Nuclear Plant Journal - Digital Versions

Nuclear Plant Journal - Digital Versions




Plant Maintenance &

Advanced Reactors Issue

September-October 2008

Volume 26 No. 5

ISSN: 0892-2055

Vermont Yankee, USA


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Nuclear Plant Journal

September-October 2008, Volume 26 No 5

Plant Maintenance &

Advanced Reactor Issue

26th Year of Publication

Nuclear Plant Journal is published by

EQES, Inc.six times a year in February,

April, June, August, October and December


The subscription rate for non-qualified

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Senior Publisher and Editor

Newal K. Agnihotri

Publisher and Sales Manager

Anu Agnihotri

Editorial & Marketing Assistant

Michelle Yong

Administrative Assistant

QingQing Zhu

Articles & Reports

Technologies of National Importance 16

By Tsutomu Ohkubo, Japan Atomic Energy Agency, Japan

Modeling & Simulation Advances Brighten Future Nuclear Power 18

By Hussein Khalil, Argonne National Laboratory

Energy & Desalination Projects 22

By Ratan Kumar Sinha, Bhabha Atomic Research Centre, India

A Plant with Simplified Design 24

By John Higgins, GE Hitachi Nuclear Energy

A Forward Thinking Design 27

By Ray Ganthner, AREVA

A Passively Safe Design 32

By Ed Cummins, Westinghouse Electric Company

A Market-Ready Design 34

By Ken Petrunik, Atomic Energy of Canada Limited, Canada

Generation IV Advanced Nuclear Energy Systems 42

By Jacques Bouchard, French Commissariat a l'Energie Atomique, France

and Ralph Bennett, Idaho National Laboratory

Innovative Reactor Designs 46

A Report by IAEA, Vienna, Austria

Guidance For New Vendors 52

By John Nakoski, U.S. Nuclear Regulatory Commission

Road Map for Future Energy 54

By John Cleveland, International Atomic Energy Agency, Vienna, Austria

Vermont's Largest Source of Electricity 61

By Tyler Lamberts, Entergy Nuclear Operations, Inc.

Industry Innovations

Intelligent Monitoring Technology 59

By Chris Demars, Exelon Nuclear


New Energy News 8

Utility, Industry & Corporation 10

New Products, Services & Contracts 12

New Documents 14

Meeting & Training Calendar 15

Journal Services

List of Advertisers 6

Advertiser Web Directory 14

On The Cover

Vermont Yankee is a nuclear site located

in Vermont. The plant is currently owned

by Entergy Nuclear Vermont Yankee, LLC,

and operated by Entergy’s nuclear business

function. The unit is a boiling water

reactor designed by General Electric Co.,

and has a net generating capacity of 587

dependable megawatts. See page 61 for

a profi le.

Mailing Identification Statement

Nuclear Plant Journal (ISSN 0892-2055) is published bimonthly in February,

April, June, August, October and December by EQES, Inc., 799 Roosevelt Road,

Building 6, Suite 208, Glen Ellyn, IL 60137-5925. The Journal is available costfree

to qualified readers worldwide. The subscription rate for non-qualified readers

is $150.00 per year. The cost for non-qualified, non-U.S. readers is $180.00. Periodicals (permit

number 000-739) postage paid at the Glen Ellyn, IL 60137 and additional mailing offices. POSTMAS-

TER: Send address changes to Nuclear Plant Journal (EQES, Inc.), 799 Roosevelt Road, Building 6,

Suite 208, Glen Ellyn, IL 60137-5925.

Nuclear Plant Journal, September-October 2008 5

List of Advertisers & NPJ Rapid Response

Page Advertiser Contact Fax/Email

19 Atomic Energy of Canada Limited Heather Smith (905) 403-7565

2 AREVA NP, Inc. Donna Gaddy-Bowen (434) 832-3840

31 Babcock & Wilcox Canada Ltd Yvette Amor (519) 621-9681

21 Bechtel Power

45 Bigge Power Constructors Andrew Wierda (510) 639-4053

37 Black & Veatch Keith Gusich (913) 458-2491

15 Ceradyne Patti Bass (714) 675-6565

41 Climax Portable Machine Tools, Inc. Debra Horn

47 Data Systems & Solutions Romain Desgeorge 33 (0) 4 76 61 17 07

49 Day & Zimmermann NPS David Bronczyk (215) 299-8395

29 Enertech Tom Schell

7 GE Hitachi Nuclear Energy Mark Marano (910) 362-5017

25 HSB Global Standards Louise Hamburger

38 Meggitt Safety Systems Jennifer Cetta (805) 584-9157

43 National Enrichment Facility Dana Starr (575) 394-0175

11 NPTS, Inc. Rebecca Broman (716) 876-8004

55 Nuclear Logistics Inc. Craig Irish (978) 250-0245

43 Power House Tool, Inc. Laura Patterson (815) 727-4835

8 Proto-Power Corporation Christopher D’Angelo (860) 446-8292

39 The Shaw Group Inc. Holly Nava (856) 482-3155

51 Thermo Fisher Scientific Tony Chapman (315) 451-9421

64 Trentec, Inc. Arlene Corkhill (714) 528-0128

13 Underwater Construction Charles Vallance (321) 779-4462

4 UniStar Nuclear Energy Mary Klett (410)470-5606

9 UniTech Services Group Steve Hofstatter (413) 543-2975

35 Urenco Enrichment Company Ltd Please e-mail

26 Westerman Companies Jim Christian (740) 569-4111

63 Westinghouse Electric Company LLC Karen Fischetti (412) 374-3244

17 WM Symposia, Inc. Mary E. Young

3 Zetec, Inc. Katina Baarslag (425) 974-2678

Information may be directly obtained from advertisers by faxing this page to the individual advertiser after completing

the bottom part of the Rapid Response Fax Form. Advertisers’ web sites are listed in the Web Directory Listings

on page 14.

Nuclear Plant Journal Rapid Response Fax Form

From the September-October 2008

issue of Nuclear Plant Journal

To: _________________________ Company: __________________ Fax: ___________________

From: _______________________ Company: __________________ Fax: ___________________

Address:_____________________ City: _______________________ State: _____ Zip: _________

Phone: ______________________ E-mail: _____________________

I am interested in obtaining information on: __________________________________________________

Comments: _____________________________________________________________________________

6 Nuclear Plant Journal, September-October 2008

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New Energy News


AmerenUE, a Missouri-based utility

subsidiary of Ameren Corporation submitted

a combined Construction and Operating

License Application (COLA) to

the U.S. Nuclear Regulatory Commission

(NRC) for a potential new nuclear power

plant in Callaway County, Missouri.

The 8,000-page application seeks

regulatory approvals to potentially build

a new 1,600-megawatt pressurized water

reactor adjacent to AmerenUE’s singleunit,

1,190-megawatt Callaway electric

generating plant which accounts for 19

percent of the company’s total generation.

Since the Callaway Plant came on line

in December 1984, it has achieved the

fourth highest generation output among

the nation’s 104 nuclear power units.

Contact: Mike Cleary, telephone: (573)

681-7137, email:

Strategy Report

The U.S. Department of Energy

(DOE) and the U.S. Nuclear Regulatory

Commission (NRC) delivered to

Congress the Next Generation Nuclear

Plant (NGNP) Licensing Strategy Report

which describes the licensing approach,

the analytical tools, the research and

development activities and the estimated

resources required to license an advanced

reactor design by 2017 and begin operation

by 2021. The NGNP represents a new

concept for nuclear energy utilization,

in which a gas-cooled reactor provides

process heat for any number of industrial

applications including electricity

production, hydrogen production, coalto-liquids,

shale oil recovery, fertilizer

production, and other applications that

meet significant industrial needs.

Visit to read the joint

Licensing Strategy Report and to learn

more about DOE’s Office of Nuclear


Contact: Angela Hill, telephone:

(202) 586-4940.

Loan Guarantee

Dominion Virginia Power submitted

to the U.S. Department of Energy

the first part of an application for a loan

guarantee as it considers a third nuclear

reactor at the North Anna Power Station

in Central Virginia.

“Today’s filing is another important

step in the process began more than seven

years ago to position ourselves to be

among the first to get a license for a new

nuclear unit,” said Mark F. McGettrick,

president and chief executive officer of

Dominion Generation.

Contact: Richard Zuercher, telephone: (804)

273-3825, email:

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Circle 107 on Reader Service Form

8 Nuclear Plant Journal, September-October 2008

Construction Agreement

After the signing of a framework

agreement, November 26, 2007 in Beijing,

in the presence of both head of States of

France and China, EDF and the Chinese

electricity producer China Guangdong

Nuclear Power Holding Company signed

the final agreements in Beijing for the

creation of a joint venture company to

be called Guangdong Taishan Nuclear

Power Joint Venture Company Limited

(TNPC). The aim of the joint venture is

to construct and operate two nuclear EPR

power stations at Taishan in the province

of Guangdong, modeled on the existing

EPR reactor built by EDF at Flamanville

in Normandy, France.

Preliminary work at the Taishan Unit 1

site started in late 2007 and the first concrete

pouring is scheduled for autumn

2009, less than two years after the one

at Flamanville 3. Some contracts have

already been signed with Areva and Alstom

for the supply of the nuclear and the

turbine equipment respectively. The first

unit should be commissioned at the end

of 2013 and the second in 2015. At the

height of construction work, over sixty

EDF experts will be on-site at Taishan.

Contact: Carole Trivi, telephone: 33

1 40 42 44 19.

123 Agreement

Statement by the Prime Minister of India

"We welcome the decision earlier

today of the Nuclear Suppliers Group to

adjust its guidelines to enable full civil

nuclear cooperation with India. This is a

forward-looking and momentous decision.

It marks the end of India's decades long

isolation from the nuclear mainstream

and of the technology denial regime. It is

a recognition of India's impeccable nonproliferation

credentials and its status as

a state with advanced nuclear technology.

It will give an impetus to India's pursuit

of environmentally sustainable economic


Contact: telephone: 43 1 2600-0,

fax: 43 1 2600-7.

Application Submitted

Progress Energy Florida, a subsidiary

of Progress Energy submitted a combined

license (COL) application with the

Nuclear Regulatory Commission (NRC)

to construct a new nuclear power plant in

Levy County, Florida.

The application, submitted to the

NRC on July 30, 2008, included the request

to build two Westinghouse AP1000

nuclear reactors at the site. Nuclear power

is a key component of Progress Energy

Florida’s balanced solution strategy to

meet Florida’s long-term energy needs.

Nuclear power, along with additional renewable

energy resources and expanded

energy-efficiency programs, is Progress

Energy Florida’s strategy to address climate

change and the need for greater fuel


Contact: telephone: (919) 546-6189.


Map To




CO2, Plastic Bead, Ultrasonic, High Pressure Water, Steam

Excavation Started

The Shandong Nuclear Power Company

with Westinghouse Electric Company

LLC and its consortium partner

The Shaw Group Inc. broke ground one

month earlier than scheduled on the Haiyang

Nuclear Power Facility in Shandong


The Haiyang facility will house two

nuclear plants, each deploying Westinghouse’s

AP1000 technology. Excavation

for the first of the two plants will take approximately

three months to create a hole

12 meters deep (39 feet) that will house

the nuclear reactor and turbine buildings.

The volume of the excavation is approximately

48,916 cubic meters or about 19.5

Olympic-size swimming pools. When

completed, a base for the plant nearly

175 meters wide (570 feet) by 250 meters

long (840 feet) will exist.

Contact: Vaughn Gilbert, telephone:

(412) 374-3896, email:

Outage management of customer equipment; long term storage

HEPA ventilation; tools and scaffolding; HP instruments



11 Licensed





(800) 344-3824

Nuclear Plant Journal, September-October 2008 9

Utility, Industry & Corporation


Utility Achievement


Constellation Energy announced

that its Calvert Cliffs Nuclear Power

Plant (CCNPP) has been awarded the

American Nuclear Society’s 2008

Utility Achievement Award for sustained

outstanding performance. Jim Spina, vice

president at CCNPP, accepted the award on

behalf of the approximately 780 employees

at Calvert Cliffs at a conference hosted by

the American Nuclear Society in Amelia

Island, Florida.

Calvert Cliffs was recognized for

demonstrating a prolonged dedication to

safe nuclear generation as evidenced by a

record high capacity factor and the highest

site generation in four of the last five


Contact: Dave Fitz, telephone: (888)


Reader Service Card & Cost-free

Subscription Cards

1. The reader service inquiries are now available

online by logging on to

2. Readers interested in cost-free subscription may access

the web site to request or

renew their subscription.

• Readers in the United States or Canada may subscribe

to the paper or digital version without any charge.

• Readers worldwide may subscribe to the digital

version without any charge. Additional subscription

charges apply for the paper version for the

international readers other than US & Canada .

Contact for details.

License Renewal

Dominion, owner of the Kewaunee

Power Station, filed an application to renew

the facility’s operating license with the U.S.

Nuclear Regulatory Commission (NRC).

The 568-megawatt nuclear unit is

licensed to operate through December 21,

2013. With a renewed license, the station

would be able to provide Wisconsin with

safe, clean and reliable electricity through

December 21, 2033.

Contact: Mark Kanz, telephone: (920)


Joint Effort

Entergy Nuclear and the Taiwan

Power Company announced a joint effort

tapping Entergy’s experience in license

renewal efforts to allow for long term

operations at Taiwan’s Kuosheng Nuclear

Power Plant.

The Institute of Nuclear Energy Research,

Taiwan, which advances nuclear

technology and assures national nuclear




An International Publication

A Digital (electronic) Version of NPJ

is Now Available!

safety, has been contracted by the Taiwan

Power Company to initiate a project

to allow for extended operation of TPC’s

Kuosheng plant, a dual unit site with boiling

water reactors constructed in the early


Contact: Mike Bowling, telephone:

(601) 368-5655, email:

New Website

Exelon Nuclear announced the launch

of a new Texas-based Web site intended

to keep the public updated and informed

about the company’s proposed Victoria

County nuclear plant. The Web site address


Contact: Bill Harris, telephone: (361)




Commissariat français à l’énergie

atomique, Cadarache, France, is the host

for the ITER project constructing an

experimental nuclear fusion reactor using

hydrogen isotopes.

The site was selected in 2005 by the

international partners of the ITER project

(India, China, South Korea, Japan, Russia,

the United States and the European Union).

It will take 10 years to build the project and

a further 20 years of scientific experiments

to prove that fusion can become a new

reliable source of energy.

Contact: Benoit Gausseron, telephone:

(212) 757-9340, email:

Award for Technology

The U.S. Department of Energy

(DOE) awarded up to $15 million to

34 research organizations as part of the

Department’s Advanced Fuel Cycle

Initiative (AFCI).

For a list of recipients please go to,


Contact: Angela Hill, telephone: (202)


10 Nuclear Plant Journal, September-October 2008



The Babcock & Wilcox Company

(B&W), a subsidiary of McDermott

International, Inc. announced that an

affiliate of B&W has entered into a

definitive agreement to acquire Nuclear

Fuel Services, Inc. (NFS) of Erwin, Tenn.,

a provider of specialty nuclear fuels and

related services. The acquisition supports

B&W’s strategic goal of being a leading

provider of nuclear manufacturing and

service businesses for government and

commercial markets.

Contact: Steve Stultz, telephone: (330)

860-6124, email:

New Facility

Day & Zimmermann, announced

its Maintenance and Modification unit has

completed the acquisition of a fabrication

and machining facility in Moss Point,

Mississippi, from Industrial Maintenance

and Machine, Inc. The facility will be

operated by DZ Atlantic, a wholly owned

subsidiary of Day & Zimmermann.

“Having a fabrication facility will allow

us to meet the needs of our existing

customers and further develop customer

relationships in other targeted industries,”

said Mike McMahon, President of Day &

Zimmermann’s Maintenance and Modification


The facility consists of a 180,000-sq.-

ft. shop situated on 20 acres of land, and

will give DZ Atlantic a significant range of

capabilities including machining, mobile

machining, structural steel production,

piping, skids, and specialty welding.

Contact: Maureen Omrod, telephone:

(215) 299-2234, email:


ENERCON, a 700-employee firm

serving energy and environmental clients

nationwide, has acquired EPIC Consulting,

Inc. of Marietta, Georgia, an environmental

and geotechnical firm specializing in highly

customized solutions primarily in energyand

environmentally-related businesses.

ENERCON Vice President John Corn

said, “EPIC is an excellent fit for ENER-

CON and will complement our environmental

and technical services. EPIC and

ENERCON’s environmental division provide

similar services but their geotechnical

expertise is a great addition to our


Contact: Peggy Striegel, telephone:

(918) 740-5584, email:

Platform for Future


Numet Engineering Ltd., a supplier

of specialized, high-reliability precision

engineered systems and equipment for

the nuclear energy & hazardous waste

management sectors has been acquired

by the ODIM Group. The company will

continue to operate as Numet Engineering

Ltd. And will continue to exclusively

focus towards the nuclear power industry.

For the ODIM Group, the Numet

acquisition brings a strong and well respected

presence to the Canadian nuclear

power sector.

Contact: Bill Potter, telephone: (705)

743-2708, email:


Exelon Corporation, recently selected

Scientech’s award-winning PMAX

software as its tool for on-line thermal

performance monitoring at its 17 nuclear

generating units. PMAX is renowned for

its ability as a software tool to assist engineers

and operators to identify megawatt

losses and reveal plant thermal performance

inefficiencies – in a real-time environment.

PMAX has been adopted worldwide

by over 300 thermal power plants (nuclear,

fossil, and combined cycle), and is now

the on-line thermal performance tool of

choice at 59 of the nation’s 104 nuclear

power plant units.

Contact: Ed Hollis, telephone: (301)

371-7485, email:

Module, Construction

Westinghouse Electric Company

and The Shaw Group Inc. signed a letter

of intent (LOI) to form a joint venture

to fabricate and assemble structural and

equipment modules for AP1000 nuclear

power plants to be built in the United

States and selected global markets in

which in-country supply is not available.

Under terms of the LOI, Westinghouse

and Shaw will each hold ownership shares

in the joint venture.

The new company, Global Modular

Solutions LLC, will construct a 600,000

sq. ft. facility in Lake Charles, Louisiana

that is scheduled to begin operation in

the late summer of 2009. When fully

operational, the facility is expected to

employ as many as 1,400 workers.

Contact: Vaughn Gilbert, telephone: (412)

374-3896, email:

Strategy & Research

Dr. Kathryn Jackson has been

appointed to the position of vice president,

Strategy, Research and Technology at

Westinghouse Electric Company. Dr.

Jackson was previously the executive vice

president of River System Operations and

Environment at Tennessee Valley Authority

(TVA), where she has served since 1998.

She holds a master’s in Industrial

Engineering Management from the

University of Pittsburgh (1983) and master’s

and doctorate degrees in Engineering

and Public Policy from Carnegie Mellon

University (1987 and 1990).

Contact: Vaughn Gilbert, telephone:

(412) 374-3896, email:

NPTS, Inc.

an Engineering, Design, and

Construction Management firm has

current and anticipated openings for the

following positions:

Licensing, USAR & Regulatory


Engineering Design (All Disciplines)

Sr. Project Managers (All


Sr. Project Planners (All Disciplines)

Power Upgrade Project Engineers

Construction Management, Planners,

Schedulers, Estimators

• Resident Engineers (All Disciplines)

• Operations Support Engineers

• Operations Training Instructors

• Procurement Specialists &


• Start-up & Commissioning


For Power Uprates, New Builds, Life

Extension, Upgrades, Modification

and Maintenance Projects

Please forward Resumes to:

NPTS, Inc.

2060 Sheridan Drive

Buffalo, New York 14221

Phone: 716.876.8066

Fax: 716.876.8004


Nuclear Plant Journal, September-October 2008 11

New Products, Services & Contracts

New Products

Ultrasonic Flaw


GE Sensing & Inspection Technologies

introduces a new family of ultrasonic

flaw detectors, providing inspectors with

a flexible platform, as inspection needs

change. The Phasor family incorporates

conventional and phased array ultrasound

technology in three upgradeable models:

Phasor CV, Phasor 16/16 Weld and Phasor

XS. The tiered platform offers inspectors

the opportunity to select the model

that best suits their specific application in

oil & gas, power generation, aerospace or


Contact: Amanda Fontaine,


Walking Robot

Zetec, Inc., the total solution

nondestructive testing (NDT) provider

for the Power Generation industry,

announced it will launch the industry’s

most flexible and functional tube sheet

walking robot at the 27th Steam Generator

NDE Workshop.

Small in size and weighing less than

35 lbs, the ZR-100 provides ultimate

flexibility in reaching all of the tubes

within the tube sheet without complex

repositioning motions. This provides

quick and efficient motion in positioning

the ZR-100 to a target zone or specific

tube. All of this is accomplished while

providing industry leading speed.

The ZR-100 can transverse across the

tube sheet at speeds of up to 5 feet per

minute for large moves and can achieve

tube-to-tube speeds during test or repair

operations of up to 4 inches/second. The

Nuclear Plant Journal’s

Product & Service Directory 2009

2009 Directory

All nuclear power industry suppliers who are not listed

in the 2008 Directory may register for the 2009 Directory

by sending an email to with complete

contact information.

