CONTACT: Amy E. Cubbage, NRR

 301-415-2875 

POLICY ISSUE

(Information)

SECY-04-0117

July 9, 2004

FOR: The Commissioners

FROM: Luis A. Reyes

Executive Director for Operations /RA/

SUBJECT: SEMIANNUAL UPDATE OF THE STATUS OF NEW REACTOR LICENSING

ACTIVITIES

PURPOSE:

This paper informs the Commission of the staff’s new reactor licensing activities since the

issuance of SECY-04-0001, “Semiannual Update of the Status of New Reactor Licensing

Activities,” dated January 2, 2004.

BACKGROUND:

In SECY-01-0188, “Future Licensing and Inspection Readiness Assessment,” dated

October 12, 2001, the staff assessed its technical, licensing, and inspection capabilities, and

described enhancements supporting new reactor licensing. The staff also committed to provide

the Commission with semiannual updates of the status of new reactor licensing activities. Four

updates have been provided so far: SECY-02-0076, dated May 8, 2002; SECY-03-0005, dated
The Commissioners -2-

January 8, 2003; SECY-03-0113, dated July 7, 2003; and SECY-04-0001, dated

January 2, 2004.

The attachment to this paper describes recent design certification review activities, early site

permit (ESP) review activities, preapplication activities for new reactor designs, and supporting

infrastructure development. Interactions with stakeholders, including international and

inter-governmental activities, are also discussed. 

DISCUSSION:

The staff continues to focus its efforts on the completion of the Westinghouse Advanced Plant

1000 (AP1000) design certification review and the review of the ESP applications filed by

Dominion Nuclear North Anna, LLC (Dominion), for the North Anna ESP site, Exelon

Generation Company, LLC (Exelon), for the Clinton ESP site and System Energy Resources,

Inc. (SERI), a subsidiary of Entergy Corporation, for the Grand Gulf ESP site. The staff is

progressing toward completion of the preapplication reviews for the General Electric (GE)

Economic and Simplified Boiling-Water Reactor (ESBWR) and the Atomic Energy of Canada,

Limited (AECL) Technologies Inc. Advanced CANDU Reactor 700 (ACR-700). The staff is

addressing regulatory infrastructure issues to support these reviews. As resources allow, the

staff is also working on long-lead-time infrastructure issues that will facilitate the review of

non-light-water-reactors (non-LWRs). Brief summaries of key new reactor licensing activities

since issuance of SECY-04-0001 are provided below.

AP1000 Design Certification

The draft safety evaluation report (DSER) was issued on June 16, 2003, with 174 open items. 

All of the DSER open items are now considered to be resolved. No additional information is

needed by the Nuclear Regulatory Commission (NRC) staff to complete the final safety

evaluation report (FSER). The staff remains on schedule to issue the FSER by

September 13, 2004.

Combined License (COL)

Three consortia have submitted proposals to the Department of Energy (DOE) in response to a

November 2003 solicitation of interest for participation in new nuclear plant licensing

demonstration projects. The first consortium consists of Dominion Resources, AECL

Technologies Inc., Hitachi America, and Bechtel. This consortium submitted a proposal to DOE

in mid-March 2004 based on the ACR-700 design at the North Anna site. A COL application for

the facility may be submitted as soon as mid-CY 2006. Based on current information, the staff

plans to budget resources for the review of a possible COL application in FY 2007.

The second consortium consists of the Tennessee Valley Authority (TVA), Toshiba, GE,

Bechtel, United States Enrichment Corporation (USEC), and Global Nuclear Fuel-Americas

LLC. This consortium submitted a proposal to DOE on April 23, 2004, to conduct a detailed

study of the potential construction of a two-unit GE-Toshiba-designed Advanced Boiling Water

Reactor (ABWR) nuclear plant at the Bellefonte site. On May 23, 2004, DOE announced that it

will fund one-half of the cost associated with the study, which will total $4.25 million over the

next 10 months. Following completion of the study in April 2005, TVA will make a decision

whether to file a COL application.
The Commissioners -3-

The third consortium, known as NuStart Energy Development LLC, consists of nine energy

companies and two nuclear reactor vendors. The energy companies are Constellation

Generation Group, Duke Energy, Electricité de France (EDF) International North America,

Entergy Nuclear, Exelon Generation, Florida Power & Light Co., Southern Company, Progress

Energy, and TVA. The nuclear reactor vendors are Westinghouse and General Electric. The

consortium submitted a proposal to DOE on April 26, 2004, and is considering Westinghouse’s

AP1000 design and GE’s ESBWR design. The Bellefonte site is one of several sites being

evaluated by the consortium. The current projection for a possible COL application to the NRC

is the 2008 time frame, with regulatory approval of the COL desired late in 2010 or in 2011.

Early Site Permits

The staff is reviewing ESP applications from three applicants: Dominion Nuclear North Anna,

LLC (Dominion), for the North Anna ESP site; Exelon Generation Company, LLC (Exelon), for

the Clinton ESP site; and System Energy Resources, Inc. (SERI), a subsidiary of Entergy

Corporation, for the Grand Gulf ESP site. Mandatory hearings associated with each of these

applications are underway. 

Construction Inspection Program (CIP)

The construction inspection framework was issued on May 11, 2004, as NUREG-1789, 

“10 CFR Part 52 Construction Inspection Framework.” The CIP team continued its efforts to

ensure staff readiness to conduct inspections by updating the agency’s Strategic Workforce

Planning (SWP) survey. The CIP team worked with the Office of Human Resources (HR) to

incorporate a series of knowledge and skill statements into the SWP survey specifically related

to construction. Interactions with the Office of Nuclear Regulatory Research (RES) and the

Office of Nuclear Materials Safety and Safeguards (NMSS) have been aimed at using the

experience and expertise of those offices to improve the development of the CIP. The CIP

team has contacted the U.S. Navy and the Federal Aviation Administration (FAA) with the

specific purpose of learning from their implementation of similar inspection activities.

Regulatory Infrastructure

The staff has proposed a rulemaking to revise 10 CFR Part 52 based on lessons learned during

the previous design certification reviews and on discussions with nuclear industry

representatives about the ESP and COL review processes. The proposed rule was published

for comment in the Federal Register on July 3, 2003 (68 FR 40025). The staff received eight

comment letters. Due to the nature of these comments and the extent of changes necessary to

address them, the staff expects to forward a revised proposed rule package to the Commission

later this year, rather than a final rule. 

In the previous semiannual status update (SECY-04-0001), the staff informed the Commission

that it had begun interactions with stakeholders to resolve generic COL issues prior to the

receipt of the first COL application. A set of 23 generic issues has been identified for

discussion. The staff plans to address these issues to facilitate the preparation and review of

future COL applications. Two of the 23 issues involve inspections, tests, analyses and

acceptance criteria (ITAAC). As discussed in the attachment, the staff is working with

stakeholders to identify acceptable emergency planning ITAAC. The staff is also working to
The Commissioners -4-

respond to the Commission’s direction in a May 14, 2004, staff requirements memorandum

(SRM) on programmatic ITAAC.

