle.<br />
Molten Salt Reactor (MSR): This is<br />
an epithermal reactor design in which the<br />
fuel is actually dissolved in the coolant.<br />
Specifically, it uses liquid fluorides of<br />
uranium and plutonium for fuel, dissolved<br />
in fluorides of lithium, beryllium, sodium<br />
or other elements. The system provides<br />
for processing the wastes and adding new<br />
fuel online, which greatly reduces the<br />
fissile material inventory and avoids the<br />
development and qualification of fuel and<br />
cladding.<br />
The Future<br />
Generation IV designs improve upon<br />
current reactors in several ways. Four<br />
of the designs are fast reactors, allowing<br />
the reactors to potentially exploit the<br />
full energy potential of uranium—both<br />
fissile and fertile isotopes. Generation<br />
III reactors extract energy from a much<br />
smaller fraction of uranium in the fuel,<br />
where as Generation IV reactors can<br />
extend the uranium resource by about a<br />
factor of 50 beyond this. Another option<br />
for Generation IV is to improve on current<br />
designs by recycling all actinides—not<br />
only the bred plutonium-239, but the<br />
other actinides found in the waste as<br />
well. This revolution in fuel utilization<br />
would also dramatically reduce the<br />
radiotoxicity and heat generated by the<br />
waste by transmuting it to shorter-lived<br />
fission products, thus making it easier to<br />
dispose.<br />
Several of the Generation IV designs<br />
are high-temperature reactors, which<br />
can generate not only electricity but<br />
also provide process heat for industrial<br />
purposes. Process heat has good potential<br />
for application to a wide range of<br />
industries, from petroleum refineries and<br />
chemical plants to large-scale hydrogen<br />
production potentially for revolutionizing<br />
transportation.<br />
One of the overarching goals of<br />
Generation IV technology, and one that<br />
is most appealing to the international<br />
community, is its potential to reduce<br />
carbon emissions. This will only be<br />
accomplished through considerable<br />
R&D, and for example, the GIF members<br />
collaborating on the SFR and VHTR have<br />
already jointly committed over $500M for<br />
the next five years. The GIF believes that<br />
Generation IV, through improved safety,<br />
economics, safety and proliferation<br />
resistance and physical protection, can<br />
help ensure nuclear energy’s long term<br />
expansion and sustained contribution to<br />
the world’s energy security.<br />
References<br />
[1] Generation IV International Forum,<br />
“Framework Agreement,” available<br />
at: http://www.gen-4.org/PDFs/<br />
Framework-agreement.pdf, 28 Feb<br />
2005.<br />
[2] Generation IV International Forum,<br />
“GIF Annual Report 2007,” available<br />
at: http://www.gen-4.org/PDFs/<br />
annual_report2007.pdf, Mar 2008.<br />
Contact: Ralph Bennett, Idaho<br />
National Laboratory, P.O. Box 1625,<br />
Idaho Falls, ID, 83415-3805; telephone:<br />
(208) 526-7708, fax: (208) 526-0876),<br />
email: Ralph.bennett@inl.gov. <br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 45
Innovative Reactor Designs<br />
A Report Based on the Recent<br />
Publications by International Atomic<br />
Energy Agency, Vienna, Austria.<br />
Introduction<br />
There is continuing interest in<br />
Member States in the development and<br />
application of small and medium sized<br />
reactors (SMRs). “Small” reactors are<br />
defined as those with an equivalent<br />
electric power less than 300 MW(e).<br />
“Medium sized” reactors are those with<br />
an equivalent electric power between<br />
300 and 700 MW(e). It is important that<br />
small or medium sized reactor does not<br />
necessarily mean small or medium sized<br />
nuclear power plant. Like any nuclear<br />
power plants, those with SMRs can be<br />
built several-at-a-site, or as twin units. In<br />
addition to this, innovative SMR concepts<br />
provide for power plant configurations<br />
with 2, 4, or more reactor modules [1, 2,<br />
and 3]. The units or modules could then<br />
be added incrementally in time taking<br />
benefits of the effects of learning, timing,<br />
construction schedule, and creating<br />
an attractive investment profile with<br />
minimum capital-at-risk.<br />
Opportunities for SMRs<br />
In the near term, deployment potential<br />
of the SMRs is based largely on their<br />
ability to fill niches where larger plants do<br />
not fit in, or to offer economic advantages<br />
related to incremental capacity increase.<br />
The applications could be industrial<br />
sites or population centres in remote<br />
off-grid locations, countries or country<br />
areas with small and medium electricity<br />
grids, investment and human resource<br />
conditions that benefit from incremental<br />
capacity addition or non-electrical<br />
applications that require proximity of a<br />
nuclear energy source to the process heat<br />
application plant [1].<br />
For the longer term, there is<br />
interest in innovative designs that<br />
promise improvements in safety,<br />
security, proliferation resistance, waste<br />
management, resource utilization,<br />
economics, product variety (e.g.<br />
desalinated seawater, process heat,<br />
district heat and hydrogen) and flexibility<br />
in siting and fuel cycles. Many innovative<br />
reactor designs have been proposed in the<br />
small-to-medium sized range, in many<br />
cases providing for multi-module plant<br />
configurations to achieve larger, often<br />
flexible, overall power station capacity<br />
[2, 3].<br />
Many of the niche advantages of SMRs<br />
are expected to be particularly attractive<br />
to some of the approximately 40 countries<br />
that have recently expressed interest in<br />
starting nuclear power programmes, for<br />
example, low investment increments and<br />
suitability for small grids. On the other<br />
hand, vendors in Argentina, China, India,<br />
Japan, the Republic of Korea, the Russian<br />
Federation, South Africa, and the USA<br />
are actively developing and promoting<br />
new SMR designs [2, 3].<br />
Progress toward<br />
Deployment<br />
For about a dozen of innovative SMR<br />
designs, current progress in developing<br />
the technology and finalizing the design<br />
suggests possible deployment within the<br />
next decade.<br />
Construction began in June 2006 in<br />
the Russian Federation on a pilot floating<br />
cogeneration plant of 300 MW(th)/70<br />
MW(e) with two water cooled KLT-40S<br />
reactors. Deployment is scheduled for<br />
2010.<br />
In July 2006, the Russian Federation<br />
and Kazakhstan created a joint venture to<br />
complete design development for a 350<br />
MW(e) VBER-350 reactor (basically a<br />
scaled-up version of the KLT-40S) for<br />
use in land-based co-generation plants<br />
[2]. The first-of-a-kind plant deployment<br />
is targeted in 2015 at the former BN-350<br />
site in Kazakhstan.<br />
Five integral PWR designs<br />
are in advanced design stages and<br />
commercialization could start around<br />
2015 [2, 3]: the 335 MW(e) IRIS design<br />
developed by International consortium<br />
led by Westinghouse of USA (currently<br />
co-owned by Toshiba Corp. of Japan) ; the<br />
330 MW(th) SMART design developed in<br />
the Republic of Korea for a co-generation<br />
plant; the prototype 27 MW(e) CAREM-<br />
25 developed in Argentina, for which<br />
construction in planned to be complete<br />
by 2011, and which is expected to further<br />
into commercial designs of 150 and<br />
300 MW(e); the 200 MW(th) NHR-200<br />
developed in China for district heating<br />
and other applications, both electrical<br />
and non-electrical; and the MASLWR<br />
of 45 MW(e) per module, developed in<br />
the USA, for multi-purpose applications<br />
and multi-modular plants of up to 540<br />
MW(e).<br />
The Advanced Heavy Water Reactor<br />
of 300 MW(e), developed in India for cogeneration<br />
plants, is considered to be built<br />
early in the next decade [2]. The reactor is<br />
being designed for operation with 233U-<br />
Pu-Th fuel and uses boiling light water<br />
coolant and heavy water moderator. All<br />
mentioned above SMRs provide for or<br />
do not exclude co-generation option<br />
with non-electric energy products being<br />
produced as well as the electricity.<br />
The 165 MW(e) PBMR, a high<br />
temperature gas cooled reactor with<br />
pebble bed fuel and direct gas turbine<br />
Brayton cycle, developed in South Africa,<br />
is- scheduled for demonstration at full<br />
size by 2012 [2]. Future configurations of<br />
this reactor will include 4 and 8-module<br />
plants. The 200 MW(e) per module<br />
HTR-PM, a high temperature gas cooled<br />
reactor with pebble bed fuel and indirect<br />
supercritical steam energy conversion<br />
cycle developed in China, is planned<br />
for a full size demonstration in 2013 [1,<br />
2]. Two-module plant configuration is<br />
foreseen for the commercial version of<br />
this reactor.<br />
Some small reactor designs<br />
incorporate an option of operation without<br />
on-site refuelling, which may help reduce<br />
the obligations of a user for spent fuel<br />
and waste management [3]. Several of<br />
such designs have a potential of being<br />
deployed as first-of-a-kind or prototype<br />
plants within the next decade [1, 3].<br />
These include [3] the ABV of 11 MW(e)<br />
and 8-year refuelling interval, which<br />
is an integral design PWR backed by<br />
(Continued on page 48)<br />
46 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
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Innovative Reactor...<br />
Continued from page 46<br />
marine reactor experience, and a couple<br />
of non water cooled reactors, which are<br />
the sodium cooled 4S reactor of 10-50<br />
MW(e) and 10-30 year refuelling interval,<br />
developed in Japan, and the lead-bismuth<br />
cooled SVBR-75/100 reactor of 101.5<br />
MW(e) and 6-9 year refuelling interval<br />
developed in the Russian Federation.<br />
The latter design is backed by operating<br />
experience of the Russian submarine<br />
reactors.<br />
Small Reactors without<br />
On-site Refuelling<br />
Small reactors without on-site<br />
refuelling are the reactors designed for<br />
infrequent replacement of well-contained<br />
fuel cassette(s) in a manner that impedes<br />
clandestine diversion of nuclear fuel<br />
material [1, 3]. Small reactors without<br />
on-site refuelling incorporate increased<br />
refuelling interval (from 5 to 15 years and<br />
more), consistent with plant economy and<br />
considerations of energy security. Small<br />
reactors without on-site refuelling are<br />
either factory fabricated and fuelled or<br />
undergo a once-at-a-time core reloading<br />
performed at the site by a dedicated<br />
service team provided by the vendor;<br />
