n, L
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NUREG-1635, Vol. 5

Review and Evaluation of the

Nuclear Regulatory Commission

Safety Research Program

A Report to the

U. S. Nuclear Regulatory Commission

Manuscript Completed: May2003

Date Published: June 2003

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

U. S. Nuclear Regulatory Commission

Washington, DC 20555-0001
TABLE OF CONTENTS

ABBREVIATIONS..................................................... v

EXECUTIVE SUMMARY .............................................. vi

1 INTRODUCTION ................................................ 1

1.1 Objective .................................................. 1

1.2 Background and Scope of RES Infrastructure Assessment Document ... 1

1.3 Scope and Structure of the ACRS Report ......... ................ 3

2 RES ADVANCED REACTOR RESEARCH INFRASTRUCTURE

ASSESSMENT DOCUMENT ...... . 5

2.1 Generic Regulatory Framework Development ............................. 5

2.2 Reactor Safety . ......................................... 7

2.2.1 Probabilistic Risk Assessment (PRA) ....... ............... 7

2.2.2 Instrumentation and Control ........... .................. 8

2.2.3 Human Factors Considerations ......... ................. 9

2.2.4 Thermal-Hydraulic Analysis .......... .................. 10

2.2.5 Neutronic Analysis ................................... 14

2.2.6 Severe Accident and Source Term Analysis ...... .......... 17

2.3 Fuel Analysis ........................................... 19

2.4 Materials Analysis ......................................... 20

2.5 Structural Analysis ......................................... 22

2.6 Consequence Analysis ....................................... 25

3 PHENOMENA IDENTIFICATION AND RANKING TABLE (PIRT)

PROCESS AND IMPLEMENTATION PLAN .......................... 27

4 IMPACT OF ADVANCED COMPUTER CAPABILITIES ON NRC

STAFF ACTIVITIES .29

5 OVERALL CONCLUSIONS AND RECOMMENDATIONS .31

6 REFERENCES .35

APPENDIX: DOE NEAR-TERM DEPLOYMENT AND GEN IV

REACTORS. .................................................. 37

... NUREG-1 635
ABBREVIATIONS

ACRS

ABWR

ACNW

ACR-700

ADS

ALWR

AP600

AP1000

APEX

APWR +

APR-1400

ASME

ATHEANA

ATWS

CAMP

CAREM

CDF

CFD

CONTAIN

COTS

CSAU

DBA

DOE

EPR

EPRI

EPZ

ESBWR

ESP

FLIRA

FLUENT

GFR

GRSAC

GT-MHR

HC-BWR

HTGR

IMR

I&C

IRIS

ISFSI

LERF

LFR

LWR

Advisory Committee on Reactor Safeguards

Advanced Boiling Water Reactor

Advisory Committee on Nuclear Waste

Advanced CANDU Reactor 700

automatic depressurization system

advanced light water reactor

Advanced Passive Reactor 600

Advanced Passive Reactor 1000

Advanced Plant Experiment

Advanced Pressurized Water Reactor Plus

Advanced Power Reactor 1400

American Society of Mechanical Engineers

A Technique for Human Event Analysis

anticipated transient without scram

Code Applications and Maintenance Program

Central Argentina de Elementos Modulares

core damage frequency

computational fluid dynamics

Containment Transient Analysis Program

commercial off the shelf

code, scaling, applicability, and uncertainty (methodology)

design basis accident

Department of Energy

European Pressurized Water Reactor

Electric Power Research Institute

emergency planning zone

European Simplified Boiling Water Reactor

early site permit

Future Licensing and Inspection Readiness Assessment

A commercial computational fluid dynamics code

Gas-Cooled Fast Reactor

Graphite Reactor Severe Accident Code

Gas Turbine Modular High Temperature Reactor

High-Conversion Boiling Water Reactor

high-temperature gas-cooled reactor

International Modular Reactor

instrumentation and control

International Reactor Innovative and Safe

independent spent fuel storage installation

large early release frequency

Lead-Cooled Fast Reactor

light water reactor

v NUREG-1 635
MAAP

MACCS

MACCS2

MASCA

MELCOR

MSR

NEI

NERAC

NRC

NRR

OECD

OSU

PARCS

PBMR

PC

PHEBUS-FP

PIRT

Plus

PRA

PRISM

RASPLAV

RELAP5

RES

RI-ISI

SCWR

SFR

SMART

SRM

SSHAC

SWR-1 000

THATCH

TRAC

TRAC-M

VHP

VHTR

Modular Accident Analysis Program

MELCOR Accident Consequence Code System

MELCOR Accident Consequence Code System 2

Follow-on program for RASPLAV

Melting of Core program

Molten Salt Reactor System

Nuclear Energy Institute

Nuclear Energy Research Advisory Committee

Nuclear Regulatory Commission

Office of Nuclear Reactor Regulation

Organization for Economic Cooperation and Development

Oregon State University

Purdue Advanced Reactor Core Simulator

Pebble Bed Modular Reactor

personal computer

international severe accident fission product research

program

Phenomena Identification and Ranking Table

Process Inherent Ultimately Safe

probabilistic risk assessment

Power Reactor Innovative Small Module

OECD experimental program for severe accident research

Reactor Excursion and Leak Analysis Program

Office of Nuclear Regulatory Research

Risk-Informed Inservice inspection

Super-critical-Water-Cooled Reactor System

Sodium-Cooled Advanced Reactor

System-Integrated Modular Advanced Reactor

staff requirements memorandum

Senior Seismic Hazard Analysis Committee

Siedwasser Reactor-1000

HTGR thermal neutronics modeling code

Transient Reactor Analysis Code

Transient Reactor Analysis Code Modernized

vessel head penetration

Very High Temperature Reactor System

NUREG-1 635 vi
EXECUTIVE SUMMARY

The objective of this report by the Advisory Committee for Reactor Safeguards (ACRS)

is to provide comments and advice on the Office of Nuclear Regulatory Research (RES)

document 'Advanced Reactor Research Infrastructure Assessment" (Reference 1),

which identifies gaps in technology that need to be filled prior to the potential

certification of a number of new reactor designs. These designs range from those

which have evolved from the current boiling and pressurized water reactor designs, i.e.,

advanced light water reactor (ALWR) designs, to those using significantly different

technology, e.g., high-temperature gas-cooled reactors (HTGRs). The range of designs

addressed in this assessment are representative of the designs vendors have

expressed an interest in submitting to NRC for certification. We note that the U.S.

nuclear industry has not constructed a new reactor for over two decades and has

focused on the operational and services aspects of the existing reactor fleet. As a

result, there are significant regulatory, management, and technical challenges to be

faced with the certification, construction, and deployment of new reactor designs,

especially those based on significantly different technology.

This ACRS report concentrates on reactor safety issues. It does not cover items

relating to nuclear materials and waste safety or to safeguards. Safeguards will be

addressed in a separate ACRS activity. We have not commented on the several

management issues that are discussed in the RES Infrastructure Assessment but

recognize that these issues are important.

Our overall conclusion is that the RES document comprehensively identifies the gaps in

the knowledge that need to be addressed in the review of the ALWR and HTGR

designs. We recognize that the applicants are responsible for providing most of this

information, but agree with the staff that the NRC should develop its own long-term

capabilities to evaluate the applicants' technical arguments that relate to regulatory

concems. In general, the RES document does not specifically distinguish between

information that should be developed by the applicant and information to be developed

by the NRC. This will be defined upon the completion of the staff's planned

Phenomena Identification and Ranking Table (PIRT) activities.

We generally agree with the identification of potential long-term RES projects that need

to be considered in the areas of probabilistic risk assessment (PRA), instrumentation

and control, materials analysis, structural analysis, and consequence analysis.

We have provided more specific advice as to the direction of the NRC staff work in the

following areas:

vii vii ~~~~NUREG-1635
I

Generic Regulatory Framework A risk-informed framework should prove useful for the

certification of the plants likely to be first deployed (i.e., ALWRs') and will likely be

mandatory for the non-LWR concepts on a longer timeframe. Thus, we recommend

that this effort be continued with a focus on addressing our concerns about the current

Option 3 framework, which is likely to serve as the basis for the generic regulatory

framework for risk-informed certification and licensing of future plant designs.

Human Factors: We believe that the topic of control room staffing will be important for

new reactors and the NRC will need to develop defensible methods for assessing the

adequacy of staffing levels.

T77ermal-Hydraulics: The TRAC-M code plays a central role in the plans for addressing

thermal-hydraulic questions for advanced reactors. There is urgency, therefore, to

complete its development and validation. Because PRA success criteria for passive

designs are sensitive to uncertainties in thermal-hydraulic code predictions, these

model uncertainties need to be quantified.

Neutronic Analysis: It is important that the NRC continue to be able to do independent

nuclear safety assessments. This will require continued support of the improvement of

its neutronic analysis capabilities. In particular, the effort to couple the thermalhydraulics code TRAC-M with the three-dimensional neutronics code Purdue Advanced

Reactor Core Simulator (PARCS) is important for the independent assessment of

nuclear safety for reactors like the Advanced Passive Reactor 1000 (AP1000) and

European Simplified Boiling Water Reactor (ESBWR) that rely on passive safety

systems. Modification of analysis methods to account for the unique features of the

Advanced CANDU Reactor (ACR-700) should be initiated now to facilitate timely

addressing of certification review issues. The NRC staff should provide explicit

direction as to the data and data quality that will be needed to support and validate

neutronic analyses for advanced reactor designs.

Severe Accident Progression and Source Term: Issues for reactor designs that are

based on LWR technology, such as the ESBWR and AP1000, can likely be addressed

through modest modifications of existing codes, but issues related to higher bumup

fuels will have to be addressed. Some of the new reactor designs incorporate cavity

flooding, which is intended to ensure that the debris produced by a severe accident is

retained in the pressure vessel. Significant effort will be needed to assess the

effectiveness of this design feature. Consequently, the NRC should continue its

involvement in the MASCA project.

It appears that a substantial effort will be required to configure NRC's accident analysis

models to treat the unique core and coolant configuration of the ACR-700. The NRC

'The term ALWRs in this report includes advanced CANDU designs that utilize a light

water coolant.