On May 3, 2004, the staff issued Review Standard (RS) 002, “Processing Applications for Early

Site Permits,” which consolidates existing ESP review guidance, updates the guidance to reflect

the ESP licensing process, and identifies the scope of the ESP review. 

The staff is also preparing to review non-LWR designs by addressing certain policy and

technical issues. SECY-03-0047, “Policy Issues Related to Licensing Non-Light-Water Reactor

Designs,” dated March 28, 2003, identified seven policy issues for Commission consideration. 

In a SRM dated June 26, 2003, the Commission approved the staff’s recommendation on four

of the issues, requested additional information on two other issues and did not approve the

staff’s recommendation on one issue. The staff is preparing a response to the Commission on

the status of the two issues requiring additional information. Further, as directed by the

Commission, the staff is continuing its interactions with external stakeholders on the issues

identified in SECY-03-0047.

CONCLUSIONS:

New reactor licensing activities are focused on the review of applications for design certification

and ESPs and preparation for a COL application. The staff continues to develop the technical

infrastructure needed to complete these reviews and the regulatory infrastructure that will make

the licensing process more efficient and effective in the future. The staff will also continue to

reach out to all stakeholders to build public confidence in these activities.

RESOURCES:

Agency resources assigned to new reactor licensing activities have risen sharply over the past

few years and are expected to continue to increase in FY 2005-2006. To develop budget

estimates for FY 2005-2006, we considered information and experience gained in

preapplication reviews, along with application schedule information provided by prospective

applicants. As new information becomes available, the staff continues to implement the

planning, budgeting, and performance management (PBPM) process to update budget

estimates and prioritize work.

Activities leading to a COL have the highest priority for new reactor licensing resources. The

staff expects to focus its efforts on certifying the AP1000 design, completing the ESP reviews,

reviewing the ACR-700 and ESBWR designs, and reviewing COL applications if they are

received. The staff will give lower priority to other design certification and preapplication

activities that are not expected to be referenced in potential COL applications. These lower

priority reviews may not receive resources commensurate with their visibility or the desires of

some prospective applicants. Projects that might be affected include the International Reactor

Innovative and Secure (IRIS) and Pebble Bed Modular Reactor (PBMR) designs and other

reactor designs which may pursue preapplication interactions with the NRC.
The Commissioners - 5 -

COORDINATION:

The Office of the General Counsel has reviewed this paper and has no legal objection. The

Office of the Chief Financial Officer has reviewed this paper for resource implications and has

no objections.

/RA/

Luis A. Reyes

Executive Director

 for Operations

Attachment: Semiannual Update of the Status of

 New Reactor Licensing Activities
The Commissioners - 5 -

COORDINATION:

The Office of the General Counsel has reviewed this paper and has no legal objection. The

Office of the Chief Financial Officer has reviewed this paper for resource implications and has

no objections.

/RA/

Luis A. Reyes

 Executive Director

 for Operations

Attachment: Semiannual Update of the Status of 

 New Reactor Licensing Activities

ADAMS ACCESSION NUMBER: ML041190618

OFFICE PM:RNRP SC:RNRP Tech Ed BC:IIPB 

NAME ACubbage LDudes (JSebrosky for) PKleene SRichards

DATE 05/10/04 05/10/04 05/05/04 05/17/04

OFFICE PMAS PD:RNRP D:DRIP RES

NAME MCase JLyons DMatthews (FGillespie for) JCraig

DATE 06/08/04 05/28/04 05/28/04 06/14/04

OFFICE OIP OCFO OGC D:NRR

NAME EBaker JFunches RWeisman JDyer-B. Sheron for:

DATE 06/07/04 06/09/04 06/17/04 06/18/04

OFFICE DEDRP EDO

NAME EMerschoff LAReyes

DATE 07/09/04 07/09/04

OFFICIAL RECORD COPY
Attachment 

Semiannual Update of the Status of 

New Reactor Licensing Activities

June 2004

INTRODUCTION

This attachment to the June 2004 update of the status of new reactor licensing activities

summarizes the status of the Advanced Plant 1000 (AP1000) design certification review, early

site permit (ESP) reviews, preapplication activities for other reactor plant designs, regulatory

infrastructure development, and stakeholder interactions.

DESIGN CERTIFICATION

Advanced Plant 1000 (AP1000)

On March 28, 2002, Westinghouse Electric Company (Westinghouse) submitted its application

for final design approval and standard design certification for the AP1000 design. The Nuclear

Regulatory Commission (NRC) staff issued its design certification review schedule on

July 12, 2002, establishing the milestones and target dates for the review.

On June 16, 2003, the AP1000 draft safety evaluation report (DSER) was issued with 174 open

items. All of the DSER open items are now considered to be resolved. No additional

information is needed by the Nuclear Regulatory Commission (NRC) staff to write the final

safety evaluation report (FSER). The staff remains on schedule to issue the FSER by

September 13, 2004. A letter to Westinghouse dated December 2, 2003, reaffirmed that the

FSER would be issued by September 13, 2004. By letter dated March 24, 2004, the staff

provided Westinghouse a reconfirmation of the AP1000 design certification review schedule.

The issuance date of the final design approval (FDA) was moved up by 6 weeks to

September 13, 2004, so that the FDA would be issued concurrent with the FSER. In addition,

it may be possible to complete the final design certification rule up to five months earlier than

the scheduled date of December 2005. These schedule improvements are contingent on timely

submittal of the final AP1000 design control document (DCD). The staff committed to provide

the final schedule for the remainder of the design certification process when the FSER is

issued. The milestones and their target due dates are listed below.

AP1000 Milestone Target Date

Issue initial requests for additional information (RAIs), not

 including security and safeguards issues

September 30, 2002 (complete)

Westinghouse response to the RAIs December 2, 2002 (complete)

Staff informs Westinghouse of RAI status February 28, 2003 (complete)

Issue draft safety evaluation report June 16, 2003 (complete)
- 2 -

AP1000 Milestone Target Date

Westinghouse to provide responses to all open items,

 and the NRC staff to determine whether these

 responses are acceptable

March 31, 2004 

(April 29, 2004 actual)

Westinghouse submits revision 11 of the AP1000 design

control document 

May 31, 2004 

(May 19, 2004 actual)

Issue final safety evaluation report September 13, 2004

Issue final design approval September 13, 2004

Complete design certification rulemaking December 2005

The staff is continually assessing resources needed to support the AP1000 design certification. 

Thus far, the resource estimates have been consistent with the estimates given in

SECY-01-0188, “Future Licensing and Inspection Readiness Assessment,” dated

October 12, 2001. 

Adjustments to the resource requirements are made using the staff’s planning, budgeting, and

performance management (PBPM) process, with consideration of other agency needs and

priorities. 

COMBINED LICENSE (COL)

Three consortia have submitted proposals to the Department of Energy (DOE) in response to a

November 2003 solicitation of interest for participation in new nuclear plant licensing

demonstration projects. The first consortium consists of Dominion Resources, Atomic Energy

of Canada, Limited (AECL) Technologies Inc., Hitachi America, and Bechtel. This consortium

submitted a proposal to DOE in mid-March 2004 based on the Advanced CANDU Reactor 700

(ACR-700) design at the North Anna site. A COL application for this facility may be submitted

as soon as mid-CY 2006. Based on the current information, the staff plans to budget resources

for the review of a possible COL application in FY 2007.

