.lill-
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A.
NUREG-1635, Vol. 7

Review and Evaluation of the

Nuclear Regulatory Commission

Safety Research Program

A Report to the

U. S. Nuclear Regulatory Commission

Manuscript Completed: March 2006

Date Published: May 2006

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

U. S. Nuclear Regulatory Commission

Washington, DC 20555-0001
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ABSTRACT

This report to the U.S. Nuclear Regulatory Commission (NRC) presents the observations and

recommendations of the Advisory Committee on Reactor Safeguards (ACRS) concerning the NRC

Safety Research Program being carried out by the Office of Nuclear Regulatory Research (RES).

These observations and recommendations focus on that portion of the NRC research program

dealing with the safety of existing nuclear reactors and advanced light water reactor designs, such

as the Economic Simplified Boiling Water Reactor (ESBWR) submitted for certification. The

research strategy for more advanced reactors that are not based on water reactor technology such

as the Generation IV reactors being studied by the Department of Energy is also discussed. In its

evaluation of the NRC research activities, the ACRS considered the programmatic justification for

the research as well as the technical approaches and progress of the work. The evaluation

identifies research crucial to the NRC missions. The ACRS also attempts to identify research that

had progressed sufficiently to meet current and anticipated regulatory needs so that it could be

curtailed in favor of more important activities. This report does not address research on the security

of nuclear power plants. Comments on such research will be reported separately. Also, the ACRS

does not comment on research activities dealing with nuclear waste issues. The Advisory

Committee on Nuclear Waste (ACNW) will report on these research activities.

iii
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TABLE OF CONTENTS

Page

ABSTRACT ......................... iii

TABLES ........................ vi

ABBREVIATIONS................................................................................................... vii

1 Introduction .

2 General Observations and Recommendations .3

3 Advanced Reactor Research .11

4 Digital Instrumentation and Control Systems .15

5 Fire Safety Research .19

6 Reactor Fuel Research .23

7 Neutronics and Criticality Safety .25

8 Human Factors and Human Reliability Research .27

9 Materials and Metallurgy .31

10 Operational Experience ........................ 39

11 Probabilistic Risk Assessment ........................ 43

12 Seismic Research ........................ 49

13 Severe Accident Research ........................ 51

14 Thermal-Hydraulics Research ........................ 57

15 References ........................ 65

V
TABLES

Page

1. Advanced Reactor Research Activities................................................................1...........12

2. Research Activities in Digital Instrumentation and Control Systems .18

3. Fire Safety Research Activities .22

4. Reactor Fuel Research Activities ........................................................ 24

5. Research Activities in Neutronics Analysis, Core Physics, and Criticality Safety .26

6. Human Factors and Human Reliability Research Activities ........................................... 30

7. Research Activities in Materials and Metallurgy .35

8. Research Activities in Operational Experience .41

9. Probabilistic Risk Assessment Research Activities .45

10. Seismic Research Activities.. 50

11 Severe Accident Research Activities .54

12. Thermal-Hydraulics Research Activities .61

vi
, X ABBREVIATIONS

ACRS

ACNW

ACR-700

ANL

ANS

ASCE

ASME

ASP

ATHEANA

ATWS

BWR

CAMP

CFR

CRDM

CSARP

DOE

ECCS

EMI

EPIX

EPR

EPRI

ESBWR

FY

GDC

GSI

HERA

HRA

HSST

l&C

IAEA

IASCC

ICET

IEC

IEEE

INL

INPO

IPEEE

IRIS

ITAAC

LANL

LERF

LOCA

LPSD

LWR

MACCS

MOX

Advisory Committee on Reactor Safeguards

Advisory Committee on Nuclear Waste

Advanced CANDU Reactor-700

Argonne National Laboratory

American Nuclear Society

American Society of Civil Engineers

American Society of Mechanical Engineers

Accident Sequence Precursor

A Technique for Human Event Analysis

Anticipated Transients Without Scram

Boiling Water Reactor

Code Applications and Maintenance Program

Code of Federal Regulations

Control Rod Drive Mechanism

Cooperative Severe Accident Research Program

Department of Energy.

Emergency Core Cooling System

Electro Magnetic Interference

Equipment Performance and Information Exchange System

Evolutionary Power Reactor,

Electric Power Research Institute

Economic Simplified Boiling Water Reactor

Fiscal Year

General Design Criterion

Generic Safety Issue;

Human Event Repository and Analyses

Human Reliability Analysis

Heavy Section Steel Technology

Instrumentation and Control

International Atomic Energy Agency

Irradiation Assisted Stress Corrosion Cracking

Integrated Chemical Effects Tests

International Electrotechnical Commission

Institute of Electrical and Electronics Engineers

Idaho National Laboratory

Institute of Nuclear Power Operations

Individual Plant Examination of Extemal Events

International Reactor Innovative and Secure

Inspections, Tests, Analyses, and Acceptance Criteria

Los Alamos National Laboratory

Large Early Release Frequency

Loss-of-Coolant Accident

Low Power and Shutdown

Light Water Reactor

MELCOR Accident Consequence Code System

Mixed Oxide

vii
ABBREVIATIONS (Cont'd)

NDE Non-Destructive Examination

NEA Nuclear Energy Agency

NEI Nuclear Energy Institute

NFPA National Fire Protection Association

NIST National Institute os Standards and Technology

NRC Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation

NSIR Office of Nuclear Security and Incident Response

OECD Organization for Economic Cooperation and Development

PARCS Purdue Advanced Reactor Core Simulator

PBMR Pebble Bed Modular Reactor

PIRT Phenomena Identification and Ranking Table

PRA Probabilistic Risk Assessment

PSF Performance Shaping Factor

PSHA Probabilistic Seismic Hazard Analysis

PTS Pressurized Thermal Shock

PUMA Purdue University Multidimensional Integral Test Assembly

PWR Pressurized Water Reactor

RCS Reactor Coolant System

RES Office of Nuclear Regulatory Research

RG Regulatory Guide

ROP Reactor Oversight Process

SDP Significance Determination Process

SMIRT Structural Mechanics in Reactor Technology

SNAP Symbolic Nuclear Analysis Package

SPAR Standardized Plant Analysis Risk Model

SRM Staff Requirements Memorandum

SSHAC Senior Seismic Hazard Analysis Committee

TRACE TRAC-RELAP Advanced Computational Engine

UNM University of New Mexico

U.S. United States

Iviii
I 1 INTRODUCTION

The Nuclear Regulatory Commission (NRC)

maintains a Safety Research Program to:

* Ensure its regulations and regulatory

processes have sound technical

bases.