such team is assumed to bring in and take<br />
away the fresh and spent fuel load and the<br />
refuelling equipment.<br />
About 30 concepts of small reactors<br />
without on-site refuelling are being<br />
analyzed or developed within national<br />
and international programmes in Brazil,<br />
India, Indonesia, Japan, Morocco, Russian<br />
Federation, Turkey, U.S.A., and Vietnam<br />
[3]. Small reactor designs without onsite<br />
refuelling are being considered for<br />
both nearer-term and longer-term water<br />
cooled, liquid metal cooled and molten<br />
PBMR single module building (PBMR, Pty, South Africa) [2]<br />
salt cooled reactor lines and some nonconventional<br />
fuel/coolant combinations.<br />
Whether for fast or for thermal<br />
neutron spectrum concepts of such<br />
reactors, the fuel discharge burn-up<br />
and the irradiation of core structures<br />
never exceeds standard practice from<br />
the conventional or typically projected<br />
designs. The refuelling interval is then<br />
extended by derating core specific power,<br />
and the power densities never significantly<br />
exceed ~100 kW(th)/litre and often are<br />
much lower. Burn-up reactivity loss is<br />
mitigated by using burnable poisons and<br />
active control rods in thermal systems and<br />
by designing for internal breeding in fast<br />
systems. Although the specific inventories<br />
of fissile materials (per unit of power<br />
and energy produced) are higher than<br />
for reactors with conventional refuelling<br />
schemes, some concepts of fast spectrum<br />
reactors without on-site refuelling are<br />
capable of self-sustainable operation<br />
on fissile materials (breeding ratio ~ 1)<br />
within a closed nuclear fuel cycle. In this,<br />
breeding option is typically excluded<br />
owing to a restricted neutron economy.<br />
Challenges for<br />
Innovative SMRs<br />
Innovative SMRs in many cases<br />
do not attempt to compete with large<br />
economy of scale plants in the established<br />
markets; they rather attempt to meet<br />
the needs of those users to whom large<br />
economy-of-scale deployments are not<br />
suited. To be competitive in anticipated<br />
alternative markets, innovative SMRs rely<br />
on approaches alternative to economy<br />
of scale. Such approaches include the<br />
economy of multiple prefabricated<br />
reactor or equipment modules, reduced<br />
design complexity resulting from the<br />
application of those design features that<br />
are most appropriate for the reactor of a<br />
given capacity, an option of incremental<br />
capacity increase with possible benefits<br />
resulting from “just in time” capacity<br />
(Continued on page 50)<br />
Potential SMR cost factor advantages (Westinghouse, USA) [1]<br />
48 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
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Innovative Reactor...<br />
Continued from page 48<br />
addition and learning curve factors,<br />
shorter construction period and, possibly,<br />
greater involvement of local labour. The<br />
effectiveness of all these approaches for<br />
the conditions of targeted markets should<br />
be demonstrated, which is a challenge of<br />
prime importance for all innovative SMRs.<br />
Many of the innovative SMR concepts<br />
incorporate design features and system<br />
configurations that are not proven in<br />
operating practice of reactors for civil<br />
nuclear power; also, many innovative<br />
SMRs are just non water cooled reactors.<br />
The designers of innovative SMRs target<br />
licensing within the currently established<br />
national regulatory framework [4] but<br />
mention that further elaboration of national<br />
regulatory norms toward technology-neutral<br />
and risk-informed approach could facilitate<br />
licensing consideration and further design<br />
improvement. In addition to incorporating<br />
many inherent safety features, some<br />
innovative SMR concepts suggest stronger<br />
reliance on passive systems of innovative<br />
design. Reliability of such systems needs<br />
to be proven to enable risk-informed<br />
qualification and licensing [5, 6].<br />
Many potential applications of SMRs<br />
may require them to be located in proximity<br />
to the users:<br />
• In industrial cogeneration applications,<br />
such as hydrogen production, they<br />
must be sited adjacent to the industrial<br />
site for delivery of process heat;<br />
• They could supply energy to cities in<br />
regions where only a local electrical<br />
grid exists;<br />
• They could produce energy products<br />
such as potable water and district<br />
heat, which cannot be transported<br />
to significant distances without a<br />
significant economic penalty.<br />
These siting considerations lead to a<br />
requirement for very high levels of safety<br />
and reliability. Co-locating a nuclear and a<br />
chemical plant on a single site may require<br />
developing additional safety rules and<br />
regulations to be applied to both of them<br />
[1].<br />
Licensing of a nuclear power plant<br />
with a reduced or eliminated emergency<br />
planning zone, which is aimed by the<br />
designers of many innovative SMRs, will<br />
benefit from risk-informed regulation being<br />
emplaced. Achieving the goal of a reduced<br />
off-site emergency planning would require<br />
both, development of a methodology to<br />
prove that such reduction is possible in<br />
the specific case of a plant design, and<br />
adjustment of the existing regulations. Riskinformed<br />
approach to reactor qualification<br />
and licensing could be of value here, once<br />
it gets established. Within the deterministic<br />
safety approach it might be very difficult to<br />
justify reduced emergency planning in view<br />
of a prescribed consideration of a postulated<br />
severe accident with radioactivity release to<br />
the environment owing to a common cause<br />
failure. Probabilistic safety assessment<br />
(PSA), as a supplement to the deterministic<br />
approach, might help justify very low core<br />
damage frequency (CDF) or large early<br />
release frequency (LERF), but it does not<br />
address the consequences and, therefore,<br />
does not provide for assessment of the<br />
source terms. Risk-informed approach that<br />
introduces quantitative safety goals, based<br />
on the probability-consequences curve,<br />
and links them to certain defence in depth<br />
levels, which could help solve the dilemma<br />
by providing for a quantitative measure<br />
for the consequences of severe accidents<br />
and by applying a rational technical and<br />
non-prescriptive basis to define a severe<br />
accident. An example of such approach is<br />
in the recently published IAEA-TECDOC-<br />
1570 “Proposal of a Technology- Neutral<br />
Safety Approach for New Reactor Designs”<br />
[7].<br />
Many small reactors without onsite<br />
refuelling incorporate substantially<br />
increased refuelling interval, ranging from<br />
~5 to 20-25 years and beyond. The operating<br />
experience for such elongated refuelling<br />
intervals is generally unavailable in civil<br />
nuclear power [1]. The known experience of<br />
marine reactors confirms the possibility of a<br />
7 to 8-year continuous operation of small<br />
reactors [3]. Therefore, the construction of<br />
a prototype would be a must for many small<br />
reactors without on-site refuelling.<br />
Conclusion<br />
In the end of 2007, of the world’s 439<br />
operating nuclear power plants, 134 were<br />
with SMRs. Of the 23 newly constructed<br />
NPPs, 9 were with SMRs [8]. In the near<br />
term, most new nuclear power reactors<br />
are likely to be evolutionary large units.<br />
But particularly in the event of a nuclear<br />
renaissance, the nuclear industry can expect<br />
an increasing diversity of customers, and<br />
thus an increasing number of customers<br />
with needs potentially best met by one or<br />
more of the innovative SMR designs now<br />
under development.<br />
References<br />
[1] INTERNATIONAL ATOMIC<br />
ENERGY AGENCY, <strong>Nuclear</strong><br />
Technology Review 2007, Attachment<br />
4: “Progress in Design and Technology<br />
Development for Innovative Small and<br />
Medium Sized Reactors”, IAEA (2007):<br />
http://www.iaea.org/About/Policy/GC/<br />
GC51/GC51InfDocuments/English/<br />
gc51inf-3-att4_en.pdf<br />
[2] INTERNATIONAL ATOMIC<br />
ENERGY AGENCY, Status of<br />
Innovative Small and Medium Sized<br />
Reactor Designs 2005: Reactors with<br />
Conventional Refuelling Schemes,<br />
IAEA-TECDOC-1485 (2006).<br />
[3] INTERNATIONAL ATOMIC<br />
ENERGY AGENCY, Status of Small<br />
Reactor Designs without On-site<br />
Refuelling, IAEA-TECDOC-1536<br />
(2007).<br />
[4] INTERNATIONAL ATOMIC<br />
ENERGY AGENCY, Safety of<br />
the <strong>Nuclear</strong> Power <strong>Plant</strong>s: Design<br />
Requirements, safety standards Series,<br />
No. NS-R-1, IAEA, Vienna (2000).<br />
[5] MARQUÈS M. et al, Methodology<br />
for the reliability evaluation of a<br />
passive system and its integration into<br />
a Probabilistic Safety Assessment,<br />
<strong>Nuclear</strong> Engineering and Design 235<br />
(2005), pp 2612-2631.<br />
[6] NAYAK, A.K., GARTIA, M.R.,<br />
ANTHONY, A., VINOD, G.,<br />
SRIVASTAV, A. AND SINHA, R.K.,<br />
Reliability Analysis of a Boiling Twophase<br />
Natural Circulation System<br />
Using the APSRA Methodology,<br />
Proceedings of International Congress<br />
on Advances in <strong>Nuclear</strong> Power <strong>Plant</strong>s<br />
(ICAPP 2007), Nice, France, May 13-<br />
18, 2007 (Paper no. 7074).<br />
[7] INTERNATIONAL ATOMIC<br />
ENERGY AGENCY, Proposal for a<br />
Technology-Neutral Safety Approach<br />
for New reactor Designs, IAEA-<br />
TECDOC 1570 (2007).<br />
[8] INTERNATIONAL ATOMIC<br />
ENERGY AGENCY, Power Reactor<br />
Information System (PRIS): http://<br />
www.iaea.org/programmes/a2/. <br />
50 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
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Guidance for New Vendors<br />
By John Nakoski, U.S. <strong>Nuclear</strong><br />
Regulatory Commission.<br />
1. What factors do the vendors new to<br />
the nuclear power industry need to take<br />
into consideration to determine if they<br />
should qualify their quality assurance<br />
program for supplying products<br />
and services to the nuclear power<br />
industry Also please describe briefl y<br />
any guidance for such vendors totally<br />
unfamiliar with the nuclear power<br />
industry.<br />
I think from the NRC perspective, we<br />
see this as a business decision. A vendor<br />
new to the industry needs to understand the<br />
requirements for quality assurance in this<br />
industry. The NRC’s quality assurance<br />
requirements are outlined in Appendix B<br />
to 10 CFR Part 50. Our quality assurance<br />
requirements are typically more stringent<br />