The second consortium consists of the Tennessee Valley Authority (TVA), Toshiba,

General Electric (GE), Bechtel, United States Enrichment Corporation (USEC), and Global

Nuclear Fuel-Americas LLC. This consortium submitted a proposal to DOE on April 23, 2004,

to conduct a detailed study of the potential construction of a two-unit GE-Toshiba-designed

Advanced Boiling Water Reactor (ABWR) nuclear plant at the Bellefonte site. On May 23,

2004, DOE announced that it will fund one-half of the cost associated with the study, which will

total $4.25 million over the next 10 months. Following completion of the study in April 2005,

TVA will make a decision whether to file a COL application.

The third consortium, known as NuStart Energy Development LLC, consists of nine energy

companies and two nuclear reactor vendors. The energy companies are Constellation

Generation Group, Duke Energy, Electricité de France (EDF) International North America,

Entergy Nuclear, Exelon Generation, Florida Power & Light Co., Southern Company, Progress

Energy, and TVA. The nuclear reactor vendors are Westinghouse and General Electric. The
- 3 -

consortium submitted a proposal to DOE on April 26, 2004, and is considering Westinghouse’s

AP1000 design and GE’s Economic and Simplified Boiling Water Reactor (ESBWR) design. 

The consortium plans to develop COL proposals for each design, and plans to select a site 12

to 14 months after the DOE award selections are announced. The Bellefonte site is one of

several sites being considered by the consortium. The current projection for a potential COL

application to the NRC is the 2008 timeframe, with regulatory approval of the COL desired late

in 2010 or 2011. 

EARLY SITE PERMITS (ESPs)

The staff received ESP applications in September and October 2003 from Dominion Nuclear

North Anna, LLC (Dominion), for the North Anna ESP site; Exelon Generation Company, LLC

(Exelon), for the Clinton ESP site; and System Energy Resources, Inc. (SERI), a subsidiary of

Entergy Corporation, for the Grand Gulf ESP site. All three applications were accepted for

docketing in late 2003, and the staff’s safety and environmental reviews of the applications are

in progress. The staff held environmental scoping meetings at all three sites. In addition, the

staff has conducted quality assurance inspections and environmental and safety audits at all

three sites as part of its review of the applications. 

In regard to the mandatory ESP hearings, Federal Register notices of hearing and opportunity

to petition for leave to intervene were published for all three applications. Subsequently,

petitions to intervene were received on all three applications. With respect to the Dominion

application, the petitioners are the Blue Ridge Environmental Defense League, the Nuclear

Information and Resource Service (NIRS), and Public Citizen. The same groups petitioned to

intervene in connection with the Exelon ESP application, along with two other groups, the

Environmental Law and Policy Center and the Nuclear Energy Information Service. Four

organizations petitioned to intervene in the SERI ESP application proceeding: the National

Association for the Advancement of Colored People, Claiborne County, Mississippi Branch;

NIRS; Public Citizen; and the Mississippi Chapter of the Sierra Club.

On March 22, 2004, the Chief Administrative Judge of the Atomic Safety and Licensing Board

Panel established an Atomic Safety and Licensing Board (ASLB) for each of the proceedings

(each with the same members). An initial prehearing conference has been scheduled by the

ASLB for June 21-23, 2004, for all three applications. Note that on March 2, 2004, the

Commission granted the applicants’ request that these proceedings be conducted in

accordance with the recently revised Rules of Practice (10 CFR Part 2).

Major remaining schedule milestones for the three applications are shown in the table below. In

all cases, the mandatory hearing and Commission decision processes are assumed to require a

total of 12 months after completion of the staff’s reviews.

ESP Milestone

Dominion

North Anna

Exelon

Clinton

SERI

Grand Gulf

Deadline for filing hearing

contentions

5/3/04 (complete) 5/3/04 (complete) 5/3/04 (complete)
- 4 -

ESP Milestone

Dominion

North Anna

Exelon

Clinton

SERI

Grand Gulf

Environmental requests for

 additional information (RAIs)

 sent to applicant

3/12/04

(complete)

5/14/04

(complete)

7/2/04

Safety/emergency planning

 RAIs sent to applicant

6/3/04

(complete)

7/27/04 9/21/04

Draft environmental impact

 statement (EIS) issued to

 EPA

10/14/04 12/27/04 2/25/05

Draft safety evaluation report

 (SER) issued

12/20/04 2/10/05 4/7/05

Final SER and EIS issued 6/16/05 8/25/05 10/21/05

PREAPPLICATION ACTIVITIES

Upcoming applications to certify the ACR-700 and ESBWR reactor designs will require

considerably more resources than the nearly completed AP1000 review. The three reactor

design certifications completed in the 1990s (ABWR, System 80+, and AP600) required

anywhere from about 65 to 120 full-time equivalent (FTE) per design for the staff’s technical

reviews over a period of about 9 years. About $5.5 million were expended on contractor efforts,

including testing required for the AP600 review. The AP1000 review has taken significant

advantage of the previous AP600 effort, with resource expenditures in line with the estimates

given in SECY-01-0188, which were 30 FTE and $1.5 million. Similar precedents do not exist

for the ACR-700 or ESBWR, so a higher level of effort (as compared to AP1000) is anticipated

for these reviews.

The most significant factor affecting resources and schedules for the ACR-700 and ESBWR

applications is the quality of the applicant’s submittal, and whether it is supported by sufficient

testing and by research and development where necessary. The level of effort required is also

affected by the nature and significance of the differences in these designs compared to

previously certified or licensed designs, the extent of the applicant’s test program, and whether

policy issues need to be addressed. Applications which do not adequately demonstrate how

the design meets regulatory requirements will take more time and effort to review.

The staff is taking steps early in the planning process to ensure that personnel with necessary

skills are available to review the applications when they are received, and to promote a

consistent understanding of schedules and priorities for these reviews. Planning for the

upcoming applications is emphasizing early and thorough identification of issues. Efforts also

include development of tools to maintain detailed status information on the hundreds of

questions and issues expected to arise in the course of the reviews. Tools are being developed

to assist technical staff to compose safety evaluation report input, that will result in a more

consistent document. These improvements will contribute to accomplishing the agency’s goal

of efficient, effective, realistic, and timely reviews.
- 5 -

Economic and Simplified Boiling-Water Reactor (ESBWR)

The General Electric ESBWR is a 1390 MWe reactor using natural circulation for normal

operation with passive safety features. This design is based on the certified advanced

boiling-water reactor (ABWR) and the simplified boiling-water reactor (SBWR) designs, which

the staff started to review in the early 1990s. On April 18, 2002, GE requested a

pre-application review of the ESBWR and proposed that the preapplication review be

conducted in two phases. During Phase 1, GE and the staff discussed the scope, schedule,

and resource estimate for the preapplication review. Phase 2 of the preapplication review is

currently underway and includes assessment of the technology basis for passive safety

systems and the analysis methodology for transients and accidents.