* Prepare for anticipated changes in the

nuclear industry that could have

safety implications.

: A.-* .

Develop improved methods to carry

out its regulatory responsibilities.

* Maintain an infrastructure of

expertise, facilities, analytical

capabilities, and' data to support

regulatory decisionmaking.

These essential missions for the research

effort were defined when the NRC was

established and there was limited experience

with the operation of light water nuclear

power plants. The need for research remains

today, despite the growth of experience with

existing power plants, because:

* Nuclear power plants age and

encounter challenges -of material

degradation not anticipated when the

plants were designed.

* The NRC considers applications for

extending licenses, uprating the

operating power levels of plants, and

new plant licenses.

* Reactor fuels are used to higher

levels of fuel bumup and new

cladding alloys for the fuels are

introduced.

* Mixed-Oxide (MOX) fuel is

considered for the disposal of excess

weapons-grade plutonium.

* The NRC evolves its regulations from

a deterministic foundation to a riskinformed basis that makes ever

greater use of best-estimate analyses

to assess safety.

* New technologies including softwarebased digital instrumentation and

control (I&C) systems are backf it into

the existing nuclear power plants.

* New water reactor designs such as

the ESBWR, which uses passive

systems, have been submitted for

certification.

There are on the horizon new power reactor

concepts that are not based on the water

reactor technologies used in the current fleet

of power reactors. The U.S. Department of

Energy is studying power reactors that use

gas cooling, liquid metal cooling, and molten

salt cooling. Reactors that use-fast rather

than thermal neutrons for fission are being

studied with the intent of development. These

new reactors -make it important for the NRC

to consider evolution of its regulatory system

from one that is specific to water reactor

technologies to one that is not specific to

particular reactor technologies, but still lead

to adequate protection of the public health

and safety. This will require substantial

research not only for the early development

of technology-neutral regulations, but also, in

the longer term, for the development of

technology-specific regulatory guidance and

plans for reviewing specific license

applications.

In this report, the Advisory Committee on

Reactor Safeguards (ACRS) presents its

observations and recommendations

concerning that portion of the NRC Safety

Research Program devoted to regulation of

existing light water reactors (LWRs) and the

certification of advanced water reactor

I NUREG-1 635
designs submitted for certification such as

the ESBWR. The ACRS also makes

observations on the need for research in

anticipation of more advanced power reactor

concepts. Observations and

recommendations on research dealing with

the security of existing nuclear power

reactors and nuclear facilities will be provided

in separate reports and are not discussed

here. The ACRS does not comment on

research activities dealing with nuclear waste

issues. The Advisory Committee on Nuclear

Waste (ACNW) will address such research

separately. -

In its review of the NRC Safety Research

Program, the ACRS considered the

programmatic justification for the research as

well as the technical approach and progress

of the work. The ACRS supports research

that:

- Provides support to the identification

and resolution of current safety and

regulatory issues.

* Provides the technical basis for the

resolution of foreseeable safety

issues.

* Develops the capabilities of the

agency to independently review risksignificant proposals and submittals

by licensees and applicants.

* Supports initiatives of the agency

such as the development of

T echnology-neutral' regulatory

systems.

* Improves the efficiency and

effectiveness of the regulatory

process.

* Maintains technical expertise within

the agency and associated facilities in

disciplines crucial to the -agency

mission and that are not readily

available from other sources.

This review of the NRC Safety Research

Program identifies some research activities

that have made valuable contributions to the

agency mission in the past, but now have

reached the point where additional research

is not needed for efficient and effective safety

regulation. This review also -identifies

research activities that could benefit by

greater collaborationwith research activities

elsewhere in the world, including

collaboration with researchers in Asia and

Europe.

General observations and recommendations

concerning NRC research activities are

presented in Chapter 2. Observations and

recommendations regarding research

activities in specific technical disciplines are

discussed in detail in Chapters 3 through 14:

* Advanced Reactor Research

* Digital Instrumentation and Control

- Systems

* Fire Safety Research

* Reactor Fuel Research

* Neutronics and Criticality Safety

* Human Factors and Human Reliability

Research

* Materials and Metallurgy

* Operational Experience

* Probabilistic Risk Assessment

* Seismic Research

* Severe Accident Research

* Thermal-Hydraulics Research

NUREG-1635 2
2 GENERAL OBSERVATIONS AND RECOMMENDATIONS

The NRC Safety Research Program is largely

focused on addressing near-term regulatory

needs of the agency. Currerit activities are

especially concentrated in three disciplines:

* Materials and Metallurgy

* Probabilistic Risk Assessment

Thermal Hydraulics

This is an appropriate focus of'the current

NRC research activities. These activities are

discussed further below along with 'other

major aspects of the research program.

The incident at the Davis-Besse Nuclear

Power Plant has emphasized, among other

things, how important it is for the agency to

have a better understanding of the corrosion

of metallic systems in the aging fleet of

currently operating nuclear power 'plants.

Aging degradation research is necessary to

ensure effective 'aging management for

plants operating for extended periods' under

license renewal and to assess the effect that

operation under extended power uprate

conditions may have on margins against

degradation. Continued challenges posed by

stress corrosion cracking of steam generator

tubes in pressurized water reactors (PWRs)

and systems within boiling water reactor

(BWR) vessels further support such focus in

the research effort.

Probabilistic risk assessment (PRA) Is the

basic technology for the risk-informed

regulatory' system envisaged by the

Commission. Research activities are focused

now largely on the application of current PRA

technology to reactor regulation through the

Reactor Oversight Process (ROP). PRA

insights are essential to develop and

implement revisions to such - central

regulations as 10 CFR 50.46. They also will

play a key role in the development ,of

"technology-neutrar regulatory systems that

will have applications to power reactors that

are not based on the LWR technology used

in the current fleet of operating plants.

The Standardized Plant Analysis Risk (SPAR)

models are fundamental tools for riskinformed regulation. A stronger commitment

should be made to the improvement of these

models and their extension on a timely basis

to include fire, seismic, and shutdown risks.

The development of these capabilities for the

SPAR models will not only provide a

regulatory capability but will also encourage

industry to more aggressively develop their

own capabilities in these areas.