than other industries. There is an added<br />
cost to meeting these requirements, and a<br />
new vendor needs to consider how best to<br />
factor that cost into its business decision.<br />
In addition to the quality assurance<br />
requirements, the NRC has regulations in<br />
place that require reporting of defects and<br />
non-compliance. These requirements are<br />
provided in 10 CFR Part 21. In terms of<br />
becoming qualified, a new vendor would<br />
need to have as a customer, an NRC<br />
licensee or an applicant with an approved<br />
quality assurance program. The NRC<br />
licensee or applicant could then conduct<br />
an audit of the new vendor’s quality<br />
assurance program to assess whether it<br />
complies with NRC requirements. If the<br />
results of the audit indicate the new vendor<br />
is in compliance, then the vendor can be<br />
added to the licensee’s or applicant’s<br />
approved suppliers list. Alternatively,<br />
if the new vendor is supplying parts or<br />
services to a vendor that is already on an<br />
NRC licensee’s or applicant’s approved<br />
suppliers list, the existing vendor can<br />
audit the new vendor and qualify the<br />
Responses to questions by Newal<br />
Agnihotri, Editor of <strong>Nuclear</strong> <strong>Plant</strong><br />
<strong>Journal</strong>.<br />
John Nakoski<br />
John A. Nakoski, Chief, Quality<br />
and Vendor Branch 2, Division of<br />
new vendor’s quality assurance program.<br />
Basically, an NRC licensee or an industry<br />
approved vendor would need to conduct<br />
an audit of the new vendors quality<br />
assurance program to assess whether it<br />
complies with NRC requirements. So,<br />
most of the burden for qualifying new<br />
vendors falls to the licensees, applicants or<br />
potential applicants. The <strong>Nuclear</strong> Utilities<br />
Procurement Issues Committee (NUPIC)<br />
has taken on the NRC licensees’ and<br />
applicants’ role of conducting these audits<br />
of the suppliers to the commercial nuclear<br />
industry in the US. Of course at the NRC,<br />
we have our regulatory oversight role. We<br />
inspect those organizations that provide<br />
basic services or basic components to the<br />
commercial nuclear industry.<br />
NUPIC is an organization that<br />
is comprised of essentially all the<br />
commercial US nuclear utilities and<br />
several international utilities. It’s an<br />
organization that shares resources to<br />
conduct audits required by Appendix B<br />
to 10 CFR Part 50 to provide reasonable<br />
assurance that vendors have an effective<br />
quality assurance program and that they<br />
comply with 10CFR Part 21.<br />
We’ve interacted with NUPIC for<br />
many years. We have been observing its<br />
processes and the implementations of<br />
its audits at selected vendors throughout<br />
the years. We have also been observing<br />
Construction Inspection and Operational<br />
Programs, Offi ce of New Reactors, U.S.<br />
<strong>Nuclear</strong> Regulatory Commission<br />
Together with Juan Peralta, Mr. Nakoski<br />
is responsible for developing and<br />
implementing the NRC’s programs for<br />
the oversight of vendors support related<br />
to new reactor construction and quality<br />
assurance programs for the design,<br />
licensing, and construction of new<br />
reactors. Mr. Nakoski has 25 years of<br />
experience in the nuclear energy arena,<br />
primarily with the NRC. He is a 1983<br />
graduate from Penn State with a B.S. in<br />
<strong>Nuclear</strong> Engineering.<br />
its periodic meetings where it discusses<br />
vendor and supply chain issues.<br />
2. Briefl y describe how USNRC<br />
implements its vendor inspection<br />
program.<br />
The NRC’s vendor inspection<br />
program for new reactors is implemented<br />
following guidance documented in our<br />
inspection manual chapter (IMC) 2507.<br />
For the current operating reactors, IMC<br />
2700 describes the vendor inspection<br />
program. These IMCs lay out the basic<br />
requirements that we follow to oversee<br />
any organization that provides safetyrelated<br />
parts or services to the nuclear<br />
power industry. Under the IMCs, we<br />
have inspection procedures that provide<br />
directions to the inspectors that they<br />
follow in planning for and conducting<br />
inspections. The inspection procedures<br />
provide guidance on reviewing vendor<br />
quality assurance, commercial grade<br />
dedication, and 10 CFR Part 21<br />
programs. In addition, our vendor<br />
inspection program includes oversight<br />
of organizations that conduct audits of<br />
vendors - organizations such as NUPIC.<br />
For new reactors, our current plan is to<br />
conduct about 10 vendor inspections and<br />
several NUPIC audit observations each<br />
year. We may perform more if necessary<br />
and have the resources available. While<br />
52 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
our vendor inspections provide us with<br />
direct insights into the performance of<br />
vendors, our observation of NUPIC audits<br />
of vendors gives us a sense of the quality<br />
of the industry’s oversight of vendors,<br />
and provides us the ability to provide<br />
feedback on how the oversight process can<br />
be enhanced. All of our vendor oversight<br />
activities, our inspections and NUPIC<br />
audit observations, are documented in<br />
publicly available inspection reports.<br />
These reports are available from the NRC<br />
website in our electronic reading room<br />
under ADAMS. We also make them<br />
available through our Quality Assurance<br />
website (http://www.nrc.gov/reactors/<br />
new-licensing/quality-assurance/vendorinsp.html).<br />
3. What is the best website link that a<br />
new vendor can click on to fi nd simple<br />
guidance for the quality assurance<br />
process required for qualifying to supply<br />
nuclear products and services to the<br />
nuclear power industry<br />
The NRC maintains information<br />
on its website that new vendors would<br />
find useful regarding the programs and<br />
requirements we follow when inspecting<br />
vendors. The website is located at:<br />
http://www.nrc.gov/reactors/new-licensing/quality-assurance.html.<br />
To get this<br />
website from the NRC’s main public<br />
(www.NRC.gov), click on the “<strong>Nuclear</strong><br />
Reactor” tab, then drop down to “New<br />
Reactor Licensing, there is a link below<br />
“Under How We Regulate” called “Quality<br />
Assurance for <strong>Nuclear</strong> Power <strong>Plant</strong>s”<br />
and there is a link to “Regulations and<br />
Standard Review Plan”, “Vendor Inspections”,<br />
Inspections for New Reactor Licensing”<br />
and “<strong>Nuclear</strong> Procurement Issues<br />
Committee and Industry Interface.”<br />
Also, presentations we’ve made during<br />
various conferences over the past several<br />
years can be found under the “<strong>Nuclear</strong><br />
Procurement Issues Committee (NUPIC)<br />
and Industry Interface ” link. We have a<br />
variety of information on the website and<br />
encourage new and existing vendors, or<br />
anyone interested in this area, to explore<br />
the site.<br />
4. Where is the new reactor licensing<br />
procedure defi ned Is this 10 CFR<br />
Part 52 What are the provisions for<br />
quality assurance in this code of federal<br />
regulation<br />
10 CFR Part 52 provides the regulatory<br />
framework for new reactor licensing.<br />
It points back to 10 CFR Part 50,<br />
for the technical and quality assurance<br />
requirements. It does point back to and<br />
specify that applicants are required to do<br />
safety-related activities under quality assurance<br />
programs that meet 10 CFR Part<br />
50, Appendix B requirements. Additional<br />
guidance for preparing new reactor applications<br />
is provided in Regulatory Guide<br />
1.206, “Combined License Applications<br />
for <strong>Nuclear</strong> Power <strong>Plant</strong>s.” When preparing<br />
license applications under 10 CFR<br />
Part 52, the information applicants use is<br />
required to be gathered under an Appendix<br />
B quality assurance process. The application<br />
itself is developed to satisfy the<br />
completeness and accuracy requirements<br />
of 10 CFR 50.9 and applications need to<br />
be submitted under oath and affirmation.<br />
5. How do you gather the list of vendors<br />
who are supplying products and services<br />
to the nuclear power plants to ensure<br />
that these vendors are qualifi ed for the<br />
supplies<br />
For new reactor construction, we<br />
have requested information from the<br />
industry through a regulatory issues<br />
summary, 2007-08, “Updated Licensing<br />
Submittal Information to Support the<br />
Design-Centered Licensing Review<br />
Approach.” So far we have received<br />
some responses from applicants and the<br />
major vendors supplying the designs.<br />
In addition, through our interface with<br />
NUPIC, we have a list of vendors that<br />
have been qualified by the current fleet<br />
of operating reactors and are supplying<br />
basic components to the currently<br />
operating fleet of power reactors. Using<br />
this information to give us confidence<br />
in the quality of products provided to<br />
nuclear power plants, the NRC conducts<br />
inspections of a sample of the vendors<br />
that have been approved by licensees and<br />
oversees the audits conducted by NUPIC<br />
of these vendors.<br />
6. If a vendor in China wants to be<br />
certifi ed, will NRC go to China<br />
It is important to recognize that the<br />
NRC is not in the process of certifying or<br />
approving vendors to supply products and<br />
services to the nuclear power industry.<br />
We inspect vendors for compliance<br />
with our regulations. Also, in today’s<br />
manufacturing arena, many of the<br />
vendors of major components are located<br />
overseas. So, if a vendor in China was<br />
selected by a licensee or applicant and<br />
put on an approved suppliers list for the<br />
construction of a new reactor, our plan is<br />
to include that vendor in the population<br />
of vendors that we may inspect. If we had<br />
concerns with the quality of the vendor<br />
regardless of where they are, domestically<br />
or internationally, that would factor<br />
into our decision on whether we should<br />
inspect a particular vendor. If we received<br />
indications from our interactions with the<br />
applicants, through NUPIC, from peer<br />
regulators in other countries, or as a result<br />
of observations of construction inspection<br />
activities by the regional staff that<br />
problems with quality existed, we would<br />
factor that into our decision. So the short<br />
answer to the question would be yes, if<br />
we determined that it was necessary or<br />
consistent with our program guidelines. I<br />
would add that over the last 18 months, we<br />
have been building an extensive interface<br />