NUREG-1 635 viii
staff should assess the needed modification of its codes and identify experimental data

needed to validate predictions of accident progression and fission product release for

the ACR-700.

The NRC staffs experience with accident analysis of gas-cooled reactors with graphite

is limited. Existing NRC codes will require substantial modification to be applicable to

the analysis of these types of reactors. It will take substantial time to develop adequate

tools. For this reason there is a need to identify the available and needed experimental

data and to begin code model development.

Fuels Increased bumup is the primary issue to be addressed with new reactor designs

(e.g. AP1000 and ESBWR) where the fuel is similar to the fuel used in currently

operating reactors. This issue is being addressed for current reactors and needs to be

given a high priority. With regard to the ACR-700 design, it is likely that the NRC can

use existing codes and databases to evaluate the fuel using data provided by the

vendors, but this needs to be confirmed. Coated-particle fuels are the focus of much of

the staffs attention in its Infrastructure Assessment, since there is little NRC experience

with these fuels. Because the design certification review for the Gas Turbine Modular

High Temperature Reactor (GTMHR) is projected to start in early 2006, we believe that

fuel modeling activities (including fission product release) that are anticipated to take

substantial time to complete should be initiated soon. In terms of priority, however, due

attention should be given to the fact that a fuel design has not yet been established in

detail.

Computer Capabilities: Although this topic is not discussed in the Infrastructure

Assessment, we recommend that the computer capabilities within RES be reassessed

in the light of the increased computational speeds now available. Faster computers

would increase efficiency and effectiveness in completing numerous tasks identified in

the assessment and in interacting with the applicant who will likely be using these

capabilities.

ix NUREG-1 635
1 INTRODUCTION

1.1 Objective

The objective of this report is to provide Advisory Committee on Reactor Safeguards

(ACRS) comments and advice to the Commission, Executive Director for Operations

(EDO), and Office of Nuclear Regulatory Research (RES) staff on the "Advanced

Reactor Research, Infrastructure Assessment' (Reference 1). This assessment

identifies gaps in technology that need to be filled before new reactor designs concepts

can be certified and deployed in the United States. Timely planning of RES activities is

needed to enable the NRC to meet its obligations for certification of these new reactor

designs.

1.2 Background and Scope of RES Infrastructure Assessment Document

In February 2001, the Commission issued a staff requirements memorandum (SRM)

entitled *Staff Readiness for New Nuclear Plant Construction and the Pebble Bed

Modular Reactor" (Reference 2), with the specific direction to assess the NRC staff's

technical, licensing, and inspection capabilities and to identify enhancements, if any,

that would be necessary to ensure that the agency can effectively carry out its

responsibilities associated with an early site permit application (ESP), a license

application, and the construction of a new nuclear power plant. In a partial response to

this SRM, the staff issued an information paper in October 2001, "Future Licensing and

Inspection Readiness Assessment (FLIRA)" (Reference 3), in which the staff committed

to develop an advanced reactor research plan that would be used to develop and guide

a comprehensive advanced reactor research program, addressing the regulatory

framework and technical challenges that might be encountered with new license

applications. As a first step in developing the program plan (in terms of objectives,

priorities, milestones, etc.), RES prepared the Infrastructure Assessment document and

its attachments to identify the gaps in knowledge. That is the assessment that the

ACRS is reviewing in this report.

In 2001 and 2002, a series of actions by the nuclear industry and the Department of

Energy (DOE) influenced the content and timing of the research that is discussed in the

Infrastructure Assessment document.

Namely:

DOE and the Nuclear Energy Institute (NEI) conducted a wide-ranging review of the

infrastructure that would be required in order to significantly increase the Nation's

electrical supply by 2020 with 30% of that supply originating from nonemissions

sources. This review concluded that, if these objectives were to become reality,

there would be a need for a significant increase in nuclear generating capacity that

could not be achieved solely by license extension and increasing the power output

1 NUREG-1635
and capacity factor of the existing LWR plants. Thus, there would be a need for

the construction of new reactors. Subsequently, DOE developed a framework that

addressed a large number of potential reactor designs, together with an

assessment of the research and development required before their deployment

(Reference 4). This review is summarized in the appendix of this report. The

summary gives a broader context to our comments on the RES Infrastructure

Assessment document.

A request for design certification for the Advanced Pressurized Water Reactor

(AP1 000) has been submitted and a review is currently under way. Pre-application

reviews have been requested for the Gas Turbine Modular High-Temperature

Reactor (GT-MHR), Advanced CANDU Reactor (ACR-700), and European

Simplified Boiling Water Reactor (ESBWR) designs, and these reviews are also

currently in progress. A similar request is expected imminently for the International

Reactor Innovative and Secure (IRIS) design. Although the request for the Pebble

Bed Modular Reactor (PBMR) design pre-application review was withdrawn by

Exelon in 2002, the Office of Nuclear Reactor Regulation (NRR) anticipates that

Exelon will resubmit this request and that this review will be restarted in late 2003.

All of these reactor design concepts, which are deemed deployable by the DOE in

the 2010-2020 time period, are within the scope of the RES Infrastructure

Assessment document. Such a deployment schedule would place a burden on RES

to conduct a significant amount of research in the 2003-2006 timeframe to support

the projected NRR design certification activities.

* In October 2002 the NRC published the draft Infrastructure Assessment document.

The document and its supporting appendices focus on identifying gaps in

knowledge that need to be addressed within the period FY2002-FY2010. The

document discusses the technical challenges of the advanced light water reactor

(ALWR) and high temperature gas reactor (HTGR) designs because the NRC needs

to be in the position of an informed reviewer of the detailed reactor designs once

they are submitted for certification. This RES document incorporates the comments

in our letter to the EDO, dated July 18, 2002 (Reference 5), on the draft Advanced

Reactor Research Plan (Reference 6), as well as inputs received during various

industry, ACRS, DOE, and NRC workshops and interactions with European and

Asian organizations with operating experience on HTGRs.

NUREG-16352 2
1.3 Scope and Structure of the ACRS Report

The Infrastructure Assessment document was organized to cover the three strategic

arenas of reactor safety, nuclear materials safety and waste safety, and safeguards.

The scope of our report is confined to the reactor safety arena and the development of

a generic regulatory framework. The nuclear materials and waste safety issues are

within the purview of the Advisory Committee on Nuclear Waste (ACNW). The topic of

safeguards will be addressed in another ACRS activity. Although of vital importance to

the success of RES work in support of the licensing of advanced reactors, the various

management issues raised in the Infrastructure Assessment document are not

addressed in this report; examples of such issues are the long-term availability and

sustainability of funding and experienced manpower not only within the NRC but also in

the industry as a whole, since the stated RES strategy depends critically on cooperative

research with outside organizations.

The Infrastructure Assessment document was divided into sections on the Generic

Regulatory Framework development, Reactor Safety subsections covering the analyses

of accidents, reactor systems, fuels, materials, structures, and offsite consequence.

The format of our report generally follows that of the RES document. Included in our

report is an additional section on the impact of advances in computer technology on

computational capability. An appendix is also attached addressing GEN IV reactor

concepts.

3 NUREG-1 635
2 RES ADVANCED REACTOR RESEARCH INFRASTRUCTURE

ASSESSMENT DOCUMENT

This section discusses the generic regulatory framework development and reactor

safety issues (accident and reactor systems analyses), and fuel, materials, structural

and consequence analyses.

2.1 Generic Regulatory Framework Development

The Infrastructure Assessment document discusses general research needs for a

"Generic Regulatory Framework Development.* Development of such a framework is

justified because the current regulatory structure is highly biased toward current LWRs

and has limited applicability to non-LWR designs, was developed without benefit of

current PRA technology, and is inconsistent in some areas. The framework would

provide a process and criteria that can be used to develop technology-neutral

regulations and reactor-specific guidance documents.

The framework is expected to possess the following attributes: (1) a hierarchical

structure with the top-level goal being to protect public health and safety, (2) a strategy

that establishes a balance between the various parameters in a defense-in-depth

philosophy; and (3) quantitative safety guidelines based on the Commission's safety

goals.

The framework is also expected to be able to meet the objectives of the Commission's

Advanced Reactor Policy Statement: (1) '...as a minimum, advanced reactors will be

required to provide the same level of protection to the public that is required for current

operating LWRs,' and (2) 'it is expected that enhanced margins of safety and

simplified, inherent, passive, or other innovative means to accomplish their safety

functions will be utilized.'

The starting point is expected to be the risk-informed Option-3 framework, which uses a

hierarchical approach that utilizes strategies that establish a balance between

prevention and mitigation. This balance involves limiting initiating event frequencies,

core damage probabilities, radionuclide releases, and public health effects. These are

intended to be the lines of defense in the defense-in-depth philosophy and, indeed,

constitute an initial working definition of defense in depth. Quantitative guidelines for

these limits would not appear in the regulations but, instead, would be in design-specific

guidance documents. To develop the limits on each of the four lines of defense, the

prompt fatality safety goal would be utilized as the top-level criterion.

5 NUREG-1 635
Comments and Recommendations on Regulatory Framework Needs

1. The Option 3 and proposed generic regulatory frameworks are still conceptual. We

believe it is important that the development of this framework be given sufficient

priority so that it is in place before certification activities for non-LWR advanced

reactors are begun.

2. A technology-neutral regulatory structure must address all of the agency's regulatory

objectives. The prompt fatality safety goal is just too limited to serve this purpose.

There is a need to define all of the agency's regulatory objectives at all levels of

frequency and consequences in a risk context and to place quantitative acceptance

guidelines on these parameters.

3. The Option 3 conceptual framework definition of defense in depth, along with the

strategy guidance, provides a good starting point for developing a definition that has

an objective quantitative basis. It still appears, however, that this will be inadequate

for certification of advanced reactors, for the following reasons:

* The Option 3 framework only addresses the prompt-fatality safety goal as the

regulatory objective (see item 2 above).

* The Option 3 framework is still LWR biased to some extent in that it deals with

'core damage probability and containment failure probability." There is a need

for a broader definition of these concepts in terms of radionuclide releases.