Suppliers listed in Nuclear Plant Journal's 2008

Directory will receive the 2009 Directory mailing

with a list of their products and services as they

appeared in the 2008 Directory.


Input Form- November 12, 2008

Ad Committment- November 12, 2008



Telephone: 630-858-6161, ext. 103

FAx: 630-858-8787




An International Publication

Product & Service Directory 2009

ZR-100 utilizes built-in Machine Vision

for secondary tube verification for all

attached tooling.

Contact: Katina Baarslag, telephone:

(425) 974-2678, email:


Inspection Time


Toshiba GE Turbine Components

(TGTC) has reduced the time required to

inspect and measure steam turbine blades

from 280 minutes to 45 minutes by using

the MAXOS non-contact measurement

system from Steintek GmbH (Greding,

Germany). The coordinate measuring

machine (CMM) used in the past to

inspect the blades was not only slow but

was unable to access hard-to-reach areas

such as dovetail hooks and fillets. The

MAXOS uses five axes to reach every

point on the blades and also generates

specific and accurate measurements of

critical areas. Resulting measurements

are reported instantly and the need

for additional manual inspection is


“The MAXOS optical scanner provides

the best possible accuracy, eliminates

the need for matt coating, and

integrates easily with our engineering

and production processes,” said Tomio

Kubota, President of TGTC. “Our trials

also demonstrated that the MAXOS is

significantly faster than the other systems

that we considered. The expertise and

professionalism that were evident during

this trial gave us the confidence to adopt

this new technology.”

Contact: NVision Inc (Southlake,

TX and Wixom, MI), telephone: (248)

268-2525, email:

ASME Renewal


TechPrecision Corporation, a

manufacturer of large-scale, highprecision

machined metal fabrications for

the alternative energy, medical, nuclear,

12 Nuclear Plant Journal, September-October 2008

defense, aerospace and other commercial

industries, announced that its wholly

owned subsidiary, Ranor, Inc. received

its renewal Certificates of Authorization

from the American Society of Mechanical

Engineers (“ASME”). The Certificates

of Authorization cover the Company’s

facilities in Westminster, Massachusetts

and are an integral part of Ranor’s

ongoing business plan to be a supplier to

the emerging nuclear renaissance.

Contact: Amanda Lleshdedaj,

telephone: (310) 477-9800, email:


Turbine Island

Alstom signed a contract worth over

200 million euros with China Guang Dong

Nuclear Power Company (CGNPC) for

the engineering and procurement of the

complete turbine island for the nuclear

power plant to be built in Taishan (southwestern

province of Guangdong). Taishan

will be China’s first EPR power plant.

This contract follows the $300

million order (including around $100

million for Alstom) booked in February

2008 and won in partnership with the

Chinese industrial group and Alstom’s

long-standing partner, Dongfang Electric

Company. This first order is for the supply

of two 1,750 MW Arabelle turbinegenerator

packages for the Taishan

nuclear plant.

Contact: Philippe Kasse, telephone:

33 1 41 49 29 82/33 08, email:

Nuclear Fuel Assemblies

AREVA has signed a contract with

Taiwan Power Company (Taipower) to

supply boiling water reactor fuel assemblies

for units 1 and 2 of the Chinshan

and Kuosheng nuclear power plants. The

award, worth more than $200 million,

is the conclusion of an invitation to bid

launched in June 2007.

The scope of work includes five

firm reload batches and three optional

reload batches for each unit. AREVA

will provide core monitoring system

assistance in addition to the fabrication

service, reload fuel design, licensing

analysis and operation support.

Contact: Laurence Pernot, telephone:

(301) 841-1694, email: Laurence.

Project Contract

SNC-Lavalin Nuclear and Murray

& Roberts announced that Pebble Bed

Modular Reactor (Pty) Ltd has awarded

their joint venture company, Murray &

Roberts SNC-Lavalin Nuclear (Pty) Ltd.

(MRSLN), a contract to provide engineering,

procurement, project and construction

management services for Phase

II of the Pebble Bed Modular Reactor

(PBMR) Demonstration Power Plant at

Koeberg, South Africa.

Phase II of the project entails

construction of a commercial scale power

plant at Koeberg near Cape Town, which

is subject to obtaining a nuclear licence

from the National Nuclear Regulator

and a positive Record of Decision on the

Environmental Impact Assessment.

Contact: Gillian MacCormack,

telephone: (514) 393-8000 ext. 7354.

Steam Generator


Studsvik has received an order for

the treatment and metal recycling of

three steam generators. The customer

is Vattenfall Ringhals in Scandinavia,

and the order is received under the

existing Memorandum of Understanding

concerning treatment of large components

signed in 2006. The steam generators are

planned to be delivered to Studsvik during

fall 2008 and the treatment is planned to

start during the first quarter 2009. The

contract value is SEK 34 million.

Contact: Magnus Groth, telephone:

46 155 22 10 86.



Nuclear Plant Journal, September-October 2008 13

New Documents


1. BWR Vessel and Internals Project,

Evaluation of RAMA Fluence

Methodology Calculational

Uncertainty, Product ID: 1016938,

Published July 2008.

This report documents the overall

calculational uncertainty associated with

the application of the Radiation Application

Modeling Application (RAMA) Fluence

Methodology to BWR reactor pressure

vessel fluence evaluations.

2. Feasibility of Direct Disposal of

Dual-Purpose Canisters in a High-

Level Waste Repository, Product ID:

1018051, Published August 2008.

A deep geologic repository at Yucca

Mountain, Nevada, has been proposed for

the disposal of commercial spent nuclear

fuel (CSNF) and other nuclear fuel and

high level radioactive waste (HLW) from

defense and nuclear weapons programs.

Atomic Energy of

Canada Limited


Babcock & Wilcox

Canada Ltd.

Bechtel Power

Bigge Power Constructors

Black & Veatch


Climax Portable

Machine Tools, Inc.

Data Systems & Solutions

The U.S. Department of Energy

(DOE) has proposed a standardized

transportation, aging and disposal (TAD)

canister for emplacement of CSNF at

Yucca Mountain.

3. Study to Identify Potential

Improvements of Operation

Tools and Support Systems–Non-

Proprietary, Product ID: 1016730,

Published August 2008.

This project analyzed safety

significant events (SSEs) in several

nuclear power plants to identify where

improvements in instrumentation and

control (I&C) and information technology

(IT) could prevent or mitigate some

of these events. This report identifies

potential improvement paths that could

enhance reliability and availability for

implementation consideration by utilities

where appropriate at their own plants.

4. Program on Technology Innovation:

Using Information Technology

to Increase Nuclear Power Plant

Performance, Product ID: 1016962,

Published August 2008.

As current nuclear power plants

(NPPs) continue to operate for the next

20–30 years, certain issues are driving the

plants to come up with new ways of doing

work. Solutions to these issues may

be possible using modern information

technology (IT). This can include the use

of both software and hardware and can

encompass traditional corporate IT systems

as well as plant instrumentation and

control (I&C) systems.

The above document may be obtained

from EPRI Order and Conference Center,

1300 West WT Harris Blvd., Charlotte,

NC 28262; telephone: (800) 313-3774,


NPJ Advertiser Web Directory

Day & Zimmermann NPS Thermo Fisher Scientific


GE Hitachi Nuclear Energy

HSB Global Standards

Meggitt Safety Systems

National Enrichment Facility

NPTS, Inc.

Nuclear Logistics Inc.

Power House Tool, Inc.

Proto-Power Corporation

The Shaw Group Inc.

Trentec, Inc.

Underwater Construction

UniStar Nuclear Energy

UniTech Services Group

Urenco Enrichment Company


Westerman Companies

Westinghouse Electric

Company LLC

WM Symposia, Inc.

Zetec, Inc.

14 Nuclear Plant Journal, September-October 2008

Meeting & Training Calendar

1. NEI International Uranium Fuel

Seminar, October 19-22, 2008, The

Westin Tarbor Center, Denver Colorado.

Contact: Nuclear Energy Institute,

Janet Schluester, telephone:

(202) 739-8098, email:

2. Technical Meeting on the International

Decommissioning Network,

October 20-24, 2008, Vienna, Austria.

Contact: International Atomic

Energy Agency, P. Dinner, email:


3. EPRI 7 International Decommissioning

& Radioactive Waste Workshop,

October 28-October 30, 2008,

Hotel Hilton Lyon, Lyon, France.

Contact: Electric Power Research

Institute, Sean Bushart, telephone:

(650) 855-2978, email:

4. 23rd Canadian Nuclear Society

Nuclear Simulation Symposium,

November 2-4, 2008, Ottawa,

Ontario, Canada. Contact: Denise

Rouben, CNS, telephone: (416) 977-

7620, email:

5. Future Power, November 4-5, 2008,

London. Contact: Nuclear Engineering

International, telephone:

44 0 208 2697 812, website: www.

6. Winter Meeting and Nuclear Technology

Expo, November 9-13, 2008,

Reno, Nevada. Contact: American

Nuclear Society, telephone: (708)


7. Technical Meeting to Maintain and

Update the Nuclear Fuel Cycle Information

System, November 12-14,

2008, Vienna, Austria. Contact: International

Atomic Energy Agency,

M. Ceyhan, email:


8. 8 International Conference on

CANDU Maintenance, November 16-

18, 2008, Metro Toronto Convention

Centre and InterContinental Toronto

Centre Hotel, Toronto, Ontario.

Contact: Denise Rouben, CNS,

telephone: (416-977-7620, email:

9. November 17-20, 2008, Las Vegas,

Nevada. Contact: Argonne National

Laboratory, Lawrence Boing,

telephone: (630) 252-6729, email:


10. 46 Semiannual Nuclear Fuel Management

Seminar, November 17-20,

2008, Atlanta, Georgia. Contact:

Christina DeLance, NAC International,

telephone: (678) 328-1281,


11. Boiler and Reactor Feedpump Turbine

Workshop, November 18-20,

2008, Nashville Marriot at Vanderbilt

University, Nashville, Tennessee.

Contact: Electric Power Research

Institute, Linda Parrish, telephone:

(704) 5952-2000.

12. The Nuclear Power Congress 2008,

December 9-10, 2008, The Ritz-

Carlton Golf Resort, Naples Florida.

Contact: Kristy Perkins, American

Conference Institute, email:



13. WM 2008 Phoenix, Waste Management

for the Nuclear Renaissance,

March 1-5, 2009, Phoenix, Arizona.

Contact: WMS Administration,

telephone: (520) 696-0399, email:

14. World Nuclear Fuel Cycle 2009,

April 22-24, 2009, Sydney, Australia.

Contact: Stuart Cloke, World

Nuclear telephone: 44 207 451 1520,


15. Annual Meeting on Nuclear Technology,

May 12-14, 2009, Congress

Center Dresden, Germany. Contact:

dbcm GmbH, telephone: 49 02241

93897 0, email:

Neutron Absorber


BORAL ® Composite


Borated Aluminum

Enriched Boron

Natural Boron Carbide


Nuclear Plant Journal, September-October 2008 15

Technologies of National


By Tsutomu Ohkubo, Japan Atomic

Energy Agency.

1. Please provide a brief description of

RMWR 300MW(e)/X.

The reduced-moderation water

reactor (RMWR) is a BWR-type reactor

being developed to ensure the sustainable

energy supply in the future through

multiple recycling of plutonium based on

the well-developed and experienced LWR

technologies. The RMWR core consists

of hexagonal fuel assemblies with MOX

fuel rods arranged in the triangular tightlattice

configuration. Therefore, it can

attain a fissile plutonium conversion ratio

or the breeding ratio over 1.0 under the

relatively hard or fast neutron spectrum.

The conceptual design of RMWR

300MWe with the passive safety

features has been accomplished in main

cooperation with Hitachi Ltd. aiming at

the electric power generation using the

small 330MWe/955MWt RMWR core

with the discharge burn-up of 65GWd/t

and the operation cycle of 25 months under

the multiple recycling situation. The core

consists of 282 hexagonal fuel bundles,

each of which has 217 fuel rods with

the outer diameter of 13.0 mm arranged

in the triangular lattice with 1.3 mm gap

width between rods. The MOX part is

shortened around 0.2 m high and two

MOX parts are piled up with an internal

blanket region, forming the double-flatcore

configuration to attain the negative

void reactivity coefficients, as shown in

the figure. Adding the upper and lower

blanket regions, the total axial length is

1.32m. The control rods are Y-shaped

ones with the follower structure above the

neutron absorber material region.

The core is cooled by the natural

circulation of the water coolant under

the same operating conditions as BWRs,

i.e. 7.2MPa and 561K. A breeding ratio

Responses to questions by Newal

Agnihotri, Editor of Nuclear Plant


of 1.01 and the negative void reactivity

coefficients are simultaneously realized

in the design. The fuel cycle concept is

a closed one and the simplified PUREX

method, in which purification processes

for Pu and U are eliminated, is considered

for the reprocessing process. Minor

actinides (MAs) could be recycled in

MOX with the enhanced proliferation

resistance, when MA recovery and MA-

MOX fuel fabrication processes are


In order to overcome what is called

the scale demerit for small reactors,

the plant systems is simplified and the

passive safety features are introduced in

the present plant system design. One of

the major passive safety features is the

natural circulation core cooling system,

and other passive safety concepts, such as

the gravity steam-water separation in the

upper plenum, the accumulator injection

system, the isolation condenser system

and the passive containment cooling

system, are also intended to be utilized

to improve the economy and to enhance

the reliability and the safety. In the

present safety system, a hybrid one with

the combination of the passive and the

active components is proposed and has

been evaluated to reduce the cost for the

reactor components.

Although no prototype for this

reactor concept has been established, a

Tsutomu Ohkubo

Tsutomu Okubo joined Japan Atomic

Energy Research Institute (JAERI, it

is now Japan Atomic Energy Agency

(JAEA) since October 2005) in 1978

and worked for advanced water reactors

design research, reactor thermalhydraulics

and safety engineering. He

is currently working on the development

of the reduced-moderation type water

reactor named FLWR as the Senior

Principal Researcher. He is a member of

the Atomic Energy Society of Japan.

large scale experimental program for the

critical heat flux in the tight-lattice rod

bundle was already conducted under the

reactor operating conditions and the core

cooling capability was demonstrated.

Some irradiation tests are necessary for

the highly enriched MOX fuel rods up

to at high burn-up. Since this reactor

concept is based on the well-developed

and experienced LWR technologies up to

now, it is expected to be realized without

serious difficulties. It would be ready for

commercialization in 2020s. This reactor

concept was also nominated as the High

Conversion BWR (HC-BWR) of the

advanced BWRs in the International

Near-Term Deployment (INTD).

2. Does your reactor include a

containment building If yes, please

describe the characteristics of your

containment building.

It has a steel containment system

to facilitate heat transfer from inside to

outside as a part of the passive containment

cooling system.

3. What has JAEA done in using nuclear

energy in applications other than

power production, including District

Heating, Seawater Desalination and


JAEA has been developing the

hydrogen production technologies using

16 Nuclear Plant Journal, September-October 2008

Very High Temperature Gas-cooled

Reactor (VHTR) and Sodium-cooled Fast

Breeder Reactor (FBR) with different

demonstration FBR will be operated

around 2025 and a commercialized FBR

will be developed before 2050.

JAEA also participates in all four GIF

VHTR Projects of hydrogen production,

fuel, material and code development.

JAEA is a world front runner of the VHTR

and hydrogen production technologies

and is willing to cooperate with foreign

organization developing the VHTRhydrogen.

Toshiba, MHI, Fuji electric and

Nuclear Fuel Industries take part in the

Next Generation Nuclear Plant (NGNP)

program. What they will achieve depends

on the budget they will acquire.

Bird’s-eye view of core and cross sectional view of fuel assembly

methods. VHTR can be also used for

desalination and district heating.

4. Who are JAEA’s partners in

producing hydrogen utilizing nuclear

energy Has a prototype already been

tested Please include a schedule for

application of hydrogen technology for

transportation in Japan.

Japanese industries such as Toshiba,

MHI etc. are working with JAEA

to develop the hydrogen production

technology using VHTR.

An experimental facility to produce

hydrogen of 30 liter /h using Iodine and

Sulfur (IS) method was constructed for

VHTR. The successful 1 week operation

was completed to confirm its chemical

process and establish the control


Though the prototype has not

been constructed, the research and

developments for hydrogen production

with the IS process and achievement

of higher efficiency than the previous

method is being planned.

Japanese Atomic Energy Commission

recently stated that VHTR-hydrogen plays

a key role to reduce CO2 emission and it

will be commercialized during 2020-2030

for application including transportation.

JAEA contributes to the fast reactor

system development out of Generation IV

reactors, especially a lot in the Sodiumcooled

Fast Reactor (SFR) program as


Japan has a national development

plan as a “key technology of national

importance” among the government,

utilities, industries and JAEA that a

Contact: Tsutomu Ohkubo, Japan

Atomic Energy Agency, 4002 Narita-

Cho, Oarai-Machi, Ibraki-Ken 311-

1393, Japan; telephone: 81-29-267-1919

ext 6480, fax: 81-29-266-3675, email:

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5. What is JAEA’s contribution to

Generation IV reactors and what is the

schedule for the industry to see some

tangible results

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Nuclear Plant Journal, September-October 2008 17

Modeling & Simulation Advances

Brighten Future Nuclear Power

By Hussein Khalil, Argonne National


Bob Hill and Jim Cahalan from the

Nuclear Engineering Division and

Andrew Siegel from the Mathematics

& Computer Science Division also


1. What applications have you currently

undertaken for design, operation, or

construction of nuclear power plants

Applications of leadership class

computers for nuclear energy R&D at

Argonne have so far focused mainly on

development and design of advanced

sodium cooled fast reactors (SFR), which

target sustainable energy generation, waste

minimization, assured passive safety,

and competitive economics. To enable

these applications we are developing

a modern computational framework

that uses advanced software tools and

computational methods for simulation

of multi-physics (neutronic, thermalhydraulic,

mechanical, etc.) phenomena

in complex reactor geometries. This

framework, named SHARP (Simulation

for High-efficiency Advanced Reactor

Prototyping), enables high-fidelity

simulation of reactor behavior taking

advantage of the enormous computing

power afforded by leadership class

computers. Its design provides flexibility

to employ less detailed (faster running)

models and to couple the different

physics modules tightly or loosely

depending on problem characteristics

and accuracy requirements. A key

goal of our development is to integrate

improved methods for characterizing the

uncertainty in predicted quantities within

the analysis framework.

To demonstrate the benefit of

leadership class computers for SFR

analysis, two computationally intensive

applications of computational fluid

dynamics (CFD) techniques are being

Responses to questions by Newal

Agnihotri, Editor of Nuclear Plant


Hussein Khalil

Hussein S. Khalil is director of

Argonne’s Nuclear Engineering Division

and is responsible for the Lab’s research

carried out using the IBM Blue Gene/P

supercomputer at Argonne’s Leadership

Class Computing Facility (see http://

• Detailed characterization of turbulent

coolant flow and heat transfer in SFR

wire-wrapped fuel pin bundles.

• Investigation of transient flow

fluctuations (thermal striping) in

the SFR upper plenum region where

the coolant discharged from fuel

assemblies mixes.

Additionally, we are performing

high-order, multigroup neutron transport

calculations for a highly detailed model

of a SFR core.

Blue Gene/P was just officially

clocked as the fastest computer in the

world dedicated to open science and is

the third fastest computer in the world

overall. It uses 163,840 parallel compute

nodes to execute at a clock rate of nearly

0.56 petaflops (1 petaflop = 10 15 floating

point operations per second) with a total

RAM of 80 terabytes. A simulation that

would take two years on a standard PC

can now be done in ten minutes. Access

to BG/P is granted using a competitive

peer reviewed process.

2. Will your system analysis cut down

the capital cost of NPP Nuclear Steam

Supply System by making the fuel and

the thermal hydraulics more effi cient

on nuclear reactor technology and

nuclear non-proliferation. He has

worked at Argonne since 1983 and

became a Senior Scientist in 2001. He

has a Ph.D. from MIT (1983) and an

MBA from the University of Chicago


Dr. Khalil is an internationally

recognized expert in nuclear reactor

physics and engineering. His research

has centered on the advancement of

reactor analysis methods and their

application for fast reactor design


The computational capabilities

under development will enable more

precise representation (modeling) of the

reactor and power plant configuration

and more accurate solution of the

equations describing reactor neutronic,

irradiation, thermal, fluid flow, and

structural/mechanical behavior. This

degree of modeling fidelity, combined

with enhanced capability for uncertainty

characterization, will make it possible

to design and operate reactors closer

to the true physical capabilities of the

fuel, materials of construction and

components. When these high fidelity

modeling capabilities are employed in the

design process, unnecessary conservatism

in reactor design and operation can be

reduced without compromising safety


3. Will your computational tools also

facilitate optimizing the usage of fuel by

providing assistance in designing and in


The high fidelity models being

integrated in SHARP allow greatly

improved characterization of fuel

operating conditions over its lifetime.