program with our peer regulators across<br />
the globe. As one example, we recently<br />
conducted coordinated inspections with<br />
our fellow regulators in Japan and Korea<br />
at specific vendors in those countries.<br />
7. What organizations other than<br />
USNRC are involved in establishing<br />
guidelines for quality for new reactor<br />
construction activities<br />
Other organizations involved in<br />
establishing guidance on quality assurance<br />
requirements include:<br />
1. <strong>Nuclear</strong> Utilities Procurement<br />
Issues Committee (NUPIC)<br />
2. <strong>Nuclear</strong> Energy Institute (NEI)<br />
3. American Society of Mechanical<br />
Engineers (ASME)<br />
4. <strong>Nuclear</strong> Industry Assessment<br />
Committee (NIAC)<br />
5. American Society of Quality<br />
(ASQ)<br />
6. Electric Power Research Institute<br />
(EPRI)<br />
Contact: John, A. Nakoski, U.S.<br />
<strong>Nuclear</strong> Regulatory Commission, MS T-7F3,<br />
Washington DC 20555; telephone: (301)<br />
415-1068, email: John.Nakoski@nrc.gov. <br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 53
Road Map for Future Energy<br />
By John Cleveland, International Atomic<br />
Energy Agency.<br />
Introduction<br />
Today nuclear energy contributes<br />
approximately 15.2% of the world’s<br />
electricity. By February 2008, there<br />
were 439 nuclear power plants (NPPs) in<br />
operation worldwide, with a total capacity<br />
of 371.7 GWe. Further, 34 units, totaling<br />
28.1 GWe, were under construction.<br />
During 2006 nuclear power produced<br />
2659.7 billion kWh of electricity. Based<br />
on information provided by its Member<br />
States, the IAEA projects that nuclear<br />
power will grow significantly, producing<br />
annually between 2760 and 2810 billion<br />
kWh by 2010, between 3120 and 3840<br />
billion kWh by 2020, and between 3325<br />
and 5040 billion kWh by 2030 [1].<br />
The vast majority of today’s nuclear<br />
power plants use water-cooled reactors.<br />
In the near term most new nuclear plants<br />
will be evolutionary water cooled reactors<br />
(Light Water Reactors (LWRs) and Heavy<br />
Water Reactors (HWRs)], often pursuing<br />
economies of scale. Other reactor types<br />
have had considerably less operational<br />
and regulatory experience and will take<br />
still some time to be widely accepted<br />
in the market. These innovative designs<br />
promise shorter construction times and<br />
lower capital costs and could help in the<br />
future to promote a new era of nuclear<br />
power.<br />
While nuclear power contributes<br />
significantly to electricity generation,<br />
most of the world’s energy consumption<br />
is for heat and transportation. Through<br />
advanced applications, nuclear energy can<br />
penetrate these energy sectors now served<br />
by fossil fuels that are characterized<br />
by price volatility, finite supply, and<br />
environmental concerns.<br />
Advanced applications of nuclear<br />
energy include seawater desalination,<br />
district heating, heat for industrial<br />
processes, and electricity and heat for<br />
hydrogen production. In addition, in<br />
the transportation sector, since nuclear<br />
electricity is generally produced in a base<br />
load mode at stable prices, nuclear power<br />
John Cleveland<br />
Mr. John Cleveland has worked at the<br />
IAEA since 1991. Until 1994 he was in<br />
charge of IAEA’s activities in technology<br />
can contribute as a carbon-free source of<br />
electricity for transportation (e.g. trains<br />
and subway systems) and for charging<br />
electric and plug-in hybrid vehicles.<br />
Due to these factors, the IAEA has<br />
carried out this study to examine the<br />
opportunities, challenges and solutions<br />
for water-cooled reactors to contribute<br />
to these advanced applications of nuclear<br />
energy [2].<br />
Seawater Desalination<br />
Water is essential for the sustainable<br />
development of society. Water scarcity<br />
is a global issue, and every year more<br />
countries are affected by growing water<br />
problems.<br />
Large-scale commercially available<br />
seawater desalination processes can<br />
generally be classified into two categories:<br />
(a) distillation processes (these are the<br />
Multi-Stage Flash – MSF, and the Multi-<br />
Effect Distillation – MED processes) that<br />
require mainly heat plus some electricity<br />
for ancillary equipment, and (b) membrane<br />
processes (Reverse Osmosis – RO) that<br />
require only electricity to provide the<br />
necessary pumping power.<br />
The desalination of seawater<br />
using nuclear energy is a feasible and<br />
demonstrated option for production of<br />
potable water. Over 200 reactor-years<br />
of operating experience on nuclear<br />
development of high-temperature gascooled<br />
reactors. Since 1994 he has been<br />
the leader of the Water-Cooled Reactors<br />
Group of the <strong>Nuclear</strong> Power Technology<br />
Development Section.<br />
Before joining the IAEA, he worked for<br />
the Babcock and Wilcox Company and at<br />
the Oak Ridge National Laboratory<br />
in the USA.<br />
Mr. Cleveland received his Masters<br />
Degree in Physics from Virginia<br />
Polytechnic Institute and State<br />
University, USA, in 1972. He has<br />
authored more than 80 technical papers<br />
and reports in the fi eld of nuclear<br />
reactor technology and safety.<br />
desalination have been accumulated<br />
worldwide, and more demonstration<br />
projects are being prepared. However,<br />
nuclear desalination today contributes<br />
only 0.1 % of the total desalting capacity<br />
worldwide [3].<br />
Table 1 (see page 56) shows the nuclear<br />
reactors used or under construction<br />
for seawater desalination. In addition to<br />
those systems shown in Table 1, other<br />
water-cooled concepts are being developed<br />
for seawater desalination. For example,<br />
the nuclear heating reactor (NHR)<br />
developed in China could provide heat for<br />
desalination, and the SMART concept,<br />
developed in the Republic of Korea, the<br />
CAREM concept of Argentina, and the<br />
KLT-40 floating power unit developed in<br />
Russia 1 , could be used for cogeneration<br />
of electricity and seawater desalination.<br />
Countries suffering from scarcity<br />
of water are generally not the holders<br />
of nuclear technology. They do not<br />
have nuclear power plants, and do not<br />
have a nuclear power infrastructure.<br />
(Continued on page 56)<br />
1 The Floating Power Unit under construction at<br />
Severodvinsk, Russia, is planned to be comissioned<br />
in 2010, and will be used for electricity and<br />
district heating. Future potential units outside of<br />
Russia could be used for electricity and seawater<br />
desalination<br />
54 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
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Road Map..<br />
Continued from page 54<br />
The utilization of nuclear energy in<br />
such countries will require infrastructure<br />
building and institutional arrangements<br />
for issues such as financing, liability,<br />
safeguards, safety, and security.<br />
District Heating<br />
District heating involves the supply<br />
of space heat and hot water through a<br />
district heating system, which consists of<br />
heat plants (usually producing electricity<br />
simultaneously) and a network of<br />
distribution pipes. Potential application of<br />
district heating is in climatic zones with<br />
relatively long and cold winters. In many<br />
countries, such as central and northern<br />
Reactor<br />
Type<br />
European countries and countries in<br />
transition economies, district heating has<br />
been widely used for decades.<br />
Coal and gas dominate the fuels used<br />
for district heating. However, several<br />
countries (Bulgaria, China, Czech Republic,<br />
Hungary, Romania, Russia, Slovakia,<br />
Sweden, Switzerland and Ukraine) have<br />
experience in nuclear district heating using<br />
water-cooled reactors, so the technical<br />
aspects can be considered well proven.<br />
In order to be able to compete with<br />
Location m 3 /day Status<br />
fossil-fuel-fired heat boilers, the capital cost<br />
per installed MW of heat production capacity<br />
for a nuclear-based system must be such<br />
that the production costs are competitive.<br />
Dedicated reactors providing district heat<br />
can potentially achieve acceptable costs,<br />
due to their lower temperature operating<br />
conditions, simple design, modularization<br />
and standardization, and advanced safety<br />
systems.<br />
New nuclear heat-producing plants<br />
must, of course, meet the user’s requirements<br />
on availability and reliability, including<br />
alternative heat-producing capacity that<br />
could serve as backup. For this purpose,<br />
heat storage allows a matching of the heat<br />
supply to the heat demand. Today there are<br />
many examples of short-term storage, for<br />
instance, on the daily scale that relies on<br />
hot water accumulator tanks. In the future,<br />
more long-term storage facilities may be<br />
realized.<br />
Table 1: Reactor types used or under construction for seawater desalination<br />
LMFR Kazakhstan (Aktau) 80,000 In service till 1999<br />
PWRs<br />
HWRs<br />
Japan<br />
Ohi 1,2,3,4<br />
Takahama<br />
Ikata 1,2,3<br />
Genkai 3,4<br />
USA (Diablo<br />
Canyon)<br />
~1500 In service<br />
Operating experience ~<br />
170 R-Ys<br />
~4500 In service<br />
India (Madras) 6,300 RO commissioned in<br />
2002<br />
MSF to be commissioned<br />
in 2008<br />
Pakistan<br />
(KANUPP)<br />
4,800 Under construction;<br />
Commissioning –in 2008<br />
Industrial Heat Process<br />
Process heat involves the supply of heat<br />
required for industrial processes from one<br />
or more centralized heat generation sites<br />
through a steam transportation network.<br />
Within the industrial sector, process heat<br />
is used for a large variety of applications<br />
with different heat requirements and<br />
with temperature ranges covering a wide<br />
spectrum. Examples of industries that<br />
consume considerable amounts of heat<br />
are:<br />
• food,<br />
• paper,<br />
• chemicals and fertilizers,<br />
• petroleum and coal processing, and<br />
• metal processing industries.<br />
The chemical and petroleum industries<br />
are the major consumers of process heat<br />
worldwide. These would be key target<br />
clients for possible applications of nuclear<br />
energy.<br />
The supply of energy for industrial<br />
processes has an essential character: all<br />
industrial users need the assurance of<br />
energy supply with a high reliability, and<br />
the heat should be produced close to the<br />
point of use. Many of the process heat<br />
users, in particular the large ones, usually<br />
are located outside urban areas, often at<br />