* There is a need to tie the 'balance' between the various prevention and

mitigation actions to the actual uncertainties associated with the assessed risk

contributions at the different lines of defense. At high confidence levels, some

lines of defense may not be appropriate.

* The frequency 'bins' in the Option 3 framework appear to be much too broad.

There is a need for acceptance criteria that are continuous at all levels of

frequency (see item 2 above).

* Certain elements of defense in depth should be independent of the results of

PRA risk assessment. These elements need to be identified and appropriate

acceptance criteria developed.

* There is likely to be a desire to develop deterministic regulations using the

design basis accident concepts. There is a need to determine if these

regulations are necessary and, if so, how they can be developed to be consistent

with top-level risk acceptance criteria.

NUREG-1 635 6
2.2 Reactor Safety

This section discusses probabilistic risk assessment, instrumentation and control,

human factors considerations, thermal-hydraulic analysis, neutronic analysis, and

severe accident and source term analysis.

2.2.1 Probabilistic Risk Assessment (PRA)

RES anticipates that future applicants will rely heavily on probabilistic arguments to

support their safety analyses for advanced reactors. These reactors may have passive

safety systems, new fuel designs, and primary coolants other than light water.

Various features of advanced reactors challenge the capabilities of current state-of-theart PRAs. For example, the reliability of passive safety systems whose performance will

exhibit slow evolutionary behavior over the course of postulated accidents or

intermediate failure states needs to be quantified. Advanced reactors will use digital

instrumentation and control (I&C) systems whose failure modes and/or responses to

low-voltage direct-current spikes, fires, or loss of cabinet cooling have not been

modeled in current PRAs. In addition, the operator's role and staffing levels in some

advanced reactors will likely be different than in current reactors. Modeling of human

performance in PRAs for advanced reactors with postulated slowly evolving or longterm accidents or with multiple reactor modules using shared control rooms will be new

areas that need to be addressed. Risk metrics and success criteria used in current

state-of-the-art PRAs, i.e., core damage frequency (CDF) and large early release

frequency (LERF), may require modification to better represent safety goals for

advanced reactors.

RES is proposing work to develop the methods, expertise, and technical basis that will

be needed to support staff review of a PRA submitted as part of a design certification

application for advanced reactors such as an HTGR or IRIS.

Comments and Recommendations on PR4 Needs

1. The proposed research is comprehensive and appropriate.

2. Limitation of the proposed research to design certification applications for an HTGR

or an IRIS reactor design seems unnecessarily narrow, given the broad range of

concepts currently under consideration for possible certification.

3. The research efforts proposed in the areas of system modeling, human reliability

analysis, internal flood/fire/seismic events, multiple modules, and risk metrics are of

particular importance in that these efforts address new issues raised by the

advanced concepts.

7 NUREG-1 635
4. Additional work is needed to develop formal methods to quantify epistemic

uncertainties in the thermal-hydraulic code predictions of success criteria.

2.2.2 Instrumentation and Control

The Infrastructure Assessment document notes some characteristics of the current

state of the art and the marketplace for nuclear l&C:

* No new nuclear plant orders have been placed for over 20 years, and development

of domestic engineering experience in nuclear l&C has slowed. Domestic l&C

vendors have focused their development effort on the chemical and petrochemical

markets.

* The extension of operating licenses for the current fleet of commercial nuclear

power plants is creating a demand for replacement of aging and obsolete I&C

components and systems, and will continue to do so even if no advanced reactors

are ever built.

* New orders for l&C systems and components, whether for replacements in existing

facilities or for the construction of advanced reactors, are likely to come from

commercial-off-the-shelf" (COTS) sources. While the safety standards for COTS

equipment and designs are good, sometimes COTS components do not specifically

meet nuclear standards.

* Development in l&C concepts appears to have advanced more in Organization for

Economic Cooperation and Development (OECD) countries than in the United

States.

* Advanced reactors are likely to incorporate levels of automation far beyond the

design concepts of the current generation of reactors. The capabilities inherent to

digital l&C lend themselves to such development.

The current NRC research plan for digital l&C for FY2001 -FY2004 is described in

SECY-01-0155 (Reference 7). Most of the work in the current plan is applicable to both

digital l&C system upgrades of operating power plants and advanced reactor designs.

The major areas of research identified in SECY-01-0155 are:

* Systems aspects of digital l&C systems

* Software quality assurance

* Risk assessment and digital I&C systems

* Emerging l&C technology and applications

NUREG-1635 8
Comments and Recommendations on to I&C

1. The analyses of l&C issues presented in the Infrastructure Assessment document

and SECY- 01-0155 are comprehensive, and the staff understands the marketing

and technical environment in which regulatory tasks must be accomplished. The

planned research leverages on a wide range of non-NRC research projects.

2. Previous ACRS conclusions and recommendations identified in past ACRS research

reports (NUREG-1635, Vol. 1 [Reference 8], and NUREG-1635, Vol. 4 [Reference

9] are still valid and should continue to be considered by the staff.

3. Regardless of the timing of the emergence of advanced reactors, the research

identified by RES will need to be performed to support replacement of l&C

components and systems in the existing fleet of nuclear reactors and to maintain a

knowledge base within the staff.

4. Work is needed to address the implementation of defense in depth and safety

margin in the digital I&C systems.

5. In addition to safety system actuation, digital I&C systems will be used for

performing diagnostic evaluations, online monitoring, and performance assessment

of equipment. A better understanding of the integration of these functions will be

needed.

6. Lessons learned from the design and operation of recently built nuclear reactors

overseas, which utilize digital l&C systems, and digital upgrades in the U.S. LWRs

should be considered for incorporation into the review of evolutionary and advanced

reactor designs.

2.2.3 Human Factors Considerations

Many of the advanced reactor concepts envision the operators having quite different

roles than they do today. Plant operations will be highly automated and emergency

operating procedures will be computer based. Furthermore, shutdowns of plants for

refueling will be much less frequent and consequently opportunities for maintenance

and testing of equipment within the reactor containment will be less frequent. Many of

the plants will have passive designs that are intended to eliminate the need for operator

intervention during upset conditions.

Regardless of these changes, the issues of human factors in the safety of advanced

nuclear power plants are largely the same as for current power plants, although the

emphasis among the various issues may be different. For advanced plants with

passive safety systems, human errors of commission may be of more concern than are

human errors of omission. As another example, reduced maintenance should reduce

the opportunities for latent errors, but reduced familiarity with maintenance procedures

9 NUREG-1 635
 I ------ i-

(due to the infrequent maintenance schedule) could lead to an increased risk of

introducing latent errors.

By and large, the staff feels that its current approach for the development of human

factors engineering will be adequate for advanced reactors. That is, investigations of

the roles of humans in advanced plants have the same intellectual foundation current

NRC human factors research. Consequently, specialized human factors research

programs peculiar to advanced reactors are not readily identified. Continuation of the

NRC staffs current efforts to stay abreast of developments in this technical community

seems to be the appropriate approach.

However, plant staffing is one human factors issue that the NRC will need to address

specifically for advanced nuclear power plants. Current requirements for staffing of

control rooms of nuclear power plants have been developed based on a combination of

tradition and experience. Advanced reactor vendors will propose reduced control room

staffing because of the passive and/or automated features of the designs. This issue

will arise in connection with many advanced designs and even with some evolutionary

designs. The NRC must have a firmly established technical basis NRC for judging the

plant staffing requirements both during normal operations and upset conditions.

Comments and Recommendations on Human Factors Considerations

1. Continuation of the NRC staffs current efforts to stay abreast of developments in

the human factors arena is appropriate.

2. Because it is likely that proposed staffing levels for some of the new advanced

reactors may be significantly lower than those for the current LWR fleet, RES should

specifically address the issue of plant staffing. The NRC will need to develop

methods that have a defensible basis for assessing the adequacy of staffing levels.

2.2.4 Thermal-Hydraulic Analysis

The Infrastructure Assessment document provides a comprehensive discussion of the

research needed for HTGRs and also identifies needed research associated with four

water-cooled reactors (AP1000, IRIS, ESBWR, and ACR-700). Research that is under

way or planned for these reactor designs is described.

High-Temperature Gas-Cooled Reactors (HTGRs)

In general, the single-phase gas flow of the HTGR under normal operation is simpler

to analyze than the two-phase flow of the LWRs. A smaller base of data and

correlations will be required and it should be straightforward to incorporate this

information into the TRAC-M code. In the near-term, the existing HTGR accident

analysis codes GRSAC and THATCH can be used. Existing computational fluid

NUREG-1635 10
dynamics (CFD) codes such as FLUENT can be used for multidimensional analysis

where required. The analysis of accidents involving air and/or water ingress will

require multicomponent analysis capability. This capability can eventually be

incorporated into TRAC-M. The staff has identified a need to perform threedimensional analysis of flow through porous and solid structures and to develop the

appropriate constitutive relationships for fuel of a particular composition and shape.

Specifically, the uncertainties associated with the fluid and material properties and

the correlations for the transport processes will need to be quantified. This is a

significant task and the staff-identified need to assess the completeness of the

available database for gas flows and transfer processes is just a start.

The staff also plans to develop a code with two different working fluids to support

the analysis of moisture-ingression accidents. Such a task will require major

resources because there are insufficient analytical capabilities and databases to

evaluate moisture penetration into fuel and graphite, the subsequent chemical

reaction and heat transfer effects, and the consequent material degradation effects.

Existing codes, however, appear likely to be adequate for predicting the cooling of

intact fuel. The staff should evaluate whether these codes are sufficient to predict

behavior should the fuel fragment or undergo significant changes in geometrical

properties.

Advanced Light Water Reactors (ALWRs)

* AP1000

A crucial safety strategy for the AP1 000 is depressurization through the

automatic depressurization system (ADS) valves and lines. Entrainment

of liquid into the ADS-4 line under stratified flow in the hot leg has a

significant influence on the AP1000 transient response during

depressurization. Although this topic is being studied, we are concerned

that important scaling issues have not been properly considered and that

the basic mechanism of entrainment may be misunderstood. This is a

particular concern if the major source of entrainment is an oscillating slug

of water, the frequency of which is dependent on the hot leg geometry and

scale.