Results of these advanced models can be

employed in models of fuel behavior (a

key component of the overall code system)

to support the optimization of fuel design

(Continued on page 20)

18 Nuclear Plant Journal, September-October 2008

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Modeling &...

Continued from page 18

and to provide the necessary assurance

of fuel integrity over its operating life

considering both normal (operational)

and abnormal conditions.

Fuel behavior models currently

available have limited predictive capacity.

They rely extensively on the results of

fuel property and irradiation tests and

post-irradiation examinations. A large

number of in-pile tests are typically

needed to encompass the fuel operating

conditions of interest, and the duration of

these tests may be several years to reach

the targeted discharge burnup.

The high cost and protracted nature

of these tests create a strong incentive to

develop computational models of fuel

behavior that have greater predictive

capability and are less dependent on

empirical testing. Advancement of such

capabilities is pursued in parallel with

the (reactor) modeling and simulation

efforts described here, with the aim of

appropriately integrating or coordinating

their application in the future.

4. Who are your global partners in this


Argonne is leading a team of U.S.

national laboratories (Idaho, Oak Ridge

and Lawrence Livermore National

Laboratories) and several universities

in the advancement of reactor modeling

and simulation capabilities centered

on the SHARP code and the effective

use of leadership class computers,

including the IBM Blue Gene/P. This

national effort is sponsored by the U.S.

Department of Energy and is carried out

in cooperation with the French Atomic

Energy Commission (CEA) and the

Japanese Atomic Energy Agency (JAEA).

Cooperative activities currently underway

include (a) joint definition of benchmark

problems that can be used to test the

existing and developmental code systems

in each country, (b) joint comparison and

assessment of benchmark results, and

(c) joint assessment and improvement

of enabling software tools, e.g., tools for

geometry description, mesh generation,

data management, solution decomposition

and parallelization and visualization of


5. Please describe your plans with

your current technology for assisting

research, design, and operation of

nuclear power plants in the next fi ve


Our current plans are focused on

continued development, testing and

integration of the SHARP code. The

development effort will be guided and

focused by applications supporting the

development of conceptual designs for

advanced reactor systems and confirmation

of their safety. Their main initial use

will be to complement experimental

measurements in the qualification of the

existing analysis tools and to investigate

design options and operating conditions

that cannot be explored reliably with

existing tools.

Although separate- and integraleffects

measurements will continue to

SFR Bundle

be needed for validation of the models

used in reactor design, the advanced

capabilities under development will make

it possible to optimize the experimental

campaigns and to support greater use of

“numerical prototyping” in the design of

reactor components and systems.

6. Please share any other details, which

you may like to bring to the attention

of our readership in the nuclear power


The code systems in use today for

reactor development and design were

initiated more than thirty years ago and

were designed to accommodate the

computing resources, tools and methods

that were available at the time. We are

targeting a vastly improved capability

that exploits advances in computers and

software tools to facilitate reactor design

optimization, provide increased assurance

of performance and safety characteristics,

and reduce the need for large scale integral

experiments to characterize or validate


In addition to the improved

ability to predict reactor behavior, we

envision a vastly superior process for

development, design and licensing of

future reactors. This process would

integrate all significant aspects of the

design to influence optimized design

choices at the conceptual stage of the

design. It would also support evolution

from the conceptual stage to the detailed

design of realizable components. Finally,

it would provide for automated transfer

of design specifications to instructions

for manufacture and assembly, enabling

the manufacture of parts and components

to close tolerances and assured fit at the

time of assembly.

7. Do have enough funding to realize

your plans in the next fi ve years

We are grateful for the sponsorship

the U.S. Department of Energy provides

for our effort to advance modeling and

simulation of nuclear reactors, as well as

for the its past and continuing investment

in high-performance computers and

the software needed to make effective

use of these computers. Our progress

on development and application of the

reactor simulation tools, centered on

the SHARP code, obviously depends

on the funding support we receive over

the next five years – not only for code

development but also for application and

validation studies and quality assurance.

We are optimistic about the prospect

for this funding, because the benefit of

this research for advancing the use of

nuclear energy is increasingly recognized

by the technical community and

policymakers. At the same time, we are

extremely interested in partnerships with

commercial organizations that can provide

additional resources for accelerating

our development and validation efforts

and bringing their products to bear on

the commercial design, licensing and

operation of nuclear power plants.

Contact: Hussein S. Khalil, Argonne

National Laboratory, 9700 S. Cass

Avenue, Bldg 208, Argonne, IL 60439;

telephone: (630) 252-1456, fax: (630)

252-4780, email:

20 Nuclear Plant Journal, September-October 2008

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Energy & Desalination Projects

By Ratan Kumar Sinha, Bhabha Atomic

Research Centre, India.

1. Is your project part of Generation

IV International Forum, International

Project on Innovative Nuclear

Reactors and Fuel Cycles (INPRO)

If so, please provide details of your

project’s involvement with the above


India is not a member of Generation-

IV International Forum. The design and

development of Advanced Heavy Water

Reactor (AHWR) has been carried out at

Bhabha Atomic Research Centre (BARC)

without any external collaboration.

The IAEA’s International Project on

Innovative Nuclear reactors and fuel

cycles (INPRO) has stipulated a set of

requirements and criteria that should be

fulfilled by the innovative nuclear reactors

and fuel cycles of the future. AHWR

served as a case study for validating these

requirements and criteria.

2. Please provide a brief description of


AHWR is a 300 MWe, vertical,

pressure tube type, boiling light water

cooled, and heavy water moderated

reactor. The reactor incorporates a

number of passive safety features and is

associated with a fuel cycle having reduced

environmental impact. At the same time,

the reactor possesses several features,

which are likely to reduce its capital and

operating costs. In the Indian context,

AHWR will serve as a platform for the

timely development and demonstration

of the reactor and fuel cycle technologies

required to be in place before large scale

thorium utilisation in the future. The

AHWR fuel cycle has, however, enough

flexibility to accommodate a large variety

of fuelling options.

The reactor uses thorium based oxide

fuel with in-situ generated Uranium-233

and Plutonium, recovered from the spent

Responses to questions by Newal

Agnihotri, Editor of Nuclear Plant


Ratan Kumar Sinha

Mr. Ratan Kumar Sinha graduated

in Mechanical Engineering in 1972

and received training in nuclear

engineering, at postgraduate level, in

the training school of Bhabha Atomic

Research Centre (BARC) Mumbai,

India. He has thirty-fi ve years of

experience in the area of development

of reactor engineering technologies for

components and systems of pressure

tube type research and power reactors.

At present he is serving as Director,

Reactor Design & Development

Group and, Director Design,

fuel of water cooled reactors, serving

as fissile materials under equilibrium

conditions. It addresses the requirement

of sustainability of nuclear fuel resource

through the use of a closed fuel cycle

along with thorium.

Incidentally, on account of the use of

thorium based fuel, with no production

of additional Plutonium and the presence

of high energy gamma emitting daughter

products of Uranium-232, the reactor is

considered to have inherent proliferation

resistant features. The production of

minor actinides, in this reactor, is reduced

by nearly one order of magnitude, in

Manufacturing & Automation Group,

BARC. His current responsibilities

include directing programmes

for new advanced reactors under

design and development at BARC

to utilise thorium. These include,

the Advanced Heavy Water Reactor

(AHWR), which produces most of its

power from thorium, and has several

innovative passive safety features.

He is also responsible for the design

and development of a Compact High

Temperature Reactor (CHTR), which

is a technology demonstrator for future

Indian High Temperature Reactors

intended for hydrogen generation.

Mr. Sinha is a nationally and

internationally recognized expert in

the area of nuclear reactor technology.

For the past four years he has been the

Chairman of the Steering Committee

of INPRO, the IAEA’s International

Project on Innovative Nuclear Reactors

and Fuel Cycles.

Mr. Sinha has received several awards

and honours. He was elected a Fellow

of the Indian National Academy of

Engineering in the year 1998. He

has been an elected member of the

Executive Committee of the Indian

Nuclear Society for the last eight years.

comparison with conventional reactors,

thus substantially reducing the burden

of managing the inventory of long-lived

radioactive waste.

In AHWR, light water at 259 ° C

enters the core through 452 feeders, each

connected to a single vertical pressure tube,

in which heat is transferred from nuclear

fuel leading to boiling of the coolant. The

steam water mixture produced in these

pressure tubes rises through tail pipes

leading to four steam drums in which

steam, at nominal conditions of 70 bar

pressure and 270 ° C, is separated and taken

to the turbine cycle. The plant is designed

22 Nuclear Plant Journal, September-October 2008

to produce 300 MWe electricity along

with 500 m 3 /day of desalinated water.

The inherent and passive safety features

of the reactor include negative void

coefficient of reactivity, full power core

heat removal using natural circulation,

shut down decay heat removal backed

up by natural circulation, a passive shut

down device to address a postulated

insider threat of disablement of the two

main shut down systems, passive cooling

of concrete structures surrounding the

main heat transport system piping, and

passive isolation of containment as

well as passive cooling of containment

environment following a postulated loss

of coolant accident.

The reactor is provided with a

double containment. A 6000 m 3 capacity

water tank located inside the primary

containment, near its top, serves as a heat

sink for a range of postulated scenarios in

which the main coolant supply to the core

and/or the cooling water to the condenser

is not available. With the help of this

heat sink and other passive features, the

reactor is designed for providing a grace

period of at least three days following

any postulated scenario affecting the

plant. Thus, even without any external

source of power, coolant and operator

actions, safety of the reactor is assured

for practically an indefinite period.

The new design features of the

reactor have been validated with the help

of several large experimental facilities. A

large Critical Facility designed to validate

the reactor physics design of AHWR has

recently been commissioned at BARC.

The safety related features of AHWR have

been subjected to a pre-licensing design

appraisal by the Indian Atomic Energy

Regulatory Board. The design of the

nuclear systems is nearly complete and is

available for initiating the construction of

the plant in the near future.

3. What has Bhabha Atomic Research

Centre done in using nuclear energy in

application other than power production,

including District Heating, Sea Water

Desalination and Transportation

The Bhabha Atomic Research Centre

has helped the domestic development of

all required technologies, materials and

hardware necessary for the Pressurised

Heavy Water Reactor programme. It

has also been engaged in providing the

inspection and maintenance support, as

needed in some critical areas for these

reactors. District heating is not a major

requirement in most of India with tropical

climate conditions. However, the Indian

programme includes 220 MWe (750

MWth) PHWRs that may be effectively

deployed for a variety of applications

requiring small/medium power reactors.

BARC has got a very active programme

in sea water desalination and its work

covers a number of technologies,

including membrane based technologies

and evaporation based technologies

for desalination and potable water

production in a cost-effective as well

as energy-efficient manner. The Indian

Madras Atomic Power Station (MAPS),

for example, is being coupled with a large

desalination plant.

4. Who are Bhabha Atomic Research

Centre’s partners in producing hydrogen

utilizing nuclear energy Has a prototype

already been tested Please include a

schedule for application of hydrogen

technology for transportation in India.

Bhabha Atomic Research Centre

has been working on the development

of technologies for producing hydrogen

using water splitting reactions. Its current

activities in this area include conventional

electrolysis, high temperature electrolysis,

and chemico-thermal processes for

hydrogen generation. Bhabha Atomic

Research Centre is one of the several

research organizations, academic

institutions and industrial partners that

have contributed towards preparation of a

national hydrogen energy road map.

5. What is Bhabha Atomic Research

Centre’s contribution to Generation IV

reactors and what is the schedule for the

industry to see some tangible results

India is not a member of Generation-

IV. However, the Advanced Heavy Water

Reactor mentioned above fulfils/exceeds

all the requirements stipulated by INPRO,

for the next generation nuclear reactors.

The design of this demonstration reactor

has reached an adequately advanced level,

and the construction of the reactor is

planned to be initiated in the near future.

6. Who is the manufacturer of forgings

for reactor pressure vessels in India

The domestic Indian nuclear

power programme is currently based

on Pressurised Heavy Water Reactors

(PHWRs) and pool type Fast Breeder

Reactors (FBRs). These reactors do not

require reactor pressure vessels. Major

components for the Indian nuclear reactor

programme have been manufactured by

several industries in the governmental

(public sector) as well as private sector in


Contact: Ratan Kumar Sinha, Bhabha

Atomic Research Centre, BARC, Mumbai,

400085; email:

Nuclear Plant Journal, September-October 2008 23

A Plant with Simplified Design

By John Higgins, GE Hitachi Nuclear


1. How does the ESBWR minimize

damage to the fuel in case of a loss of

coolant accident (LOCA)

With the ESBWR, the fuel remains

covered and well-cooled through all

operational events including the unlikely

event of a LOCA. This ensures that the

fuel temperature remains at or below the

fuel’s normal operating temperature. The

ESBWR builds on the outstanding safety

record of the world’s established BWR


2. How has the ESBWR improved the

reactor water chemistry to minimize

affect on the fuel and on reactor

internals during normal operation and

during accident conditions

The ESBWR operates well within

the industry-established BWR water

chemistry guidelines (specifically the

limits on feedwater iron levels), which

effectively precludes the buildup of iron

oxide deposits on fuel elements and

reactor internals.

One of the goals of maintaining

good BWR water chemistry during plant

operation is to minimize the potential for

developing intergranular stress corrosion

cracking (IGSCC) on reactor internals.

For the ESBWR, the potential for IGSCC

resistance is addressed through the use

of significantly improved materials, such

as Type 316 Nuclear Grade stainless

steel and stabilized nickel-base niobium

modified Alloy 600 and Alloy 82. The

ESBWR design has significantly reduced

the number of welds needed, and along

with the use of improved materials, the

potential for cracking is substantially


3. What fuel and fuel cladding material

design enhancements have been made

in ESBWR to ensure minimum damage

Responses to questions by Newal

Agnihotri, Editor of Nuclear Plant


John Higgins

John Higgins serves as Vice President,

ESBWR Projects, for GE Hitachi

Nuclear Energy. Higgins joined the

company in 2005 as the Project Manager

responsible for a joint Department of

Energy initiative to advance the design

of the next-generation boiling water

reactor technology, the ESBWR. In

of the fuel during normal operation, and

during accident scenarios

The ESBWR takes advantage of years

of operating experience with BWRs and

offers improvements to address typical

fuel cladding problems. Global Nuclear

Fuel - a joint venture of GE, Hitachi and

Toshiba formed to produce BWR fuel -

will supply ESBWR fuel incorporating

the following features :

• Debris filtration devices to trap

debris material before it reaches the

fuel rods in order to prevent debris


• Pellet cladding interaction (PCI)

resistant fuel rod technology to

prevent PCI failures

• Corrosion resistant cladding to

prevent fuel failures due to build-up

or chemical intrusion events

4. What innovative fuel cycles have

been used in ESBWR to maximize fuel

effi ciency

Building on years of operating

experience and advanced fuel design, the

ESBWR core design provides numerous

2006, Higgins assumed additional

responsibilities for overall deployment

planning for the ESBWR, and in

2007, he was promoted to his current

position. In 2008, Higgins assumed

overall management responsibility

for the global ESBWR business,

responsible for completing the NRC

certifi cation process, fi nalizing the

detailed design, establishing the

advanced modularization requirements

and construction methods, and

commercialization of the technology.

Higgins is a degreed engineer with

30 years of professional experience

supporting both nuclear and fossil

projects. During his career, Higgins has

accumulated a broad base of experience

that includes licensed merchant marine

offi cer, nuclear start-up engineer, and

business unit manager.

options for our customers that will help

to minimize outage lengths, support

high discharge exposure and reduce

enrichment requirements for fuel cycles.

ESBWR cores can support a wide range

of refueling cycle intervals ranging from

12 to 24 months.

5. How has ESBWR ensured a longer

cable life to ensure a longer plant life

To ensure a longer cable life, the

ESBWR utilizes a comprehensive quality

assurance (QA) program, a disciplined

electrical design regimen, the most

stringent nuclear industry standards

and rigorous qualification testing. GEH

participates in the development of

consensus nuclear power industry cable

standards, which are based on research

and testing specifically for nuclear

power plant applications. The ESBWR

electrical engineering team continues

to utilize disciplined practices and

state-of-the art design tools to build on

GEH’s nuclear legacy. GEH’s QA and

equipment qualification programs include

evaluation of life-limiting mechanisms,

24 Nuclear Plant Journal, September-October 2008

special material selection and rigorous

proof testing of cable performance.

The qualification proof testing includes

condition monitoring, flammability,

radiation exposure, simulation of the

60-year life span, mechanical stress and

LOCA testing.

6. What is the plant life of ESBWR

The design life for the ESBWR plant

and all its major components is 60 years.

7. What enhancements have been made

in the control station design to ensure

improved human-system interface

By utilizing experienced plant

operators and human factors engineering

concepts, the ESBWR control room was

developed, designed and evaluated using

an integrated top-down design process

that uses state-of-the-art methods and

technology. The control room was

developed to meet the review criteria

detailed in industry standards, including

Standard Review Plan Chapter 18 from

NUREG-0800 and also NUREG 0711.

Significant ESBWR control room

enhancements include:

• A wide display panel

• Alarm filtering and prioritization

• Computerized procedures

Plant and system automation

• Video workstations

• Advanced trending

• Comprehensive human factors


8. How has the current instrumentation

and control system in the ESBWR been

upgraded from the previous GE Hitachi

Nuclear Energy designs to ensure a

reliable plant operation with longer

plant life

The ESBWR Distributed Control

and Information System (DCIS) has

four divisions and is designed with no

single failure points that could affect

the performance of support systems.

In addition, DCIS contains sufficient

redundancy so that even in the unlikely

event of a failure while maintenance is

being performed, there would be no impact

on safety system functions. ESBWR

automation systems ensure consistent

and conservative plant operation, either

remotely or control room dispatched, for

plant functions such as pulling control

rods to critical, heat-up/pressurization,

turbine roll and synchronization, and

power operation. Key control systems

are triply redundant to improve the safety

and reliability of the plant.

9. How has information technology

been used to survey and self-diagnose

problems in the systems, structure and

components of the ESBWR

Major ESBWR plant components

are fully instrumented to support on-line

monitoring for equipment degradation and

maintenance. Examples of parameters

included in the on-line condition

monitoring include:

• Flow rate, suction pressure, discharge

pressure and speed for pumps

• Current, voltage, power and running

hours for motors

• Flow rates, differential pressure and

inlet/outlet temperatures for heat


Similarly, large rotating machines

(or small inaccessible machines) like

feedwater pumps and drywell cooling

(Continued on page 26)

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Nuclear Plant Journal, September-October 2008 25

A Plant...

Continued from page 25

fans are equipped with instrumentation to

support high-speed vibration monitoring

and other condition evaluation techniques.

The alarms from the advanced condition

monitoring are integrated into the plant

displays and alarm system.

The DC Power Supply utilizes the

latest proven technology to monitor battery

voltage and provides for a “battery

maintenance” feature that maintains

batteries at full charge. Uninterruptible

Power Supply (UPS) systems have input

voltage electronic switching that protect

from grid-induced spikes that could trip

the safety-related DC power from their


10. How does ESBWR handle unstable

and disruptive phenomena, such as

water hammer

The passive safety design of the ES-

BWR has an enhanced design capability

to mitigate disruptive phenomena. In the

case of water hammer, an improved system

layout and enhanced features mitigate

the probability of water hammer and

potential consequences. Those improved

features include various system design

and layouts, such as surge tanks, automatic

air release/vacuum valves installed

at high points in system piping and at the

pump discharge, proper valve actuation

times to minimize water hammer, procedural

requirements ensuring proper line

filling prior to system operation and after

maintenance operations, and the use of

a check valve at each pump discharge to

prevent backflow into the pump.

11. Is GE Hitachi Nuclear Energy

exploring options to manufacture reactor

pressure vessels given the fact that there

are very few manufacturers in the world

to meet the required demand

Because of GEH’s continued

involvement in building and uprating

nuclear plants around the world, we have

maintained a robust manufacturing and

supply chain, which serves us well as

we engage in the nuclear renaissance.

However we recognize that the demand

is increasing, and as we have done in the

past, we continually explore additional

options for manufacturing reactor pressure

vessels and other large components.

12. How does the economy of the

ESBWR compare with its previous


As GEH’s next evolution of advanced

BWR technology, the ESBWR offers a

simplified design providing improved

safety, excellent economics, better plant

security, a broad seismic design envelope

and operational flexibility that increases

plant availability.

Contact: Ned Glascock, GE Hitachi

Nuclear Energy, 3901 Castle Hayne

Road, Wilmington, NC 28402; email:

26 Nuclear Plant Journal, September-October 2008

A Forward Thinking Design

By Ray Ganthner, AREVA.

1. What were AREVA’s objectives

in introducing the EPR to the global


World-wide energy demand is

increasing at an accelerating pace. At

the same time, there are environmental

challenges to consider. The world needs

more CO2-free nuclear energy to provide

certainty of energy supply to the economy.