In support of the preapplication review, GE submitted topical reports regarding the ESBWR test

and analysis program description, the ESBWR test program, the SBWR test program, the

qualification of the TRACG computer code analysis method, the application of the TRACG code

to the ESBWR design, ESBWR scaling, and the ESBWR design description. The staff plans to

issue a safety evaluation addressing the application of the TRACG computer code for ESBWR

loss-of-coolant accident (LOCA) analyses in July 2004. GE plans to submit additional

information in mid-CY 2004 in support of the ESBWR preapplication review, including the

application of TRACG for ESBWR anticipated transients without scram (ATWS) and for

stability, and topical reports related to severe accident analyses and probabilistic risk

assessment (PRA) methods. It is anticipated that GE will submit a design certification

application for the ESBWR design in mid-CY 2005.

The staff continues to develop infrastructure to support a future design certification application. 

These activities include the development of methods, tools, and data to be used for

independent safety assessments or confirmatory calculations. The staff has initiated programs

to add modeling capabilities to NRC’s thermal-hydraulic system analysis code TRACE.

Modeling capabilities are important to assess the safety features of the ESBWR design,

including the tight coupling between the reactor vessel and the containment under certain

accident scenarios. The coupled TRACE and CONTAIN (NRC’s containment analysis code)

codes will be the primary analytic tools used to perform confirmatory analyses of the ESBWR

integral response to off-normal conditions. The staff is using the coupled TRACE/CONTAIN

code system for independent calculations in its TRACG code review. 

Advanced CANDU Reactor 700 (ACR-700)

The ACR-700 is an advanced CANDU (Canada Deuterium Uranium) design that utilizes

horizontal fuel channels passing through a heavy-water moderator tank. As with other CANDU

designs, the ACR-700 is designed to be refueled during power operation. Other features of the

reactor system, coolant pumps, U-tube steam generators, and pressurizer are similar to

pressurized water reactor (PWR) designs in the United States.

The ACR-700 will have features that make it significantly different from operating CANDU

reactors. The ACR-700 utilizes light water as coolant within the fuel channels, whereas

operating CANDU reactors utilize heavy water. The ACR-700 will be designed to have a

negative void reactivity coefficient so that if boiling occurs within the fuel channels, the reactor

power will decrease. The negative void coefficient for ACR-700 would be achieved by using
- 6 -

slightly enriched uranium in the fuel and neutron-absorbing dysprosium elements in the fuel

assemblies. Natural uranium fuel is used in operating CANDU reactors. The reactor core will

be smaller than operating CANDU reactors with fewer fuel channels.

AECL Technologies Inc. provided its suggested plan for NRC’s preapplication review of the

ACR-700 in a letter dated September 26, 2002, and an amended plan on December 18, 2002. 

The December 18, 2002, letter expanded the scope of the preapplication review as originally

identified in the September 26, 2002, plan. 

The ACR-700 preapplication review is being conducted in two phases. Phase 1 was a series of

familiarization meetings designed to provide the staff with a general overview of the ACR-700

design. Phase 1 of the ACR-700 preapplication review has been completed and the ACR-700

preapplication review is currently in Phase 2. The objective of Phase 2 is to provide more

information about the ACR-700 design to facilitate the staff’s review of 12 focus topics, and to

provide feedback to AECL Technologies Inc. prior to their application for standard design

certification.

In addition, given limited NRC resources, AECL Technologies Inc. has requested that priority be

given to the following key focus topics during Phase 2 of the preapplication review:

Focus Topic #1 - Class 1 pressure boundary design

Focus Topic #3 - Computer codes and validation adequacy

Focus Topic #8 - On-power fueling

Focus Topic #9 - Confirmation of negative void reactivity

The phase 2 technical meetings with AECL Technologies Inc. on the on-power fueling, Class 1

pressure boundary design and pressure tubes, PRA methodology, and design basis accident

and severe accidents have been conducted and requests for additional information (RAIs) have

been developed for these focus topics. Results of the preapplication review, including

responses to RAIs, will be documented in a report to AECL Technologies Inc. The report will

highlight significant issues or safety problems, based on the staff’s preliminary review of the

design, that could affect the ability of the ACR-700 to receive design certification. Completion

of the ACR-700 preapplication review is expected by September 30, 2004, in anticipation of

AECL Technologies Inc.’s plan to submit a design certification application on March 31, 2005,

and a possible COL application for the North Anna site in mid-CY 2006. 

The staff briefed the future plant subcommittee of the Advisory Committee for Reactor

Safeguards (ACRS) on January 13, 2004, to provide information on the preapplication review

scope for the ACR-700 design and to present a design overview of the ACR-700. The

presentation included the staff’s approach, plans, and schedule for the preapplication review of

the key focus topics.

The Office of Nuclear Regulatory Research (RES) led a series of Phenomena Identification and

Ranking Table (PIRT) panel meetings in February 2004, on ACR-700 thermal-hydraulics,

neutronics, and severe accident issues. The objective of the PIRT process is to identify

potentially important safety issues and to provide the technical basis for resolution of such

issues as they apply to the ACR-700 design. The PIRTs will form the basis for subsequent

activities and will be used to provide guidance to assess the adequacy of the experimental data

base, the code modeling, and the code validation and assessment.
- 7 -

In addition, RES is developing a comprehensive research plan to support the review of the

ACR-700 design certification application. The structure of the plan will be similar to that

presented in SECY-03-0059, “NRC’s Advanced Reactor Research Program.” The research

activities contained in this plan will be reviewed by the Advanced Reactor Steering Committee

and the Advanced Reactor Technical Advisory Group (TAG).

The staff met with representatives of the Canadian Nuclear Safety Commission (CNSC), on

March 2, 2004, to discuss the details of the ACR-700 quality assurance (QA) program

implementation review plan. QA review planning is part of ongoing international collaboration

with CNSC on the review of the ACR-700 design. The agenda included discussions on the

review objectives, the definitions of “safety-related” in Canada and in the U.S., systems and

processes controlled by the International Organization for Standardization (ISO), and a

discussion on the documents needed by CNSC and NRC prior to the QA review. Other topics

discussed included Part 21 (“Reporting of Defects and Noncompliance”), commercial-grade

dedication, and QA issues associated with research that is conducted in another country in

support of a design certification.

During the week of March 15-19, 2004, the staff met with CNSC to discuss significant

regulatory differences between the Canadian regulatory QA standard and the U.S.

requirements, and the approach for reviewing the ACR-700 QA program. The staff also met

with AECL Technologies Inc. to discuss the QA controls applied to the design and testing of the

ACR-700. Other activities included fact-finding visits to the Stern Laboratory, the Chalk River

National Laboratories, and the Sheridan Park AECL facilities in Ottawa, Canada.