Upper plenum entrainment in the AP1 000 has so far been approached theoretically

using correlations that are not specifically validated for AP1 000 conditions. The

Advanced Plant Experiment (APEX) facility at Oregon State University is being

modified to better model the AP1000 and the data from testing will be used to

assess the prediction capability of the TRAC-M code.

Low-pressure choked flow tests are being conducted at Purdue University using the

PUMA facility in order to extend the empirical database to cover the regimes of

11 NUREG-1 635
operation of the ALWRs. The results of these experiments will be used to improve

and validate the choked flow models of TRAC-M.

The current RES approach is to perform experiments and to develop corresponding

models for use in TRAC-M at a later date. It would be more effective to develop

models that are compatible with the TRAC-M framework as part of the work scope

for these experimental studies.

* IRIS

Most post-accident phenomena in IRIS should be amenable to prediction

using existing codes such as TRAC-M, MELCOR, and CONTAIN. The

staff has identified four phenomena that may require experimental

verification. Plans for such tests should be prepared and justified after

code runs and assessments of uncertainty show the need for specific

tests.

* ESBWR

The staff plans to develop the TRAC-M and CONTAIN codes to analyze

the ESBWR. The staff has identified several key phenomena that need to

be modeled, particularly those involving natural circulation, and anticipates

that three-dimensional modeling capability will be required. This is a

significant upgrade that will require suitable assessment, perhaps against

new data that may be provided by the integral tests planned for the PUMA

facility. The staff is planning to use a CDF code such as FLUENT to

analyze boron mixing following an anticipated transient without scram

(ATWS) event. The staff should assess the need to validate these

predictions against a broader database than is currently available.

* ACR-700

The staff plans to assess the still fledgling TRAC-M code to determine its

applicability to ACR-700. The horizontal orientation and geometrical

arrangement of tubes in this reactor are quite different than in current U.S.

LWRs. Thus the large database that has been developed for assessing

current LWR designs is mostly inapplicable. The staff needs to carefully

assess the existing Canadian database and decide how well it will support

the certification process for the ACR-700 design. The staff has identified

several key phenomena that require modeling, but has yet to determine

the extent of the need for new code development or experiments.

NUREG-1 635 12
TRAC-M Development and Uncertainty Issues

Several of the current RES research efforts are likely to have applicability to advanced

reactors as well as to current LWRs. For example, the research program at Penn State

University on reflood heat transfer, in response to needs identified by RES, has already

produced important data. These research results need to be analyzed and used to

develop analytical methods to be incorporated into the TRAC-M code (not a trivial

matter). An assessment should then be performed of the adequacy of this database for

evaluating similar phenomena in ALWRs.

Thermal-hydraulic codes play a central role in the analysis of transients and accidents

in all present reactors and will continue to do so for future reactors. The RES plan for

addressing future thermal-hydraulic issues of a system nature is based primarily on the

development of the thermal-hydraulic system code TRAC-M. This code will combine

the features needed for all current generation LWRs as well as any features specifically

identified as needed for ALWRs and HTGRs. The development and implementation of

TRAC-M needs to be accelerated to the point where it is the current tool of use. Having

the code available is important, but this is just the first requirement; the agency needs

extensive experience with its qualification. Much needs to be learned about how much

effort it will take to adapt the code to future designs. While there is general agreement

that this can be done, it will be uncertain how much effort is needed until the specific

qualification work is undertaken.

The current trend toward the reduction of excessive conservatism and the use of best

estimate codes requires explicit consideration of the uncertainties associated with the

predictions of the code. The need to assess uncertainties associated with thermalhydraulic calculations has motivated the development of various methods (such as

nonparametric probabilistic evaluation of epistemic and aleatory uncertainty, and

response surface methods) to evaluate the relative importance of the various sources

of uncertainty. Use of these evaluation methods will require that the parameters in the

code and the accuracy of the thermal-hydraulic models in the code be characterized by

uncertainty ranges and associated frequency distributions. These methods should be

included as integral parts of the codes so that Monte Carlo methods for assessing the

effect of the uncertainties can be applied routinely.

Although the nonparametric probabilistic and response surface methods can address

uncertainty of the modeled phenomena, there is also a need to address the overarching

uncertainty associated with phenomena that are not modeled or are incorrectly

modeled. This may require evaluation and identification of levels of conservatism

needed in order to achieve the desired margin of safety. The code scaling applicability,

and uncertainty (CSAU) evaluation methodology can identify the importance of these

areas of uncertainty, but stops short of identifying methods for evaluating the

uncertainties.

13 NUREG-1635
The assessment of the performance of passive safety systems may be particularly

sensitive to uncertainties in the codes because the driving forces in natural circulation

are much smaller than in a forced circulation system. Therefore, more accurate

predictions of the balance between the competing effects of natural and forced

circulation are needed. For instance, the outcomes of PRA analyses of emergency

cooling systems for current reactors are largely determined by the availability and

reliability of active components such as pumps. In the new designs utilizing passive

safety features, the successful prediction of emergency cooling systems will be

assessed by codes the results of which are sensitive to uncertainties in predicting flows

and heat transfer. The first priority is, therefore, to quantify the uncertainties in these

codes and evaluate the probability of success. Formal methods to incorporate this

model uncertainty into the PRA need to be developed.

Comments and Recommendations on Thermal-Hydraulic Analysis

1. The TRAC-M code is viewed as the major tool to be used by the agency for

predicting the thermal-hydraulic behavior of advanced reactors. Data are being

developed that need to be analyzed and used to develop analytical methods that will

be incorporated into the TRAC-M code. There is a need to speed up the code's

development and range of applicability. The staff also needs extensive experience

with the use of the TRAC-M code before the code can be relied upon for making

regulatory decisions.

2. Development of improved models of thermal-hydraulic phenomena for incorporation

into the TRAC-M code should be an integral part of the RES programs related to the

investigation of the phenomena.

3. Formal methods need to be developed to quantity epistemic uncertainties in the

code predictions of success criteria for use in PRAs.

2.2.5 Neutronic Analysis

The neutronic analysis of the reactor core is fundamental to the safety of nuclear power

plants. Power reactor cores must be stable, reliable, and predictable. The safety

analysis of the core requires detailed databases and computer codes for analyzing both

local details of neutronic behavior and core-wide behavior.

The NRC staff has done a good job of maintaining its neutronic-analysis capabilities at

or near the state of the art. It is essential that the staff continue to maintain these

capabilities at a high technical level. There must be no doubts in the public's collective

mind about the nuclear safety of power reactor cores.

Existing reactors have been designed conservatively based on neutronic analyses that

are quite simplified relative to the capabilities that now exist. They have also been

licensed using simpler analytic methods that entail substantial conservative margins.

NUREG-1635 14
Even within the operating reactor community there is a growing desire and need to take

advantage of the technical developments in neutronic analyses that have taken place

over the last two decades to remove unnecessarily conservative margins in core design

and operation. As an example, this need has become manifest as plants use reactor

fuel to higher levels of bumup. Experiments have shown that high-bumup fuel cannot

sustain as much energy from reactivity insertion events once thought. But, improved

neutronic analyses have shown that energy inputs from the more probable reactivity

insertion events are much smaller than had been thought based on older analyses.

There is, then, pressure from the industry on the staff to accept operation of cores with

higher bumup fuel based on the predictions of the improved analysis tools.

This pressure on the staff to adopt more advanced neutronic analysis capabilities in the

regulatory process will continue to grow as advanced reactors with cores designed

using the more sophisticated analysis tools are proposed for certification. The licensing

process needs to evolve in a way that takes advantage of the technical advances and

abandons unnecessarily conservative approaches of the past to neutronic safety.

Though the NRC staff has done a remarkable job of maintaining its databases and

computational resources, advanced reactors under consideration today do pose some

challenges. For purposes of discussion, three groups of advanced reactors are

considered. The first group consists of advanced reactors such as the AP 000 and the

ESBWR that involve modest evolutions from the current reactor technologies and that

depend on passive safety systems. Safety analyses of these reactors demand close

coupling between models of neutronics and thermal-hydraulics within the reactor

coolant system. Steps being taken by the staff to couple the three-dimensional

neutronics code PARCS to the TRAC-M code appear adequate to meet the regulatory

safety analysis needs for these reactors, though the challenges of implementation of

the analyses should not be underestimated. Thus, we recommend that the

development of the combined TRAC-M/PARCS computer code should continue to be

supported.

The ACR-700 is a second type of advanced reactor for consideration. This reactor is

similar to the CANDU reactors which have positive void coefficients that affect the

neutronics in ways that are usually considered undesirable. Modest enrichments of the

fuel currently proposed for the ACR-700 and the use of light water as the coolant are

intended to ameliorate the tendency for positive void coefficients. Still, the interplay

between neutronics and the unfamiliar thermal- hydraulics of a core configured

horizontally will challenge the analytic tools available to the staff for performing safety

assessments. Modifications of these tools to account for the core composition, core

configuration, and reactor coolant system thermal-hydraulics of the ACR-700 are longlead-time activities that need to be initiated now if the regulatory process of certification

is to proceed in a timely manner.

The gas-cooled high-temperature reactors make up the third group of advanced

reactors for discussion. Gas-cooled reactors involve substantial departures from

is NUREG-1 635
I

today's LWR technology. These reactors received much of the attention in the NRC

staff's Infrastructure Assessment because the reactors present challenges that are at or

even beyond the current state of the art in neutronic analysis. The moving annular core

of the PBMR with its stochastic fuel densities and unpredicted temperature distributions

is just one example of the analysis challenges that may be presented to the NRC staff

by gas-cooled advanced reactors.

The staff has presented an appropriate research program to confront these challenges.

The staff does have some philosophical issues to confront as advanced reactors based

on novel technologies are proposed. Historically, the NRC staff has done almost

completely independent analyses of reactor neutronics, even to the point of using

independently developed analytic tools. The question arises, therefore, should this

practice continue, or should the NRC staff confine itself to a review of analyses done by

the applicant? We believe that independent analyses should continue to be the

practice. Independent analyses may become more important as technically advanced

tools are increasingly used to study novel concepts. Neutronic analysis is simply too

central to nuclear safety to take a lesser approach.