The'quality of PRA results depends on good

phenomenological models and there are

important areas where such models still need

further development. Approximate and often

bounding risk analyses done for individual

plants suggest that the risk of core damage

as a result of events initiated by fires can be

comparable to risks from other accidents

initiated during normal operations. It is

important to know if similar results would also

be obtained using fire risk assessments of

sophistication comparable to the risk

assessments'possible for normal operations.

Such a finding would have ramifications on

both regulatory attentions and licensee

attentions to safety. The ACRS continues to

believe that' based on the potential risk

significance of fires, fire safety research

merits -strong consideration in the NRC

research program.' The collaboration with

Electric Power Research Institute (EPRI) is

providing a good understanding of the current

state-of-the-art methodology for fire risk

assessment. This work provides a basis for

determining;'-the need for further

development.' -

Thermal hydraulics'is a fundamental feature

of safety analyses of nuclear power plants.

3 NUREG-1 635
The NRC allows licensees to do either

bounding or best-estimate analysis of plant

thermal hydraulics for design basis accidents.

Confirmatory review of licensee analyses

requires that the agency have high quality

thermal-hydraulic analytical tools. Need for

such tools is even greater for the analysis of

advanced light water reactors that rely on

passive systems to achieve safe

configurations following accidents.

NRC has consolidated several models of the

thermal-hydraulic transient analysis codes

into a single code called TRACE. The

TRACE code should be subjected to an

independent technical review to assess its

range of validity. The TRACE code then

should be at a point at which it can be used

as the primary thermal-hydraulic tool for

regulatory analyses. -A plan should be

developed for its integration into the

regulatory -process. This integration will

require strong support from the management

of the NRC user organizations since such a

change in the short run will create additional

burden on the staff.

The potential for blockage of sump screens

by debris dispersed into the sumps during

depressurization of the reactor coolant

system during an accident remains an

unresolved issue. The complexity of the

interactions between fibrous and particulate

debris, as well as the chemical interactions

that can occur among debris materials and

solutes in the coolant, make-predictions of

blockage and consequently screen size

requirements difficult Research needed to

reach a prompt resolution of this issue should

receive the required resources.

International Collaboration

Reactor, safety, is an -international

undertaking. It is important that there not be

great differences in safety regulations among

the nations making major use of nuclear

power generation. The NRC research is

making good use of collaborations with other

countries on reactor safety research. Much of

this collaboration has been in the nature of

information exchange. Such information

exchanges are important and should continue

to be encouraged and supported. They

provide access to information and a kind of

peer review that might not otherwise be

obtained. However, there are other important

cases where NRC has gone farther and

formally partnered with other countries to

leverage resources for experimental

investigations of important reactor safety

research issues. Such collaborations are

especially noteworthy in the disciplines of

reactor fuel research and in severe accident

research. The combined resources of the

partners in these collaborations are yielding

higher quality and more extensive results

than would be possible in research programs

sponsored by individual countries.

Other areas of NRC research could benefit

from more extensive collaborations. Such

areas include fire safety research and

thermal-hydraulics research. The benefits of

such collaborations become more apparent

as NRC moves to more realistic analyses

which may require validation by costly largescale, integral tests. Collaborations of this

type may become even more important in the

future as new types of reactors are proposed

for certification internationally. To be effective

and efficient in dealing with future challenges,

NRC should look for opportunities to increase

significantly collaboration with other countries.

The ongoing collaborative efforts are very

extensive with European countries. More

collaboration Iwith Asian countries having

active nuclear power plant programs should

be pursued.

Support for Future Ucensing Activities

There has been a recent resurgence in

interest in the use of nuclear reactors for the

generation of electrical power., Innovative

reactor designs are being suggested to

NUREG-1 635 4
sustain uranium resources and to generate

electrical energy at much greater efficiency.

The U.S. Department of Energy is studying

very high temperature .gas reactors,

supercritical water reactors, sou1ium-cooled

reactors, lead-bismuth cooled reactors, and

molten salt cooled reactors. Some of these

reactors will use fast neutrons rather than

moderated neutrons for fission. These

reactors use technologies quite different than

those used for the currently operating fleet of

reactors. The current regulatory framework is

not well suited for the licensing, regulation, or

monitoring of such different reactor

technologies. Several years ago, it appeared

that a substantial portion of NRC resources

might need to be devoted to the development

of the capabilities to address these very

advanced reactor technologies. Today, this

is not the case. NRC advanced reactor

research resources are focused on

addressing issues associated with advanced

water reactors such as the ESBWR and the

EPR.

This seems to be an appropriate use of

NRC's limited resources for advanced reactor

safety research. Very advanced reactor

concepts have not reached a sufficient state

of development that productive use of

regulatory research resources can be made.

However, work should continue on the

development of a technology-neutral

framework for regulation, although the

development of technology- specific guides

can be delayed until it is clearer which

alternate reactor technologies will be of the

greatest interest.

Development of the framework is not only

important for the licensing of non-light-water

reactors, but also may provide insights that

are useful in developing a more efficient

regulatory program for advanced reactors of

all types.

There are some indications that certifications

may be sought for advanced designs with

minimal experimental study of plant response

under accident conditions. NRC needs to

provide clear guidance on its expectations for

the experimental validation of computer

models used in the licensing of advanced

reactors that do not use familiar technologies.

Development of such guidance is an area of

advanced reactor research that can be

pursued at relatively low cost, but which can

play an important role in timely and efficient

licensing of advanced reactors with new

technologies.

Opportunities for Independent Research

In recent years, a strong effort has been

made to ensure that NRC research is

supportive of the needs of the line

organizations. Focusing NRC research

entirely on the immediate needs of the line

organizations does, however, entail an

important risk. This focus reduces the

opportunities for independent thought by the

research staff and the opportunities to

conduct research that could make more

dramatic improvements in the regulatory

process, for example, in the tools that support

it at a time when there is a rapid increase in

workload. The risk is magnified by the

diversion -,of so much research talent to

address issues of security of nuclear facilities.

There is the further risk of a loss of prestige

in the research program focused as it is on

issues of implementation. This could

eventually lead to a loss in the credibility of

the technical basis that underlies regulatory

decisions.

It is important that NRC research stay abreast

of technological developments that can

enhance safety. Areas where developments

in the larger technical community can be

important to the NRC include reactor fuels,

corrosion and materials degradation, manmachine interfaces, technologies for

monitoring component performance,

inspection techniques, and virtual facility

inspections. Where NRC can adopt or adapt

5 NUREG-1 635
developments in other industries, safety can

be improved and the efficiencies of NRC

reviews enhanced.