The ongoing collaboration with CNSC is expected to facilitate the staff's review of the computer

codes and analysis tools used in the ACR-700 safety analysis and the adequacy of their

validation to support the upcoming ACR-700 design certification review. On April 23, 2004, the

staff participated in a meeting with CNSC to discuss the details of CNSC's independent analysis

to confirm the nature of the coolant void reactivity for the ACR-700 design. The CNSC and

NRC staff agreed to collaborate on future nuclear analysis efforts to evaluate the coolant void

reactivity and other inherent feedback characteristics of the ACR-700 core design.

International Reactor Innovative and Secure (IRIS)

On July 11, 2002, Westinghouse requested the preapplication review of the IRIS design. IRIS

is a 1000 MWt integral light-water reactor with all reactor coolant piping and heat transport

systems located inside the reactor vessel. The IRIS integral vessel is larger than a traditional

PWR pressure vessel, but the size of the IRIS containment is a fraction of the size of

corresponding loop reactors. 

The first in a series of technical meetings regarding preapplication review of the IRIS design

was held on April 15, 2004. Westinghouse described two objectives for the preapplication

review. The first objective is NRC agreement on the scope of the IRIS testing program, which

will be the focus of NRC discussions with Westinghouse over the next several months. The

meeting was focused on providing background and conducting introductory discussions with the

staff supporting this objective.
- 8 -

Westinghouse’s second preapplication review objective is NRC agreement on a procedure to 

approve the design with no emergency response planning requirements. Westinghouse noted

that this objective is consistent with Department of Energy objectives for Generation IV reactor

designs. The interest in eliminating emergency response requirements comes from

international partners in the IRIS project. The staff informed Westinghouse that certifying a

design with no emergency response requirements involves a Commission policy decision. 

Westinghouse plans to submit information addressing this topic in spring 2005. The staff will

keep the Commission informed as additional information on this topic is received, identifying

policy issues and proposed solutions as necessary.

Westinghouse presently plans to request certification of the IRIS design in early 2006. Minimal

resources have been budgeted for the preapplication review, however, the staff is not currently

budgeted to review the design certification application.

Framatome ANP Activities including SWR 1000

The SWR1000 is a 1253 MWe boiling-water reactor incorporating passive safety features. 

There have been no formal interactions between Framatome ANP and the NRC staff regarding

the SWR 1000 for the past several months. While Framatome ANP stated in an April 5, 2004,

letter that they were continuing to plan for submittal of materials supporting the SWR 1000

preapplication review, the company is presently focused on fulfilling its contract to build a

European Pressurized Reactor (EPR) at the Olkiluoto site in Finland. Framatome ANP is also

considering pursuing preapplication review of other reactor designs, such as the EPR or a

gas-cooled design.

Gas Turbine Modular Helium Reactor (GT-MHR)

The General Atomics Company (GA) GT-MHR design is an approximately 300 MWt helium

reactor design based on high-temperature gas-cooled reactor (HTGR) technology. The

GT-MHR design uses helium to cool the ceramic-coated fuel particles contained in fuel

compacts inserted in graphite fuel elements.

There have been no formal interactions between GA and the NRC staff regarding the GT-MHR

in the last 12 months. GA indicated to the staff that they planned to terminate GT-MHR

preapplication review interactions with the NRC and refocus their gas reactor design and

development effort on the Very High Temperature Reactor (VHTR) project. NRC expects to be

formally notified soon of GA’s decision to terminate GT-MHR preapplication interaction

activities.
- 9 -

Pebble Bed Modular Reactor (PBMR)

The PBMR is a 110 MWe modular high-temperature helium-cooled reactor. By letter dated

February 18, 2004, PBMR Pty., LTD, notified the NRC that it intends to apply for design

certification of the PBMR once the detailed design for a PBMR demonstration plant to be built in

South Africa is sufficiently completed. In order to make the staff’s design certification

application review process efficient and effective, PBMR Pty. has requested that discussions be

initiated with the NRC to plan the scope and content of a PBMR preapplication review. 

PBMR Pty. indicates that preapplication review planning could begin in the second quarter of

CY 2004. Staff technical reviews of selected topics could begin in early CY 2005. PBMR Pty.

envisions preapplication review activities lasting 1.5 to 2 years before a formal design

certification application would be filed. PBMR Pty. has indicated that the testing program for the

PBMR demonstration plant would be used to support PBMR design certification review

activities.

RES has informed PBMR Pty. that NRC resources are not currently available for infrastructure

development, or for reviewing HTGR reactor designs. Available resources are being allocated

to reactor designs that have the potential for being granted a COL in the near term. However,

RES in coordination with the Office of Nuclear Reactor Regulation (NRR) and other agency

offices, are planning to meet with PBMR Pty, and other industry representatives interested in

pursuing HTGR reactor designs, to discuss their initiatives and plans for submitting a COL

application. Following that meeting, the staff plans to inform the Commission on the prospects

of a COL application(s), and seek Commission guidance on whether to proceed or delay

planning for preapplication or design certification reviews of HTGR designs.

REGULATORY INFRASTRUCTURE

10 CFR Part 52 Update

The staff has proposed rulemaking to revise 10 CFR Part 52 based on lessons learned during

the previous design certification reviews and on discussions with nuclear industry

representatives about the ESP and COL review processes. The proposed rule was published

for comment in the Federal Register on July 3, 2003 (68 FR 40025). The staff received eight

comment letters on the proposed rule and is in the process of preparing a revised rule package. 

One of the comments relates to how the regulations identify the applicability of requirements in

10 CFR Part 50, and other parts, to the 10 CFR Part 52 licensing processes. In the proposed

rule, the staff attempted to address this applicability issue by adding Section 52.5 which

identifies 10 CFR Part 50 provisions that the staff believed were applicable to the licensing

processes in 10 CFR Part 52. Commenters have pointed out that some of these provisions are

not applicable to all of the 10 CFR Part 52 licensing processes. In addition, some subparts in

10 CFR Part 52 contain very broad applicability statements such as, “Applications filed under

this subpart [for design certifications] will be reviewed according to the applicable standards set

out in 10 CFR Part 50 and its appendices and 10 CFR Part 100 as they apply to applications for

construction permits for nuclear power plants.” After considering comments on this issue, the

staff concluded that, in order to ensure that all necessary changes related to the applicability of

requirements outside of 10 CFR Part 52 were captured accurately, it needed to do a top to

bottom review of 10 CFR to identify all provisions where conforming language needs to be
- 10 -

added to clarify applicability to 10 CFR Part 52. The staff believes that the resulting revised rule

package will provide needed clarification on the applicability of the relevant technical and

administrative requirements to each of the 10 CFR Part 52 licensing processes. However, the

staff was concerned that external stakeholders would not have the opportunity to review and

comment on the staff’s proposals if they were published as a final rule. Therefore, the staff has

recommended that the revised 10 Part 52 rulemaking package be published as a proposed

rule, again, to ensure all interested stakeholders have the opportunity to comment on the

applicability of technical and administrative requirements to each of the 10 CFR Part 52

licensing processes. Consequently, the staff expects to forward a proposed rule package to the

Commission later this year, rather than a final rule package. Additional resources required to

support this effort will be reviewed by the staff through implementation of the PBPM process.