An overarching issue is the availability of data to validate the predictions of neutronic

analyses. In the past, there have been abundant data for such validations and it is likely

that data can still be obtained for code validation using lead test assemblies for reactors

and fuel that are modest departures from existing technologies. For the more novel

designs, neither the NRC staff nor the reactor designers have access to experimental

facilities to provide validation data that are prototypic. Indeed, the staff has already

identified challenges in obtaining suitably prototypic data for the ACR-700. The staff

must insist that advanced reactor designs be supported by prototypic experimental

data. Though modem tools of neutronic analysis are very sophisticated, they are

neither transparent nor immune to error. Therefore, we recommend that the staff

provide explicit direction to the applicant on the experimental data and data quality that

it thinks are essential to support and validate neutronic analyses of reactor designs

submitted for certification.

Comments and Recommendations on Neutronic Analysis

1. The maintenance and improvement of neutronic analysis capabilities at the NRC

should continue to be supported so that the NRC can continue to perform

independent assessments of reactor neutronics.

2. The effort to couple the thermal-hydraulics code TRAC-M with the three-dimensional

neutronics code PARCS will be essential for certification of reactors such as

API 000 that depend on passive safety systems.

3. Modifications to analysis methods to account for the different features of the ACR700 should be initiated now to facilitate anticipated certification review.

NUREG-1635 16
4. NRC staff should provide explicit direction to the applicant as to the data and data

quality that will be needed to support and validate advanced reactor designs

submitted for certification.

2.2.6 Severe Accident and Source Term Analysis

Risk associated with any nuclear power plant is dominated by the frequencies and

consequences of severe accidents that involve substantial releases of radioactivity from

the fuel. The issues of accident progression, fuel degradation, fission product release

and transport within the reactor coolant system and fission product behavior within the

reactor containment or confinement are discussed in this section.

As with neutronic analysis, it is convenient to consider three groups of advanced

reactors for the discussions of accident progression and accident source terms. The

first of these groups consists of reactors whose designs evolved from current reactors.

This group includes the API 000 and the ESBWR. Accident analysis and fission

product behavior codes developed by the NRC for the current generation of power

reactors appear applicable with modest modification to this group of reactors.

For instance, two issues have emerged for these reactors that necessitate additional

experimental data and specialized analyses. One of these is the degradation and

fission product release from fuel taken to high levels of bumup (>60 GWd/t). The NRC

is considering an extension or follow-on to the current PHEBUS-FP program to obtain

data on degradation and fission product release from high-bumup fuel. We certainly

encourage these efforts and recommend that a program to provide the needed data be

defined.

A second issue has to do with the arguments by some advanced reactor designers that

core debris produced by a severe accident can be retained within the reactor pressure

vessel. The necessary heat removal is to be achieved by flooding the reactor cavity.

This safety strategy was proposed for the AP600 design. The strategy was not credited

in the staff's safety case because of concerns over the heat transfer model and

chemical interactions within the core debris.

The NRC has been an instigator of the international cooperative research programs

RASPLAV and MASCA. These programs have provided important information on the

convective heat transfer from core debris in the lower plenum of a reactor vessel.

These programs have also shown that chemical interactions within core debris and

between core debris and reactor vessel materials are more complicated and more

aggressive than envisaged by the advanced reactor designers. The NRC needs to

develop a quantitative understanding of these chemical processes for review of

proposed reactor designs. The MASCA experimental studies should be continued and

appropriate models incorporated into the NRC's reactor accident models.

17 NUREG-1 635
The ACR-700 is a second category of advanced reactor for discussion of severe

accident concerns. It appears that substantial effort will be required to modify the

NRC's accident analysis models to treat this reactor, which has a unique horizontal

configuration of the core and a complicated reactor coolant system. The NRC does not

have a body of data on the degradation of such a reactor core design under severe

accident conditions comparable to the data available for fuel configurations found in

current U.S. power reactors. It is not evident what accident progression data will be

available from the reactor designers. Early examinations suggest that accident

progression could be rapid when significant pressure tube deformation occurs.

Furthermore, issues of energetic interactions, of molten core debris with coolant or

moderator may be more problematic for the ACR-700 than these processes are thought

to be for current U.S. reactors. The NRC needs to investigate modification of its

accident analysis codes to address the ACR-700 and identify experimental data needed

to validate predictions of accident progression, fuel-coolant interactions, and fission

product release. It should be the responsibility of the applicant to provide the data

identified in this effort.

The third class of advanced reactors consists of HTGRs with graphite moderator. It is

evident that some severe accidents in such reactors will involve the ingression of air.

The NRC staff's experience with accident progression in such reactors is limited.

Similarly, the NRC staff has little experience with either the release of fission products

from fuels or the transport of these fission products within the coolant system and

containment for such reactors.

Existing NRC accident analysis tools and fission product release and transport models

would require very substantial modification to be applicable to these gas-cooled

reactors. It will take substantial time to develop adequate tools for the staff's analysis of

accident progression and source term predictions. Consequently, there is a need to

initiate these efforts. It is not readily apparent, however, what data are available from

well-scaled tests for validation of the modified models. The staff will need to address

this issue as part of the effort to update the tools.

Comments and Recommendations on Severe Accidents and Source Terms

1. Accident progression and source term analyses for new reactors such as the

ESBWR and AP1000 can be addressed through modest modifications of existing

codes. Issues associated with the use of high-bumup fuels will have to be

addressed.

2. Some of the new reactor designs incorporate cavity flooding, which the designers

argue will retain the debris produced by a severe accident in the pressure vessel.

Research will be needed to assess these arguments. The NRC should continue its

involvement in the MASCA work.

NUREG-1 635 18
3. A substantial effort will be required to configure the NRC staffs accident analysis

models to treat the unique core and coolant configuration of the ACR-700. The

NRC staff needs to investigate what modifications are needed for its codes and to

identify the experimental data that should be supplied by the applicant to validate

predictions of accident progression and fission product release.

4. The NRC has limited experience with severe accident analysis and fission product

release for gas-cooled reactors with graphite moderators and structures. Existing

NRC codes will require substantial modification to be applicable to the analysis of

these types of reactors. It will take substantial time to develop adequate tools and

to identify the available and needed new experimental data before code model

development can begin.

2.3 Fuel Analysis

Advanced reactor designs under active consideration today use fuels that can be

grouped, as in discussions of neutronics and severe accident analyses, into three

classes. The first of these classes comprises somewhat prosaic fuels little different

than those used in the current LWRs, though possibly with improved cladding alloys.

The primary issue associated with this class of fuel will be increased levels of bumup.

The NRC currently has a confirmatory research program to ensure the safety of fuel to

bumups of 62 GWd/t. It is imperative that this research continue to confirm the

adequacy of regulatory decisions made to date concerning high-bumup fuel. As stated

in the previous section, it is also important that a companion research program on

degradation of high-bumup fuels under severe accident conditions be continued.

Ucensees and designers are interested in using fuels to bumups in excess of 62

GWd/t. The NRC has decided that licensees will have the responsibility to provide

technical justification for using fuels to the higher levels of bumup. Regardless, RES

will still have to provide the agency with tools to review and independently evaluate

these technical justifications since currently available fuel behavior models cannot be

reliably extrapolated to such high bumup levels. A crucial issue for the staff to consider

is the need for experimental data on fuel behavior under upset conditions when fuel has

been used to bumups on the order of 75 GWd/t. In light of the change in fuel physics

that has been observed as bumups exceed -50 GWd/t, it seems likely that

experimental data would be needed to justify use of fuel to bumups in excess of 62

GWd/t.

A second class of fuel for advanced reactors is the fuel to be used in the ACR-700

design. This fuel is a slightly enriched U0 2 fuel which is horizontally configured in the

reactor. The NRC staff has minimal experience with this fuel and fuel configuration. It

does appear, however, that the fuel is not a radical departure from designs that the

NRC staff has considered in the past. It may well be possible for NRC to extrapolate its

existing fuel behavior codes and databases to evaluate the ACR-700 fuels. The

confidence with which this can be done will depend on experimental data available from

19 NUREG-1 635
the developers of the ACR-700 concerning fuel behavior, especially under upset

conditions and conditions of reactivity insertion events.

The third class of advanced reactor fuel is the coated-particle fuel to be used in gascooled reactors, such as the PBMR and the GT-MHR. Because little experience is

available within the NRC for the review of coated-particle fuel, it may be necessary now

to initiate long-term efforts to develop these capabilities. Development of these

capabilities will have to begin with analysis methods using data for fuel prepared and

tested in the past, largely overseas. We caution that analysis methods now available in

the literature are for highly idealized circumstances and may not be adequate for the

safety analysis of coated-particle fuels during normal operations and upset conditions.

Comments and Recommendations on Fuel Analysis

1. As with currently licensed reactors the primary issue to be addressed with ALWR

fuels is increased bumup. This issue is being addressed for current reactors and

needs to be given a high priority.

2. It may be possible for the NRC to extrapolate the use of existing codes to evaluate

ACR-700 fuel. The confidence with which this can be done will depend on the

experimental data provided by the developers of the ACR-700.

3. Coated-particle fuels were the focus of much of the staff's attention in its Advanced

Reactor Research Infrastructure Assessment since there is little NRC experience

with these fuels. Because of the long-term nature of the associated research,

efforts to develop the NRC staff's capability to address the review of these types of

fuels should be initiated.

2.4 Materials Analysis

The Infrastructure Assessment document identifies critical gaps in the materials

behavior knowledge needed for transitioning from the relatively well-understood LWR

systems to HTGR systems are identified in. The document focuses on ensuring an

adequate understanding of the behavior of metallic and graphite HTGR component

which have critical structural, barrier, or retention functions under normal and off-normal

conditions. Potential degradation mechanisms in high-temperature helium for these

components are thoroughly discussed and include, for instance, graphite swelling, the

effect of coolant impurities, and creep-fatigue interactions in metallic components.