AREVA’s objective was to develop and

offer a design that would most effectively

meet the demand for a reliable source

of power generation. The EPR is an

evolutionary design based on mature,

yet greatly enhanced, technology that

improves safety and performance. That’s

why we call it the evolutionary power

reactor. The EPR has many innovative

design features; but they are all based

on proven technologies to provide the

confidence and certainty of design the

public and plant operators demand.

2: What innovative fuel cycles have

been used in the EPR to maximize fuel

effi ciency

The EPR design provides enhanced

and flexible fuel utilization. The EPR has

increased thermal margin compared with

existing plants. The linear heat rate has

been reduced and the coolant flow per

assembly has been increased compared

with a typical Pressurized Water Reactor

plant. Therefore, the EPR has improved

flexibility in designing fuel cycles from

12 months to 24 months. Using our

proven gadolinium burnable absorber and

axial blankets, coupled with a new heavy

neutron reflector around the core, we are

able to design extremely efficient cores

that minimize uranium requirements.

This is important with the price of

Uranium being much higher than it was

only a few years ago.

Responses to questions by Newal

Agnihotri, Editor of Nuclear Plant


Ray Ganthner

Ray Ganthner is AREVA NP Inc.

senior vice president, New Plants

Deployment. He joined the company

in 1980 and is currently responsible

for certifi cation of advanced reactor

designs for deployment in North

America, including light water reactors

and advanced high temperature gas


3. What fuel and fuel cladding material

design enhancements have been made

in EPR to ensure minimum damage of

the fuel during normal operation, and

during accident scenarios

AREVA utilizes our latest, most

advanced cladding material, M5 TM , for

EPR fuel. This cladding is already in use

in many operating reactors world wide.

The testing and operating history has

verified that this cladding has superior

mechanical properties and significantly

reduces cladding oxidation as compared

with standard zirconium or zircoloy-4

material. The experience of this advanced

material developed by AREVA increases

the certainty of optimum fuel performance

in the most challenging operating

environments. Additionally, the plant is

designed to operate in the range where

Ganthner became Manager of Group

and New Projects in 1991 and

among his notable achievements was

completion of the Bellefonte and WNP-1

nuclear power plants. In 1994, Ganthner

was named vice president of engineering

and project services, where he was

responsible for commercial nuclear

power plant products and services.

Ganthner’s executive responsibilities

were expanded to include business

development in 1996.

In 1997, Ganthner was called upon to

sponsor an international team to develop

advanced nuclear fuel designs for

introduction in the U.S. and Europe, and

in 2000, he returned to the commercial

nuclear power plant business where he

was responsible for plant systems and

analysis, and engineering programs

with offi ces in Virginia, North Carolina

and Massachusetts. Ganthner holds a

Bachelor of Science in Naval Science

from the U.S. Naval Academy and a

Master of Business Administration from

Lynchburg College.

there are significant margins and thus the

fuel starts out with a better margin in the

case of any operational transients or in the

unlikely event of an accident transient.

4. How has EPR improved the reactor

water chemistry to minimize affect on

the fuel and on reactor internals during

normal operation and during accident


The EPR is designed to be compatible

with the latest water chemistry limits

specified by the Electric Power Research

Institute. All materials in contact with

the reactor coolant are selected to be

compatible with these requirements.

Furthermore, pH control of the reactor

coolant is made easier by the use of

enriched B10 for soluble reactivity

control, which decreases the amount of

(Continued on page 28)

Nuclear Plant Journal, September-October 2008 27

A Forward...

Continued from page 27

boric acid required as compared with

most previous plant designs.

5. How do the EPR’s active and

passive safety systems, including onsite

and offsite emergency power sources,

minimize damage to the fuel in case of a

loss of coolant accident (LOCA)

The EPR concept first and foremost

is to design in prevention of fuel damage,

then mitigation. Each of the EPR’s

four independent safety trains has the

designed capacity to provide the full

safety function. Each of the four systems

has its own dedicated emergency power

source supplied by a separate diesel

generator. The EPR’s safety margins are

approximately a factor of 100 better than

the regulatory requirements. These four

safety systems are activated by automatic

digital protection systems and can also be

controlled by reactor operators. During an

anomalous event you don’t want to rely on

the laws of physics and the engineering

alone, but you also want to have control

of what is happening inside your plant. In

addition, the lower power density of EPR

fuel compared to other designs provides

greater safety margins.

As an example of the EPR’s forwardthinking

design philosophy when it

comes to safety, the EPR design led the

industry by providing the extra margin of

safety against airplane crash now being

proposed in the recent NRC rulemaking.

Its “double walled” containment and four

physically separated safety trains provide

the certainty of protection against a

potentially severe threat to containment


6. What is the plant life of EPR

The EPR is designed for a plant

life of 60 years. But even longer plant

life is possible largely due to the use of

more advanced materials and welding

techniques. The metallurgical properties

of Inconel 690 greatly improve the life

of steam generator tubes. Reactor vessel

materials, weld materials, and even the

location of welds all work together to

optimize and probably eventually achieve

actual plant lifetimes in the unprecedented

range of 60 to100 years.

7. What enhancements have been made

in the steam generator to ensure a longer

plant life How long are the EPR steam

generators expected to last

The EPR steam generators are

designed to last the entire plant design

life of 60 years. This is due to the

significant enhancements in materials

and fabrication techniques incorporated

into all of AREVA’s steam generators

over the last 20 years. Thermallytreated

alloy-690 tubing with full depth

hydraulic expansion in the tube sheets

virtually eliminates the potential for

stress-corrosion cracking observed in

many of the current generation plants

that used mill-annealed alloy-600 tubing.

Tube support plates are fabricated using

410 SS, which has been proven to reduce

fouling. Anti-vibration bars made of 405

stainless steel are meticulously installed

in such a way that unwanted tube wear is

virtually eliminated.

8. How does the EPR ensure a longer

cable life to facilitate a longer plant life

Cable technology has improved since

existing plants were built, and longer

life cables are available from various

manufacturers. We are working with these

manufacturers to develop even longer life

cables that are compatible with the design

life of the EPR.

9. How has the current instrumentation

and control system in the EPR been

upgraded from the previous AREVA

designs to ensure a reliable plant

operation with longer plant life

As in previous AREVA designs, the

EPR I&C system design pays specific

attention to safety and ensuring a high

level of operational flexibility in order

to meet the needs of reliable electric

generation. The notable upgrade in the

EPR I&C system is the TELEPERM XS

digital control equipment. Digital I&C

systems offer improved reliability over

analog systems, and do not suffer the same

types of degradation problems that occur

with analog systems to support longer

life. In addition, I&C systems implement

advanced functionality, such as partial

trips, to respond to a plant disturbance

while maintaining operation. The overall

design of I&C systems and associated

equipment complies with requirements

imposed by the process, nuclear safety

and operating conditions.

10. What enhancements have been made

in the control station design to ensure

improved human-system interface

A great deal of consideration was

given at the design stage by human-factor

engineers for enhancing the reliability

of operators’ actions during operation,

testing and maintenance phases. The Main

Control Room (MCR) is the centralized

location used by the operators to supervise

and control plant processes. The MCR

is ergonomically designed using stateof-the-art

human factors principles. It

will be equipped with information rich

screen-based indication and controls for

both safety-related and non-safety related

functions, computer-based procedures,

and alarm display screens.

The MCR provides the operator

with a clear understanding of the plant

status including severe accident. The

enhanced human system interface (HSI)

elements will provide significantly

more information to the operator in a

more efficient way versus conventional

displays. These upgrades are expected

to increase situation awareness, without

creating information overload. The

increased automation will help to

minimize operator error and assists in

(Continued on page 30)

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A Forward...

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error detection and recovery capability.

With a well-designed system

overview, the decision making process is

made easier because the “data collection”

mode required when using conventional

panels is minimized. Alarm displays and

computerized procedures will have ties

to the indications and controls that the

operator requires to make procedure step


The MCR is equipped with:

• Two screen-based workstations for

the operators

• A screen-based workstation for

presenting information to the shift

supervisor and the safety engineer

• An additional workstation for a

third operator to monitor auxiliary


• An auxiliary panel to bring the plant

to cold shutdown using safety grade

displays and control

• Large plant overview panels that give

information on the status and main

parameters of the plant

11. How has information technology

been used to survey and self-diagnose

problems in the systems, structure, and

components in EPR

I&C systems, along with specialized

diagnostic systems, provide advanced

capabilities for the collection and storage

of information regarding plant equipment.

This information can be transferred

to business management systems for

analysis to support a wide variety of

operational and maintenance objectives.

12. How does the EPR handle unstable

and disruptive phenomena, such as

water hammer

Unstable or undesired disruptive

phenomena are handled at the engineering

and design stage by specific design rules

set to eliminate the problem. For example,

geometries that could lead to rapid steam

collapse or rapid valve movements are

avoided, limiting the potential for water

hammer. Flow assisted corrosion is

limited by employing design limits on

liquid velocity and water chemistry or by

specification of more robust materials,

for example, stainless steel or chromiummolybdenum


13. What enhancements have been made

in the designs and construction of EPR

to control fi re and smoke in the plant

affecting safety critical systems

The U.S. EPR is a robust design

with increased safety margin with respect

to fire safe shutdown capability. The

physical separation of safety system

trains and the redundancy of safety

systems minimize the possible effect of

smoke and fire on critical safety systems.

We designed redundant safety systems to

exceed regulatory requirements. The four

train safety concept means you can have

one train in maintenance, one train may

be affected by a fire, and the remaining

train or trains required for safe shutdown

are still available. Since each safety

train is independent and located within a

physically separate building, propagation

of fire between divisions is eliminated.

14. How does the economy of the EPR

compare with its previous designs

The EPR original design objective

was to make the plant at least 10 percent

more economic to operate than existing

plants. We think we have achieved that by

increasing the power level, reducing the

numbers of components, and eliminating

unnecessary maintenance activities. With

the four independent operating trains,

online maintenance has been made

possible. As a result, the plant’s output in

megawatt hours is higher, so fixed costs

are spread over more megawatts. The

EPR has a higher thermal efficiency and

projected lifetime availability between 92

and 95 percent.

15. Is AREVA exploring options to

manufacture reactor pressure vessels

given the fact that there are very few

manufactures in the world to meet the

required demand

The demand for all these new reactors

around the world is a challenge for heavy

component manufacturers, and AREVA

is involved with the global supply of

these components. In fact, we’ve been

consistently investing in manufacturing to

make sure we are ready when the expected

demand for more nuclear energy is finally

realized. We’ve been in the process of

upgrading and expanding all of our heavy

component shops. We’ve completed two

large expansions of our manufacturing

plant at Chalon, France and recently

acquired a large steel forging plant in

France. We’re also looking at building

a large component manufacturing plant

in the USA. A part of our strategy is

to continuously evaluate the global

marketplace and the forging business to

determine whether we need to develop

in-house ultra-heavy forging capability

and if so, when this would make the

most business sense. We’re completely

committed to the expansion of clean

nuclear energy, so we look at everything

involved. AREVA’s tremendous domestic

and global resources and our EPR design

currently under construction in Finland

and France, together provide a significant

cost and schedule certainty for more

nuclear energy to become a reality.

Contact: Susan M. Hess, AREVA NP

Inc., 3315 Old Forest Road, Lynchburg,

VA 24501; telephone: (434) 832-2379,

fax: (434) 382-2379, email:




30 Nuclear Plant Journal, September-October 2008

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A Passively Safe Design

By Ed Cummins, Westinghouse Electric


1. How does AP1000 minimize damage

to the fuel in case of a loss of coolant

accident (LOCA)

The AP1000 Passive Core Cooling

Systems together with other safety

features is designed to protect the fuel in

case of a LOCA.

Regarding Accident scenarios,

the AP1000 meets the U. S. NRC

deterministic-safety and probabilistic-risk

criteria with large margins. The safety

analysis is documented in the AP1000

Design Control Document (DCD) and

Probabilistic Risk Assessment (PRA).

Results of the PRA show a very low core

damage frequency (CDF) that is 1/100 of

the CDF of currently operating plants.

The Advisory Council on Reactor

Safeguards (ACRS) and the U.S. NRC

have scrutinized the AP1000 Passive

Safety Systems and ruled that they meet

the U.S. NRC core cooling criteria, and

other safety criteria such as Three Mile

Island lessons learned.

2. How has AP1000 improved the

reactor water chemistry to minimize

affect on the fuel and on reactor

internals during normal operation and

during accident conditions

Zinc addition; a soluble zinc

compound is added to the coolant as a

means to reduce radiation fields within

the primary system and to reduce the

potential for crud-induced power shift

(CIPS). The zinc used may be either

natural zinc or zinc depleted of 64Zn.

3. What fuel and fuel cladding material

design enhancements have been made

in AP1000 to ensure minimum damage

of the fuel during normal operation, and

during accident scenarios

The use of ZIRLO cladding material;

ZIRLO cladding material combines

neutron economy (low absorption crosssection);

high corrosion resistance to

Responses to questions by Newal

Agnihotri, Editor of Nuclear Plant


Ed Cummins

Ed Cummins has spent his 32-year

Westinghouse career in a variety of

assignments in project management,

coolant, fuel, and fission products; and

high strength and ductility at operating

temperatures. ZIRLO is an advanced

zirconium based alloy that has the same

or similar properties and advantages as

Zircaloy-4 and was developed to support

extended fuel burn up.

Regarding accident scenarios,

the AP1000 meets the U. S. NRC

deterministic-safety and probabilistic-risk

criteria with large margins. The safety

analysis is documented in the AP1000

Design Control Document (DCD) and

Probabilistic Risk Assessment (PRA).

Results of the PRA show a very low core

damage frequency (CDF) that is 1/100 of

the CDF of currently operating plants.

4. What innovative fuel cycles have

been used in AP1000 to maximize fuel

effi ciency

The AP1000 is designed to use an 18

month or 16/20 month alternating cycle

for optimum economics.

5. How has AP1000 ensured a longer

cable life to ensure a longer plant life

The AP1000 instrumentation and

control systems are designed in accordance

with guidance provided in applicable

portions of the following and

other related standards: IEEE 383-1974,

engineering management and new plant


In March of 2000, Westinghouse initiated

development of the AP1000 plant

designed to be competitive with natural

gas fi red combined cycle plants. He

is currently Vice President, Nuclear

Power Plant Regulatory Affairs and

Standardization, responsible for the

licensing and commercialization of the


Mr. Cummins holds a Bachelor of

Science Degree from the U.S. Naval

Academy, a Master of Science Degree

in Engineering Applied Science from

the University of California, Davis,

Livermore and a Master of Business

Administration from Duquesne


“IEEE Standard for Type Test of Class IE

Electric Cables, Field Splices, and Connections

for Nuclear Power Generating


6. What is the plant life of AP1000

The AP1000 has a 60 year design


7. What enhancements have been made

in the control station design to ensure

improved human-system interface

Use of digital Instrumentation and

Control systems.

8. How has the current instrumentation

and control system in AP1000

been upgraded from the previous

Westinghouse designs to ensure a

reliable plant operation with longer

plant life

Use of digital Instrumentation and

Control systems with rigorous adherence

to NRC developed Human Factors

Engineering guidance.

9. How has information technology

been used to survey and self-diagnose

problems in the systems, structure, and

components in AP1000

Design Reliability Assurance Program

(D-RAP); the AP1000 D-RAP is

32 Nuclear Plant Journal, September-October 2008

implemented as an integral part of the

AP1000 design process to provide confidence

that reliability is designed into the

plant and that the important reliability

assumptions made as part of the AP1000

probabilistic risk assessment (PRA) will

remain valid throughout plant life. The

PRA quantifies plant response to a spectrum

of initiating events to demonstrate

the low probability of core damage and

resultant risk to the public. PRA input

includes specific values for the reliability

of the various structures, systems, and

components in the plant that are used to

respond to postulated initiating events.

10. How does AP1000 handle unstable

and disruptive phenomena, such as

water hammer

The AP1000 is designed to minimize

phenomena such as water hammer by

incorporating industry lessons learned.

The layout of the startup feed water

piping and the main feed water line

include features to minimize the potential

for water hammer.

The potential for water hammer,

stratification, and striping is additionally

reduced by the use of separate startup

feed water piping and nozzles for each

steam generator. The startup feed water

nozzle is located at an elevation that is the

same as the main feed water nozzle and

is rotated circumferentially away from

the main feed water nozzle. A startup

feed water spray system independent

of the main feed water feed ring is used

to introduce startup feed water into the

steam generator.

11. How does the economy of AP1000

compare with its previous designs

The AP1000 is designed to be

simpler, with less systems and equipment,

and thus more economic.

12. What enhancements have been

made in the designs and construction of

AP1000 to control fi re and smoke in the

plant affecting safety critical systems

Separation and fire areas. As

presented in the AP1000 Design Control

Document (DCD); fire areas are three

dimensional spaces designed to contain

a fire that may exist within them. They

are separated by fire barriers, fire barrier

penetration protection, and other devices,

such as those within the heating and air

conditioning ducts that isolate a fire to

within the fire area.

13. What enhancements have been made

in the steam generator to ensure a longer

plant life

Use of Alloy 690 tubes; Nickelchromium-iron

alloy in various forms is

used for parts where high velocities could

otherwise lead to erosion/corrosion to

help increase component life.

14. How long are the AP1000 steam

generators expected to last

The AP1000 plant is being designed

to meet the ALWR utility requirements

specified in Volume III of the ALWR

Utility Requirements Document (URD).

The URD states that for Plant Design

Life, “The plant shall be designed to

operate for 60 years without necessity

for an extended refurbishment outage.

The plant shall be designed to permit

expeditious component replacement for

obsolescence and failure over a lifetime

of 60 years.”

15. Does AP1000, having a passive

safety system, still need an onsite and

offsite emergency power

No, not for safety. To minimize the

challenges to the passive safety systems,

the AP1000 design does include nonsafety

connections to the site power grid

and 2 non-safety diesel generators.

Contact: Scott Shaw, Westinghouse

Nuclear, 4350 Northern Pike, Monroeville,

PA 15146; telephone: (412) 374-6737,


Nuclear Plant Journal’s

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An International Publication

Published in the United States

Nuclear Plant Journal, September-October 2008 33

A Market-Ready Design

By Ken Petrunik, Atomic Energy of

Canada Limited.


Atomic Energy of Canada Limited’s

(AECL’s) newest CANDU ® (CANada

Deuterium Uranium) reactor, the ACR-

1000 ® (Advanced CANDU Reactor ® ),

is a 1200 MWe-class Generation III+

nuclear power plant with a 60-year design

life, including a mid-life pressure-tube

replacement. It is a light-water-cooled,

heavy-water-moderated pressure-tube

reactor, with low-enriched uranium fuel

(LEU), which has evolved from the

well-established CANDU line. It retains

proven CANDU design features while

incorporating innovations and state-ofthe-art

technologies to enhance safety,

operation, maintenance, performance and


A key strategy in designing the ACR-

1000 was to expand the Instrumentation

and Control (I&C) and Information

Technology (IT) systems by designingin

and integrating operations and

maintenance (O&M) functions. SMART

CANDU modules allow on-line health

monitoring of systems and components.

Maximum use of modularization and

‘open-top’, parallel construction—which

have already been demonstrated at the

Qinshan Phase III CANDU units, both

delivered under budget and ahead of

schedule—are key to AECL’s ACR-1000

new-build project model.

AECL is currently focusing on nearterm

opportunities to build ACR-1000

plants in Canada. CANDU reactors, now

operating successfully on four continents,

have already demonstrated that the

technology can be easily localized in

other countries—due to a core comprised

of a large number of small, identical fuel

channel components. Recent offshore

new-build projects have also proven that

nuclear power plants can be built on time

and on budget.

Responses to questions by Newal

Agnihotri, Editor of Nuclear Plant


Ken Petrunik

Ken Petrunik, PhD is President,

CANDU Reactor Division, Atomic

Energy of Canada Limited (AECL).

He also holds the AECL corporate

position of Executive Vice-President

and Chief Operating Offi cer. Dr

Petrunik has spent more than 30

1. How do the economics of ACR-1000

compare with those of other Generation

III+ reactors

The ACR-1000 has evolved from

the CANDU 6 design, and has attractive

economics. It is designed to achieve lower

specific capital cost, shorter construction

schedule, higher plant capacity factor,

lower operating cost, increased operating

life and enhanced ease of operation.

The ACR-1000’s economics are fully

competitive with numbers published in

the literature for other Generation III+


2. What is the status of ACR-1000


ACR technology had extensive

pre-project review from the United

States Nuclear Regulatory Commission

(USNRC, 2001-02), the Canadian Nuclear

Safety Commission (CNSC, 2003-06)

and, more recently, by the UK regulator

(2007-08). Findings were positive. On

April 1, 2008, AECL and CNSC signed

a Memorandum of Understanding for

performing a pre-project design review

on ACR-1000, which will be conducted

in two phases through 2009.

