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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 />

critical inspection applications for<br />

the power generation industry<br />

Snoqualmie, Washington, USA • Deep River, Ontario • Quebec, Canada<br />

Paris, France • Seoul, Korea • Beijing, China<br />

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 />

for six issues per year. The additional air<br />

mail cost for non-U.S. readers is $30.00.<br />

Payment may be made by American Express<br />

® , Master Card ® , VISA ® or check<br />

and should accompany the order. Checks<br />

not drawn on a United States bank should<br />

include an additional $45.00 service fee.<br />

All inquiries should be addressed to<br />

<strong>Nuclear</strong> <strong>Plant</strong> <strong>Journal</strong>, 799 Roosevelt<br />

Road, Building 6, Suite 208, Glen Ellyn,<br />

IL 60137-5925; Phone: (630) 858-6161,<br />

ext. 103; Fax: (630) 858-8787.<br />

*Current Circulation:<br />

Total: 12,000<br />

Utilities: 4,600<br />

*All circulation information is subject to<br />

BPA Worldwide, Business audit.<br />

Authorization to photocopy articles is<br />

granted by EQES, Inc. provided that<br />

payment is made to the Copyright<br />

Clearance Center, 222 Rosewood Drive,<br />

Danvers, MA 01923; Phone: (978) 750-<br />

8400, Fax: (978) 646-8600. The fee code<br />

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 />

Confidence<br />

built on<br />

experience.<br />

How do you keep your nuclear deployment<br />

plans on track to succeed Choosing a<br />

development partner with a proven track<br />

record is the perfect way to get started. For<br />

decades, GE Hitachi <strong>Nuclear</strong> Energy (GEH) has<br />

been developing nuclear plants on schedule<br />

and on budget while minimizing risk. Our<br />

approach combines GEH experience with<br />

innovation to give you more options – from<br />

proven ABWR Generation-III solutions to<br />

the simplicity of the evolved ESBWR design.<br />

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 />

most comprehensive software package available, we support today’s<br />

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 />

Chicago, IL • 630.357.0156<br />

www.protopower.com<br />

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 />

2. Readers interested in cost-free subscription may access<br />

the web site subscribe.NPJOnline.com to request or<br />

renew their subscription.<br />

• Readers in the United States or Canada may subscribe<br />

to the paper or digital version without any charge.<br />

• Readers worldwide may subscribe to the digital<br />

version without any charge. Additional subscription<br />

charges apply for the paper version for the<br />

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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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 />

2009 Directory<br />

All nuclear power industry suppliers who<br />

are not listed in the 2008 Directory may register<br />

for the 2009 Directory by sending an email to<br />

npj@goinfo.com with complete contact information.<br />

Suppliers listed <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 />

Published in the United States<br />

<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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A Market...<br />

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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reference plants. The move to digitalbased<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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A Market...<br />

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


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Improving outage performance and applying the benefits and industry<br />

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