There is an existing experience base in many of these areas, and the designers are

expected to develop additional data. However, because the degradation phenomena

impact the codes and standards for the design and fabrication of the reactor as well as

the safety margins under normal and accident conditions, we agree with the staffs

argument that the NRC must have an independent research capability in the most

critical of these areas. Previous experience has shown that characterization of these

material degradation phenomena requires a long-term effort. Thus, it is necessary to

NUREG-1 635 20
initiate such efforts relatively soon. We agree that prioritization should be given initially

to (a) evaluating the completeness of the existing databases needed to formulate

design codes, (b) initiating the construction of the experimental facilities needed to

investigate the potential degradation of proposed materials in HTGR environments, and

(c) finalizing effective collaborative agreements with overseas organizations. These

tasks will lead to a definition of additional data needed for regulatory decisions, and will

ensure that the experimental infrastructure is in place to provide timely information.

The ACR-700 design is sufficiently different from the current LWRs that research is

needed in several areas. The staff identified the following concerns: (a) the effect of

the ACR-700 environment on component fatigue and creep life, (b) irradiation and

embrittlement of the pressure tube material, (c) the performance of the large number of

dissimilar metal welds in the header system, and (d) component material behavior

under severe accident conditions. We agree with these concerns and believe that their

resolution should be given priority since the ACR-700 is currently undergoing preapplication review by the staff and relevant data will be required before certification.

For ALWRs like the AP I000 and ESBWR, the materials issues are similar to those

associated with the current LWRs and hence are being addressed by the RES

programs dealing with the current operating plants. We note that RES has ongoing

programs on technical basis for reevaluation of the pressurized thermal shock

screening criteria, steam generator tube degradation, and vessel head penetration

cracking and degradation. This work should continue and is of direct relevance to most

of the ALWR design concepts. The development of risk-informed regulations for both

the current LWR fleet and the future reactors will require knowledge of the probabilities

of material failuren and an assessment of associated uncertainties. There is a

considerable amount of international interest in the following reports of this topic:

* The relevance and quality of environmentally assisted cracking and corrosion

databases upon which structural integrity decisions are made

* Databases to support the factors of improvement for cracking-resistant" alloys such

as 690, 52, and 316L

* The development of prediction models which lead to proactive life management

capabilities

Although the development of databases and proaction is primarily the applicant's

responsibility, we recommend that RES continue to participate in these activities so as

to understand the assumptions inherent to the applicant's assertions of materials

integrity. Materials degradation issues have been at the core of many of the current

LWR fleet problems, and in order to ensure adequate regulation and control in the

future it is important that active participation in the research activities for current and

advanced water-cooled reactors be continued.

21 NUREG-1 635
Finally, the Infrastructure Assessment document correctly identifies a need for an

improved inservice inspection program for both ALWRs and HTGRs. This need is

prompted by the integral nature of many of the advanced reactor designs (leading to

diminished accessibility for inspection) and the anticipated longer operating cycles

between scheduled, short-duration refueling outages when inspections would normally

occur to quantify damage and the impact on safety. This task is primarily the

responsibility of the applicant, but RES should conduct sufficient research to remain

aware of the technical issues.

Comments and Recommendations on Materals Analysis

1. RES should undertake research activities to identify potential materials degradation

processes in HTGRs. Tasks in the near-term (i.e., FY2002-FY2004) include (a)

evaluating the completeness of the existing databases needed to formulate design

codes, (b) initiating the construction of the experimental facilities needed to

investigate the potential degradation of proposed materials in HTGR environments,

(c) initiating research on environmental effects on cracking (fatigue, stress

corrosion) and creep, and, (d) finalizing effective collaborative agreements with

overseas organizations. This prioritization is appropriate given that these lead tasks

are long-term items that prepare the foundation for future work in support of

projected design certification.

2. The proposed work outlined for the ACR-700, which involves somewhat different

materials and environments than current PWRs, is appropriate and will provide a

technical basis for NRC assessment of industry proposals and supporting data.

3. We concur with the assessment that the ongoing studies to address materials

issues for operating reactors should continue. RES should remain involved in

international collaborations to develop and analyze materials degradation databases

in order to assess the probability of failure of 'degradation-resistant materials.

2.5 Structural Analysis

The Infrastructure Assessment document section on structural analysis primarily

addresses the analysis, aging, and inspection of containment/confinement structures

and the response of those structures and the primary system to external events. The

document also recognizes that while acceptance requirements for current LWRs are

primarily deterministic, there is a strong interest in performance-based and riskinformed acceptance criteria.

The areas in which the Infrastructure Assessment document indicates that research will

be needed include (1) seismic hazard assessment, (2) nonlinear seismic analysis of

reactor vessel and core support structures, (3) the analysis of the soil-structure

interaction of deeply embedded or buried structures during seismic events, (4) the

effects of high temperature on properties of concrete, (5) the use of modular

NUREG-1 635 22
construction techniques for safety-related structures, and (6) risk-informed inservice

inspection methodologies for containment and associated structures.

Current NRC guidance for determination of seismic hazards states that both the

Lawrence Livermore National Laboratory and the EPRI probabilistic seismic hazard

methodologies are acceptable for determination of the safe shutdown earthquake for

nuclear power plants. For some sites, the estimates from the two methodologies can

be significantly different, which can lead to difficulties in cases where quantitative

acceptance criteria must be met. RES proposes to use a set of guidelines developed

jointly by the NRC, DOE, and the EPRI Senior Seismic Hazard Analysis Committee

(SSHAC) to update the two currently accepted seismic hazard assessment

methodologies for the Central and Eastem U.S. to obtain more consistent seismic

hazard assessment.

Research is planned to develop an independent NRC staff capability to evaluate the

seismic integrity of advanced reactors. The applicability of existing finite element

analysis codes to analyze nonlinear configurations such as the nonductile graphite core

reflectors and supports in HTGRs will be assessed. In designs with long fuel tubes like

AP1000 and IRIS, the seismic margin may be controlled by the fuel design, and

nonlinear static and dynamic structural analyses may be needed to assess seismic

margins.

For some new reactor designs, the entire reactor building and a significant portion of

the steam generator building will be partially or completely embedded below grade.

Soil-structure interactions will have a significant effect on the seismic response of such

structures, and models of soil-structure interaction in existing computer codes (which

have generally been developed for structures with relatively shallow embedments) will

be reviewed to determine their applicability. Shake table tests will also be performed to

provide experimental verification of analytical results.

The Infrastructure Assessment document notes that concrete structures in HGTR

designs will be subjected to sustained high temperatures. Substantial information is

available from the literature and previous work at the Sandia and Oak Ridge National

Laboratories. The available information will be collected and reviewed to identify gaps

in the data.

Although used in Japan, modular construction has not been used in the U.S. for nuclear

power plants. The modules of interest are concrete-filled steel plate wall and

foundation modules. The designers of the GT-MHR, AP1000, and the IRIS reactors

have proposed using modular construction techniques in order to shorten the

construction schedule, reduce costs, and improve the quality of construction. Technical

issues that must be addressed to accept use of modular construction include

assessment of the strength and ductility of the module and of the joints and

connections needed to join the modules, and determination of appropriate seismic

damping values for the resulting structures. The proposed research will be used to

23 NUREG-1 635
develop evaluation criteria that will facilitate review of reactors that use modular

construction.

The Infrastructure Assessment document notes that inservice inspection requirements

for new reactors will be largely based on risk-informed approaches. Indeed, the ASME

has formed a task group to develop methodologies for risk-informed inservice

inspection of containments. The staff plans to actively participate in this ASME Code

activity while independently developing methodologies for risk-informed inservice

inspection of containments. The work the staff intends to do to support this effort

includes compiling data on degradation mechanisms for structures, developing

appropriate inspections strategies for these degradation mechanisms, and defining risk

categories based on the potential consequences of failure. We address this proposal in

the following comments and recommendations.

Comments and Recommendations on Structural Analysis

1. The Infrastructure Assessment document appears to address the potential areas of

concern for the structural analysis of the advanced reactor design bases and

seismic events.

2. We support the effort to implement the SSHAC guidance and associated

methodology to obtain new probabilistic seismic hazard estimates to provide a

sounder technical basis for assessing seismic risk. This is important not only for

advanced reactors but also for current reactors.

3. The seismic design and analysis of advanced reactors will involve extensions of

current technology. Although the primary responsibility for the development of this

technology lies with the nuclear industry, a modest program in this area seems

warranted to help ensure that the staff maintains an awareness of the state of the

art and the limitations of the available technologies.

4. The potential degradation of containment or confinement structures is an important

area of concern for the NRC. Enough is known about the design of the proposed

HTGRs to realize that the temperatures will be high enough to require an extension

of current design codes and a reassessment of the aging of concrete structures.

The RES Infrastructure Assessment document properly addresses this as a

potential area of concern.

5. Risk-informed inservice inspection (RI-ISI) clearly has provided a more rational

approach to the inservice inspection of piping, and it is reasonable to expect that it

can be beneficial in developing inspection programs for containment. RES played

an active role in the initial development of RI-ISI for piping systems. This was

appropriate since it was a new approach and neither the NRC nor the industry had

prior experience. Although the extension to containment systems will involve new

concepts and approaches, we believe that it is not necessary for the staff to

NUREG-1 635 24
independently develop this methodology. It seems sufficient that the staff

participate actively in the ASME Code activity. The staffs other research efforts in

this area should be sufficient to maintain an awareness of the necessary expertise

and technology.

2.6 Consequence Analysis

Licensing of advanced reactors will require an evaluation of the offsite consequences of

severe accidents for such reactors. Specifically, a technical justification for a

recommendation to the Commission on the size of the emergency planning zone (EPZ)

will be required. The supporting calculations will need to be consistent with those

currently utilized for choosing a 10-mile EPZ for the present LWRs, which follow the

criteria established in NUREG-0654, Criteria for Preparation and Evaluation of

Radiological Emergency Response Plans and Preparedness in Support of Nuclear

Power Plants," and NUREG-0396, "Planning Basis for the Development of State and

Local Govemment Radiological Emergency Response Plans in Support of Light Water

Nuclear Power Plants.'