The key submission for this preproject

design review is the 3,000-page,

years with AECL, leading the company

through design, licensing, construction

and commissioning of CANDU power

stations around the world. The teams he

assembled were instrumental in bringing

in all of our recent new-build reactor

projects into service on time or ahead

of schedule, and on budget. Dr Petrunik

introduced open top construction and

modularization technology to CANDU

power plants, and also led the fi rst use

in Canada of authorized electronic

documentation for AECL projects, the

model for future projects. More recently,

in his role of Chief Operating Offi cer,

he has further deepened his already

excellent relationships with customers

and governments, working to develop

markets for AECL’s market-ready ACR-

1000 and world leading CANDU 6.

20-chapter, Generic Safety Case Report

(GSCR), submitted on June 30, 2008.

This report provides an integral picture

of the ACR-1000 safety design and

bounding safety analysis. Being in the

format of a Preliminary Safety Analysis

Report (PSAR), it is comprehensive and


3. Is AECL exploring options to

manufacture reactor pressure vessels

given the fact that there are very few

manufactures in the world to meet the

required demand

The ACR-1000 and all CANDU

reactors are pressure-tube reactors. Thus,

they do not have high-pressure reactor

vessels typical of light water reactors,

or the associated supply difficulty with

heavy forgings. The only large forgings

for ACR-1000 are related to the steam

generators, for which there are alternate

suppliers. There is a robust supply

chain for pressure tubes with alternative

suppliers in North America and overseas,

with recent supply availability clearly

demonstrated in refurbishment projects.

4. What fuel and fuel cladding material

design enhancements have been made in

(Continued on page 36)

34 Nuclear Plant Journal, September-October 2008

Our enrichment technology

provides a whole new momentum.

Our uranium enrichment technology is a

revolutionary force in the nuclear fuel cycle. As the

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As an independent energy and technology group

with global leadership in centrifuge technology,

Urenco is ideally placed to give a reliable and

flexible source of support to the nuclear industry.

Enriching the future

A Market...

Continued from page 34

ACR-1000 to ensure minimum damage

of the fuel during normal operation, and

during accident scenarios

Reference fuel for the ACR-1000

is the 43-element CANFLEX-ACR

(CANDU FLEXible) bundle, which

incorporates 42 elements with 11.5 mm

OD, 2.4% enriched LEU and one 20-mmdiameter

central element with burnable

neutron absorbers (BNA). Sheath material

is Zircaloy-4.

ACR-1000 fuel acceptance criteria

for normal operation were used to

systematically evaluate any potential

damage mechanisms that could affect

fuel robustness. This ensures that fuel

cannot be damaged in fulfilling design

requirements for normal operation.

Design changes, listed below, help to

minimize fuel damage during normal

operation and accidents:

• More highly subdivided 43-element

CANFLEX-ACR fuel bundle,

lowering fuel element ratings and

reducing the power-related damage


• Fuel pellet geometry optimized to

minimize sheath strains and fission

gas pressure

• CANLUB interlayer thickness

increased to improve resistance to

damage due to power ramp failures

• Fuel sheath thickness defined to

maintain its intrinsic collapsibility

• Fuel bundle endplate geometry

modified to improve irradiated fuel

bundle strength during refuelling


• Use of CANFLEX-ACR fuel

bundle with AECL’s patented flowenhancing

sheath appendages,

providing increased margin to dryout

in postulated accident conditions

• Central fuel element containing

BNAs to control the coolant void

reactivity, thus minimizing potential

for fuel damage in the case of a

postulated large-break loss-ofcoolant

accident (LOCA)

5. How does ACR-1000 minimize

damage to the fuel in case of a loss-ofcoolant


The ACR-1000 design has

incorporated some new features to

minimize fuel damage that might occur

during a postulated large-break Loss-of-

Coolant Accident:

• Reduced core lattice pitch (distance

between the fuel channels), reducing

the coolant void reactivity (CVR)

during a postulated large-break


• Increased calandria-tube diameter,

resulting in reduced moderatorto-fuel

ratio, which reduces the

moderator volume and, hence,

reduces the CVR

• Enhanced fuel design, with the centre

element containing zirconia with

BNAs, further reducing the CVR

All of the above features combine

to give a small negative CVR value for

nominal end-of-life conditions, such that

the power transient during a large-break

LOCA is benign.

Changes to the fuel design make the

fuel less susceptible to failure during a

LOCA. As above (Question 4), the more

subdivided CANFLEX-ACR fuel bundle

lowers fuel element ratings and reduces

the power-related damage mechanisms

while fuel pellet geometry minimizes

sheath strains and fission-gas pressure,

ACR-1000 Four Unit Layout

reducing the likelihood of fuel failures

during power transients.

Finally, the ACR-1000 design has

retained the two independent fast-acting

reactor shutdown systems, which are the

well-established means of limiting the

reactivity transient during a postulated

large-break LOCA in traditional CANDU

reactors. As a result of all of these

enhancements, calculations show that

during a postulated large-break LOCA,

there will be no fuel failures in the ACR-

1000 reactor design.

6. What innovative fuel cycles have

been used in ACR-1000 to maximize fuel

effi ciency and to minimize concerns of


The reference fuel for the ACR-

1000 has a uniform 2.4% enrichment.

The ACR-1000 uses the advanced

CANFLEX ® fuel bundle, developed

as the optimal carrier for CANDU

advanced fuel cycles. Development is

underway to increase enrichment and

burnup, to further improve economics.

In addition, Recovered Uranium (RU)

from conventional reprocessing can be

burned efficiently in the ACR-1000, with

the addition of fissile LEU or plutonium

(Pu). The reactor can operate with a

full core of 2.4% LEU, or with RU plus

fissile to 2.4% Heavy Element (HE). The

on-power refuelling capability permits

switching back and forth between the two

fuel types, without any hardware changes

to the safety/control systems.

Additionally, spent ACR-1000 fuel

with a residual fissile content of about

1%, opens the possibility of its re-use

in existing CANDU reactors. The ACR-

1000 is also amenable to thorium fuel

cycles. The simplest case, feasible in the

short term, is the Once-Through Cycle

(OTT). This is easy to implement, with

no reprocessing required, to achieve a

burnup of about 21,000 MWd/TeHE.

This cycle also creates a “reservoir” of

Uranium-233 (233U) for future use. In

the longer term, a closed-cycle option

offers burnups to 40,000 MWd/TeHE.

Spent fuel is reprocessed to recycle 233U,

and burnup can be tailored by adding Pu

to fresh bundles.

Proliferation-resistance results from

a combination of technical design features,

operational modalities, institutional

arrangements and safeguards measures.

In CANDU technology, these features are

strongly linked and self-enforced, with

the result that their combination is greater

than the sum of the parts. CANDU technology

has always incorporated intrinsic

proliferation-resistance features—derived

from the fundamental physics of naturaluranium

or LEU-fuelled reactors.

While these inherent barriers

minimize the attractiveness of CANDU

technology as a target for proliferation,

external measures provide verification

(Continued on page 38)

36 Nuclear Plant Journal, September-October 2008


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Continued from page 36

and deterrence through timely detection.

International Atomic Energy Agency

(IAEA) safeguards have been successfully

incorporated in CANDU reactors for

decades, and have evolved over time.

7. What enhancements have been made

in the control station design to ensure

improved human-system interface

Improvements in computer technology—particularly

digital communications

and distributed systems—provided

a significant opportunity to improve the

human system interface in the ACR-1000

Main Control Room (MCR), as follows:

• Enlarged main operator console,

with more computer display stations,

allowing for control and monitoring

at the console instead of at the standto-operate

panels of the past designs;

automating standard manual control

sequences reduces the chance of

human error.

• Improved main operator console and

shift interrogation console, providing

work stations for monitoring safety

and production functions, and for

administrative functions; large

work areas for paperwork and

documentation with easy-access

document storage

• Large-screen displays and a small

section of hardwired backup panels

providing plant overview information

for situation awareness.

• Automated safety system testing,

which can be initiated from the

main operator console and reduces

operator workload

• Highly-effective CANDU Alarm

Message List System (CAMLS),

filtering the alarm message stream to

ensure only pertinent alarms appear

• Seismically-qualified MCR, allowing

operator to remain there following

a seismic event and handle it using

familiar interfaces

8. How has information technology

been used to survey and self-diagnose

problems in the systems, structure, and

components in ACR-1000 How does

this ensure reliable operation and longer

plant life

From smart sensors to increased

process and large equipment diagnostic

monitoring and assessments, the new

digital technologies will enhance

ACR-1000 diagnostic and prognostic

or condition-monitoring capabilities,

including smart sensors and control

elements, vibration-monitoring and

neutronics analysis from the CANDU

6 reference design. ACR 1000 will be

incorporating significantly different

designs and levels of integration than the




reference plants. The move to digitalbased

systems, not just in the areas of

I&C but in communications and other

areas as well, will greatly impact the

functionality performance of the plant

industrial network systems:

Network Design Topologies

• The new “distribution system”

• Operator Support

• Operator rounds Logs

• Video surveillance

(Continued on page 40)


38 Nuclear Plant Journal, September-October 2008


Leading Maintenance Solutions

Industry leadership is something Shaw shares with our clients. As a leader in nuclear

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technology for new nuclear plants and standardized modular construction.




A Market...

Continued from page 38

Health Physics Support

• Personnel tracking Dosimetry Video


• Portable monitoring


• Paging Telephone Cell phone

• Radio


• Access Personnel tracking

• Video surveillance

AECL has developed a suite of

Operations and Maintenance (O&M)

support applications, known as

SMART CANDU, to assist the O&M

organization. These track the health

of key systems and components and

provide diagnostic tools to identify and

correct problems before they result in a

loss of performance. SMART CANDU

applications combine process, chemistry

and inspection data to provide up-todate

assessments of the current status

of key plant systems and components.

For example, ChemAND (Chemistry

Analysis and Diagnostic) monitors water

chemistry and ThermAND monitors heat

transfer systems and components.

Data are stored in a Life-of-Plant

Historian, where they can be easily

retrieved and displayed to compare the

current plant status with past behavior.

The impact of plant operating conditions

on the future performance of critical

components in the system can be further

assessed using one of the embedded

analytical models that are interfaced

with the plant data. This enables

engineering staff to track, for example,

thermal performance, fatigue usage, the

performance of pump/motor sets and

the results of inspection campaigns, and

to predict the impact of plant operating

conditions on steam generator fouling,

activity transport and steam generator


Field tests at domestic CANDU

utilities have demonstrated that these

features greatly reduce the time required

to diagnose problems and allow plant staff

to operate in a more proactive mode.

Thus, these new tools help to optimize

ongoing operation and maintenance while

allowing informed decision-making and

planning for the future. This ensures

reliable plant operation and leads to

longer plant-life.

9. How has the current instrumentation

and control system in the ACR-1000

been upgraded from the previous

AECL designs to ensure reliable plant

operation with longer plant life

The ACR 1000 plant design uses

a distributed control system (DCS) to

perform plant monitoring and control

functions previously implemented using

centralized digital control computers,

analog control devices and relay logic.

The control strategies for the DCS control

programs are based on previous CANDU

designs but are implemented on a new

hardware platform taking advantage of

advances in computer technology and

supplementing this process with new

techniques and analyses. These new

techniques allow system designers to

take advantage of new features possible

in a DCS application, and ensure the

DCS achieves significant capital and

operating cost reductions and improved

safety through high operational and

safety reliability, reduced I&C system


In previous CANDUs, plant control

was performed by centralized control

computers (DCC), analog devices

and relay logic. System control was

performed by dual redundant computers

that executed a set of control programs for

monitoring, annunciation, and control of

plant systems. In a second level, control

devices such as analog controllers and

programmable logic controllers (PLCs)

handled lower-level control functions.

The control and instrumentation

design used in the ACR 1000 plant has

separated the computer control system

from the plant information system in

recognition of the fact that the controls are

less subject to change and more sensitive

to the risk of change. The computer

information systems and human-machine

interaction systems, on the other hand,

must be flexible, expandable and easy to

upgrade to exploit evolving technology.

The primary advantages of this

evolutionary DCS design are as follows:

• The significant elimination of C&I

hardware components, wiring,

cabling and wire terminations

achieves significant capital and

operating cost savings

Plant safety will be enhanced

because the distributed architecture

of the group control functions

makes simultaneous loss of all these

functions due to component failures


• Improved software design tools,

software reviewability and simplified

operating environment will contribute

to reduced software errors.

• Elimination of intrusive hardware

maintenance activities to modify

functionality will also improve plant


The DCS design concept provides

very high reliability and fault tolerance,

minimizing the need to provide separate

local control or hardwired backup. Faulttolerant

features include channelization,

redundancy and fail-safe outputs. Use of

a single hardware platform for high- and

low-level controls reduces maintenance

errors by ensuring familiarity of the

maintenance personnel with a single

control system.

Digital protection systems first

formed part of the CANDU 6 product.

The systems were called Programmable

Digital Comparators (PDCs). The PDCs

formed part of the shutdown systems in

the reactors. They provided much of the

process-related reactor trip coverage,

increasing the potential for more complex

trips. Using computer capabilities, it

was possible to add self-checking and

monitoring to the equipment. The

actuation of the safety functions for the

two shutdown systems in ACR-1000 will

also be software-based, using proven

methods from past and current reactor


For further information on the ACR-

1000, see Nuclear Plant Journal

Volume 26 No.1, January-February


*CANDU ® , Advanced CANDU

Reactor ® , ACR-1000 ® and CANFLEX ®

are registered trademarks of Atomic

Energy of Canada Limited (AECL).


ACR and ChemAND are also

AECL trademarks.

Contact: Heather Smith, AECL,

2251 Speakman Drive, Mississauga,

Ontario. L5K 1B2 Canada; telephone:

(905) 823-9060 ext 7541, fax: (905) 403-

7565, email:

40 Nuclear Plant Journal, September-October 2008

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Generation IV Advanced Nuclear

Energy Systems

By Jacques Bouchard, French

Commissariat à l’Energie Atomique,

France and Ralph Bennett, Idaho

National Laboratory.

Generation IV first stepped on to the

international scene in 2001 when nine

countries joined together on a mission to

develop and implement the next wave of

safe and sustainable nuclear reactors, and

created the Generation IV International

Forum (GIF) to oversee it. Seven years of

important changes in energy, environment

and public acceptance have given the

GIF a renewed sense of purpose. During

those years an R&D program, with a

framework covering technical and legal

aspects, was created to meet the challenge

of expanding nuclear energy throughout

the 21st century.

The GIF countries’ pledge

to cooperate comes at a particularly

urgent time. Worldwide greenhouse gas

emissions grew 70 percent between 1970

and 2004, and if current energy practices

remain unchecked such emissions

will have a devastating effect on the

planet. The dramatic effect on climate

of increased carbon emissions poses a

problem that transcends national borders

and politics. Safe, efficient nuclear

energy must be a part of a serious effort

to stabilize greenhouse gas levels.

Making a significant difference

in carbon emissions would require

a large expansion of nuclear power.

According to Princeton University’s

Carbon Mitigation Initiative, increasing

the number of nuclear power plants to

1000 worldwide—more than double

what it is today—could avoid one billion

tons of carbon emissions per year by

2055. Generation IV aims to develop

reactors and their associated fuel cycles

that assure their long term sustainability

and allow them to address more than just

electricity generation, thereby setting the

stage for sustained expansion through the


Jacques Bouchard

Jacques Bouchard is Special Adviser

to the Chairman of the French

Commissariat à l’Energie Atomique.

Mr. Bouchard has also served as

chairman of the Generation IV

International Forum since 2006.

A Framework for R&D


The effort towards a new generation

of nuclear energy systems started in July

2001, when nine countries signed the GIF

Charter. In doing so, France, Argentina,

Brazil, Canada, Japan, the Republics of

Korea and South Africa, the United States

and United Kingdom signaled their mutual

interest in new nuclear systems. Since

then, Switzerland, Euratom (representing

the nations of the Euratom Treaty), China

and Russia have all become members of

the GIF.

To date, nine of the members have

also acceded to a Framework Agreement, 1

which allows its signatories to formally

participate in the development of

Generation IV nuclear systems. Under

that agreement, System Arrangements

provide the framework for collaboration

on each type of reactor. These

arrangements allow for cooperation

with industry, academia and even other

governments to accomplish the R&D.

Each member finances its own research

Ralph Bennett

Ralph Bennett, PhD, is Director

of International and Regional

Partnerships, Idaho National

Laboratory. In 1979, he earned a

Ph.D. in nuclear engineering at MIT.

He is also the Technical Director of the

Generation IV International Forum.

and development, chooses which systems

it will work on, and shares and protects

the intellectual property they develop


Nuclear Power through

the Generations

The conventional paradigm for

the history of nuclear reactors has been

to separate different types of nuclear

designs into “generations.” Generation I,

dating from the 1950s and 60s, includes

early prototypes in a number of countries.

Generation II, the first commercial power

plants, date from the 70s and 80s and

include Pressurized Water Reactors

and Boiling Water Reactors—designs

generally utilizing water for coolant and

slightly enriched uranium for fuel, almost

all of which are still operating today.

Most nuclear power plants being built

now are categorized as Generation III—

water-cooled reactors with more refined

designs than their Generation II ancestors.

This third generation has evolved to

be both safer and more efficient, but

42 Nuclear Plant Journal, September-October 2008

is nevertheless focused on electricity

generation and only a limited recycle of

the plutonium generated during one cycle

through the core.

Worldwide projections of increased

demand for electricity and new

imperatives to reduce carbon emissions

have lent special urgency to the promise

of next generation systems. New

markets, technical innovations and a

rising acceptance of nuclear energy

have produced the conditions needed

for a revolution in nuclear technology.

Economic competitiveness, improved

safety, conservation of uranium resources

and minimalization of waste, increased

physical protection of the plants and

added resistance to threats of nuclear

proliferation are the new challenges

posed to Generation IV reactors, which

will ensure the sustainable development

of nuclear energy.

Goals for Generation IV

Six different Generation IV nuclear

reactor systems are currently being

advanced. They were identified by an

international group of over 100 experts

who examined more than 130 proposals

sent by specialists from around the

world. The GIF took a top-down

approach to choosing which designs were

most promising versus the challenges

of sustainability, safety, economics,

proliferation resistance and physical

protection. Further considerations

included estimated R&D costs and time

horizons. Though the final six systems

selected have different strengths, each

one was chosen for its unique potential

to contribute to the new face of nuclear

energy and advance toward the following

eight goals:

Sustainability–1: Generation IV

nuclear energy systems will provide

sustainable energy generation that meets

clean air objectives and promotes longterm

availability of systems and effective

fuel utilization for worldwide energy


Sustainability–2: Generation IV

nuclear energy systems will minimize

and manage their nuclear waste and

notably reduce the long-term stewardship

burden in the future, thereby improving

protection for the public health and the


Economics–1: Generation IV nuclear

energy systems will have a clear lifecycle

cost advantage over other energy


Economics–2: Generation IV nuclear

energy systems will have a level of

financial risk comparable to other energy


Safety and Reliability–1: Generation

IV nuclear energy systems operations

will excel in safety and reliability.

Safety and Reliability–2: Generation

IV nuclear energy systems will have a

very low likelihood and degree of reactor

core damage.

Safety and Reliability–3: Generation

IV nuclear energy systems will eliminate

the need for offsite emergency response.

Proliferation Resistance and

Physical Protection–1: Generation IV

nuclear energy systems will increase the

(Continued on page 44)




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Nuclear Plant Journal, September-October 2008 43

Generation IV...

Continued from page 43

assurance that they are a very unattractive

and the least desirable route for diversion

or theft of weapons-usable materials, and

provide increased physical protection

against acts of terrorism.

A short overview of each system


Sodium-Cooled Fast Reactor

(SFR): The GIF is currently devoting

much of its effort to this system. It

uses liquid sodium for coolant, thereby

gaining a high power density and lower

coolant volume fraction. It features a

closed fuel cycle, which is needed for fuel

breeding and/or actinide management.

The layout is flexible, with a pool layout

(shown) or a compact loop layout. Either

could be adjusted to produce a small-,

medium- or large-sized reactor. The

SFR can be economically competitive in

electricity markets with innovations to

reduce capital costs. The SFR is more

efficient than thermal-spectrum reactors

with open fuel cycles, with its potential

to use both fissile and fertile isotopes

of uranium. GIF has been taking a

streamlined approach to developing SFR,

by building upon technologies that are

already being deployed throughout the

world and advancing their performance.

Progress in developing the SFR is well

underway, with advances being made in

fuel technology in France, compact heat

exchangers in the United States, and

design innovations underway in Japan 2 .

Very High Temperature Reactor

(VHTR): The GIF is also devoting much

of its effort to this system. It is a heliumcooled

thermal reactor that can achieve

an outlet temperature approaching 900

degrees Celsius. The ceramic fuel of

the VHTR has a high degree of passive

safety, and the high temperature gives it

a high thermal efficiency approaching

50%. The high temperature also allows

the VHTR to be applied to hydrogen

production and other high temperature

process heat applications, as well as low

temperature heat applications such as

water desalination, thereby addressing

non-electric energy needs. The primary

areas of research involve fuels, high

temperature materials, and hydrogen

production processes, and virtually all

of the GIF members are collaborating on

this system.