Construction Inspection Program (CIP) Development

The CIP team, composed of staff from the NRR and each of the four regions, completed the

development of the construction inspection framework. The framework describes the

organization and philosophy for the inspection activities to be conducted during licensing and

construction of plants under 10 CFR Part 52. The CIP framework is the result of more than

6 months of interaction with internal and external stakeholders. The construction inspection

framework was issued on May 11, 2004, as NUREG-1789, “10 CFR Part 52 Construction

Inspection Framework.” Over the next 6 months, the team will develop three Inspection Manual

Chapters, the first tier of documents that will provide specific guidance for implementing the

framework of the CIP described in NUREG-1789. 

The CIP team continued its efforts to ensure staff readiness to conduct inspections by updating

the agency’s Strategic Workforce Planning (SWP) survey. The CIP team worked with the

Office of Human Resources (HR) to incorporate a series of knowledge and skill statements into

the SWP survey specifically related to construction. When implemented, the updated survey

will help to identify any construction experience, whether or not it was specifically related to the

construction of a commercial nuclear power plant. The information being gathered through the

survey will be used to identify any gaps in the construction knowledge and skills of inspectors

and to develop the scope of any training that may be needed to prepare the inspection staff to

conduct construction inspection activities. The team has already identified the need to provide

training to the technical staff who will be developing, revising, and updating the construction

inspection guidance. The training will ensure that individuals tasked with authoring construction

inspection guidance are well versed in the requirements of Part 52 licensing and the role and

importance of ensuring the successful completion of inspections, tests, analyses, and

acceptance criteria (ITAAC) by a licensee.

The team has initiated interactions with RES and the Office of Nuclear Materials Safety and

Safeguards (NMSS). The team is working with RES to develop guidance for selecting

construction inspection samples. When complete, the sampling selection methodology will be

incorporated into the framework NUREG as an appendix. Discussions with NMSS are aimed at

understanding the lessons learned from inspections associated with the preapplication review of

the proposed repository at Yucca Mountain in areas such as training and qualifying inspectors

and the transition from the preapplication to the application phase.
- 11 -

Other Federal agencies, including the U.S. Navy and the Federal Aviation Administration (FAA),

have been contacted with the specific purpose of learning from their implementation of similar

inspection activities. The team has asked the Navy for information about the oversight roles

and responsibilities of naval reactor inspectors during fabrication, especially when modular

construction is being used. Additionally, the team has requested an opportunity to observe

large scale fabrication of naval vessels where modular construction techniques are being used. 

Interactions with the FAA are focused on understanding the FAA’s surveillance and oversight of

the manufacture of airplane parts by foreign suppliers. The lessons learned from these

interactions will be used in the development of NRC procedures for the inspection of licensee

quality assurance and corrective action programs during construction.

The NRC continues to evaluate the need to expand its oversight of nuclear component

suppliers if new plant construction becomes a reality. Expanded oversight will be necessary to

deal with the nuclear component suppliers that are new to the industry, some of whom will be

foreign-based. It is also anticipated that new designs will use modular construction. The staff

believes that these new suppliers engaged in modular construction activities will need to be

inspected to ensure that the components are designed and fabricated to 10 CFR Part 50,

Appendix B QA requirements. The NRC staff continues to interact with the Nuclear

Procurement Issues Committee (NUPIC), which provides oversight of nuclear industry

suppliers, to discuss the Committee’s possible role in the oversight of foreign as well as

domestic suppliers to new nuclear plant construction.

Our regulatory counterparts in other countries continue to provide insights about construction

activities which the team evaluates for incorporation into the CIP. For example, before a recent

NRC staff visit to Taiwan, the team provided the staff with specific areas of interest, including

any experience and outcomes related to the selection of oversight individuals and scheduling of

oversight activities.

ESP Review Standard

As discussed in the previous update, the NRC staff has developed a review standard for ESP

applications to provide guidance to the NRC staff on the process and criteria for reviewing an

ESP application. The standard, designated Review Standard (RS) 002, “Processing

Applications for Early Site Permits,” consolidates existing guidance, updates the guidance to

reflect the ESP licensing process, and identifies the scope of the ESP review. The standard

also informs stakeholders regarding the information the staff expects to be provided in an ESP

application. 

The draft review standard was released for interim use and public comment in December 2002. 

Two additional sections, addressing QA and accident analysis, were released on April 11, 2003. 

The staff has responded to all comments received on the document and has placed the

comments and staff responses on the NRC’s new reactor licensing Web page.
- 12 -

The staff has also incorporated comment responses into the draft RS-002 as appropriate. In

accordance with Commission direction in the staff requirements memorandum (SRM) for

SECY-02-0199, “Denial of Petition for Rulemaking To Use Information From Prior Licensing

Actions as Resolved Information for Early Site Permit and Combined License Applications,” the

staff submitted RS-002 to the Commission in December 2003 via SECY-03-0227. On

March 15, 2004, the Commission issued the SRM for SECY-03-0227. The SRM approved

issuance of RS-002, subject to certain required changes, such as updating the document to

reflect the new 10 CFR Part 2. The document was revised in accordance with the SRM and

was issued on May 3, 2004.

The staff has also developed an interim staff guidance (ISG) process for review of ESP

applications, similar to the one already developed for license renewal. The ESP ISG process

will be used to develop and issue interim guidance as necessary during reviews of ESP

applications. The process should reduce the frequency of revisions to RS-002 while ensuring

timely guidance is available to the staff. As appropriate, approved ISGs would be captured in a

future revision to RS-002. On December 3, 2003, the staff requested by letter that

stakeholders comment on the draft ISG process instruction. The Nuclear Energy Institute (NEI)

provided a set of comments, which the staff considered in development of the final process

instruction. The process instruction was issued for staff and stakeholder use on April 20, 2004.

Combined License (COL) Issues

In the previous semiannual status update (SECY-04-0001), the staff informed the Commission

that it had begun interactions with stakeholders to resolve generic COL issues prior to the

receipt of the first COL application. A set of 23 generic issues has been identified for

discussion. Some of the generic issues will be resolved by the CIP framework document and

the 10 CFR Part 52 rulemaking. The status of interactions on some of the other COL generic

issues is discussed below.

Programmatic ITAAC

SECY 04-0032, “Programmatic Information Needed for Approval of a Combined License

Without Inspections, Tests, Analyses and Acceptance Criteria (ITAAC),” was issued on

February 26, 2004. The paper requested the Commission’s approval of a staff proposal

regarding the level of programmatic information needed for approval of a COL without ITAAC

for any particular program. This SECY paper responded to the Commission’s direction in a

September 11, 2002, SRM regarding SECY-02-0067, “Inspections, Tests, Analyses and

Acceptance Criteria (ITAAC) for Operational Programs (Programmatic ITAAC).” In

SECY-04-0032, the staff proposed to categorize 15 operational programs into five different

categories depending on whether or not programmatic ITAAC are associated with the program,

and what the basis was for having or not having ITAAC. Two of the categories have

programmatic ITAAC, and three of the categories do not. The paper also provided an example

of the information needed at the COL stage in order to issue a COL without ITAAC. The

example included final safety analysis report (FSAR)-level information and procedures directly

related to implementation of the fire protection program.
- 13 -

In a May 14, 2004, SRM, the Commission approved the staff’s proposed categorization of

programs, but disapproved the remainder of the staff’s proposal including the staff

recommendation concerning procedure-level information. In the SRM, the Commission

directed the staff to reflect the Commission’s decision on this issue in the Statements of

Consideration of the final Part 52 amended rule. The Commission also directed the staff to

complete its work on the information necessary for the COL application for each of the

programs for which the staff had previously assumed ITAAC would be necessary by

December 31, 2005, and present its results to the Commission.