One mechanism for RES to interact with the

larger technical community is by sharing its

own research plans. This has been done for

research into digital instrumentation and

control. Investigators did creditable reviews of

the state-of-the-art, presented them at

appropriate professional society meetings as

a kind of public peer review, and developed

from these state-of-the-art reviews a research

plan that is well directed to address agency

needs. Sharing research plans with a larger

technical community is a strategy that would

benefit other NRC research activities. Such

interactions also help provide visibility for and

help sustain the prestige of the NRC research

program.

A Vision for the Future

Nuclear energy will remain an important and

perhaps growing component in the mix of

energy generating technologies used in this

Country. There is the potential that many new

reactors could be built in the next 15 to 20

years. It is unlikely that agency resources of

either manpower or funds will experience a

similar growth. Indeed, the experience level

of the NRC staff is likely to decrease due to

retirements just when the new plant licensing

activities accelerate. A portion of the

research program needs to be devoted to the

development of a regulatory infrastructure for

regulatory work in the next 20 years that

supports a staff with less experience dealing

with more tasks. Computerization will be

undoubtedly an important element of such an

infrastructure. The ACRS can foresee, for

example, a time when regulatory staff have

routine access to superior analysis tools for

systems analysis, phenomenological

analysis, and risk assessment. Development

of such validated and verified tools for routine

use by non-specialists will require a research

program that is not tied exclusively to the

near-term issues of the regulatory process.

Appropriate attention will have to be paid to

the agency's analytical tools, its access to

facilities, and its ability to provide recently

recruited staff with a sound understanding of

past safety decisions. Availability of good

infrastructure will enhance safety and allow

for much more efficient and effective NRC

review of new reactor designs and licensing

applications based on realistic evaluations of

safety.

Observations and Recommendations on

Specific Research Activities

NRC research has made substantial progress

since the last ACRS 'report, NUREG-1635,

Vol. 6, on the research program. -This

progress has occurred despite the diversion

of substantial research talent in the agency to

address issues of reactor security that are not

reviewed here. Notable accomplishments of

the research program in recent years include:

* Multidisciplinary review of pressurized

thermal shock criteria

* Performance of high-burnup fuel

during reactivity transients

* Embrittlement of zirconium alloy

cladding when taken to high burnup.

The ACRS applauds these high technical

quality research accomplishments. The

ACRS is, however, disappointed at the pace

with which these important research results

are being used to modify regulations.

Other major observations and

recommendations concerning the NRC

research activities are summarized below and

also discussed in more detail in individual

Chapters.

* NUREG-1635 6
Advanced Reactor Research

Highest priority should be given to those

research activities that support the ESBWR

design certification process. The irriportance

of tasks associated with the ACR-700 or a

related design with higher power depends on

whether the certification review for such a

reactor is resumed.

Digital Instrumentation and Control Systems

Software-based digital electronic systems are

inevitable for both current and-: more

advanced design nuclear power plants. The

staff has developed a research plan that

addresses the challenges associated with the

use of digital technology that will face the

agency in the next five years.

The ACRS has recently reviewed and

reported -favorably on the research plan for

digital -systems. The ACRS was impressed by

the technical quality in the development of the

research plan, the scope and content of the

plan, and the prioritization-of activities in the

plan. The ACRS -recommends a number of

improvements to an already quality research

plan, including addition of an explicit element

to the plan to study the acceptability of

international standards in comparison to

Institute of Electrical and Electronics

Engineers (IEEE) standards for meeting

regulatory requirements concerning digital instrumentation and control systems. This

study will be an important element of efforts

to develop a multi-national design approval

process.

Fire Safety Research

There have been a number of important

accomplishments by NRC research-in the

area of fire protection since the last ACRS

report on NRC safety research program in

2004. Fire safety research continues to merit

emphasis in the NRC research program.

RES, in cooperation with EPRI has taken

some important steps to consolidate the fire

PRA research and development activities,

conducted over the past few years, into a

single state-of-the-art methodology for fire

risk assessment.

There are a variety of methods that can be

used to model the progression of fires. Some

of these have been used in fire protection

programs for non-nuclear facilities for many

years. The ranges of applicability of these

methods have not been well studied or

documented. In cooperation with EPRI, a

program is in progress to verify and validate

a set of fire progression modeling tools. The

accuracies of these tools are being examined

for different fire conditions and applications

by comparison with benchmark tests.

RES has worked closely with the U.S.

industry in undertaking generic fire risk

research activities. Fire risk is, however, an

issue of world-wide concern. RES has not

aggressively sought collaborations with the

international community to advance NRC

capabilities for- fire risk assessment.

Collaborations with other countries especially

in experimental studies may be essential to

leverage resources of all partners sufficiently

to achieve fire risk assessment capabilities

commensurate with what can now be done

for risk from normal plant operations.

Reactor Fuel Research

The NRC research on reactor fuel has been

concentrated in recent years on the

confirmation of regulatory decisions that allow

:licensees to take light water reactor fuels to

burnups of nominally 62 GWd/t. The

research on high-bumup fuel is reaching a

substantial level of maturity.. Some major

confirmatory experiments remain to be done - notably experiments on reactivity insertion to

be done in a water loop at the CABRI reactor.

Since the last ACRS report on NRC safety

research program, plans for these

7 7 NUREG-1635
experiments have been revised so the

experiments which are part of an international

collaborative effort now better meet the

agency needs. It is important that this work

that is so well coordinated both with agency

needs and with international partners be

taken to completion. Still major findings of

the research effort can be -reduced to

regulatory practice now. This reduction to

regulatory practice needs to be initiated and

pursued aggressively.

It is evident that high-bumup fuel research

will soon achieve results that are adequate

for agency needs. The NRC has made clear

that it will expect the nuclear industry to

provide necessary safety analyses and

experimental data should the industry want to

take fuel to bumups that exceed the current

regulatory maximum. NRC needs to make

these expectations more explicit, particularly

its expectations for the experimental data

needed to support the analyses of highbumup fuel behavior under accident

conditions.

Neutronics and Criticality Safeti

The neutronics and criticality safety research

program is small but appears adequate to

ensure that the NRC has capabilities to meet

immediately foreseen regulatory needs.

In the future, more innovative core designs

for advanced reactors may be submitted to

the NRC. Confirmatory analyses of reactor

core physics will be an essential part of the

regulatory process for these advanced

reactors. The capabilities now available to the

NRC in the area of core physics may well be

stretched. It will be useful to the agency to

understand these future needs. If long-term

development activities are identified, such as

those that might be needed for analysis of the

PBMR, additional research may be needed in

this area.