considerable distances. This makes joint<br />
siting of nuclear reactors and industrial<br />
users of process heat not only viable, but<br />
also desirable in order to drastically reduce<br />
the heat transportation costs.<br />
The nuclear process heat supply has<br />
to be reliable. As an example, the average<br />
steam supply availabilities for chemical<br />
processing and oil refineries are 92% and<br />
above.<br />
There is experience in providing<br />
process heat for industrial purposes with<br />
nuclear energy in Canada, Germany,<br />
Norway, Switzerland, and India. New<br />
plant designs that can provide heat, or both<br />
heat and electricity, are being designed in<br />
Russia, the Republic of Korea, Canada, and<br />
other countries.<br />
Current water cooled reactors can<br />
provide process heat up to about 300ºC,<br />
and some future innovative water cooled<br />
reactor designs 2 have potential to provide<br />
heat up to approximately 550ºC.<br />
Although nuclear industrial process<br />
heat applications have significant potential,<br />
it has not been realized to a large extent.<br />
In fact, currently only the Goesgen reactor<br />
in Switzerland and the RAPS–2 reactor in<br />
India continue to provide industrial process<br />
heat, whereas other nuclear process heat<br />
systems have been discontinued after<br />
successful use. Among the reasons cited for<br />
closure of these units, one is availability of<br />
cheaper alternate energy sources.<br />
For potential future application of<br />
nuclear process heat, an important example<br />
2<br />
Specifically Super-critical Water Cooled Reactors,<br />
being developed within the Generation-IV<br />
International Forum, could be deployed by around<br />
2025-2030.<br />
56 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
is the use of nuclear energy for oil sand<br />
open-pit mining and deep-deposit extraction<br />
in Canada. Alberta’s oil sand deposits are<br />
the second largest oil reserves in the world,<br />
and have emerged as the fastest growing,<br />
soon to be dominant, source of crude oil in<br />
Canada. Currently, the majority of oil sand<br />
production is through open-pit mining,<br />
which is suitable for bitumen extraction<br />
when the oil sand deposits are close to<br />
the surface. The ore, a mixture of bitumen<br />
and sand, is removed from the surface by<br />
truck and shovel operation. The ore is then<br />
mixed with hot water to form a slurry that<br />
eventually undergoes a separation process<br />
to remove bitumen from the sand.<br />
The thermal energy required for the<br />
open-pit mining process is in the form of<br />
hot water at a relatively low temperature<br />
(around 70°C), and the rest is dry process<br />
steam at around 1.0 to 2.0 MPa. The oil<br />
extraction facilities require electrical<br />
power as well. The steam and electricity<br />
requirements can be met by water cooled<br />
reactors.<br />
To increase production capacity, oil<br />
companies are developing new technologies<br />
to extract bitumen from deep deposits.<br />
Among them, Steam-Assisted Gravity<br />
Drainage (SAGD), which uses steam<br />
to remove bitumen from underground<br />
reservoirs, appears to be the most promising<br />
approach. Recently, this in-situ recovery<br />
process has been put into commercial<br />
operation.<br />
Overall, for both extraction<br />
methodologies (open pit mining and<br />
SAGD), a significant amount of energy is<br />
required to extract bitumen and upgrade it<br />
to synthetic crude oil as the feedstock for<br />
oil refineries. Currently, the industry uses<br />
natural gas to provide this energy. As oil<br />
sand production continues to expand, the<br />
energy required for production becomes a<br />
great challenge with regard to economic<br />
sustainability, environmental impact<br />
and security of supply. Therefore, the<br />
opportunity for nuclear reactors to provide<br />
an economical, reliable and virtually zeroemission<br />
source of energy (both electricity<br />
and steam) for the oil sands becomes a<br />
realistic option.<br />
Energy for Transportation<br />
Transportation represents approximately<br />
20% of the world’s energy consumption.<br />
In the United States, transportation<br />
is the fastest growing energy sector.<br />
The Organization for Economic Co-operation<br />
and Development International Energy<br />
Agency projects that global primary energy<br />
demand will grow by 50% by 2030, with<br />
70% of that growth coming from developing<br />
countries, especially China. Half of<br />
that increase will be for electricity production<br />
and 20% for transportation.<br />
It is clear that if means are found for<br />
nuclear energy to power a significant part<br />
of the transportation sector, it could have a<br />
significant impact on global environmental<br />
sustainability. Two ways this could<br />
occur would be through the advancement<br />
transportation systems based on electricity,<br />
such as trains, subways, electric and plugin<br />
hybrid vehicles charged with nuclear<br />
generated electricity, and of vehicles fuelled<br />
with hydrogen produced by nuclear energy.<br />
Following are some examples.<br />
A) Electricity for plug-in hybrid<br />
electric vehicles<br />
The potentially large market demand<br />
for electricity for powering plug-in hybrid<br />
electric vehicles is eminently suited to<br />
current and evolutionary water-cooled<br />
nuclear power plants. Because nuclear<br />
plants generally operate at base load<br />
conditions, provide electricity at stable<br />
and predictable prices, and produce clean<br />
electricity, they are especially well suited<br />
to play a near term role in powering the<br />
transportation sector, while helping to<br />
reduce greenhouse gasses from this sector.<br />
Hybrid vehicles are commercially<br />
available today. Almost all use regenerative<br />
braking to charge an on-board battery<br />
for locomotive power. With these battery<br />
systems, vehicles can be designed to allow<br />
the gasoline engine to turn off when the vehicle<br />
is stopped or during cruising.<br />
Overall energy use for hybrids is<br />
about 40% less than that for conventional<br />
vehicles, with an equivalent reduction in<br />
greenhouse gas emissions (CO 2<br />
, CH 4<br />
, and<br />
N 2<br />
O).<br />
Plug-in hybrid electric vehicles<br />
extend this technology by allowing the<br />
drive battery to be charged externally. In<br />
this way, the vehicle can be driven in an<br />
all-electric mode for a certain distance<br />
with no power from the gasoline engine.<br />
This can provide significant savings in<br />
terms of petroleum usage and emissions,<br />
especially since the majority of miles<br />
driven are for short commutes. These<br />
emission reductions materialize only if the<br />
source of external electricity is clean and<br />
carbon free, of course. Importantly, plugin<br />
hybrid manufacturers have announced<br />
targets of 20 to 40 miles on a single charge.<br />
One developer recently unveiled a plugin<br />
hybrid demonstration vehicle which<br />
uses a combination of ultra-capacitors and<br />
batteries for energy storage and has an allelectric<br />
range of 40 miles.<br />
In this study, a simplified model of<br />
potential growth in usage of plug-in hybrid<br />
electric vehicles, which assumed that all<br />
automobiles and light trucks in the US would<br />
be plug-in hybrid vehicles by 2035, showed<br />
that 200-250 GW of electricity would be<br />
needed for overnight charging in the U.S.<br />
This would replace 280 million gallons of<br />
fuel per day with the corresponding large<br />
reduction in production of greenhouse<br />
gasses from the transportation sector. New<br />
electricity generation capacity at this scale<br />
would also require new transmission and<br />
distribution lines and substations. A similar<br />
analysis for Japan suggests the need for 35<br />
GW of electricity for overnight charging,<br />
which is within the capacity of spare power<br />
at night.<br />
Aside from the need for increases in<br />
generating and transmission capacity, other<br />
barriers will need to be overcome before<br />
there is widespread adoption of plug-in<br />
hybrid electric vehicles:<br />
• Conversion of automobile technology<br />
from conventional gasoline-powered<br />
vehicles to electric and plug-in hybrid<br />
vehicles;<br />
• Public acceptance of plug-in hybrid<br />
vehicles;<br />
• Structuring of electricity pricing<br />
mechanisms to provide low-price<br />
electricity during off-peak demand<br />
periods to encourage use of nuclear<br />
power plants for base load generation;<br />
• Provision of other incentives (e.g., tax<br />
benefits) for adoption of vehicles that<br />
produce less greenhouse gases and<br />
reduce reliance on petroleum fuels.<br />
A key technology need is development<br />
of lighter, less expensive, reliable batteries<br />
having a factor of 5 to 10 greater energy<br />
storage capacity that would support longer<br />
all-electric distances. Lithium-ion batteries<br />
are the main focus of current research and<br />
development.<br />
B) Hydrogen for transportation<br />
Hydrogen for transportation is<br />
receiving significant attention around the<br />
(Continued on page 58)<br />
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Road Map...<br />
Continued from page 57<br />
world because of high petroleum prices<br />
and unreliable oil supplies. Two ways of<br />
hydrogen utilization in transportation are<br />
currently being taken into consideration –<br />
internal combustion engine (ICE) vehicles<br />
and fuel cell (FC) vehicles. While ICE<br />
vehicles represent current technology<br />
with modest modifications, fuel cell<br />
vehicles are in a stage of intensive R&D<br />
and prototype testing.<br />
Car manufacturers are focusing<br />
more effort on fuel cell vehicles than on<br />
hydrogen ICE vehicles. Many prototypes<br />
have been introduced, some of them in<br />
small series (tens of cars). Current trends<br />
are mainly focused on hybridization, such<br />
as combining fuel cells with Nickel metal<br />
hydride (NiMH) batteries, ultra capacitors,<br />
or other types of electric storage.<br />
Although this increases the complexity<br />
of the vehicle, thus increasing the cost, it<br />
brings advantages with regard to covering<br />
power peaks during acceleration, when<br />
the electric motor draws high current<br />
from the fuel cell, and also increases the<br />
driving range, because hybrid vehicles<br />
optimize fuel consumption, and also the<br />
use of braking recuperation.<br />
It is not only important to have technical<br />
problems solved, public acceptance<br />
is also important. For this purpose, hydrogen<br />
fuelled buses have been successful.<br />
Currently there are about 60 of them<br />
serving on a daily basis in different cities<br />