Gas-Cooled Fast Reactor (GFR):

A fast-spectrum thermal reactor using

helium coolant with an outlet temperature

of 850 degrees Celsius. It is attractive

because of its high efficiency and

minimization of transuranic waste.

Supercritical Water Reactor

(SCWR): The SCWR design uses

water above its critical point condition

(374°C, 22.1 MPa) as the coolant. This

avoids the need for steam generators

and considerably reduces the size of the

turbine generator. It is a flexible design,

configurable as a fast or thermal reactor.

Its thermal efficiency may exceed 45%,

and its lower capital cost favors the

economical production of electricity.

Lead-Cooled Fast Reactor (LFR):

This fast reactor uses molten lead or

lead/bismuth as a coolant and has a

high degree of safety since the coolant

is less chemically reactive than sodium.

It operates at a temperature higher than

the SFR, which may allow its use for

44 Nuclear Plant Journal, September-October 2008

hydrogen production as well as electricity.

It has a closed fuel cycle.

Molten Salt Reactor (MSR): This is

an epithermal reactor design in which the

fuel is actually dissolved in the coolant.

Specifically, it uses liquid fluorides of

uranium and plutonium for fuel, dissolved

in fluorides of lithium, beryllium, sodium

or other elements. The system provides

for processing the wastes and adding new

fuel online, which greatly reduces the

fissile material inventory and avoids the

development and qualification of fuel and


The Future

Generation IV designs improve upon

current reactors in several ways. Four

of the designs are fast reactors, allowing

the reactors to potentially exploit the

full energy potential of uranium—both

fissile and fertile isotopes. Generation

III reactors extract energy from a much

smaller fraction of uranium in the fuel,

where as Generation IV reactors can

extend the uranium resource by about a

factor of 50 beyond this. Another option

for Generation IV is to improve on current

designs by recycling all actinides—not

only the bred plutonium-239, but the

other actinides found in the waste as

well. This revolution in fuel utilization

would also dramatically reduce the

radiotoxicity and heat generated by the

waste by transmuting it to shorter-lived

fission products, thus making it easier to


Several of the Generation IV designs

are high-temperature reactors, which

can generate not only electricity but

also provide process heat for industrial

purposes. Process heat has good potential

for application to a wide range of

industries, from petroleum refineries and

chemical plants to large-scale hydrogen

production potentially for revolutionizing


One of the overarching goals of

Generation IV technology, and one that

is most appealing to the international

community, is its potential to reduce

carbon emissions. This will only be

accomplished through considerable

R&D, and for example, the GIF members

collaborating on the SFR and VHTR have

already jointly committed over $500M for

the next five years. The GIF believes that

Generation IV, through improved safety,

economics, safety and proliferation

resistance and physical protection, can

help ensure nuclear energy’s long term

expansion and sustained contribution to

the world’s energy security.


[1] Generation IV International Forum,

“Framework Agreement,” available


Framework-agreement.pdf, 28 Feb


[2] Generation IV International Forum,

“GIF Annual Report 2007,” available


annual_report2007.pdf, Mar 2008.

Contact: Ralph Bennett, Idaho

National Laboratory, P.O. Box 1625,

Idaho Falls, ID, 83415-3805; telephone:

(208) 526-7708, fax: (208) 526-0876),


Nuclear Plant Journal, September-October 2008 45

Innovative Reactor Designs

A Report Based on the Recent

Publications by International Atomic

Energy Agency, Vienna, Austria.


There is continuing interest in

Member States in the development and

application of small and medium sized

reactors (SMRs). “Small” reactors are

defined as those with an equivalent

electric power less than 300 MW(e).

“Medium sized” reactors are those with

an equivalent electric power between

300 and 700 MW(e). It is important that

small or medium sized reactor does not

necessarily mean small or medium sized

nuclear power plant. Like any nuclear

power plants, those with SMRs can be

built several-at-a-site, or as twin units. In

addition to this, innovative SMR concepts

provide for power plant configurations

with 2, 4, or more reactor modules [1, 2,

and 3]. The units or modules could then

be added incrementally in time taking

benefits of the effects of learning, timing,

construction schedule, and creating

an attractive investment profile with

minimum capital-at-risk.

Opportunities for SMRs

In the near term, deployment potential

of the SMRs is based largely on their

ability to fill niches where larger plants do

not fit in, or to offer economic advantages

related to incremental capacity increase.

The applications could be industrial

sites or population centres in remote

off-grid locations, countries or country

areas with small and medium electricity

grids, investment and human resource

conditions that benefit from incremental

capacity addition or non-electrical

applications that require proximity of a

nuclear energy source to the process heat

application plant [1].

For the longer term, there is

interest in innovative designs that

promise improvements in safety,

security, proliferation resistance, waste

management, resource utilization,

economics, product variety (e.g.

desalinated seawater, process heat,

district heat and hydrogen) and flexibility

in siting and fuel cycles. Many innovative

reactor designs have been proposed in the

small-to-medium sized range, in many

cases providing for multi-module plant

configurations to achieve larger, often

flexible, overall power station capacity

[2, 3].

Many of the niche advantages of SMRs

are expected to be particularly attractive

to some of the approximately 40 countries

that have recently expressed interest in

starting nuclear power programmes, for

example, low investment increments and

suitability for small grids. On the other

hand, vendors in Argentina, China, India,

Japan, the Republic of Korea, the Russian

Federation, South Africa, and the USA

are actively developing and promoting

new SMR designs [2, 3].

Progress toward


For about a dozen of innovative SMR

designs, current progress in developing

the technology and finalizing the design

suggests possible deployment within the

next decade.

Construction began in June 2006 in

the Russian Federation on a pilot floating

cogeneration plant of 300 MW(th)/70

MW(e) with two water cooled KLT-40S

reactors. Deployment is scheduled for


In July 2006, the Russian Federation

and Kazakhstan created a joint venture to

complete design development for a 350

MW(e) VBER-350 reactor (basically a

scaled-up version of the KLT-40S) for

use in land-based co-generation plants

[2]. The first-of-a-kind plant deployment

is targeted in 2015 at the former BN-350

site in Kazakhstan.

Five integral PWR designs

are in advanced design stages and

commercialization could start around

2015 [2, 3]: the 335 MW(e) IRIS design

developed by International consortium

led by Westinghouse of USA (currently

co-owned by Toshiba Corp. of Japan) ; the

330 MW(th) SMART design developed in

the Republic of Korea for a co-generation

plant; the prototype 27 MW(e) CAREM-

25 developed in Argentina, for which

construction in planned to be complete

by 2011, and which is expected to further

into commercial designs of 150 and

300 MW(e); the 200 MW(th) NHR-200

developed in China for district heating

and other applications, both electrical

and non-electrical; and the MASLWR

of 45 MW(e) per module, developed in

the USA, for multi-purpose applications

and multi-modular plants of up to 540


The Advanced Heavy Water Reactor

of 300 MW(e), developed in India for cogeneration

plants, is considered to be built

early in the next decade [2]. The reactor is

being designed for operation with 233U-

Pu-Th fuel and uses boiling light water

coolant and heavy water moderator. All

mentioned above SMRs provide for or

do not exclude co-generation option

with non-electric energy products being

produced as well as the electricity.

The 165 MW(e) PBMR, a high

temperature gas cooled reactor with

pebble bed fuel and direct gas turbine

Brayton cycle, developed in South Africa,

is- scheduled for demonstration at full

size by 2012 [2]. Future configurations of

this reactor will include 4 and 8-module

plants. The 200 MW(e) per module

HTR-PM, a high temperature gas cooled

reactor with pebble bed fuel and indirect

supercritical steam energy conversion

cycle developed in China, is planned

for a full size demonstration in 2013 [1,

2]. Two-module plant configuration is

foreseen for the commercial version of

this reactor.

Some small reactor designs

incorporate an option of operation without

on-site refuelling, which may help reduce

the obligations of a user for spent fuel

and waste management [3]. Several of

such designs have a potential of being

deployed as first-of-a-kind or prototype

plants within the next decade [1, 3].

These include [3] the ABV of 11 MW(e)

and 8-year refuelling interval, which

is an integral design PWR backed by

(Continued on page 48)

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marine reactor experience, and a couple

of non water cooled reactors, which are

the sodium cooled 4S reactor of 10-50

MW(e) and 10-30 year refuelling interval,

developed in Japan, and the lead-bismuth

cooled SVBR-75/100 reactor of 101.5

MW(e) and 6-9 year refuelling interval

developed in the Russian Federation.

The latter design is backed by operating

experience of the Russian submarine


Small Reactors without

On-site Refuelling

Small reactors without on-site

refuelling are the reactors designed for

infrequent replacement of well-contained

fuel cassette(s) in a manner that impedes

clandestine diversion of nuclear fuel

material [1, 3]. Small reactors without

on-site refuelling incorporate increased

refuelling interval (from 5 to 15 years and

more), consistent with plant economy and

considerations of energy security. Small

reactors without on-site refuelling are

either factory fabricated and fuelled or

undergo a once-at-a-time core reloading

performed at the site by a dedicated

service team provided by the vendor;

such team is assumed to bring in and take

away the fresh and spent fuel load and the

refuelling equipment.

About 30 concepts of small reactors

without on-site refuelling are being

analyzed or developed within national

and international programmes in Brazil,

India, Indonesia, Japan, Morocco, Russian

Federation, Turkey, U.S.A., and Vietnam

[3]. Small reactor designs without onsite

refuelling are being considered for

both nearer-term and longer-term water

cooled, liquid metal cooled and molten

PBMR single module building (PBMR, Pty, South Africa) [2]

salt cooled reactor lines and some nonconventional

fuel/coolant combinations.

Whether for fast or for thermal

neutron spectrum concepts of such

reactors, the fuel discharge burn-up

and the irradiation of core structures

never exceeds standard practice from

the conventional or typically projected

designs. The refuelling interval is then

extended by derating core specific power,

and the power densities never significantly

exceed ~100 kW(th)/litre and often are

much lower. Burn-up reactivity loss is

mitigated by using burnable poisons and

active control rods in thermal systems and

by designing for internal breeding in fast

systems. Although the specific inventories

of fissile materials (per unit of power

and energy produced) are higher than

for reactors with conventional refuelling

schemes, some concepts of fast spectrum

reactors without on-site refuelling are

capable of self-sustainable operation

on fissile materials (breeding ratio ~ 1)

within a closed nuclear fuel cycle. In this,

breeding option is typically excluded

owing to a restricted neutron economy.

Challenges for

Innovative SMRs

Innovative SMRs in many cases

do not attempt to compete with large

economy of scale plants in the established

markets; they rather attempt to meet

the needs of those users to whom large

economy-of-scale deployments are not

suited. To be competitive in anticipated

alternative markets, innovative SMRs rely

on approaches alternative to economy

of scale. Such approaches include the

economy of multiple prefabricated

reactor or equipment modules, reduced

design complexity resulting from the

application of those design features that

are most appropriate for the reactor of a

given capacity, an option of incremental

capacity increase with possible benefits

resulting from “just in time” capacity

(Continued on page 50)

Potential SMR cost factor advantages (Westinghouse, USA) [1]

48 Nuclear Plant Journal, September-October 2008

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addition and learning curve factors,

shorter construction period and, possibly,

greater involvement of local labour. The

effectiveness of all these approaches for

the conditions of targeted markets should

be demonstrated, which is a challenge of

prime importance for all innovative SMRs.

Many of the innovative SMR concepts

incorporate design features and system

configurations that are not proven in

operating practice of reactors for civil

nuclear power; also, many innovative

SMRs are just non water cooled reactors.

The designers of innovative SMRs target

licensing within the currently established

national regulatory framework [4] but

mention that further elaboration of national

regulatory norms toward technology-neutral

and risk-informed approach could facilitate

licensing consideration and further design

improvement. In addition to incorporating

many inherent safety features, some

innovative SMR concepts suggest stronger

reliance on passive systems of innovative

design. Reliability of such systems needs

to be proven to enable risk-informed

qualification and licensing [5, 6].

Many potential applications of SMRs

may require them to be located in proximity

to the users:

• In industrial cogeneration applications,

such as hydrogen production, they

must be sited adjacent to the industrial

site for delivery of process heat;

• They could supply energy to cities in

regions where only a local electrical

grid exists;

• They could produce energy products

such as potable water and district

heat, which cannot be transported

to significant distances without a

significant economic penalty.

These siting considerations lead to a

requirement for very high levels of safety

and reliability. Co-locating a nuclear and a

chemical plant on a single site may require

developing additional safety rules and

regulations to be applied to both of them


Licensing of a nuclear power plant

with a reduced or eliminated emergency

planning zone, which is aimed by the

designers of many innovative SMRs, will

benefit from risk-informed regulation being

emplaced. Achieving the goal of a reduced

off-site emergency planning would require

both, development of a methodology to

prove that such reduction is possible in

the specific case of a plant design, and

adjustment of the existing regulations. Riskinformed

approach to reactor qualification

and licensing could be of value here, once

it gets established. Within the deterministic

safety approach it might be very difficult to

justify reduced emergency planning in view

of a prescribed consideration of a postulated

severe accident with radioactivity release to

the environment owing to a common cause

failure. Probabilistic safety assessment

(PSA), as a supplement to the deterministic

approach, might help justify very low core

damage frequency (CDF) or large early

release frequency (LERF), but it does not

address the consequences and, therefore,

does not provide for assessment of the

source terms. Risk-informed approach that

introduces quantitative safety goals, based

on the probability-consequences curve,

and links them to certain defence in depth

levels, which could help solve the dilemma

by providing for a quantitative measure

for the consequences of severe accidents

and by applying a rational technical and

non-prescriptive basis to define a severe

accident. An example of such approach is

in the recently published IAEA-TECDOC-

1570 “Proposal of a Technology- Neutral

Safety Approach for New Reactor Designs”


Many small reactors without onsite

refuelling incorporate substantially

increased refuelling interval, ranging from

~5 to 20-25 years and beyond. The operating

experience for such elongated refuelling

intervals is generally unavailable in civil

nuclear power [1]. The known experience of

marine reactors confirms the possibility of a

7 to 8-year continuous operation of small

reactors [3]. Therefore, the construction of

a prototype would be a must for many small

reactors without on-site refuelling.


In the end of 2007, of the world’s 439

operating nuclear power plants, 134 were

with SMRs. Of the 23 newly constructed

NPPs, 9 were with SMRs [8]. In the near

term, most new nuclear power reactors

are likely to be evolutionary large units.

But particularly in the event of a nuclear

renaissance, the nuclear industry can expect

an increasing diversity of customers, and

thus an increasing number of customers

with needs potentially best met by one or

more of the innovative SMR designs now

under development.




Technology Review 2007, Attachment

4: “Progress in Design and Technology

Development for Innovative Small and

Medium Sized Reactors”, IAEA (2007):





Innovative Small and Medium Sized

Reactor Designs 2005: Reactors with

Conventional Refuelling Schemes,

IAEA-TECDOC-1485 (2006).


ENERGY AGENCY, Status of Small

Reactor Designs without On-site

Refuelling, IAEA-TECDOC-1536




the Nuclear Power Plants: Design

Requirements, safety standards Series,

No. NS-R-1, IAEA, Vienna (2000).

[5] MARQUÈS M. et al, Methodology

for the reliability evaluation of a

passive system and its integration into

a Probabilistic Safety Assessment,

Nuclear Engineering and Design 235

(2005), pp 2612-2631.

[6] NAYAK, A.K., GARTIA, M.R.,



Reliability Analysis of a Boiling Twophase

Natural Circulation System

Using the APSRA Methodology,

Proceedings of International Congress

on Advances in Nuclear Power Plants

(ICAPP 2007), Nice, France, May 13-

18, 2007 (Paper no. 7074).


ENERGY AGENCY, Proposal for a

Technology-Neutral Safety Approach

for New reactor Designs, IAEA-

TECDOC 1570 (2007).


ENERGY AGENCY, Power Reactor

Information System (PRIS): http://

50 Nuclear Plant Journal, September-October 2008

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Guidance for New Vendors

By John Nakoski, U.S. Nuclear

Regulatory Commission.

1. What factors do the vendors new to

the nuclear power industry need to take

into consideration to determine if they

should qualify their quality assurance

program for supplying products

and services to the nuclear power

industry Also please describe briefl y

any guidance for such vendors totally

unfamiliar with the nuclear power


I think from the NRC perspective, we

see this as a business decision. A vendor

new to the industry needs to understand the

requirements for quality assurance in this

industry. The NRC’s quality assurance

requirements are outlined in Appendix B

to 10 CFR Part 50. Our quality assurance

requirements are typically more stringent

than other industries. There is an added

cost to meeting these requirements, and a

new vendor needs to consider how best to

factor that cost into its business decision.

In addition to the quality assurance

requirements, the NRC has regulations in

place that require reporting of defects and

non-compliance. These requirements are

provided in 10 CFR Part 21. In terms of

becoming qualified, a new vendor would

need to have as a customer, an NRC

licensee or an applicant with an approved

quality assurance program. The NRC

licensee or applicant could then conduct

an audit of the new vendor’s quality

assurance program to assess whether it

complies with NRC requirements. If the

results of the audit indicate the new vendor

is in compliance, then the vendor can be

added to the licensee’s or applicant’s

approved suppliers list. Alternatively,

if the new vendor is supplying parts or

services to a vendor that is already on an

NRC licensee’s or applicant’s approved

suppliers list, the existing vendor can

audit the new vendor and qualify the

Responses to questions by Newal

Agnihotri, Editor of Nuclear Plant


John Nakoski

John A. Nakoski, Chief, Quality

and Vendor Branch 2, Division of

new vendor’s quality assurance program.

Basically, an NRC licensee or an industry

approved vendor would need to conduct

an audit of the new vendors quality

assurance program to assess whether it

complies with NRC requirements. So,

most of the burden for qualifying new

vendors falls to the licensees, applicants or

potential applicants. The Nuclear Utilities

Procurement Issues Committee (NUPIC)

has taken on the NRC licensees’ and

applicants’ role of conducting these audits

of the suppliers to the commercial nuclear

industry in the US. Of course at the NRC,

we have our regulatory oversight role. We

inspect those organizations that provide

basic services or basic components to the

commercial nuclear industry.

NUPIC is an organization that

is comprised of essentially all the

commercial US nuclear utilities and

several international utilities. It’s an

organization that shares resources to

conduct audits required by Appendix B

to 10 CFR Part 50 to provide reasonable

assurance that vendors have an effective

quality assurance program and that they

comply with 10CFR Part 21.

We’ve interacted with NUPIC for

many years. We have been observing its

processes and the implementations of

its audits at selected vendors throughout

the years. We have also been observing

Construction Inspection and Operational

Programs, Offi ce of New Reactors, U.S.

Nuclear Regulatory Commission

Together with Juan Peralta, Mr. Nakoski

is responsible for developing and

implementing the NRC’s programs for

the oversight of vendors support related

to new reactor construction and quality

assurance programs for the design,

licensing, and construction of new

reactors. Mr. Nakoski has 25 years of

experience in the nuclear energy arena,

primarily with the NRC. He is a 1983

graduate from Penn State with a B.S. in

Nuclear Engineering.

its periodic meetings where it discusses

vendor and supply chain issues.

2. Briefl y describe how USNRC

implements its vendor inspection


The NRC’s vendor inspection

program for new reactors is implemented

following guidance documented in our

inspection manual chapter (IMC) 2507.

For the current operating reactors, IMC

2700 describes the vendor inspection

program. These IMCs lay out the basic

requirements that we follow to oversee

any organization that provides safetyrelated

parts or services to the nuclear

power industry. Under the IMCs, we

have inspection procedures that provide

directions to the inspectors that they

follow in planning for and conducting

inspections. The inspection procedures

provide guidance on reviewing vendor

quality assurance, commercial grade

dedication, and 10 CFR Part 21

programs. In addition, our vendor

inspection program includes oversight

of organizations that conduct audits of

vendors - organizations such as NUPIC.

For new reactors, our current plan is to

conduct about 10 vendor inspections and

several NUPIC audit observations each

year. We may perform more if necessary

and have the resources available. While

52 Nuclear Plant Journal, September-October 2008

our vendor inspections provide us with

direct insights into the performance of

vendors, our observation of NUPIC audits

of vendors gives us a sense of the quality

of the industry’s oversight of vendors,

and provides us the ability to provide

feedback on how the oversight process can

be enhanced. All of our vendor oversight

activities, our inspections and NUPIC

audit observations, are documented in

publicly available inspection reports.

These reports are available from the NRC

website in our electronic reading room

under ADAMS. We also make them

available through our Quality Assurance

website (


3. What is the best website link that a

new vendor can click on to fi nd simple

guidance for the quality assurance

process required for qualifying to supply

nuclear products and services to the

nuclear power industry

The NRC maintains information

on its website that new vendors would

find useful regarding the programs and

requirements we follow when inspecting

vendors. The website is located at:

To get this

website from the NRC’s main public

(, click on the “Nuclear

Reactor” tab, then drop down to “New

Reactor Licensing, there is a link below

“Under How We Regulate” called “Quality

Assurance for Nuclear Power Plants”

and there is a link to “Regulations and

Standard Review Plan”, “Vendor Inspections”,

Inspections for New Reactor Licensing”

and “Nuclear Procurement Issues

Committee and Industry Interface.”