Human Factors and Human Re/iabilit'

Research

As new reactor designs, likely dependent on

a higher degree of automation than the

current fleet, are introduced, the need for

revised guidance and tools for the NRC staff

in human factors and human reliability

analysis will increase. RES- has initiated a

project to develop regulatory guidance and

analytical techniques to review human factors

for advanced nuclear power plants. The

ACRS views this five-year project essential

for preparing the staff in reviewing advanced

reactor designs.

The quantification of human reliability

continues to be a challenge in risk

assessments. Human reliability modeling

introduces large uncertainties in probabilistic

risk assessments. The NRC staff needs

guidance in its review of the human reliability models used by the industry in risk-informed

licensing applications. Progress has been

made with the publication of NUREG-1,792,

.Good Practices for Implementing Human

Reliability -Analysis (HRA)." Still, further

guidance is needed for reviewers of licensing

applications.

Materials and MetallurcN

The NRC is investing heavily in the better

understanding of materials degradation

issues in the currently operating fleet of

nuclear power plants. Such investment is

justified in view of significant agency

regulatory activities that aging degradation

research supports.

The current program is well focused on

improving the agency's ability to

independently evaluate licensees' efforts to

prevent, detect, and mitigate environmentally

assisted stress corrosion cracking.

-NUREG-1 635 -8
The nuclear industry and the NRC have often

been surprised by unexpected material

degradation problems. As a result, they have

responded to such problems in a reactive

mode which has proven to be inefficient. The

Proactive Materials Degradation Assessment

project is an effort to identify potential

material degradation problems before they manifest in operating nuclear power plants.

The ACRS admires the vision of this

undertaking and supports its continuation.

The ACRS looks forward to reviewing the

initial results of this ongoing effort soon and

learning whether the admirable goal of this

project is, in fact, feasible.

RES needs to reevaluate the need for

continued research into heavy section steel

components. This research may be justified

if there is a clear need for NRC to develop its

capabilities in the area of probabilistic fracture

mechanics (PFM) so that it can evaluate

licensees' applications. If this is the case, the

research needs to be clearly focused on this

objective and not the research that the

industry should perform to meet its

responsibilities to ensure reactor pressure

vessel integrity. It appears now, however, that

it is NRC that is advancing the state-of-the-art

and making available information that allows

licensees to reduce conservatism in their

analyses.

Operational Experence

The ACRS is supportive of the research

activities in the area of operational

experience and recommends that these

activities be continued. In light of the limited

resources allocated to these tasks, RES has

done a commendable job in producing

outputs in well-documented and thorough

fashion. Tasks that are currently in the 2005

research plan related to operational

experience should remain funded and should

be continued for the foreseeable future.

Probabilistic Risk Assessment

Altogether the scope and the number of

activities in the NRC's PRA research program

is quite impressive. The ACRS cautions,

however, that NRC should not allow its work

in such- a crucial technology as risk

assessment to become totally devoted to the

support of line activities. Methods

development -is still important. As an

example, the ACRS notes that considerable

research is being reported in the literature

regarding Binary Decision Diagrams as tools

for solving large fault tress without resort to

cutoff frequencies as is now done. The staff

needs to review the literature concerning

Binary Decision Diagrams and evaluate the

need to adopt this technology. The growing

importance of the SAPHIRE code and the

SPAR models in the regulatory process

warrants such an investigation.

Seismic Research

Seismic hazard analysis and structural

response are not areas where NRC must

maintain state-of-the-art expertise. Such

expertise is available to the NRC on a

contractual basis. As ACRS noted in its

previous report on NRC safety research

program, research activities at the agency

can be confined to support needed updates

to regulatory guides and collaborative work

with the international community to stay

abreast of developments in other countries.

The current research program is, indeed,

largely focused on needs of the regulatory

process and a few important international

collaborations.

Severe Accident Research

The ACRS is very supportive of the strategy

NRC has developed to maintain and update

its capabilities for severe accident analyses.

The leveraging of resources through

international collaborative experimental

research is especially important. The planned

9 NUREG-1 635
extensions and continuations of current

collaborations are well worth the investment.

Thermal-Hydraulics Research

Highest priority should be given to the

integration of TRACE code into the regulatory

process. As this integration progresses, the

research staff can continue its efforts to

improve and further develop TRACE on a

"time available" basis. The ACRS is

concerned now that efforts to improve

TRACE lack prioritization and defensible

organization., Prioritization of technical

improvements might be aided substantially by

commissioning a detailed peer review of

TRACE. To do this, the staff will have to have

available code documentation of outstanding

scope and quality. Such high quality code

documentation will also be needed if the code

is to become part of the regulatory process.

Code documentation, then, is a task that

ought to take precedence in the thermalhydraulic research effort.

. . .

'

. .

, .

. . .

:

NUREG-1 635 ., 10
3 ADVANCED REACTOR RESEARCH

The agency is already engaged in various

activities related to a number of new plant

designs, including ESBWR, PBMR, IRIS, and

ACR-700; The staff has begun its review of

ESBWR design certification application. It is

anticipated that requests for design

certification'reviews will be received for EPR,

and PBMR. Of these, the ESBWR, ACR-700,

IRIS, and EPR can be certified in all

likelihood under the current requirements in

10 CFR Part 52 using the design basis

accidents as they are' now defined.

Nevertheless, there will be the need for NRC

to verify the thermal-hydraulic assessments

made by the! applicants.: for the various

designs. This will require review and approval

of the computer codes that were used by the

applicants for assessing the design basis

accidents. Confirmatory analyses will require

that design-specific versions of the computer

codes TRACE and CONTAIN be available to

the staff for audit calculations and

independent assessment of separate effects

and integral system experiments.' Highest

priority should be given to those research

activities that make such tools available for

the ESBWR design certification review. This

includes tasks Y6857, Y6898, N6018, and

Y6804. The importance of tasks associated

with the ACR-700 or a related design with

higher power, Y6831, Y6812, Y6899, Y6489,

Y6748 and Y6933, depends on whether the

certification process for such a reactor is

resumed.

Certification reviews for designs such as the

PBMR and the 4S that do not use water

reactor technology will be more challenging.