including London, Hamburg, Madrid,<br />
Stuttgart, Stockholm, Porto, Amsterdam,<br />
Barcelona, Luxembourg, Reykjavik and<br />
Perth.<br />
The lack of the hydrogen infrastructure<br />
makes fleet customers important for<br />
early hydrogen transportation markets.<br />
It is much easier to build one centralized<br />
filling station near a city bus operator<br />
or dispatch service than to service the<br />
distributed market for personal cars.<br />
Motorcycles, scooters and electric<br />
bikes represent a smaller, but interesting,<br />
market opportunity. Such means of transportation<br />
are significant in many Asian<br />
countries, where the pollution is growing<br />
and causing health problems.<br />
Hydrogen Production<br />
As an alternative path to the current<br />
fossil fuel economy, a hydrogen economy<br />
is envisaged in which hydrogen would<br />
play a major role in energy systems<br />
and serve all sectors of the economy,<br />
substituting for fossil fuels. Hydrogen<br />
as an energy carrier can be stored in<br />
large quantities, unlike electricity, and<br />
converted into electricity in fuel cells,<br />
with only heat and water as by-products.<br />
It can also fuel combustion turbines and<br />
reciprocating engines to produce power<br />
with near-zero emission of pollutants.<br />
The current worldwide hydrogen<br />
production is roughly 50 million tonnes per<br />
year. Although current use of hydrogen in<br />
energy systems is very limited, its future<br />
use could become enormous, especially if<br />
fuel-cell vehicles would be deployed on a<br />
large commercial scale.<br />
Today, hydrogen is used mainly in<br />
petroleum refineries and the chemical industry.<br />
In the United States, for example,<br />
these uses represented 93% of hydrogen<br />
consumption in 2003.<br />
The U.S., Japan, and other nations<br />
are exploring ways to produce hydrogen<br />
using nuclear energy. While some consideration<br />
is given to hydrocarbon reforming<br />
techniques, such as steam-methane reforming,<br />
much of the work is focused on<br />
means of splitting water by electrolytic,<br />
thermo-chemical, and hybrid processes.<br />
Considerable efforts have concentrated<br />
on high-temperature processes such as<br />
high-temperature steam electrolysis and<br />
the sulphur–iodine and calcium-bromine<br />
cycles. These processes operate at higher<br />
temperatures (>750°C) than can be<br />
achieved by water-cooled reactors. Advanced<br />
reactors such as the very high<br />
temperature gas cooled reactor (VHTGR)<br />
can generate heat at these temperatures,<br />
but first demonstration of hydrogen production<br />
with gas cooled reactors is not<br />
expected until around 2015 (in Japan) to<br />
2020 (in the USA).<br />
Current and evolutionary water cooled<br />
reactors can produce outlet temperatures<br />
in the range of ~300-350°C. Supercritical<br />
water cooled reactors (SCWRs),<br />
being developed within the Generation-<br />
IV International Forum, can achieve<br />
temperatures of ~550°C. Examples of<br />
processes for hydrogen production within<br />
these temperature ranges follow.<br />
A. Steam Reforming of Dimethyl<br />
Ether (~300°C)<br />
Toshiba of Japan has proposed that<br />
steam reforming of dimethyl ether (DME),<br />
a derivative from fossil fuels or biomass,<br />
could be used to produce hydrogen with<br />
300°C heat from water cooled reactors.<br />
DME is synthesized from natural gas<br />
from small or medium-sized gas fields,<br />
coal seam gas, and natural gas with a large<br />
CO 2<br />
fraction. DME is usually produced<br />
by a partial oxidation process of natural<br />
gas without emitting CO 2,<br />
as shown by<br />
the following formula:<br />
2CH 4<br />
+ O 2<br />
-->CH 3<br />
OCH 3<br />
+ H 2<br />
O<br />
The DME reforming reaction is as<br />
follows:<br />
(1/2)CH 3<br />
OCH 3<br />
+ (3/2)H 2<br />
OCO 2<br />
+3H 2<br />
–24.4 kJ/ (H 2<br />
mol)<br />
The produced hydrogen fraction is<br />
high at temperatures of 285-300°C. Specifically,<br />
Toshiba has developed, together<br />
with Shizuoka University, a DME reforming<br />
catalyst that gives 98% conversion of<br />
DME to hydrogen at 285°C. The catalyst<br />
is Cu-Zn/Al 2<br />
O 3<br />
powder [4].<br />
With 40 MW of heat supply about<br />
108 kg H 2<br />
/year of hydrogen production<br />
is possible, which is of the same scale as<br />
the largest hydrogen plant in the world.<br />
To date, the demonstrated production rate<br />
is 4.10 kg H 2<br />
/day.<br />
B. Low temperature electrolysis<br />
Hydrogen production processes<br />
based on reforming of methane not only<br />
use fossil resources (CH 4<br />
), but also produce<br />
CO 2<br />
. <strong>Nuclear</strong> energy can be used<br />
for splitting water to produce hydrogen<br />
without using fossil resources and without<br />
producing CO 2<br />
. Although the energy<br />
requirements for hydrogen production<br />
by low-temperature water electrolysis are<br />
relatively high, it is a presently available<br />
technology for hydrogen production. Water<br />
electrolyzers can be decoupled from the<br />
power plant. Therefore, electrolysers can<br />
be used for decentralized hydrogen production.<br />
C. Steam reforming of methane with<br />
a membrane reformer system (500 to<br />
600°C)<br />
A conventional steam methane reforming<br />
(SMR) system for hydrogen production<br />
involves introducing a mixture of methane<br />
and steam into a nickel-based catalyst bed<br />
in the steam reformer, where the SMR reaction<br />
proceeds at 750 to 800°C. The re-<br />
(Continued on page 60)<br />
58 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
Intelligent Monitoring Technology<br />
By Chris Demars, Exelon <strong>Nuclear</strong>.<br />
Summary:<br />
Exelon has developed and deployed<br />
over 500 plant equipment computer<br />
models to identify early degradation<br />
which has resulted in avoided losses of<br />
approximately $600K in 2 months of<br />
operation.<br />
The Exelon Centralized Performance<br />
Monitoring (CPM) pilot project was<br />
formally established in June of 2007. A<br />
team of two employees augmented with<br />
summer intern assistance was established<br />
to develop approximately 500 plant<br />
equipment models.<br />
With InStep’s experience in the<br />
<strong>Nuclear</strong> Industry and data historian<br />
specialty experience they were able to<br />
develop an extremely effective and easy<br />
to use anomaly detection tool.<br />
The application allows a user to<br />
quickly assemble and train a group of<br />
related plant process computer points<br />
in a model that when deployed will<br />
constantly monitor those points for other<br />
than normal behavior. The software<br />
package can then be configured to alert<br />
an individual to parameter relationship<br />
changes that should be investigated for<br />
potential adverse equipment conditions<br />
that could otherwise lead to failure.<br />
Safety:<br />
Early detection of equipment failures<br />
prevents the hazardous environment that<br />
<strong>Nuclear</strong> Energy Institute’s Top Industry<br />
Practice (TIP) Award highlight the<br />
nuclear industry’s most innovative<br />
techniques and ideas.<br />
This was a 2008 NEI Process Award<br />
winner.<br />
The team members who participated<br />
included: Chris Demars, Project<br />
Manager, Exelon <strong>Nuclear</strong>; Dave Miller,<br />
Exelon <strong>Nuclear</strong>; Mike Rog, Exelon<br />
<strong>Nuclear</strong>; Bill Bielke, InStep Software;<br />
Sean Gregerson, InStep Software.<br />
often accompanies rotating equipment<br />
failures or the release of industrial gases<br />
and process fluids, and improves nuclear<br />
and radiological safety through early<br />
detection and improved management of<br />
equipment degradation.<br />
Specific examples include the recent<br />
condensate pump failure avoidance. Lead<br />
time for a replacement pump is 4–6 weeks,<br />
and during that time a backup pump would<br />
not be available which reduces plant<br />
margin and impacts safety. The coupling<br />
failure would have also challenged<br />
personnel safety due to the accessibility of<br />
the area the pump is installed in. Overall<br />
safety is also improved due to the reduced<br />
scope and frequency of equipment repair<br />
challenges.<br />
The cost saving methodology that<br />
the centralized performance monitoring<br />
pilot has employed is to conservatively<br />
calculate the cost of the worst case<br />
scenario(s) that may have occurred<br />
without early detection of a degraded<br />
condition and to then multiply the<br />
worst case cost by a probability factor<br />
to obtain avoided cost. The following<br />
three recent early detections examples<br />
demonstrate that method and allow an<br />
annual approximation of avoided cost<br />
based on two months of monitoring with<br />
approximately 30 models deployed for<br />
each unit in the fleet.<br />
1. Condensate pump motor coupling<br />
seizure - $500K<br />
The condensate pump model alerted<br />
due to two bearing oil temperatures that<br />
were not within the predicted pattern of<br />
allowable values. The temperatures of the<br />
two bearing were well within accepted<br />
operating levels but were approximately<br />
4 °F outside “normal behavior” as<br />
defined by the multi-dimensional cluster<br />
based technology applied in the models.<br />
The cause was found to be an improperly<br />
assemble coupling that was seizing and<br />
approaching mechanical failure.<br />
Failure of the coupling would have<br />
resulted in damage to both the motor and<br />
pump with a replacement lead time of 4<br />
to 6 weeks. Replacement cost, expediting<br />
fees and craft overtime is estimated at<br />
Chris Demars<br />
Chris Demars has over 28 years of<br />
experience in nuclear power generation<br />
management. His diverse experience<br />
includes project management, various<br />
program recovery management<br />
positions (work management,<br />
engineering, operations, unit restart),<br />
engineering, operations and nuclear<br />
station corrective action program<br />
development and implementation,<br />
initial and accelerated license operator<br />
training (lead instructor), training<br />
program development/implementation,<br />
of on-line work management and<br />
engineering work management<br />
processes. He has a Bachelor of Science<br />
in <strong>Nuclear</strong> Engineering Technology.<br />
$700K. The probability of this failure<br />
scenario is estimated at 0.70, or $490K.<br />
Online loss of the pump with a failure<br />
of the standby pump to start would have<br />
resulted in a power reduction of 34% for<br />
12 hours or ~$100K. The probability of<br />
this failure scenario is estimated at 0.10<br />
or $10K.<br />
2. Service water temperature<br />