The RES Infrastructure Assessment document foresees that the staff will perform

offsite consequence analysis of advanced reactors using NRC's Level 3 evaluation

code MACCS2. The mix of radionuclides and the chemical forms in the releases from

severe accidents in advanced reactors may be different than in the releases during

accidents in LWRs. A Level 2 analysis will provide estimates of the radionuclides

produced. RES plans to evaluate MACCS2 to ensure that design-specific biologically

important radionuclides, dose conversion factors, and factors that account for their

chemical form are appropriately modeled in MACCS2.

Comment on Consequence Analysis

The RES plan will enable the staff to perform consequence analysis of advanced

reactors and support developing recommendations to the Commission on the size of

the EPZ.

25 NUREG-1 635
3 PHENOMENA IDENTIFICATION AND RANKING TABLE (PIRT) PROCESS

AND IMPLEMENTATION PLAN

A PIRT process, similar to that used by RES in the review of advanced reactor designs

in the early 1990s (MHTR, PIUS, PRISM, CANDU-3), is outlined in the RES document.

This will be used to identify and prioritize the additional research needed for the

development of a risk-informed regulatory framework for advanced reactors. The

implementation of the advanced reactor research activities will reflect the outcome of

these task prioritization activities and budget and timing constraints. For instance, longlead time tasks such as rulemaking, codes, and standard development activities should

be completed in advance of the start of formal design certification.

Numerous management questions are raised in the RES document, including, the need

for collaborative projects with international and domestic partners, the risks associated

with noncompletion of task objectives, longevity of funding resources and related

issues. All of these implementation questions are valid and pertinent and need to be

addressed.

Comments and Recommendations on PIRT activities and Plan Implementation

1. The PIRT process that RES has outlined is appropriate for defining the prioritization

of tasks necessary to formulate a risk-informed regulatory framework for advanced

reactors.

2. RES has identified valid research plan implementation issues, which need to be

addressed.

27 NUREG-1 635
4 IMPACT OF ADVANCED COMPUTER CAPABILITIES ON NRC STAFF

ACTIVITIES

Computing power is expanding at extraordinary rates. Modem personal computers

(PCs) perform at speeds > 4 billion calculations per second (4-gigaflops) that exceed

the capability of supercomputers just a few years ago. However, these speeds pale

compared to those available in modem multiprocessor systems. Currently, there are 47

systems in the world with speeds that exceed 1000 gigaflops (1 -teraflop). Six months

ago, there were only 23 such systems. IBM is now developing a 360-teraflop computer

for DOE. Although such powerful systems are currently available primarily at

Government laboratories and university supercomputer centers, Linux PC clusters with

speeds of 200-400 gigaflops now cost less than $1 00K and are becoming widely

available. Indeed, gigaflop machines are becoming standard tools for engineering

analysis by commercial firms.

The availability of such computing power makes possible calculations of far greater

complexity and opens new possibilities for visualizing information. We expect that new

reactor designs will take advantage of virtual reality tools to optimize the layout of the

plants. Clearly, enormous increases will be possible in the scale of problems that can

be analyzed by computational fluid dynamics models; truncation, approximation, and

recalculation of PRAs need no longer be a significant consideration; core designers will

find ever more ingenious ways to eke out more power from existing reactors; and

enormous increases in the sophistication of sensitivity and uncertainty analyses will be

possible.

We see little evidence that the NRC staff is preparing for the potential impact of these

changes. The staff should assess the possibilities that quantum increases in computing

power can have on the way it does analyses and develop a plan to use the new tools to

increase the efficiency and effectiveness of the agency. The potential challenge the

agency faces is that new reactors will be designed using software that runs only on

these advanced computing resources. Without similar calculational tools (both software

and hardware), the NRC will be hard pressed to carry out its safety assurance mandate

without imposing unnecessary and even anachronistic burdens on applicants and

licensees.

Recommendation on Computer Capabilities

The NRC staff needs to consider the impact the enormous increase in computer

capabilities that are occurring will have on the way engineering analyses (thermalhydraulic, fire, core neutronics, and PRA analyses, etc.) are done and assess how this

capability can be used to increase the efficiency and effectiveness of the Agency's

work.

29 NUREG-1 635
5 OVERALL CONCLUSIONS AND RECOMMENDATIONS

The Advanced Reactor Research Infrastructure Assessment document is

comprehensive and reflects a good understanding of the issues, existing state of the

art, and past and ongoing research results and activities as they pertain to future

reactor designs. It identifies the increments in technology necessary for the

development of a risk-informed and performance-based regulatory structure for some

specific ALWR and HTGR design concepts.

We generally agree with the RES assessment of development needs and planned

activities in several of the technology areas (i.e. probabilistic risk assessment,

instrumentation and control, materials analysis, structural analysis, and consequence

analysis) and offer supportive comments on details. In other technology areas, we

have more specific comments on the focus that RES proposes. These later comments

are expanded upon in Section 2 of this report, and include the following:

1. The research needs for a technology-neutral "Generic Regulatory Framework are

discussed in the general terms of a hierarchical structure based on the preliminary

framework under development for "Option 3." This framework should prove useful

for the certification of the more likely plants to be first deployed (i.e., ALWRs which

will be probably licensed under the current 10 CFR Part 52 rule), and will be

mandatory for the non-LWR concepts on a longer timeframe. This development will

take some time. Our concerns with the current risk-informed Option 3 concept

should be addressed.

2. Proponents of advanced reactor designs are proposing fewer numbers of control

room staff than are used in today's LWR fleet. In the area of "Human Factors," we

believe that the nuclear industry will be challenging existing requirements for control

room staffing. It is important, therefore, that the NRC staff undertake activities to

ensure that its positions on control room staffing have firm and defensible basis.

3. In the area of "Thermal-Hydraulicsn, much reliance is placed on the TRAC-M code to

meet the licensing challenges posed by features of the various advanced water

reactor designs. It is recommended, therefore, that priority be given to quickly

code's developing this code and extending its range of applicability and, just as

importantly, putting it to use. Because PRA success criteria for passive designs are

sensitive to uncertainties in thermal-hydraulic code predictions, these model

uncertainties need to be quantified.

4. The maintenance and improvement of neutronic analysis capabilities at NRC should

continue to be supported so that the NRC can continue to do fully independent

nuclear safety assessments. Completion of the effort to couple the thermalhydraulics code TRAG-M with the three-dimensional neutronics code PARCS will be

essential for certification of reactors such as AP1000 that represent only modest

departures from current reactor technologies. Modifications to analysis methods to

31 NUREG-1635
- -

account for the different features of the ACR-700 should be initiated now to facilitate

timely completion of the anticipated certification review.

Work to develop neutronic analysis of gas-cooled reactors will be challenging given

the quite limited experience at the NRC in this area and the significant challenges

associated with, for example, stochastic fuel pellet movements and unpredictable

temperature distributions. The NRC staff should provide explicit direction to the

applicant as to the data and data quality that will be needed to support and validate

advanced reactor designs submitted for certification.

5. The needs in the area of Severe Accident and Source Term Analysiso for new

reactors such as the ESBWR and AP1000 can be addressed through modifications

of existing codes. Issues related to higher bumup fuels will have to be addressed.

Some of the new reactor designs have features, such as cavity flooding, that the

designers argue will result in severe accident debris being retained in the pressure

vessel. Research will be needed for the NRC staff to evaluate these arguments and

become knowledgeable on this issue. Consequently, the NRC staff should continue

its involvement in the MASCA work.

It appears that a substantial effort will be required to configure NRC's accident

analysis models to treat the unique core and coolant configuration of the ACR-700.

The NRC staff needs to investigate what modifications to its codes are needed and

to identify experimental data that may be needed to validate predictions of accident

progression and fission product release.

The NRC staff's experience with severe accident analysis for gas-cooled reactors

with graphite structure and moderator is quite limited. Existing NRC codes will

require substantial modification to be applicable to the analysis of this type of

reactor. It will take substantial time to develop adequate tools and there is a need to

identify the available and needed new experimental data and to begin code model

development.

6. In the area of Fuels Analysis," the advanced reactors under active consideration

utilize fuel designs that can be grouped into three classes: (1) fuel little different

than used in currently operating reactors, (2) the fuel that will be used in the ACR700, and (3) coated-particle fuels to be used in gas-cooled reactors such as the

PBMR and GT-MHR.

The primary issue to be addressed with the first class of fuel is the potential impacts

of increased bumup. This issue is being addressed for current reactors and needs

to be given a high priority. It may be possible for the NRC to use existing codes and

databases to evaluate ACR-700 fuel using data provided by the ACR-700

developers, but this needs to be confirmed.

NUREG-1 635 32
Coated-particle fuelswere the focus of much of the staff's attention in the Advanced

Reactor Research Infrastructure Assessment since there is little NRC experience

with these fuels. Given that design certification activities for the GT-MHR are

projected to start in early 2006, it is necessary that efforts be initiated to develop

NRC's capability to address the review of these types of fuels.

7. The NRC staff needs to consider the impact that the enormous increase in computer

capabilities that are occurring may have on the way engineering analyses in the

areas of thermal-hydraulics, fire, core-neutronics modeling, PRA etc. are done, and

to assess how this capability can be used to increase the efficiency and

effectiveness of the Agency.

33 NUREG-1635
6 REFERENCES

1. Advanced Reactor Research Infrastructure Assessment, Rev. 2, Office of Nuclear

Regulatory Research, October 2002.

2. Revised Memorandum to William D. Travers, Executive Director for Operations,

from Annette L. Vietti-Cook Secretary, Subject: 'Staff Requirements-COMJSM-00-

0003-Staff Readiness for New Nuclear Plant Construction and the Pebble Bed

Reactor," February 13,2001.

3. Policy Issue Information for he Commissioners, from William D. Travers, Executive

Director for Operations, Subject: "Future Licensing and Inspection Readiness

Assessment, October 12, 2001.

4. A Technology Roadmap for Generation IV Nuclear Energy Systems, Technical

Roadmap Report, Review Version 7a, dated September 20, 2002, submitted to the

Nuclear Energy Research Advisory Committee by the Subcommittee on Generation

IV Technology Planning.