Although significant efforts were undertaken

in the past to license such non-LWR designs

under the current regulatory system designed

for light water reactors, it would be far more

appropriate, effective, and efficient to have

the "technology-neutral-framework' for

certification of such designs. For timely

application to these reactor types (and

possibly even more unusual designs in later

years), the development of the technology- neutral framework needs to be given high

priority and provided sufficient resources to

complete the job in 2006 and to allow two

years for rulemaking.'. High priority,'.' then,

should be given to the tasks N6205 and

Y6487 that will develop a technology-neutral

framework- for the 'regulation of advanced

nuclear power plants.

The Commission has expressed a' desire for

"enhanced safety" for new reactor designs. To

ensure that new designs have reached

enhanced levels of' safety, the NRC will

require each of the applicants for design certification to submit a full-scope PRA with

consideration of uncertainties. The staff must

be prepared to review these PRAs, to

validate the results and to compare the results

with acceptance criteria for "enhanced

safety." This evaluation will include

undoubtably a complete Level-2 evaluation of

accident source terms since LERF (large early

release frequency)' will no longer be an

appropriate safety metric. To review and

independently assess the Level-2 analyses of

source terms, the regulatory organizations will

need design-specific versions of the MELCOR

computer code. There is, then, the potential

need to develop MELCOR versions specific

for the PBMR and 4S designs. Development

of such code versions will take time. Second

priority should be given then to tasks K6703,

Y6801, and Y6619. Again, the importance of

developing an accident progression model for

ACR-700 depends on resumption of its

certification process.

11 11 NUREG-1635
Table 1. Advanced Reactor Research Activities

Job Code Title Comment

Y6857 ESBWR Input Deck Analysis of DBAs in ESBWR using the

Development TRACE code; This project should have high

priority.

Y6898 ESBWR Design Certification Support for review of PRA for ESBWR;

_ _ Report This project should have high priority.

N6018 Separate Effects Separate effects tests in support of TRACE

Experiments model development for ESBWR;

This project should have high -priority

Y6804 ESBWR Containment Analysis of experiments with CONTAIN and

Support MELCOR. This is a high priority task for

_ _ ^ESBWR __ design certification review.

Y6489 PRA forACR-700 Support for review of ACR-700 PRA. This

project can be deferred until certification

D__ -application becomes active again.

Y6899 ACR-700 Design Certification Support for review of PRA for ACR-700. This

Support project can be deferred until the certification

-_ application becomes active again.

Y6748 Review ACR-700 Support Support for thermal hydraulics review of ACR700. This project can be deferred until the

certification application becomes active again.

Y6831 Methods Development for TRAC code upgrades needed for ACR-700

ACR-700 certification calculations. This project can be

deferred until the certification application

becomes active again.

Y6812 ACR-700 Input Model Develop RELAP5 and TRAC-M input models

Development for ACR-700. This project can be deferred

until the certification application becomes

active again.

Y6933 Evaluate Severe Accident Analysis of risk dominant sequences for ACRPhenomena In ACR-700 700. This project can be deferred until the

certification application becomes active again.

K6703 Coop. Agreement with Improve NRC's knowledge and information on

Center for Advanced Nuclear advanced reactors. This project is useful but

Energy Systems can have a second level priority.:

Y6619 Advanced Reactor PRA Develop knowledge needed to review

Development advanced reactor PRAs. Second priority work

for non-LWR design certifications.

I NUREG-1 635 12
Table 1. Advanced Reactor Research Activities

(Continued)

Job Code Title Comment

Y6801 Advanced Reactor/Severe Develop a version of MELCOR code for

Accident Code advanced reactors. This project can have a

Development second level priority.

Y6755 Materials Evaluations for Research materials engineering issues for

Advanced LWR Reactors advanced LWRs especially effect of coolant

environment on fatigue and in-service

inspection and monitoring. This project can

have a second level priority.

N6205 Assistance for Development Development of a technology-neutral

of a Regulatory Structure for regulatory framework. This project should

New Plant Licensing have high priority.

Y6487 Advanced Reactor Development of a regulatory framework for

Regulatory Framework advanced reactors. This project should have

Development high priority.

Y6741 Environmental Effect on Develop understanding of the properties of

Containment concrete in high temperature gas cooled

reactors. This project can have a low priority.

13 NUREG-1 635
:

I . I

I , I

; 7
4 DIGITAL INSTRUMENTATION AND CONTROL SYSTEMS

Software-based digital electronic systems are

inevitable for both current and advanced

design nuclear power plants. Already such

software-based digital electronics appear ever

more frequently'in systems for plant control.

Eventually, they will appear in safety systems. The reliability of digital systems especially when using commercial, "off the shelf"

hardware and software has become an issue

because they cannot be comprehensively tested. The quality of the requirements for the

software cannot be assessed fully through testing. Quality in the software-based systems is achieved through the control of the process of software development. Particular attention

has to be given to the requirements'f6r the

system software. 'Failure to specify adequate requirements has often been found to be the

root cause of digital system failures. Review

and approval of licensee applications to

incorporate software-based digital systems in

its facility is, then, time-consuming for both

the regulator and the licensee. New failure

modes that arise in digital systems need to be

recognized. Such failures can depend on the

operational state of the system at the time of

failure.' Indeed, testing and maintenance as

well as normal operationsbof digital systems

can 'create the opportunities for their own

unique kinds of failures.

Security of digital systems has become a

major concern and there needs' to be

regulatory guidance and acceptance criteria

for the security aspects of digital systems. .Codes,'' Standards, and regulations must

prompt the' designer of digital safety systems to avoid system communications outside of

the controlled areas of the plant and the use

of wireless technology must be carefully evaluated to prevent interception, interdiction,

or interference in communications to digital

systems.

Current licensing guidelines provide information on what -to review in digital

systems. They do not'necessarily provide sufficient guidance on how to review

submittals or the acceptance criteria to apply. The NRC staff needs a firm technical basis for

deciding when review of submittals is

adequate and when confirmatory analyses are

necessary. The situation will get worse with

time. Digital systems in nuclear power plants

are expected to become more numerous. The

complexity of these systems will increase.

There is the potential for the consolidation of

what are now discrete analog safety systems into a single digital system. At the same time, there is interest both within the agency and on

the part of licensees to adopt risk-informed

techniques for the review of digital software

systems. NRC lacks the technical basis to

support risk-informed reviews of digital systems. Currently, the ability, to model the

reliability of software-based digital systems in

PRAs is very limited. Without quantitative risk

information, a much less defensible,

qualitative, ugraded approach" to the review of

digital systems is likely to emerge.