Also, presentations we’ve made during

various conferences over the past several

years can be found under the “Nuclear

Procurement Issues Committee (NUPIC)

and Industry Interface ” link. We have a

variety of information on the website and

encourage new and existing vendors, or

anyone interested in this area, to explore

the site.

4. Where is the new reactor licensing

procedure defi ned Is this 10 CFR

Part 52 What are the provisions for

quality assurance in this code of federal


10 CFR Part 52 provides the regulatory

framework for new reactor licensing.

It points back to 10 CFR Part 50,

for the technical and quality assurance

requirements. It does point back to and

specify that applicants are required to do

safety-related activities under quality assurance

programs that meet 10 CFR Part

50, Appendix B requirements. Additional

guidance for preparing new reactor applications

is provided in Regulatory Guide

1.206, “Combined License Applications

for Nuclear Power Plants.” When preparing

license applications under 10 CFR

Part 52, the information applicants use is

required to be gathered under an Appendix

B quality assurance process. The application

itself is developed to satisfy the

completeness and accuracy requirements

of 10 CFR 50.9 and applications need to

be submitted under oath and affirmation.

5. How do you gather the list of vendors

who are supplying products and services

to the nuclear power plants to ensure

that these vendors are qualifi ed for the


For new reactor construction, we

have requested information from the

industry through a regulatory issues

summary, 2007-08, “Updated Licensing

Submittal Information to Support the

Design-Centered Licensing Review

Approach.” So far we have received

some responses from applicants and the

major vendors supplying the designs.

In addition, through our interface with

NUPIC, we have a list of vendors that

have been qualified by the current fleet

of operating reactors and are supplying

basic components to the currently

operating fleet of power reactors. Using

this information to give us confidence

in the quality of products provided to

nuclear power plants, the NRC conducts

inspections of a sample of the vendors

that have been approved by licensees and

oversees the audits conducted by NUPIC

of these vendors.

6. If a vendor in China wants to be

certifi ed, will NRC go to China

It is important to recognize that the

NRC is not in the process of certifying or

approving vendors to supply products and

services to the nuclear power industry.

We inspect vendors for compliance

with our regulations. Also, in today’s

manufacturing arena, many of the

vendors of major components are located

overseas. So, if a vendor in China was

selected by a licensee or applicant and

put on an approved suppliers list for the

construction of a new reactor, our plan is

to include that vendor in the population

of vendors that we may inspect. If we had

concerns with the quality of the vendor

regardless of where they are, domestically

or internationally, that would factor

into our decision on whether we should

inspect a particular vendor. If we received

indications from our interactions with the

applicants, through NUPIC, from peer

regulators in other countries, or as a result

of observations of construction inspection

activities by the regional staff that

problems with quality existed, we would

factor that into our decision. So the short

answer to the question would be yes, if

we determined that it was necessary or

consistent with our program guidelines. I

would add that over the last 18 months, we

have been building an extensive interface

program with our peer regulators across

the globe. As one example, we recently

conducted coordinated inspections with

our fellow regulators in Japan and Korea

at specific vendors in those countries.

7. What organizations other than

USNRC are involved in establishing

guidelines for quality for new reactor

construction activities

Other organizations involved in

establishing guidance on quality assurance

requirements include:

1. Nuclear Utilities Procurement

Issues Committee (NUPIC)

2. Nuclear Energy Institute (NEI)

3. American Society of Mechanical

Engineers (ASME)

4. Nuclear Industry Assessment

Committee (NIAC)

5. American Society of Quality


6. Electric Power Research Institute


Contact: John, A. Nakoski, U.S.

Nuclear Regulatory Commission, MS T-7F3,

Washington DC 20555; telephone: (301)

415-1068, email:

Nuclear Plant Journal, September-October 2008 53

Road Map for Future Energy

By John Cleveland, International Atomic

Energy Agency.


Today nuclear energy contributes

approximately 15.2% of the world’s

electricity. By February 2008, there

were 439 nuclear power plants (NPPs) in

operation worldwide, with a total capacity

of 371.7 GWe. Further, 34 units, totaling

28.1 GWe, were under construction.

During 2006 nuclear power produced

2659.7 billion kWh of electricity. Based

on information provided by its Member

States, the IAEA projects that nuclear

power will grow significantly, producing

annually between 2760 and 2810 billion

kWh by 2010, between 3120 and 3840

billion kWh by 2020, and between 3325

and 5040 billion kWh by 2030 [1].

The vast majority of today’s nuclear

power plants use water-cooled reactors.

In the near term most new nuclear plants

will be evolutionary water cooled reactors

(Light Water Reactors (LWRs) and Heavy

Water Reactors (HWRs)], often pursuing

economies of scale. Other reactor types

have had considerably less operational

and regulatory experience and will take

still some time to be widely accepted

in the market. These innovative designs

promise shorter construction times and

lower capital costs and could help in the

future to promote a new era of nuclear


While nuclear power contributes

significantly to electricity generation,

most of the world’s energy consumption

is for heat and transportation. Through

advanced applications, nuclear energy can

penetrate these energy sectors now served

by fossil fuels that are characterized

by price volatility, finite supply, and

environmental concerns.

Advanced applications of nuclear

energy include seawater desalination,

district heating, heat for industrial

processes, and electricity and heat for

hydrogen production. In addition, in

the transportation sector, since nuclear

electricity is generally produced in a base

load mode at stable prices, nuclear power

John Cleveland

Mr. John Cleveland has worked at the

IAEA since 1991. Until 1994 he was in

charge of IAEA’s activities in technology

can contribute as a carbon-free source of

electricity for transportation (e.g. trains

and subway systems) and for charging

electric and plug-in hybrid vehicles.

Due to these factors, the IAEA has

carried out this study to examine the

opportunities, challenges and solutions

for water-cooled reactors to contribute

to these advanced applications of nuclear

energy [2].

Seawater Desalination

Water is essential for the sustainable

development of society. Water scarcity

is a global issue, and every year more

countries are affected by growing water


Large-scale commercially available

seawater desalination processes can

generally be classified into two categories:

(a) distillation processes (these are the

Multi-Stage Flash – MSF, and the Multi-

Effect Distillation – MED processes) that

require mainly heat plus some electricity

for ancillary equipment, and (b) membrane

processes (Reverse Osmosis – RO) that

require only electricity to provide the

necessary pumping power.

The desalination of seawater

using nuclear energy is a feasible and

demonstrated option for production of

potable water. Over 200 reactor-years

of operating experience on nuclear

development of high-temperature gascooled

reactors. Since 1994 he has been

the leader of the Water-Cooled Reactors

Group of the Nuclear Power Technology

Development Section.

Before joining the IAEA, he worked for

the Babcock and Wilcox Company and at

the Oak Ridge National Laboratory

in the USA.

Mr. Cleveland received his Masters

Degree in Physics from Virginia

Polytechnic Institute and State

University, USA, in 1972. He has

authored more than 80 technical papers

and reports in the fi eld of nuclear

reactor technology and safety.

desalination have been accumulated

worldwide, and more demonstration

projects are being prepared. However,

nuclear desalination today contributes

only 0.1 % of the total desalting capacity

worldwide [3].

Table 1 (see page 56) shows the nuclear

reactors used or under construction

for seawater desalination. In addition to

those systems shown in Table 1, other

water-cooled concepts are being developed

for seawater desalination. For example,

the nuclear heating reactor (NHR)

developed in China could provide heat for

desalination, and the SMART concept,

developed in the Republic of Korea, the

CAREM concept of Argentina, and the

KLT-40 floating power unit developed in

Russia 1 , could be used for cogeneration

of electricity and seawater desalination.

Countries suffering from scarcity

of water are generally not the holders

of nuclear technology. They do not

have nuclear power plants, and do not

have a nuclear power infrastructure.

(Continued on page 56)

1 The Floating Power Unit under construction at

Severodvinsk, Russia, is planned to be comissioned

in 2010, and will be used for electricity and

district heating. Future potential units outside of

Russia could be used for electricity and seawater


54 Nuclear Plant Journal, September-October 2008

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a touchscreen (shown here)

that is easy to read and use, as

well as an ergonomic arm that

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any height or angle. Supplied

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And with our digital control system, if very nearly can be.

This innovative upgrade eliminates control system obsolescence and

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> the single source 800.448.4124


Road Map..

Continued from page 54

The utilization of nuclear energy in

such countries will require infrastructure

building and institutional arrangements

for issues such as financing, liability,

safeguards, safety, and security.

District Heating

District heating involves the supply

of space heat and hot water through a

district heating system, which consists of

heat plants (usually producing electricity

simultaneously) and a network of

distribution pipes. Potential application of

district heating is in climatic zones with

relatively long and cold winters. In many

countries, such as central and northern



European countries and countries in

transition economies, district heating has

been widely used for decades.

Coal and gas dominate the fuels used

for district heating. However, several

countries (Bulgaria, China, Czech Republic,

Hungary, Romania, Russia, Slovakia,

Sweden, Switzerland and Ukraine) have

experience in nuclear district heating using

water-cooled reactors, so the technical

aspects can be considered well proven.

In order to be able to compete with

Location m 3 /day Status

fossil-fuel-fired heat boilers, the capital cost

per installed MW of heat production capacity

for a nuclear-based system must be such

that the production costs are competitive.

Dedicated reactors providing district heat

can potentially achieve acceptable costs,

due to their lower temperature operating

conditions, simple design, modularization

and standardization, and advanced safety


New nuclear heat-producing plants

must, of course, meet the user’s requirements

on availability and reliability, including

alternative heat-producing capacity that

could serve as backup. For this purpose,

heat storage allows a matching of the heat

supply to the heat demand. Today there are

many examples of short-term storage, for

instance, on the daily scale that relies on

hot water accumulator tanks. In the future,

more long-term storage facilities may be


Table 1: Reactor types used or under construction for seawater desalination

LMFR Kazakhstan (Aktau) 80,000 In service till 1999




Ohi 1,2,3,4


Ikata 1,2,3

Genkai 3,4

USA (Diablo


~1500 In service

Operating experience ~

170 R-Ys

~4500 In service

India (Madras) 6,300 RO commissioned in


MSF to be commissioned

in 2008



4,800 Under construction;

Commissioning –in 2008

Industrial Heat Process

Process heat involves the supply of heat

required for industrial processes from one

or more centralized heat generation sites

through a steam transportation network.

Within the industrial sector, process heat

is used for a large variety of applications

with different heat requirements and

with temperature ranges covering a wide

spectrum. Examples of industries that

consume considerable amounts of heat


• food,

• paper,

• chemicals and fertilizers,

• petroleum and coal processing, and

• metal processing industries.

The chemical and petroleum industries

are the major consumers of process heat

worldwide. These would be key target

clients for possible applications of nuclear


The supply of energy for industrial

processes has an essential character: all

industrial users need the assurance of

energy supply with a high reliability, and

the heat should be produced close to the

point of use. Many of the process heat

users, in particular the large ones, usually

are located outside urban areas, often at

considerable distances. This makes joint

siting of nuclear reactors and industrial

users of process heat not only viable, but

also desirable in order to drastically reduce

the heat transportation costs.

The nuclear process heat supply has

to be reliable. As an example, the average

steam supply availabilities for chemical

processing and oil refineries are 92% and


There is experience in providing

process heat for industrial purposes with

nuclear energy in Canada, Germany,

Norway, Switzerland, and India. New

plant designs that can provide heat, or both

heat and electricity, are being designed in

Russia, the Republic of Korea, Canada, and

other countries.

Current water cooled reactors can

provide process heat up to about 300ºC,

and some future innovative water cooled

reactor designs 2 have potential to provide

heat up to approximately 550ºC.

Although nuclear industrial process

heat applications have significant potential,

it has not been realized to a large extent.

In fact, currently only the Goesgen reactor

in Switzerland and the RAPS–2 reactor in

India continue to provide industrial process

heat, whereas other nuclear process heat

systems have been discontinued after

successful use. Among the reasons cited for

closure of these units, one is availability of

cheaper alternate energy sources.

For potential future application of

nuclear process heat, an important example


Specifically Super-critical Water Cooled Reactors,

being developed within the Generation-IV

International Forum, could be deployed by around


56 Nuclear Plant Journal, September-October 2008

is the use of nuclear energy for oil sand

open-pit mining and deep-deposit extraction

in Canada. Alberta’s oil sand deposits are

the second largest oil reserves in the world,

and have emerged as the fastest growing,

soon to be dominant, source of crude oil in

Canada. Currently, the majority of oil sand

production is through open-pit mining,

which is suitable for bitumen extraction

when the oil sand deposits are close to

the surface. The ore, a mixture of bitumen

and sand, is removed from the surface by

truck and shovel operation. The ore is then

mixed with hot water to form a slurry that

eventually undergoes a separation process

to remove bitumen from the sand.

The thermal energy required for the

open-pit mining process is in the form of

hot water at a relatively low temperature

(around 70°C), and the rest is dry process

steam at around 1.0 to 2.0 MPa. The oil

extraction facilities require electrical

power as well. The steam and electricity

requirements can be met by water cooled


To increase production capacity, oil

companies are developing new technologies

to extract bitumen from deep deposits.

Among them, Steam-Assisted Gravity

Drainage (SAGD), which uses steam

to remove bitumen from underground

reservoirs, appears to be the most promising

approach. Recently, this in-situ recovery

process has been put into commercial


Overall, for both extraction

methodologies (open pit mining and

SAGD), a significant amount of energy is

required to extract bitumen and upgrade it

to synthetic crude oil as the feedstock for

oil refineries. Currently, the industry uses

natural gas to provide this energy. As oil

sand production continues to expand, the

energy required for production becomes a

great challenge with regard to economic

sustainability, environmental impact

and security of supply. Therefore, the

opportunity for nuclear reactors to provide

an economical, reliable and virtually zeroemission

source of energy (both electricity

and steam) for the oil sands becomes a

realistic option.

Energy for Transportation

Transportation represents approximately

20% of the world’s energy consumption.

In the United States, transportation

is the fastest growing energy sector.

The Organization for Economic Co-operation

and Development International Energy

Agency projects that global primary energy

demand will grow by 50% by 2030, with

70% of that growth coming from developing

countries, especially China. Half of

that increase will be for electricity production

and 20% for transportation.

It is clear that if means are found for

nuclear energy to power a significant part

of the transportation sector, it could have a

significant impact on global environmental

sustainability. Two ways this could

occur would be through the advancement

transportation systems based on electricity,

such as trains, subways, electric and plugin

hybrid vehicles charged with nuclear

generated electricity, and of vehicles fuelled

with hydrogen produced by nuclear energy.

Following are some examples.

A) Electricity for plug-in hybrid

electric vehicles

The potentially large market demand

for electricity for powering plug-in hybrid

electric vehicles is eminently suited to

current and evolutionary water-cooled

nuclear power plants. Because nuclear

plants generally operate at base load

conditions, provide electricity at stable

and predictable prices, and produce clean

electricity, they are especially well suited

to play a near term role in powering the

transportation sector, while helping to

reduce greenhouse gasses from this sector.

Hybrid vehicles are commercially

available today. Almost all use regenerative

braking to charge an on-board battery

for locomotive power. With these battery

systems, vehicles can be designed to allow

the gasoline engine to turn off when the vehicle

is stopped or during cruising.

Overall energy use for hybrids is

about 40% less than that for conventional

vehicles, with an equivalent reduction in

greenhouse gas emissions (CO 2

, CH 4

, and

N 2


Plug-in hybrid electric vehicles

extend this technology by allowing the

drive battery to be charged externally. In

this way, the vehicle can be driven in an

all-electric mode for a certain distance

with no power from the gasoline engine.

This can provide significant savings in

terms of petroleum usage and emissions,

especially since the majority of miles

driven are for short commutes. These

emission reductions materialize only if the

source of external electricity is clean and

carbon free, of course. Importantly, plugin

hybrid manufacturers have announced

targets of 20 to 40 miles on a single charge.

One developer recently unveiled a plugin

hybrid demonstration vehicle which

uses a combination of ultra-capacitors and

batteries for energy storage and has an allelectric

range of 40 miles.

In this study, a simplified model of

potential growth in usage of plug-in hybrid

electric vehicles, which assumed that all

automobiles and light trucks in the US would

be plug-in hybrid vehicles by 2035, showed

that 200-250 GW of electricity would be

needed for overnight charging in the U.S.

This would replace 280 million gallons of

fuel per day with the corresponding large

reduction in production of greenhouse

gasses from the transportation sector. New

electricity generation capacity at this scale

would also require new transmission and

distribution lines and substations. A similar

analysis for Japan suggests the need for 35

GW of electricity for overnight charging,

which is within the capacity of spare power

at night.

Aside from the need for increases in

generating and transmission capacity, other

barriers will need to be overcome before

there is widespread adoption of plug-in

hybrid electric vehicles:

• Conversion of automobile technology

from conventional gasoline-powered

vehicles to electric and plug-in hybrid


• Public acceptance of plug-in hybrid


• Structuring of electricity pricing

mechanisms to provide low-price

electricity during off-peak demand

periods to encourage use of nuclear

power plants for base load generation;

• Provision of other incentives (e.g., tax

benefits) for adoption of vehicles that

produce less greenhouse gases and

reduce reliance on petroleum fuels.

A key technology need is development

of lighter, less expensive, reliable batteries

having a factor of 5 to 10 greater energy

storage capacity that would support longer

all-electric distances. Lithium-ion batteries

are the main focus of current research and


B) Hydrogen for transportation

Hydrogen for transportation is

receiving significant attention around the

(Continued on page 58)

Nuclear Plant Journal, September-October 2008 57

Road Map...

Continued from page 57

world because of high petroleum prices

and unreliable oil supplies. Two ways of

hydrogen utilization in transportation are

currently being taken into consideration –

internal combustion engine (ICE) vehicles

and fuel cell (FC) vehicles. While ICE

vehicles represent current technology

with modest modifications, fuel cell

vehicles are in a stage of intensive R&D

and prototype testing.

Car manufacturers are focusing

more effort on fuel cell vehicles than on

hydrogen ICE vehicles. Many prototypes

have been introduced, some of them in

small series (tens of cars). Current trends

are mainly focused on hybridization, such

as combining fuel cells with Nickel metal

hydride (NiMH) batteries, ultra capacitors,

or other types of electric storage.

Although this increases the complexity

of the vehicle, thus increasing the cost, it

brings advantages with regard to covering

power peaks during acceleration, when

the electric motor draws high current

from the fuel cell, and also increases the

driving range, because hybrid vehicles

optimize fuel consumption, and also the

use of braking recuperation.

It is not only important to have technical

problems solved, public acceptance

is also important. For this purpose, hydrogen

fuelled buses have been successful.

Currently there are about 60 of them

serving on a daily basis in different cities

including London, Hamburg, Madrid,

Stuttgart, Stockholm, Porto, Amsterdam,

Barcelona, Luxembourg, Reykjavik and


The lack of the hydrogen infrastructure

makes fleet customers important for

early hydrogen transportation markets.

It is much easier to build one centralized

filling station near a city bus operator

or dispatch service than to service the

distributed market for personal cars.

Motorcycles, scooters and electric

bikes represent a smaller, but interesting,

market opportunity. Such means of transportation

are significant in many Asian

countries, where the pollution is growing

and causing health problems.

Hydrogen Production

As an alternative path to the current

fossil fuel economy, a hydrogen economy

is envisaged in which hydrogen would

play a major role in energy systems

and serve all sectors of the economy,

substituting for fossil fuels. Hydrogen

as an energy carrier can be stored in

large quantities, unlike electricity, and

converted into electricity in fuel cells,

with only heat and water as by-products.

It can also fuel combustion turbines and

reciprocating engines to produce power

with near-zero emission of pollutants.

The current worldwide hydrogen

production is roughly 50 million tonnes per

year. Although current use of hydrogen in

energy systems is very limited, its future

use could become enormous, especially if

fuel-cell vehicles would be deployed on a

large commercial scale.

Today, hydrogen is used mainly in

petroleum refineries and the chemical industry.

In the United States, for example,

these uses represented 93% of hydrogen

consumption in 2003.

The U.S., Japan, and other nations

are exploring ways to produce hydrogen

using nuclear energy. While some consideration

is given to hydrocarbon reforming

techniques, such as steam-methane reforming,

much of the work is focused on

means of splitting water by electrolytic,

thermo-chemical, and hybrid processes.

Considerable efforts have concentrated

on high-temperature processes such as

high-temperature steam electrolysis and

the sulphur–iodine and calcium-bromine

cycles. These processes operate at higher

temperatures (>750°C) than can be

achieved by water-cooled reactors. Advanced

reactors such as the very high

temperature gas cooled reactor (VHTGR)

can generate heat at these temperatures,

but first demonstration of hydrogen production

with gas cooled reactors is not

expected until around 2015 (in Japan) to

2020 (in the USA).