Emergency Planning (EP) ITAAC

On January 29, 2004, the staff provided a letter to NEI with draft proposed emergency planning

ITAAC to initiate discussion with industry and public stakeholders. The NRC staff, in

consultation with the Department of Homeland Security/Federal Emergency Management

Agency (DHS/FEMA), crafted the EP ITAAC, which address 15 of the 16 EP planning

standards in 10 CFR 50.47(b) and NUREG-0654/FEMA-REP-1, Rev. 1, “Criteria for Preparation

and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of

Nuclear Power Plants.” (The excluded planning standard, in section 50.47(b)(13), pertains to

the development of general plans for recovery and reentry, and was determined not to be

applicable to EP ITAAC.) 

The staff issued a Federal Register notice on March 10, 2004 (69 FR 11464), seeking public

comment on the draft EP ITAAC and announcing an April 27, 2004, public workshop to discuss

the issue. The April 27, 2004, EP ITAAC workshop was transcribed and facilitated. The

workshop addressed four areas for which the staff requested feedback: (1) EP issues that

should be resolved before the COL is issued, (2) State and local government participation,

(3) concurrent EP ITAAC review, and (4) the nature of EP ITAAC for existing sites. NEI also

presented a proposal on how to risk-inform the development of EP ITAAC. The comment

period for the Federal Register notice ended on May 27, 2004. The staff is continuing to

discuss EP ITAAC with stakeholders. The main issues that have been identified are the role of

risk information in the development of EP ITAAC and whether it is appropriate to have ITAAC

associated with offsite actions. The staff expects to review the comments and to develop a

SECY paper during the summer of 2004 and is targeting providing the Commission with a

SECY paper in the fall of 2004. 

COL Application Guidance

In a January 12, 2004, letter, NEI stated that it has begun development of guidance for

preparing COL applications and requested the staff to provide feedback in five areas. NEI

expects to complete Revision 0 of NEI 04-01, “COL Application Guidance,” in December 2004. 

The staff provided feedback in the five areas that NEI requested during a public meeting on

January 29, 2004. The staff expects to meet with NEI this year to discuss the development of

NEI 04-01. In 2005, NEI expects to address NRC comments, incorporate additional COL

application and process guidance, and seek NRC endorsement of NEI 04-01, Revision 1. NEI

expects that the process for obtaining NRC endorsement of NEI 04-01 will be similar to that

which led to NRC’s endorsement of NEI 95-10, “Industry Guideline for Implementing License

Renewal,” in Regulatory Guide (RG) 1.188. NEI believes that such a schedule will support

preparation and submittal of a COL application in the 2006-2007 timeframe.
- 14 -

The staff has budgeted resources to support these COL generic review activities. The

resources may need to be revised depending on the timing and the scope of the work involved.

Advanced Reactor Steering Committee (ARSC)

The ARSC, a joint RES and NRR management team, continues to review activities associated

with advanced reactor preapplication reviews, design certification reviews, and advanced

reactor research infrastructure development. The steering committee, with the support of the

TAG reviews research activities necessary to support design certification reviews, including the

agency’s independent assessments of new reactor designs and the formulation of the technical

bases for applicable regulatory requirements. Recent ARSC meetings have focused on the

staff’s plans to address non-light-water reactor (non-LWR) policy issues and potential policy

issues related to the ACR-700 preapplication review. In addition, recent TAG meetings have

focused on independent accident analysis capability for the ACR-700.

Non-LWR Policy Issues

SECY-03-0047, “Policy Issues Related to Licensing Non-Light-Water Reactor Designs,” dated

March 28, 2003, identified seven policy issues for Commission consideration. Some of these

issues are also relevant to the licensing of future LWR designs. In an SRM dated

June 26, 2003, the Commission:

(1) approved the staff’s recommendation on four of the issues

• Issue 2: the development of a policy statement or description on

defense-in-depth for nuclear power plants

•

• Issue 4: the use of a probabilistic approach in the identification of events to be

considered in the design, for the safety classification of structures, systems, and

components (SSCs), and to replace the single failure criterion

•

• Issue 5: the use of scenario-specific source terms in licensing decisions

•

• Issue 7: no change to the emergency preparedness requirements

(2) requested additional information on two other issues

• Issue 1: staff should provide further details on the options for, and associated

impacts of, requiring that modular reactor designs account for the integrated risk

posed by multiple reactors

• Issue 6: staff should pursue the development of containment functional

performance standards and then submit options

(3) disapproved Issue 3, the staff’s recommendation to proactively participate in

development of and endorsing international codes and standards where such codes and

standards have been identified by applicants or preapplicants for use in their submittals

or by staff as needed to fill gaps on the NRC’s non-LWR infrastructure.
- 15 -

The staff is preparing a response to the Commission on the status of the two issues requiring

additional information. Further, as directed by the Commission, the staff is continuing its

interactions with external stakeholders on the issues identified in SECY-03-0047.

The SECY-03-0047 discussion on the integrated risk for modular reactors (i.e., the cumulative

effect on risk to the population around a site of adding a large number of small reactors to

produce power equivalent to the power of a large reactor) notes that the three ESP applications

describe the possible construction and operation of multiple small reactors at existing sites. 

Advanced Reactor Framework Development 

The staff is developing a framework for a regulatory structure for future plant licensing. 

Development of this framework was discussed in SECY-03-0059, “NRC’s Advanced Reactor

Research Program,” dated April 18, 2003. In addition, in SECY-03-0047, the staff noted that

the issues (except for the issue on International Codes and Standards) are being addressed

through this framework. The framework is intended to provide the technical basis for improving

the effectiveness and efficiency of future plant licensing in the longer term (beyond the

advanced designs currently in the preapplication stage). 

The framework is intended to be technology-neutral, risk-informed and performance-based. 

The framework anchors the regulatory structure to high level safety goals and provides the

guidelines and criteria to define the content of technology-neutral requirements. These

technology-neutral requirements would serve as the technical basis for formulation of

technology-neutral regulations. Another aspect of the framework is guidance for implementing

the technology-neutral framework in conjunction with the technology-neutral requirements on a

technology-specific basis. This guidance is intended to be used to generate technology-specific

RGs for specific candidate reactor designs. This will be accomplished by translating the high

level, technology-neutral regulations into technology-specific guidance.

High-Temperature Gas-Cooled Reactor (HTGR)

On October 30, 2003, a PIRT panel completed the identification of phenomena important to the

failure and fission product transport of coated fuel particles and fuel elements used in HTGRs. 