'If the use of digital protection systems and

control systems becomes as widespread as

now predicted, review of digital systems as

partof ITAAC (Inspections, Tests, Analyses, and Acceptance Criteria) may eventually become a burdensome, time-consuming aspect of the licensing process. Methods and

tools to facilitate confirmation that "as built'

systems conform'to accepted designs are

going to be needed. As use of digital systems becomes more extensive in nuclear facilities,

NRC may find it necessary to reconsider its

current positions on defense-in-depth and

diversity in instrumentation and control

systems.

15 NUREG-1 635
The nuclear industry is not a major user of

digital technology relative to many other

industries. Yet, the consequences of failure of

digital systems in nuclear power plants are

likely to be less acceptable to the public than

are failures of such systems in other

industries even when consequences are

significant. Greater rigor in the review of

digital systems is necessary for nuclear

applications of these systems. It is expected

then that NRC will have to ablaze new paths"

in this area through research. In particular, the

usual industrial practice" of 'separately

considering hardware and software reliabilities

may not be adequate for nuclear systems and

a more integrated or systems approach may

be needed.

The staff has developed a research program

plan'that addresses these challenges that will

face the agency in the next five years. Critical

reviews of the state of the art in several areas

were completed, documented and presented

before audiences in professional societies.

Recommendations made to the NRC' by

independent bodies, including the National

Academy of Sciences were considered in the

development of the plan. Inputs from the

program offices at NRC (NRR, NSIR, and

NMSS) were also obtained. The research plan

is well directed toward meeting the agency

needs and is intended to provide:

* Improved technical guidance for

review of digital systems

* Technical support for developing

improved acceptance criteria for

assessing the safety and security of

the systems

* Tools and methodologies for

improved review of digital systems

* Technical bases for including models

of digital systems in PRAs

The research plan has six major elements:

* Systems aspects of digital technology

* Risk assessment of digital systems

* Emerging digital technology with

application to nuclear facilities

* Software quality assurance

* Security aspects of digital systems

* Advanced nuclear power plant digital

systems

Within each of these major'elements of the

plan, there are a number of subelement. The

staff has prioritized work on the subelement

basis. Now, the' major focus of the work is on

collection of data on the failure modes of

digital systems, including international

experience with digital system failures,

software quality assurance, environmental

stressors on digital systems, modeling digital

systems in PRAs and cyber security of digital

systems. Within the general element of

emerging digital technologies applicable to

nuclear facilities, attentions are on system

diagnosis, prognosis and on-line monitoring

as well as wireless technology. Research on

digital systems for advanced nuclear power

plants was given a low priority. Perhaps,

future new orders for advanced plants

(AP1000, ESBWR, etc.) may create new

regulatory demands and cause this priority to

be re-evaluated.

The ACRS has 'recently 'reviewed and

reported favorably on the research plan for

digital systems. The ACRS was impressed by

the technical quality in the development of the

research plan, the scope and content of the

plan, and the prioritization of activities in the

plan. Indeed, it would help better

understanding of other research programs if

they were also based on such thorough

planning efforts. The ACRS recommends the

NUREG-1 635 '16
following to further improve an already quality

research plan:

* The plan is currently focused very

much on the software aspects of

digital systems. Eventually, the

research will have to be expanded to

recognize the entire system of

interest. Though the-- focus on

software is appropriate now, the plan

should reflect the need for expansion

in scope in the longer term.

* There should be an explicit element

of the plan to study the acceptability

of international standards in

comparison to IEEE standards (such

as IEC 60780 in comparison to IEEE

'323) -for meeting regulatory

requirements concerning digital

instrumentation and control systems.

This study will be an 'important

element of efforts to develop a multinational design approval process.

* As data on digital system failures are

collected and analyzed, the research

staff should prepare episodic papers

or presentations to professional

societies of their interpretations and

:lessons learned" for peer review by

the larger digital system reliability

community.

17 NUREG-1 635
Table 2. Research Activities in Digital Instrumentation and Control Systems

Job Code Title Comment

N6116 Secure Network Design Develop technical guidance for mitigating cyber

Techniques vulnerabilities in secure networks

N6095 Assignment Robert Support analysis of digital systems failures and

Edwards consequences

Y6962 Emerging Technologies Conduct periodic surveys of the state of the art for a

wide range of technology issues in the I&C field

Y6873 International Cooperative Search for opportunities to collaborate in the safety

Research Program on assessment of digital systems

Digital I&C

N6010 COMPSYS OECD/NEA international program to develop database on digital systems failures

K6472 Risk Importance of Digital Develop methods to include digital systems in PRAs

Systems

Y6332 Digital Systems Risk Develop a PRA method for modeling failures of digital l&C systems.

Y6591 Software Reliability Code Large-scale validation of NRC methodology for

Measurements predicting software reliability in digital systems

N6080 Interactions with Industry Development of standards on EM1/RFI

on Standards

Y6475 Wireless Confirmatory research on effects of wireless

communications

N6113 Security of Digital Study in laboratory digital systems generically

Platforms qualified for nuclear safety applications

N6114 Site-specific Protocol Study power plant implementation of digital systems

Analysis generically qualified for nuclear safety applications

N6124 Digital System Qualify safety of a digital system using a process

Dependability developed in NRC research

Performance

W6851 Review Guidance for Support for response to public comments on draft

Lightning regulatory guide; Program completed.

Y6924 SPACE Engineering Evaluate the use of the RETRAN tool for review of

Workstation for Review of TELEPERM-based digital instrumentation and

TXC Applications control upgrades

Y6349 Halden Environmentally Despite the name this is research on COS operating

Assisted Cracking experience, ranking software engineering practices

(The title of this program is and testing digital reliability assessment methods

amazingly misleading!)

NUREG-1635 :18
5 FIRE SAFETY RESEARCH

The fire safety research program can be

divided into three technical areas:

* Fire Risk Assessment

* Fire Modeling

Fire Testing

Each of these areas is discussed below.