controller failure – $30K.<br />
The main turbine vibration model<br />
alerted due to a small step change in<br />
vibration on the number 11 bearing. The<br />
vibration level was not significant enough<br />
to cause an alarm of any normal plant<br />
monitoring systems. The cause of the<br />
step change was a change in generator<br />
hydrogen temperature which is cooled<br />
by stator water cooling that is cooled by<br />
service water. This particular nuclear unit<br />
has not removed or blocked the stator<br />
water cooling temperature turbine trip<br />
and was susceptible to a trip during the<br />
temperature changes that were caused<br />
by the failed controller. Trip of the main<br />
turbine would have resulted in a loss of<br />
generation for 24 hours or $600K. The<br />
probability of a turbine trip is estimated<br />
at 0.050 or $30K.<br />
(Continued on page 62)<br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 59
Road Map...<br />
Continued from page 58<br />
formed gas is supplied to a shift converter,<br />
where carbon monoxide and water are converted<br />
into carbon dioxide and additional<br />
hydrogen. The reformed gas is then passed<br />
to a pressure swing adsorption separator to<br />
separate the hydrogen.<br />
A membrane reformer system, composed<br />
of a steam reformer equipped with<br />
catalytic membrane modules with a palladium-based<br />
alloy and a separate nickelbased<br />
catalyst, can perform the reforming<br />
reaction, the shift reaction, and the hydrogen<br />
separation process simultaneously at<br />
temperatures of 500 to 600°C [5].<br />
In 2004-2005 Tokyo Gas Company<br />
demonstrated the operation of a methanecombusting<br />
membrane reformer system<br />
at a hydrogen fuelling station for fuel cell<br />
vehicles in downtown Tokyo. The system<br />
performance, efficiency, and long-term<br />
reliability were confirmed by producing<br />
>99.99% hydrogen at 3.6 kg/h for more<br />
than 3,000 hours with hydrogen production<br />
efficiency of about 80. SCWRs could<br />
provide heat at the temperatures needed<br />
for steam-methane membrane reformer<br />
systems.<br />
D. Thermo-chemical and Hybrid<br />
Processes (500 to 600°C)<br />
Thermo-chemical and hybrid<br />
thermo-electrochemical cycles have the<br />
potential for hydrogen production by<br />
water-splitting with higher efficiencies<br />
than low-temperature water electrolysis.<br />
Although over 200 thermo-chemical and<br />
hybrid electro-thermo-chemical reaction<br />
cycles for producing hydrogen have been<br />
identified [7], only about eleven of them<br />
have maximum reaction temperatures<br />
below 600°C. These lower-temperature<br />
cycles can reduce the thermal burden,<br />
mitigate demands on materials, and<br />
potentially be coupled with nearer-term<br />
nuclear reactors.<br />
Five of these cycles have recently<br />
been the subject of active research. They<br />
include a family of copper-chloride<br />
cycles (530° - 550°C) [8], an active metal<br />
(potassium-bismuth) cycle (475 - 675°C)<br />
[9], a magnesium-chloride cycle (500°C)<br />
known as the Reverse Deacon Cycle [10],<br />
a U-Eu-Br heavy-element halide cycle,<br />
and a hybrid sulphur-based cycle [11].<br />
Development work on such cycles has<br />
generally been limited to small laboratory<br />
scale testing.<br />
Conclusions<br />
While there are very important opportunities<br />
for deployment of nuclear energy<br />
into advanced applications, challenges<br />
and difficulties should not be overlooked.<br />
In particular, competition will drive the<br />
choice of energy sources for each application.<br />
Policies internalising the cost of<br />
carbon and other pollutants are needed to<br />
fully realize the benefits of nuclear energy<br />
in alleviating the risk of climate change.<br />
Advanced applications of nuclear energy,<br />
due to their ability to provide energy products<br />
economically and without producing<br />
greenhouse gases, can play an important<br />
role in enhancing public acceptance of<br />
nuclear energy.<br />
References<br />
[1] INTERNATIONAL ATOMIC ENERGY<br />
AGENCY, Energy, Electricity and<br />
<strong>Nuclear</strong> Power Estimates for the<br />
Period up to 2030, Reference Data<br />
Series No. 1 (2007 Edition)<br />
[2] INTERNATIONAL ATOMIC EN-<br />
ERGY AGENCY, Advanced Applications<br />
of Water-Cooled <strong>Nuclear</strong> Power<br />
<strong>Plant</strong>s, (IAEA TECDOC-1584, Vienna,<br />
2008)<br />
[3] INTERNATIONAL ATOMIC<br />
ENERGY AGENCY, Status of <strong>Nuclear</strong><br />
Desalination in IAEA Member States,<br />
TECDOC-1542, IAEA, Vienna<br />
(2006)<br />
[4] YAMADA, K., MONIWA, S., MAKI-<br />
NO, S., YOKOBORI, S., SEGAWA,<br />
N., FUKUSHIMA, K., and TAKEI-<br />
SHI, K., “Hydrogen Production with<br />
Steam Reforming of Dimethyl Ether<br />
at the Temperature Less Than 573 K”,<br />
in Proceedings of International Congress<br />
on Advances in <strong>Nuclear</strong> Power<br />
<strong>Plant</strong>s, No. 5138 (2005)<br />
[5] TASHIMO, M. et. al., “Advanced<br />
Design of Fast Reactor-Membrane<br />
Reformer (FR-MR)”, Proceedings<br />
of Second Information Exchange<br />
Meeting on <strong>Nuclear</strong> Production of<br />
Hydrogen, Argonne USA (2003).<br />
[6] UCHIDA, S. et. al., “Concept of<br />
Advanced FR-MR”, 15th World<br />
Hydrogen Energy Conference, Paper<br />
No. 30D-08, Yokohama Japan (2004).<br />
[7] CARTY, R.H., MAZUMDER, M.M.,<br />
SCHREIBER, J.D., PANGBORN,<br />
J.B., Thermochemical Hydrogen<br />
Production, GRI-80-0023, Institute of<br />
Gas Technology, Chicago, IL 60616<br />
(June 1981).<br />
[8] SERBAN, M., LEWIS, M.A., and<br />
BASCO, J.K., Kinetic Study for the<br />
Hydrogen and Oxygen Production<br />
Reactions in the Copper-Chlorine<br />
Thermochemical Cycle, 2004 AIChE<br />
Spring National Meeting, Conference<br />
Proceedings, 2004 AIChE Spring National<br />
Meeting, Conference Proceedings,<br />
pp. 2690-2698 (2004).<br />
[9] MILLER, W.E., MARONI, V.A. and<br />
WILLIT, J.L., DOE Patent Case<br />
Number S-104650 (2006).<br />
[10] SIMPSON, M.F., HERRMANN,<br />
S.D., and BOYLE, B.D., A Hybrid<br />
Thermochemical Electrolytic Process<br />
for Hydrogen Production Based<br />
on the Reverse Deacon Reaction,<br />
International <strong>Journal</strong> of Hydrogen<br />
Energy, 31 (Aug. 2006) 1241 - 1246.<br />
[11] NAKAGIRI, T. et. al., “A new<br />
thermo-chemical and electrolytic<br />
hybrid hydrogen production process<br />
for FBR”, Paper 1021, GENES4/<br />
ANP2003, Kyoto (Sep. 2003).<br />
Acknowledgements<br />
The IAEA appreciates the contributions<br />
of the following persons to this study: B.M.<br />
Misra (Consultant to IAEA, India); S. Kuran<br />
(Atomic Energy of Canada Ltd., Canada);<br />
L. Janik (<strong>Nuclear</strong> Research Institute Řež,<br />
Czech Rep.); D.S. Shukla (Bhabha Atomic<br />
Research Centre, India); M. Hori (<strong>Nuclear</strong><br />
Systems Association, Japan); T. Chirica<br />
(Societatea Nationala <strong>Nuclear</strong>electrica SA,<br />
Romania); V. Polunichev (Experimental<br />
Machine Design Bureau OKBM, Russian<br />
Federation); C. Halldin (OKG AB,<br />
Sweden); M. C. Petri (Argonne National<br />
Laboratory, USA, and Chairman of this<br />
activity); R. Uhrig (Univ. of Tennessee,<br />
USA); and E. Bertel (Organization for<br />
Economic Co-operation and Development<br />
- <strong>Nuclear</strong> Energy Agency).<br />
Contact: John Cleveland, International<br />
Atomic Energy Agency, P.O. Box 100,<br />
Vienna, A-140, Austria; telephone: 43-1-<br />
2600-22819, fax: 43-1-2600-29598, email:<br />
j.cleveland@iaea.org.<br />
<br />
60 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
Vermont's Largest Source of<br />
Electricity<br />
By Tyler Lamberts, Entergy <strong>Nuclear</strong><br />
Operations, Inc.<br />
Vermont Yankee nuclear power<br />
station is the largest in-state source of<br />
electricity. It provides about a third of the<br />
electricity used by Vermonters from its<br />
site on the Connecticut River in the town<br />
of Vernon.<br />
The plant was planned and<br />
constructed at a time when New England<br />
was heavily dependent on imported oil<br />
for electric generation. As oil supplies<br />
for New England grew more unstable and<br />
as the environmental degradation caused<br />
by fossil-fired pollution was becoming<br />
apparent, New England was among the<br />
first regions in the country to invest in<br />
nuclear plants as an alternative to fossilfueled<br />
power plants.<br />
Central Vermont Public Service and<br />
Green Mountain Power Corporation<br />
were the original lead utilities in the<br />
joint ownership of the 540 megawatt<br />
plant. After considering several Vermont<br />
sites, including the eastern shore of Lake<br />
Champlain, the 102 acre Vernon site on<br />
the western shore of the Connecticut<br />
River was selected. The site was chosen<br />
for its available land, sound bedrock,<br />
electric transmission lines, cooling water<br />
and its proximity to an active rail line<br />
for receiving large components and for<br />
shipping spent fuel.<br />
In 1972, after a four-year construction<br />
and federal licensing, the plant was<br />
connected to New England’s 345kv grid<br />
in time to position the state well against<br />
the 1974 Arab embargo on oil shipments<br />
to the United States.<br />
With Vermont Yankee reliably on<br />
line, fossil-fired power plants in the<br />
northeast were gradually edged out of the<br />
role of baseload electric generators – a<br />
major step in reducing air pollution in the<br />
region.<br />
In the late 1990’s, Vermont Yankee’s<br />
utility owners decided that the plant would<br />
fare better in every respect as part of a<br />
fleet of plants owned and operated by a<br />
utility specializing in nuclear generation.<br />
In 2001, Entergy was the high bidder for<br />
the Vermont Yankee plant. In 2002, the<br />
Vermont Public Service Board considered<br />
Entergy’s expertise and experience in the<br />
nuclear energy field, and approved the<br />
purchase of Vermont’s most valuable<br />
and reliable generating asset as being in<br />
the long-term best interest of the state of<br />
Vermont.<br />
Stakeholder Benefits<br />
As a condition of the sale, Entergy<br />
committed to supply the plant’s electricity<br />
to the utilities that formerly owned the<br />
plant at capped prices through to the<br />
end of the license term in 2012. Recent<br />
estimates by the Vermont Department<br />
of Public Service show that Vermonters<br />
are likely to save more than $665 million<br />
on their electric rates thanks to that<br />
agreement.<br />
In Entergy’s first year of Vermont<br />
Yankee ownership, it doubled the plant’s<br />
community contribution level including<br />
a large donation for restoration of a<br />
downtown theatre as a community<br />
cultural arts center.<br />