Current and evolutionary water cooled

reactors can produce outlet temperatures

in the range of ~300-350°C. Supercritical

water cooled reactors (SCWRs),

being developed within the Generation-

IV International Forum, can achieve

temperatures of ~550°C. Examples of

processes for hydrogen production within

these temperature ranges follow.

A. Steam Reforming of Dimethyl

Ether (~300°C)

Toshiba of Japan has proposed that

steam reforming of dimethyl ether (DME),

a derivative from fossil fuels or biomass,

could be used to produce hydrogen with

300°C heat from water cooled reactors.

DME is synthesized from natural gas

from small or medium-sized gas fields,

coal seam gas, and natural gas with a large

CO 2

fraction. DME is usually produced

by a partial oxidation process of natural

gas without emitting CO 2,

as shown by

the following formula:

2CH 4

+ O 2

-->CH 3


+ H 2


The DME reforming reaction is as


(1/2)CH 3


+ (3/2)H 2


+3H 2

–24.4 kJ/ (H 2


The produced hydrogen fraction is

high at temperatures of 285-300°C. Specifically,

Toshiba has developed, together

with Shizuoka University, a DME reforming

catalyst that gives 98% conversion of

DME to hydrogen at 285°C. The catalyst

is Cu-Zn/Al 2

O 3

powder [4].

With 40 MW of heat supply about

108 kg H 2

/year of hydrogen production

is possible, which is of the same scale as

the largest hydrogen plant in the world.

To date, the demonstrated production rate

is 4.10 kg H 2


B. Low temperature electrolysis

Hydrogen production processes

based on reforming of methane not only

use fossil resources (CH 4

), but also produce

CO 2

. Nuclear energy can be used

for splitting water to produce hydrogen

without using fossil resources and without

producing CO 2

. Although the energy

requirements for hydrogen production

by low-temperature water electrolysis are

relatively high, it is a presently available

technology for hydrogen production. Water

electrolyzers can be decoupled from the

power plant. Therefore, electrolysers can

be used for decentralized hydrogen production.

C. Steam reforming of methane with

a membrane reformer system (500 to


A conventional steam methane reforming

(SMR) system for hydrogen production

involves introducing a mixture of methane

and steam into a nickel-based catalyst bed

in the steam reformer, where the SMR reaction

proceeds at 750 to 800°C. The re-

(Continued on page 60)

58 Nuclear Plant Journal, September-October 2008

Intelligent Monitoring Technology

By Chris Demars, Exelon Nuclear.


Exelon has developed and deployed

over 500 plant equipment computer

models to identify early degradation

which has resulted in avoided losses of

approximately $600K in 2 months of


The Exelon Centralized Performance

Monitoring (CPM) pilot project was

formally established in June of 2007. A

team of two employees augmented with

summer intern assistance was established

to develop approximately 500 plant

equipment models.

With InStep’s experience in the

Nuclear Industry and data historian

specialty experience they were able to

develop an extremely effective and easy

to use anomaly detection tool.

The application allows a user to

quickly assemble and train a group of

related plant process computer points

in a model that when deployed will

constantly monitor those points for other

than normal behavior. The software

package can then be configured to alert

an individual to parameter relationship

changes that should be investigated for

potential adverse equipment conditions

that could otherwise lead to failure.


Early detection of equipment failures

prevents the hazardous environment that

Nuclear Energy Institute’s Top Industry

Practice (TIP) Award highlight the

nuclear industry’s most innovative

techniques and ideas.

This was a 2008 NEI Process Award


The team members who participated

included: Chris Demars, Project

Manager, Exelon Nuclear; Dave Miller,

Exelon Nuclear; Mike Rog, Exelon

Nuclear; Bill Bielke, InStep Software;

Sean Gregerson, InStep Software.

often accompanies rotating equipment

failures or the release of industrial gases

and process fluids, and improves nuclear

and radiological safety through early

detection and improved management of

equipment degradation.

Specific examples include the recent

condensate pump failure avoidance. Lead

time for a replacement pump is 4–6 weeks,

and during that time a backup pump would

not be available which reduces plant

margin and impacts safety. The coupling

failure would have also challenged

personnel safety due to the accessibility of

the area the pump is installed in. Overall

safety is also improved due to the reduced

scope and frequency of equipment repair


The cost saving methodology that

the centralized performance monitoring

pilot has employed is to conservatively

calculate the cost of the worst case

scenario(s) that may have occurred

without early detection of a degraded

condition and to then multiply the

worst case cost by a probability factor

to obtain avoided cost. The following

three recent early detections examples

demonstrate that method and allow an

annual approximation of avoided cost

based on two months of monitoring with

approximately 30 models deployed for

each unit in the fleet.

1. Condensate pump motor coupling

seizure - $500K

The condensate pump model alerted

due to two bearing oil temperatures that

were not within the predicted pattern of

allowable values. The temperatures of the

two bearing were well within accepted

operating levels but were approximately

4 °F outside “normal behavior” as

defined by the multi-dimensional cluster

based technology applied in the models.

The cause was found to be an improperly

assemble coupling that was seizing and

approaching mechanical failure.

Failure of the coupling would have

resulted in damage to both the motor and

pump with a replacement lead time of 4

to 6 weeks. Replacement cost, expediting

fees and craft overtime is estimated at

Chris Demars

Chris Demars has over 28 years of

experience in nuclear power generation

management. His diverse experience

includes project management, various

program recovery management

positions (work management,

engineering, operations, unit restart),

engineering, operations and nuclear

station corrective action program

development and implementation,

initial and accelerated license operator

training (lead instructor), training

program development/implementation,

of on-line work management and

engineering work management

processes. He has a Bachelor of Science

in Nuclear Engineering Technology.

$700K. The probability of this failure

scenario is estimated at 0.70, or $490K.

Online loss of the pump with a failure

of the standby pump to start would have

resulted in a power reduction of 34% for

12 hours or ~$100K. The probability of

this failure scenario is estimated at 0.10

or $10K.

2. Service water temperature

controller failure – $30K.

The main turbine vibration model

alerted due to a small step change in

vibration on the number 11 bearing. The

vibration level was not significant enough

to cause an alarm of any normal plant

monitoring systems. The cause of the

step change was a change in generator

hydrogen temperature which is cooled

by stator water cooling that is cooled by

service water. This particular nuclear unit

has not removed or blocked the stator

water cooling temperature turbine trip

and was susceptible to a trip during the

temperature changes that were caused

by the failed controller. Trip of the main

turbine would have resulted in a loss of

generation for 24 hours or $600K. The

probability of a turbine trip is estimated

at 0.050 or $30K.

(Continued on page 62)

Nuclear Plant Journal, September-October 2008 59

Road Map...

Continued from page 58

formed gas is supplied to a shift converter,

where carbon monoxide and water are converted

into carbon dioxide and additional

hydrogen. The reformed gas is then passed

to a pressure swing adsorption separator to

separate the hydrogen.

A membrane reformer system, composed

of a steam reformer equipped with

catalytic membrane modules with a palladium-based

alloy and a separate nickelbased

catalyst, can perform the reforming

reaction, the shift reaction, and the hydrogen

separation process simultaneously at

temperatures of 500 to 600°C [5].

In 2004-2005 Tokyo Gas Company

demonstrated the operation of a methanecombusting

membrane reformer system

at a hydrogen fuelling station for fuel cell

vehicles in downtown Tokyo. The system

performance, efficiency, and long-term

reliability were confirmed by producing

>99.99% hydrogen at 3.6 kg/h for more

than 3,000 hours with hydrogen production

efficiency of about 80. SCWRs could

provide heat at the temperatures needed

for steam-methane membrane reformer


D. Thermo-chemical and Hybrid

Processes (500 to 600°C)

Thermo-chemical and hybrid

thermo-electrochemical cycles have the

potential for hydrogen production by

water-splitting with higher efficiencies

than low-temperature water electrolysis.

Although over 200 thermo-chemical and

hybrid electro-thermo-chemical reaction

cycles for producing hydrogen have been

identified [7], only about eleven of them

have maximum reaction temperatures

below 600°C. These lower-temperature

cycles can reduce the thermal burden,

mitigate demands on materials, and

potentially be coupled with nearer-term

nuclear reactors.

Five of these cycles have recently

been the subject of active research. They

include a family of copper-chloride

cycles (530° - 550°C) [8], an active metal

(potassium-bismuth) cycle (475 - 675°C)

[9], a magnesium-chloride cycle (500°C)

known as the Reverse Deacon Cycle [10],

a U-Eu-Br heavy-element halide cycle,

and a hybrid sulphur-based cycle [11].

Development work on such cycles has

generally been limited to small laboratory

scale testing.


While there are very important opportunities

for deployment of nuclear energy

into advanced applications, challenges

and difficulties should not be overlooked.

In particular, competition will drive the

choice of energy sources for each application.

Policies internalising the cost of

carbon and other pollutants are needed to

fully realize the benefits of nuclear energy

in alleviating the risk of climate change.

Advanced applications of nuclear energy,

due to their ability to provide energy products

economically and without producing

greenhouse gases, can play an important

role in enhancing public acceptance of

nuclear energy.



AGENCY, Energy, Electricity and

Nuclear Power Estimates for the

Period up to 2030, Reference Data

Series No. 1 (2007 Edition)


ERGY AGENCY, Advanced Applications

of Water-Cooled Nuclear Power

Plants, (IAEA TECDOC-1584, Vienna,



ENERGY AGENCY, Status of Nuclear

Desalination in IAEA Member States,

TECDOC-1542, IAEA, Vienna





SHI, K., “Hydrogen Production with

Steam Reforming of Dimethyl Ether

at the Temperature Less Than 573 K”,

in Proceedings of International Congress

on Advances in Nuclear Power

Plants, No. 5138 (2005)

[5] TASHIMO, M. et. al., “Advanced

Design of Fast Reactor-Membrane

Reformer (FR-MR)”, Proceedings

of Second Information Exchange

Meeting on Nuclear Production of

Hydrogen, Argonne USA (2003).

[6] UCHIDA, S. et. al., “Concept of

Advanced FR-MR”, 15th World

Hydrogen Energy Conference, Paper

No. 30D-08, Yokohama Japan (2004).



J.B., Thermochemical Hydrogen

Production, GRI-80-0023, Institute of

Gas Technology, Chicago, IL 60616

(June 1981).

[8] SERBAN, M., LEWIS, M.A., and

BASCO, J.K., Kinetic Study for the

Hydrogen and Oxygen Production

Reactions in the Copper-Chlorine

Thermochemical Cycle, 2004 AIChE

Spring National Meeting, Conference

Proceedings, 2004 AIChE Spring National

Meeting, Conference Proceedings,

pp. 2690-2698 (2004).

[9] MILLER, W.E., MARONI, V.A. and

WILLIT, J.L., DOE Patent Case

Number S-104650 (2006).


S.D., and BOYLE, B.D., A Hybrid

Thermochemical Electrolytic Process

for Hydrogen Production Based

on the Reverse Deacon Reaction,

International Journal of Hydrogen

Energy, 31 (Aug. 2006) 1241 - 1246.

[11] NAKAGIRI, T. et. al., “A new

thermo-chemical and electrolytic

hybrid hydrogen production process

for FBR”, Paper 1021, GENES4/

ANP2003, Kyoto (Sep. 2003).


The IAEA appreciates the contributions

of the following persons to this study: B.M.

Misra (Consultant to IAEA, India); S. Kuran

(Atomic Energy of Canada Ltd., Canada);

L. Janik (Nuclear Research Institute Řež,

Czech Rep.); D.S. Shukla (Bhabha Atomic

Research Centre, India); M. Hori (Nuclear

Systems Association, Japan); T. Chirica

(Societatea Nationala Nuclearelectrica SA,

Romania); V. Polunichev (Experimental

Machine Design Bureau OKBM, Russian

Federation); C. Halldin (OKG AB,

Sweden); M. C. Petri (Argonne National

Laboratory, USA, and Chairman of this

activity); R. Uhrig (Univ. of Tennessee,

USA); and E. Bertel (Organization for

Economic Co-operation and Development

- Nuclear Energy Agency).

Contact: John Cleveland, International

Atomic Energy Agency, P.O. Box 100,

Vienna, A-140, Austria; telephone: 43-1-

2600-22819, fax: 43-1-2600-29598, email:

60 Nuclear Plant Journal, September-October 2008

Vermont's Largest Source of


By Tyler Lamberts, Entergy Nuclear

Operations, Inc.

Vermont Yankee nuclear power

station is the largest in-state source of

electricity. It provides about a third of the

electricity used by Vermonters from its

site on the Connecticut River in the town

of Vernon.

The plant was planned and

constructed at a time when New England

was heavily dependent on imported oil

for electric generation. As oil supplies

for New England grew more unstable and

as the environmental degradation caused

by fossil-fired pollution was becoming

apparent, New England was among the

first regions in the country to invest in

nuclear plants as an alternative to fossilfueled

power plants.

Central Vermont Public Service and

Green Mountain Power Corporation

were the original lead utilities in the

joint ownership of the 540 megawatt

plant. After considering several Vermont

sites, including the eastern shore of Lake

Champlain, the 102 acre Vernon site on

the western shore of the Connecticut

River was selected. The site was chosen

for its available land, sound bedrock,

electric transmission lines, cooling water

and its proximity to an active rail line

for receiving large components and for

shipping spent fuel.

In 1972, after a four-year construction

and federal licensing, the plant was

connected to New England’s 345kv grid

in time to position the state well against

the 1974 Arab embargo on oil shipments

to the United States.

With Vermont Yankee reliably on

line, fossil-fired power plants in the

northeast were gradually edged out of the

role of baseload electric generators – a

major step in reducing air pollution in the


In the late 1990’s, Vermont Yankee’s

utility owners decided that the plant would

fare better in every respect as part of a

fleet of plants owned and operated by a

utility specializing in nuclear generation.

In 2001, Entergy was the high bidder for

the Vermont Yankee plant. In 2002, the

Vermont Public Service Board considered

Entergy’s expertise and experience in the

nuclear energy field, and approved the

purchase of Vermont’s most valuable

and reliable generating asset as being in

the long-term best interest of the state of


Stakeholder Benefits

As a condition of the sale, Entergy

committed to supply the plant’s electricity

to the utilities that formerly owned the

plant at capped prices through to the

end of the license term in 2012. Recent

estimates by the Vermont Department

of Public Service show that Vermonters

are likely to save more than $665 million

on their electric rates thanks to that


In Entergy’s first year of Vermont

Yankee ownership, it doubled the plant’s

community contribution level including

a large donation for restoration of a

downtown theatre as a community

cultural arts center.

Overall, Vermont Yankee’s operation

represents about $200 million of economic

activity per year in the region through its

payroll, taxes and local purchases.

Extended Power Uprate

It is Entergy’s goal, as owner and

operator of Vermont’s largest generating

asset, to maintain the plant’s favorable

economics so as to continue to serve

the region. The previous utility owners

had found the plant to be an excellent

Tyler Lamberts

Tyler Lamberts graduated in June,

2008 with a degree in Marketing from

Oregon State University. Tyler currently

works for OSU Conference Services in

Corvallis, Oregon.

candidate for a power uprate, but were

not in a position to make the substantial

investment as they were leaving the

generation end of the utility business.

After Entergy conducted its own 10-

month in-house engineering evaluations,

the company moved forward with a

full 20-percent extended power uprate

initiative. The Vermont Public Service

Board approved the uprate in March

of 2004 and the Nuclear Regulatory

Commission followed suit two years

later after a review by the Atomic Safety

and Licensing Board and the Advisory

Committee on Reactor Safeguards.

According to the NRC, their staff review

of Vermont Yankee’s uprate petition was

the most extensive uprate review to-date

involving more than 9,000 hours of NRC

staff time.

Entergy’s uprate power ascension

program implemented over three months

in the Spring of 2006 was notable for

its deliberate and incremental approach

that involved several hold points for

plant performance data analyses and for

communicating the results with General

Electric, the Nuclear Steam Supply

System designer, and state and federal


Of particular interest during

the ascension was the steam dryer

performance. Similar boiling water

reactors ascending to uprate power

levels had experienced unexpected dryer

degradation due to changes in steam line

acoustics in the increased steam flow.

Acoustic data collected by several dozen

(Continued on page 62)

Nuclear Plant Journal, September-October 2008 61

Vermont's Largest...

Continued from page 61

monitors on the steam piping was fully

analyzed and compared with predictions

before ascending to the next power level.

In the first operating cycle following

the uprate, Vermont Yankee posted a

reliable breaker to breaker run of 549

days and inspections of the dryer in the

subsequent refueling outage found no

flaws related to the new uprate steam flow

and verified the accuracy of engineering


Dry Cask Storage

Another initiative underway at

Vermont Yankee is construction of a dry

fuel storage pad to allow Vermont Yankee

to remain in service beyond 2008. The

Vermont legislature and the Vermont

Public Service Board approved the project

in April 2006.

In August 2007, local contractors

completed a 1,050 cubic yard, 12-hour

Intelligent Monitor...

Continued from page 59

3. Reactor Feed Pump (RFP) lube

oil cooler temperature controller failure -


The nuclear unit was recovering from

the effects of a transformer failure induced

voltage transient that caused some system

isolations and momentary power losses.

There was no significant plant transient.

Shortly after the transient the RFP bearing

cooling models for all 3 pumps went into

alert. The plant was notified the following

day that one of the controllers did not

recover form and initial transient and was

continuing to cycle significantly. The station

determined that the controller for the C RFP

oil cooler had failed and was able to stabilize

temperatures manually until the controller

was replaced.

The worst case scenario is bearing

damage due to rapid over heating and loss of

the RFP. The physical damage is estimated

at $100K with a probability of 0.10 and lost

generation of 33% for 24 hrs or $200K with

a probability of 0.050.

The total avoided costs for the 2 month

period is $550K. If detected failures of a

continuous concrete pour for the ten

thousand square foot pad.

License Renewal

In January of 2006, Entergy filed a 20-

year license renewal request with the NRC

to extend license expiration from 2012 to

2032. The federal review is progressing

well. In 2007, NRC staff issued the final

Site Environmental Impact Statement

and the draft Safety Evaluation Report.

Also in 2007, the Advisory Committee

on Reactor Safeguards sub-committee

recommended proceeding with the full

committee review of Vermont Yankee


In 2008, the Atomic Safety and

Licensing Board will hear several

contentions brought by interveners and a

state review process on Vermont Yankee

license renewal will get underway.

With the uprate, dry cask and license

renewal initiatives in place, Vermont

Yankee will continue as an economical

and reliable source of electricity and

a vital component of New England’s

diversified energy mix.

similar magnitude continue to be revealed

by the centralized performance monitoring

technology, an annualized avoidance of

$3.3M can be expected. Avoidance of a

failure of a generation critical component

could also easily exceed this amount but

the cost avoidance calculation methods ate

conservative and follow methods similar

to those in an EPRI technical paper on

intelligent monitoring case studies.

In addition to online monitoring the

technology is being employed to assist in

trouble shooting by focusing on discreet time

frames and re-playing the plant conditions

through the program to detect additional

anomalies. The technology is also being

promoted for increased monitoring when

returning equipment and systems to service

after maintenance. These two areas have the

potential to increase the annualized savings

from improved equipment reliability.


Work continues with the software vendor

InStep to improve current productivity in

investigating and acknowledging alerts that

are generated by the software models.

Additional efficiency gains are in

progress relative to the integration of CPM

into the Exelon model for performance

Community Partnership

The employees at Vermont

Yankee play a vital role in neighboring

communities by routinely supporting

educational, civic and cultural projects

and events. Over the years, they have

volunteered their time as guest speakers

at local schools, sponsored child daycare

and learning centers, constructed

playgrounds and taken an active role in

local robotic competitions. Employees

have also contributed to the education

system as coaches, referees and mentors.

Each year, employees participate in

company-sponsored events such as the

Brattleboro Fourth of July Celebration

and the Winter Carnival. They also give

their time, expertise and efforts to Habitat

for Humanity. Vermont Yankee is also

one of the founding sponsors and ongoing

contributors to the local food drive called

Project Feed the Thousands.

Contact: Rob Williams, Vermont

Yankee, P.O. Box 7002, 185 Old Ferry

Road Brattleboro, VT 05302-7002;

phone: (802) 258-4181; fax: (802) 258-

2150; e-mail:

and equipment condition monitoring.

The monitoring that is being performed

by individual system managers can be

optimized, standardized and integrated more

effectively when considered together with

all of the station monitoring activities that

are performed by the various departments.

Opportunities also exist to increase the

number of sensors that are available for

modeling and realize additional efficiencies

to eliminate more time consuming, labor

intensive and in many cases, less effective



The use of this and similar intelligent

monitoring technology within a centralized

group monitoring a fleet of generating

stations would apply across the industry. The

recently evolved cluster based monitoring

technology can also be implemented on a

smaller scale at single units or a few units

with similar result.

Contact: Chris Demars, Exelon Nuclear,

200 Exelon Way, KSA-2-N, Kennett Square,

PA 19348; telephone: (610) 765-5427,

pager: (800) 672-2285 PIN 0338, email:

62 Nuclear Plant Journal, September-October 2008


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