The panel identified and defined factors associated with design and manufacture, as well as

phenomena associated with plant operations and postulated accidents, including heatup,

reactivity insertion, and air and water ingress accidents. The final report is complete and was

sent to international HTGR fuel technology experts for comment. Publication of the PIRT report

as a NUREG/CR is expected by the end of this fiscal year.

The PIRT report can be used by the staff to (1) identify key attributes of gas-cooled reactor fuel

manufacture which may need regulatory oversight, (2) provide a reference for the review of

vendor fuel qualification test plans, (3) provide insights for TRISO fuel safety margin test plans,

(4) assist in defining data needed to develop HTGR fuel performance and fission product

transport models, (5) support the development of NRC's independent reactor fuel performance

code and fission product transport models, (6) support the development of NRC’s independent

models for mechanistic source term calculations, and (7) provide insights for the review of

applicant fuel safety analyses.
- 16 -

Additionally, a review and evaluation of the current American Society of Mechanical Engineers

(ASME) Code and Code cases for applicability to HTGR metallic components was completed

and published in NUREG/CR-6816 in June 2003. A review of existing information on metal

component behavior in HTGR environments was conducted and the results were published in

NUREG/CR-6824 in July 2003. An experimental apparatus was designed to supply simulated

HTGR coolant with controlled levels of impurities to creep and fatigue test chambers. This

experimental coolant loop was assembled, and initial creep tests of alloys Inconel 617 and

Incoloy 800H at temperatures between 750 °C and 950 °C (1380 °F to 1740 °F) were

completed on March 31, 2004. 

Staff and contractor effort also developed a draft materials specification standard for nucleargrade graphite. The draft was revised by the American Society for Testing Materials (ASTM)

committee, and there will be a ballot version for the June 2004 ASTM committee meeting. 

Efforts are also underway with ASME to develop a code that incorporates the best available

methods and design criteria for graphite. In February 2004, ASME formed a Special Working

Group to develop codes and standards for nuclear-grade graphite. However, the NRC’s budget

for this work was eliminated for 2004, and further results anticipated from the materials testing

program and improvements to the Code may not be available to assist the staff in evaluating

the integrity and failure modes of components used in HTGR designs by licensees.

Instrumentation and Control

A review of lessons learned from the design, construction, and licensing of advanced digital

instrumentation, control, and protections systems used in foreign evolutionary reactor designs

was completed in March 2004. In this review, the staff evaluated areas where revisions to the

current regulatory guidance may be needed. A research program for development of new risk

models for advanced digital instrumentation and control and reactor protection systems is the

next phase of this work. This research program would support possible inclusion of the new

models in risk-informed licensing. This research is intended to provide reliability and availability

models for advanced digital safety systems expected to be used in new reactor designs. Due

to resource limitations this research will not begin before FY 2005.

Structural and Seismic Accomplishments

The staff is sponsoring a research program at Brookhaven National Laboratory (BNL) to

develop a technical basis to support the safety evaluation of deeply embedded and/or buried

(DEB) structures proposed for advanced reactor designs (e.g., PBMR, GT-MHR). The overall

objective of this research is to investigate the applicability of existing seismic soil-structure

interaction (SSI) computer codes to DEB structures and to recommend any necessary

modifications to the computer codes. For the PBMR and GT-MHR new reactor designs

submitted to the NRC for preliminary review, the entire reactor building and a significant portion

of the gas turbine generator building will be partially or completely embedded below grade. SSI

effects and passive earth pressures for these types of deeply embedded structures will have a

significant influence on the predicted seismic response of the plant structure and components.

Research performed by foreign research and development (R&D) organizations and regulators

will also be reviewed for applicability and to determine gaps where additional research is

needed.
- 17 -

On February 16-20, 2004, meetings were held between NRC, its contractor, and several

Japanese organizations (i.e., Japan Nuclear Energy Safety (JNES), Nuclear Power Engineering

Corporation (NUPEC), Tokyo University, Tokyo Electric Power Corporation (TEPCO), Kajima

Corporation, and Tokyo Gas). The purpose of these meetings was to exchange information

and to gather Japanese data on the seismic performance of deeply embedded or buried

structures to develop the technical basis for the safety evaluation of deeply embedded/buried

structures that have been proposed for some advanced reactor designs. The NRC staff will

analyze information collected during the trip and follow up on a proposal to join an international

collaborative test program planned by JNES.

Organizational Structure

In March 2004, the Regulatory Effectiveness Assessment and Human Factors Branch in RES

was reorganized and renamed the Advanced Reactor and Regulatory Effectiveness Branch

(ARREB). RES staff involved in new reactors activities were consolidated into a New Reactors

Section in ARREB. This reorganization focuses RES’s new reactor efforts, and provides a point

of contact within RES for new reactor activities. ARREB coordinates research applicable to

current reactors and advanced reactors. It will lead the preapplication review for non-LWRs,

and will support NRR in the design certification review and the preapplication review of

advanced light-water reactor designs. 

ADDITIONAL STAKEHOLDER INTERACTIONS

International Cooperation

In addition to the international cooperation efforts with the CNSC as described in the ACR-700

section of this paper, the staff continues to seek opportunities to interact with and, where

appropriate, initiate cooperative programs with other agencies and organizations. Such

organizations include the Japan Atomic Energy Research Institute (JAERI), and the National

Nuclear Safety Administration (NNSA), the regulatory authority of the People’s Republic of

China, in the area of HTGR technology. 

The NRC has completed a revised cooperative research agreement with NNSA, designating

the Institute of Nuclear Energy Technology (INET) as the representative on HTGR Technology

exchange with NRC. The agreement with NNSA is scheduled to be signed in the summer of

2004, and includes the HTR-10 (high temperature gas-cooled test reactor) reactor safety

experiments. In addition, a bilateral meeting with NNSA is scheduled in July on the design

certification process.

NRC is also supporting meetings sponsored by the International Atomic Energy Agency (IAEA),

including various Coordinated Research Projects dealing with advances in fuel technology for

HTGR and the effects of near-field earthquakes on nuclear facilities. 
- 18 -

Department of Energy

On March 24, 2004, the staff met with representatives of DOE’s Office of Nuclear Energy,

Science, and Technology to discuss nuclear energy research, development, and demonstration

activities. During this meeting, DOE briefed the NRC staff on the “Department of

Energy/Nuclear Power Industry Strategic Plan for Light Water Reactor Research and

Development,” released in February 2004. DOE co-developed the strategic plan with the

nuclear power industry to establish a framework that would allow both parties to plan the

nuclear energy research and development agenda considered essential to achieving national

energy goals. The strategic plan addresses both the near-term deployment of new plants and

research and development for existing plants, such as advanced generation technologies,

technologies for long-term operation, security, and high-burnup fuel.

DOE also briefed the NRC staff on the status of its COL solicitation, several research focus

topics, and the status of their assessment of the constructability of new reactors. Regarding

COL solicitation, DOE has indicated that it expected to make additional award selections later

this year.