Overall, Vermont Yankee’s operation<br />
represents about $200 million of economic<br />
activity per year in the region through its<br />
payroll, taxes and local purchases.<br />
Extended Power Uprate<br />
It is Entergy’s goal, as owner and<br />
operator of Vermont’s largest generating<br />
asset, to maintain the plant’s favorable<br />
economics so as to continue to serve<br />
the region. The previous utility owners<br />
had found the plant to be an excellent<br />
Tyler Lamberts<br />
Tyler Lamberts graduated in June,<br />
2008 with a degree in Marketing from<br />
Oregon State University. Tyler currently<br />
works for OSU Conference Services in<br />
Corvallis, Oregon.<br />
candidate for a power uprate, but were<br />
not in a position to make the substantial<br />
investment as they were leaving the<br />
generation end of the utility business.<br />
After Entergy conducted its own 10-<br />
month in-house engineering evaluations,<br />
the company moved forward with a<br />
full 20-percent extended power uprate<br />
initiative. The Vermont Public Service<br />
Board approved the uprate in March<br />
of 2004 and the <strong>Nuclear</strong> Regulatory<br />
Commission followed suit two years<br />
later after a review by the Atomic Safety<br />
and Licensing Board and the Advisory<br />
Committee on Reactor Safeguards.<br />
According to the NRC, their staff review<br />
of Vermont Yankee’s uprate petition was<br />
the most extensive uprate review to-date<br />
involving more than 9,000 hours of NRC<br />
staff time.<br />
Entergy’s uprate power ascension<br />
program implemented over three months<br />
in the Spring of 2006 was notable for<br />
its deliberate and incremental approach<br />
that involved several hold points for<br />
plant performance data analyses and for<br />
communicating the results with General<br />
Electric, the <strong>Nuclear</strong> Steam Supply<br />
System designer, and state and federal<br />
regulators.<br />
Of particular interest during<br />
the ascension was the steam dryer<br />
performance. Similar boiling water<br />
reactors ascending to uprate power<br />
levels had experienced unexpected dryer<br />
degradation due to changes in steam line<br />
acoustics in the increased steam flow.<br />
Acoustic data collected by several dozen<br />
(Continued on page 62)<br />
<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 61
Vermont's Largest...<br />
Continued from page 61<br />
monitors on the steam piping was fully<br />
analyzed and compared with predictions<br />
before ascending to the next power level.<br />
In the first operating cycle following<br />
the uprate, Vermont Yankee posted a<br />
reliable breaker to breaker run of 549<br />
days and inspections of the dryer in the<br />
subsequent refueling outage found no<br />
flaws related to the new uprate steam flow<br />
and verified the accuracy of engineering<br />
analyses.<br />
Dry Cask Storage<br />
Another initiative underway at<br />
Vermont Yankee is construction of a dry<br />
fuel storage pad to allow Vermont Yankee<br />
to remain in service beyond 2008. The<br />
Vermont legislature and the Vermont<br />
Public Service Board approved the project<br />
in April 2006.<br />
In August 2007, local contractors<br />
completed a 1,050 cubic yard, 12-hour<br />
Intelligent Monitor...<br />
Continued from page 59<br />
3. Reactor Feed Pump (RFP) lube<br />
oil cooler temperature controller failure -<br />
$20K.<br />
The nuclear unit was recovering from<br />
the effects of a transformer failure induced<br />
voltage transient that caused some system<br />
isolations and momentary power losses.<br />
There was no significant plant transient.<br />
Shortly after the transient the RFP bearing<br />
cooling models for all 3 pumps went into<br />
alert. The plant was notified the following<br />
day that one of the controllers did not<br />
recover form and initial transient and was<br />
continuing to cycle significantly. The station<br />
determined that the controller for the C RFP<br />
oil cooler had failed and was able to stabilize<br />
temperatures manually until the controller<br />
was replaced.<br />
The worst case scenario is bearing<br />
damage due to rapid over heating and loss of<br />
the RFP. The physical damage is estimated<br />
at $100K with a probability of 0.10 and lost<br />
generation of 33% for 24 hrs or $200K with<br />
a probability of 0.050.<br />
The total avoided costs for the 2 month<br />
period is $550K. If detected failures of a<br />
continuous concrete pour for the ten<br />
thousand square foot pad.<br />
License Renewal<br />
In January of 2006, Entergy filed a 20-<br />
year license renewal request with the NRC<br />
to extend license expiration from 2012 to<br />
2032. The federal review is progressing<br />
well. In 2007, NRC staff issued the final<br />
Site Environmental Impact Statement<br />
and the draft Safety Evaluation Report.<br />
Also in 2007, the Advisory Committee<br />
on Reactor Safeguards sub-committee<br />
recommended proceeding with the full<br />
committee review of Vermont Yankee<br />
application.<br />
In 2008, the Atomic Safety and<br />
Licensing Board will hear several<br />
contentions brought by interveners and a<br />
state review process on Vermont Yankee<br />
license renewal will get underway.<br />
With the uprate, dry cask and license<br />
renewal initiatives in place, Vermont<br />
Yankee will continue as an economical<br />
and reliable source of electricity and<br />
a vital component of New England’s<br />
diversified energy mix.<br />
similar magnitude continue to be revealed<br />
by the centralized performance monitoring<br />
technology, an annualized avoidance of<br />
$3.3M can be expected. Avoidance of a<br />
failure of a generation critical component<br />
could also easily exceed this amount but<br />
the cost avoidance calculation methods ate<br />
conservative and follow methods similar<br />
to those in an EPRI technical paper on<br />
intelligent monitoring case studies.<br />
In addition to online monitoring the<br />
technology is being employed to assist in<br />
trouble shooting by focusing on discreet time<br />
frames and re-playing the plant conditions<br />
through the program to detect additional<br />
anomalies. The technology is also being<br />
promoted for increased monitoring when<br />
returning equipment and systems to service<br />
after maintenance. These two areas have the<br />
potential to increase the annualized savings<br />
from improved equipment reliability.<br />
Productivity/Efficiency:<br />
Work continues with the software vendor<br />
InStep to improve current productivity in<br />
investigating and acknowledging alerts that<br />
are generated by the software models.<br />
Additional efficiency gains are in<br />
progress relative to the integration of CPM<br />
into the Exelon model for performance<br />
Community Partnership<br />
The employees at Vermont<br />
Yankee play a vital role in neighboring<br />
communities by routinely supporting<br />
educational, civic and cultural projects<br />
and events. Over the years, they have<br />
volunteered their time as guest speakers<br />
at local schools, sponsored child daycare<br />
and learning centers, constructed<br />
playgrounds and taken an active role in<br />
local robotic competitions. Employees<br />
have also contributed to the education<br />
system as coaches, referees and mentors.<br />
Each year, employees participate in<br />
company-sponsored events such as the<br />
Brattleboro Fourth of July Celebration<br />
and the Winter Carnival. They also give<br />
their time, expertise and efforts to Habitat<br />
for Humanity. Vermont Yankee is also<br />
one of the founding sponsors and ongoing<br />
contributors to the local food drive called<br />
Project Feed the Thousands.<br />
Contact: Rob Williams, Vermont<br />
Yankee, P.O. Box 7002, 185 Old Ferry<br />
Road Brattleboro, VT 05302-7002;<br />
phone: (802) 258-4181; fax: (802) 258-<br />
2150; e-mail: rwill23@entergy.com. <br />
and equipment condition monitoring.<br />
The monitoring that is being performed<br />
by individual system managers can be<br />
optimized, standardized and integrated more<br />
effectively when considered together with<br />
all of the station monitoring activities that<br />
are performed by the various departments.<br />
Opportunities also exist to increase the<br />
number of sensors that are available for<br />
modeling and realize additional efficiencies<br />
to eliminate more time consuming, labor<br />
intensive and in many cases, less effective<br />
monitoring.<br />
Transferability:<br />
The use of this and similar intelligent<br />
monitoring technology within a centralized<br />
group monitoring a fleet of generating<br />
stations would apply across the industry. The<br />
recently evolved cluster based monitoring<br />
technology can also be implemented on a<br />
smaller scale at single units or a few units<br />
with similar result.<br />
Contact: Chris Demars, Exelon <strong>Nuclear</strong>,<br />
200 Exelon Way, KSA-2-N, Kennett Square,<br />
PA 19348; telephone: (610) 765-5427,<br />
pager: (800) 672-2285 PIN 0338, email:<br />
Christopher.demars@exeloncorp.com. <br />
62 http://subscribe.npjonline.com http://www.NPJOnline.com <strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, September-October 2008
WESTINGHOUSE HAS SOME<br />
simple ideas,<br />
TO ACCOMPLISH<br />
great things.<br />
PETE SENA<br />
Site Vice President<br />
Beaver Valley Power Station<br />
FENOC<br />
DAVE BALAS<br />
Manager of Global Outage Support<br />
Westinghouse<br />
JIM LASH<br />
Senior Vice President<br />
FENOC – Operations<br />
WESTINGHOUSE ELECTRIC COMPANY LLC<br />
When our Alliance Partner, FirstEnergy <strong>Nuclear</strong> Operating Company<br />
(FENOC), set a long-term goal to reduce outage dose at its Beaver Valley<br />
Power Station, they asked Westinghouse to assist.<br />
The Westinghouse team, led by Customer 1 st leader Dave Balas, worked<br />
with Pete Sena, Jim Lash and the Beaver Valley Power Station to apply<br />
Customer 1 st tools. As a result, the plant achieved a 39 percent dose<br />
reduction during the fall 2007 Unit 1 outage, advancing Beaver Valley<br />
from the industry’s fourth to second quartile in outage dose performance.<br />
During Beaver Valley’s spring 2008 outage, results were even better as<br />
Unit 2 advanced from fourth to first quartile in outage dose performance.<br />
Improving outage performance and applying the benefits and industry<br />
lessons learned are just a few ways that Westinghouse nuclear technology<br />
is strengthening performance at the world’s leading nuclear power plants.<br />
Check us out at www.westinghousenuclear.com<br />
Committed to customer success.<br />
“ Unit 1 achieved its lowest outage<br />
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Westinghouse outage windows.”<br />
—PETE SENA
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