5. Letter to William D. Travers, Executive Director for Operations, from George E.

Apostolakis, Chairman, Subject: 6Draft Advanced Reactor Research Plan,' July 18,

2002.

6. Memorandum to John T. Larkins, Executive Director, ACRS, from Farouk Eltawila,

Director, Division of Systems Analysis and Regulatory Research, Subject: Draft

Advanced Reactor Research Plan," March 28, 2002.

7. Policy Issue Information for the Commissioners, from William D. Travers, Executive

Director for Operations, Subject: NRC Research Plan for Digital Instrumentation

and Control," August 15, 2001.

8. Review and Evaluation of the Nuclear Regulatory Commission Safety Research

Program, A Report to the U.S. Nuclear Regulatory Commission, NUREG-1 635, Vol.

1, June 1998.

9. Review and Evaluation of the Nuclear Regulatory Commission Safety Research

Program, A Report to the U.S. Nuclear Regulatory Commission, NUREG-1635, Vol.

4, May 2001.

10.A Roadmap to Deploy New Nuclear Power Plants in the United States by 2010,

Volume I, Summary Report, prepared for the U.S. Department of Energy, Office of

Nuclear Energy, Science and Technology and its Nuclear Energy Research

Advisory Committee Subcommittee on Generation IV Technology Planning, October

21, 2001.

35 NUREG-1 635
APPENDIX

DOE NEAR-TERM DEPLOYMENT AND GEN IV REACTORS

The RES Infrastructure Assessment document addresses some of the DOE advanced

reactor designs (AP1 000, ESBWR, PBMR, ACR-700, GT-MHR, IRIS) and these are

discussed in previous sections. However, the RES document does not discuss the

wider range of design concepts that DOE considers for near-term deployment (i.e. by

2015) or the Generation IV" designs which, in DOE's best case assessment, may be

deployable by 2020-2025.

In 2000 DOE's Office of Nuclear Energy, Science, and Technology initiated a review of

the electrical generation deployment needed to provide increased capacity and to

continue to provide 30% of the electrical supply from emissions-free sources. The role

of nuclear power obviously would play a significant role in achieving the latter goal. The

initial challenge was to identify the reactor designs that would satisfy the estimated

additional 50,000 MWe nuclear capacity required on the grid by 2020, with the first

reactors being deployed around 2010. Consequently, DOE's Nuclear Energy Research

Advisory Committee (NERAC) conducted a study to identify the actions needed by

government and industry to overcome the technical and regulatory barriers to new plant

construction. The participants in this assessment of potential reactor designs included

international representatives from reactor vendors, utilities, national laboratories and

academia. The results of this study were documented in an October 2001 report titled

OA Roadmap to Deploy New Nuclear Power Plants in the U.S. by 201 0 (Reference 10).

Six designs were found to be at least possibly deployable in the U.S. by 2010, provided

new plant orders were placed by 2003. These U.S. Near-Term Deployment plants are

as follows:

Advanced Boiling Water Reactor (ABWR)

Advanced Pressurized Water Reactor 1000 (AP1000)

European Simplified Boiling Water Reactor (ESBWR)

Gas Turbine Modular High Temperature Reactor (GT-MHR)

Pebble Bed Modular Reactor (PBMR)

Siedewasser Reactor-1000 (SWR-1000)

DOE identified other designs which had significant international support, and were

deployable by 2015 or earlier. These International Near-Term Deployment plants are

as follows:

Advanced Boiling Water Reactors

Advanced Boiling Water Reactor II (ABWR II)

European Simplified Boiling Water Reactor2 (ESBWR)

High-Conversion Boiling Water Reactor (HC-BWR)

Siedewasser Reactor-10002 (SWR-1000)

2These reactors are also identified as U.S. Near-Term Deployment plants

37 NUREG-1635
I

Advanced Pressure Tube Reactors

Advanced Pressurized Water Reactor 600 (AP600)

Advanced Pressurized Water Reactor 10003 (AP1000)

Advanced CANDU Reactor 700 (ACR-700)

Advanced Pressurized Water Reactors

Advanced Power Reactor 1400 (APR-1400)

Advanced Pressurized Water Reactor Plus (APWR+)

European Pressurized Water Reactor (EPR)

Integral Primary System Reactors

Central Argentina de Elementos Modulares (CAREM)

International Modular Reactor (IMR)

International Reactor Innovative and Secure (IRIS)

System-Integrated Modular Advanced Reactor (SMART)

Modular High-Temperature Gas-Cooled Reactors

Gas Turbine Modular High-Temperature Reactor' (GT-MHR)

Pebble Bed Modular Reactor' (PBMR)

It is noteworthy that the ABWR design and the AP600 design (which is the precursor to

the AP1 000) have been certified under 10 CFR Part 52. The other DOE U.S. NearTerm Deployment plants and the ACR-700 and IRIS reactor designs are addressed

within the RES Infrastructure Assessment document. (Note: the pr-application review

for the SWR-1 000 is currently expected to begin in mid-2004.)

In addition to these Near-Termn Deployment reactor designs, DOE considered a number

of more advanced design concepts (GEN IV designs) which addressed the entire fuel

cycle from ore extraction to final waste disposal (e.g., actinide management) and also

address the potential interactions between higher temperature nuclear reactor designs

and, for instance, other energy conversion systems such as hydrogen production.

Nearly 300 potential reactor designs water-cooled, gas-cooled, or liquid metal-cooled

cores were evaluated by DOE against specific sustainability, safety, and reliability, and

economics goals.

The resulting Generation IV Roadmap' (Reference 4) issued.in 2002 defined the

necessary research and development to support a generation of innovative GEN IV

reactor designs believed to be deployable by either 2020 or 2025:

Deployable by 2020

Sodium-Cooled Fast Reactor (SFR)

Very High Temperature Reactor (VHTR)

3

Ibid.

NUREG-1 635 38
Deployable by 2025

Gas-Cooled Fast Reactor (GFR)

Lead-Cooled Fast Reactor (LFR)

Molten Salt Reactor (MSR)

Supercritical-Water-Cooled Reactor (SCWR)

These GEN IV reactor design concepts are not addressed in the RES Infrastructure

Assessment document because they are in a preliminary stage of development and

there are some significant areas of research and development still to be addressed.

There are currently no vendor requests for NRC reviews of these designs.

Given the extensive set of potentially deployable reactors over the next two decades,

RES has a daunting task to chart a workable regulatory research plan that covers all

eventualities. Thus, the plan is expected to be a living document and will need to be

updated to reflect the listed domestic and international concepts in some coherent way.

The objective of the plan, given the significant uncertainties as to the course and timing

of domestic deployment efforts by industry, should be to position the agency to review

applications as they are received. RES will need to proactively acquire an early

indication of concepts that are moving towards the regulatory interface, and to put in

place mechanisms to revise plans and redirect funding in an appropriate and

responsive manner. Technologically neutral plans may be useful, but are likely to fall

short of the specificity needed by NRR to efficiently support licensing actions.

39 NUREG-1635
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1 REPORT NUMBER

1102. EAssigned by NRC, Add Vol, Supp, Rev..

iuRl:t 1102, and Addendu Numbrs IU any.)

3201 32M BIBLIOGRAPHIC DATA SHEET

(See kinatruns oan he awwrse)

2. TITLE AND SUBTITLE NUREG-1635, Volume 5

Review and Evaluation of the Nuclear Regulatory Commission 3. DATE REPORT PUBLISHED

Safety Research Program MONTH | O

A Report to the U. S. Nuclear Regulatory Commission June 2003

4. FIN OR GRANT NUMBER

S. AUTHOR(S) 6. TYPE OF REPORT

Annual

7. PERIOD COVERED 7rxkAvw Dates)

Fiscal Year 2003

8. PERFORMING ORGANIZATION -NAME AND ADDRESS (of NRC, proAd Amn. Off= orRegion. US. AiudearRegultwy omnwssi and ing address, 9contraolor, provide nam and mam ing euss)

Advisory Committee on Reactor Safeguards

U. S. Nuclear Regulatory Commission

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION -NAME AND ADDRESS (fARC, gx 'Swmesabmv'1f on&;&dor, prol NRCOWwon. Offimc orRegson, US. NudearnReguatory Cmnisson.

san maifing address.)

Same as above

1U. bUrrLztlN MIgT NUI tI

1 1. ABSTRACT (200 words or lass)

The objective of this report by the Advisory Committee for Reactor Safeguards (ACRS) is to provide comments and advice on

the Office of Nuclear Regulatory Research (RES) document "Advanced Reactor Research Infrastructure Assessment"

(Reference 1), which identifies gaps in technology that need to be filled prior to the potential certification of a number of new

reactor designs. These designs range from those which have evolved from the current boiling and pressurized water reactor

designs. i.e., advanced light water reactor (ALWR) designs, to those using significantly different technology, e.g., high-temperature gas-cooled reactors (l-ITGRs). The range of designs addressed in this assessment are representative of the

designs vendors have expressed an interest in submittingg to NRC for certification. We note that the U.S. nuclear industry has

not constructed a new reactor for over two decades and has focused on the operational and services aspects of the existing

reactor fleet. As a result, there are significant regulatory, management, and technical challenges to be faced with the

certification, construction, and deployment of new reactor designs, especially those based on significantly different technology.

This ACRS report concentrates on reactor safety issues. It does not cover items relating to nuclear materials and waste safety

or to safeguards. Safeguards will be addressed in a separate ACRS activity. We have not commented on the several

management issues that are discussed in the RES Infrastructure Assessment but recognize that these Issues are important.

12. KEY WORDSIDESCRIPTORS (wist ads orphrses At wifassist researchers in Acing the repor.) 13. AVAILABILITY STATEMENT

Unlimited

Advanced Reactor Research Reactor Operations 14. SECURflYCLASSIFICA1ION

Nuclear Reactors Safety Engineering (mis page)

Nuclear Reactor Safety Safety Research Unclassified

Reactor Designs

(ThisRep-od)

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2.89)
Federal Recycling Program
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