Fire Risk Assessment: The nuclear industry

has made substantial progress over the past

thirty years in the development and

standardization of internal events risk

assessment. Progress in the development of

the methods of fire risk assessment has been

much slower. Only a few nuclear power plants currently have full-scope fire risk

assessments. The requirements placed by the NRC on the industry for performing

Individual Plant Examinations of External

Events (IPEEE) permitted the use of CofraoyTsigfHmclTie simplified and qualitative techniques. Most Bat Fer analyses of fire risk at nuclear power plants

were performed with these less quantitative TheH anMecral ra barrier techniques.. s e a nu fplats tovide .a fire barrerbetwen two tan fsaesudw

As the NRC moves from deterministic fir ro nipeton t n aants. regulations to risk-informed and performance- qusinraedegdngtefe.eitac based regulations, the need for quality risk cpblt fteesses R odce information increases greatly. It is expected ers that many nuclear power plants will transition at t Omga Pont iin.Athe from their current fire protection programs to cniuaintse aldt etacpac the risk-informed, performance-based fire c

protection programs that meet the barerr nt p andtpride requirements of 10 CFR 50.48(c) and the p a schdl fo a referenced 2001 Edition of National Fire

Protection Association (NFPA) standard,

NFPA 805, "Performance-Based Standard for

Fire Protection for Light-Water Reactor

Electric Generating Stations." This is only possible if a full-scope fire risk assessment is

performed for each transitioning nuclear

power plant. NRC will need appropriate standard to assess the quality of such fire risk

assessments and inspectors will need tools

19 NUREG-1635

I 1111110
and the knowledge to assess the validity of

changes to the licensing basis made at the

plants.

RES in cooperation with EPRI

has taken some important steps to

consolidate the fire PRA research and

development activities, conducted over the

past few years, into a single state-of-the-art

methodology for fire risk assessment. In

2005, the final NUREG/CR-6850,

'EPRI/NRC-RES Fire PRA Methodology for

Nuclear Power Facilities," was issued. This

document provides a structured framework

for the overall fire risk assessment as well as

specific recommended practices to address

key aspects of the analysis. While -the

primary objective of the project was to

consolidate state-of-the-art methods, in many'

areas the newly documented methods

represent a significant advancement over

those previously documented. Although

some 'utilities have used parts of the

improved approach, no utility has completed

a fire risk assessment using the methodology

and submitted the'assessment for critical

peer review.

Areas of fire risk analysis where further

development in methodology is needed have

been recognized by RES. These include

spurious equipment; actuations, post-fire

human reliability analysis, aging effects, and

low power and shutdown fire risk.

Fire Modeling: Deterministic criteria for fire

protection are typically very conservative in

their treatment of fire progression. Fire risk

assessment, on therother hand, requires a

realistic assessment of fire progression.

There are a variety of methods that can be

used to model the progression of fires. Some

of these have been used in fire protection

programs for non-nuclear facilities for many

years. The ranges of applicability of these

methods have not been well studied or

documented. In cooperation with EPRI, a

Project (Y6688) is in progress to verify and

validate a set of fire progression modeling

tools. The accuracies of these tools are being

examined for different fire conditions and

applications by comparison with benchmark

tests performed by National Institute of

Standards and Technology (NIST) The

phenomena identification and ranking table

(PIRT) process is being used by RES to

identify potential limitations of the fire'

progression modeling tools. Preliminary draft

of multi volumes NUREG-1826, 'Verification'

and Validation of Selected Fire Models for

Nuclear Power Plant Applications," was

issued for Public Comment in January 2006.

Fire Testing: Confirmatory testing is another

critical element of the fire safety program.'

During the past year, tests were performed

at the Omega Point Test Facility on the

Hemyc and MT electrical raceway fire barrier

systems (see side column). The test results

indicated that these fire barrier systems are

not capable of satisfying regulatory

requirements. It is somewhat distressing that

confirmatory testing of these fire barriers did

not occur until sixteen years after problems

were identified with a similar fire barrier

material, Thermo-lag, and five years after

inspection teams raised specific concerns

about the Hemyc and MT fire barriers. 'The

results of these tests provide further evidence

of the continuing value of NRC's confirmatory

testing program.

There have been a number of important

accomplishment by NRC research in the area

of fire protection since the last ACRS report

on NRC safety research program in 2004.

Fire safety research continues to -merit'

emphasis in the NRC research program.

Approximate, and often bounding risk

analysis, performed for individual plants

indicate that the risk of core damage from

fire- initiated events is comparable to or

greater than the risk from other accidents

initiated during normal operations. It is

important to know whether the same'

conclusion would be drawn if fire risk

NUREG-1 635 20
assessments were performed using tools of

comparable sophistication as those used for

assessing risk of accidents initiated 'by

internal events. Conclusions based on more

realistic fire risk assessments could have

ramifications on both regulatory attention and

licensee attention to safety. In the interim,

risk-informed regulatory decisions are being

made with an incomplete understanding of

the impact of fire on risk.

RES has worked closely with the U.S.

industry in undertaking -generic fire risk

research activities. Fire risk isj however, an

issue of world-wide concem. France, for

example, has recently initiated a fire

research program 'in a multi-volume test

facility. RES has not aggressively sought

collaborations with the international

community to advance NRC capabilities for

fire risk assessment. Collaborations with

other -countries especially in experimental

studies may be ~ essential to' leverage

resources of all partners sufficiently to

achieve fire risk assessment capabilities

commensurate with what can'nobw be done

for risk from normal plant operations.

21 NUREG-1 635
Table 3. Fire Safety Research Activities m.

Job Code Title Comment

N6107 10 CFR 50.48C - related Technical Develop fire PRA methods, tools, and

Activities data. Perform demonstration studies.

This is a collaborative effort between

NRC and EPRI.

N6108 Fire Risk Assessment and Risk Improve fire PRA approaches. Develop

Applications test plan to address spurious equipment

actuation issues.

N6134 LPSD Level I & Fire Risk Supports NRC staff in the development

Standard of industry standards.

Y6651 Effects of Switchgear Aging on Assess the aging of medium voltage

Energetic Faults switch gear as it affects the potential for

energetic electrical faults. Such faults are

thought to contribute significantly to fire

initiation. The work addresses how aging

affects fire risk.

Y6688 Fire Model Benchmarking and Benchmark fire model computer codes

Validation against fire experiments performed by

NIST. Such validation is necessary to

ensure that appropriate tools are, used for

regulatory applications.

Y6817 Fire Protective Wrap Test Hemyc and MT fire wrap materials.

Performance Testing These important tests conducted in 2005

showed there to be significant issues

associated with these fire barrier

materials.

Y6955 Fire Incident Records Exchange Collect and analyze international fire

events data. This is a long-term

collaborative effort with OECD.

NUREG-1 635 .22
6 REACTOR FUEL RESEARCH

Reactor fuel is an important element of safety

technology. -NRC must maintain expertise in

the area of reactor fuel because of both the

importance to safety- and because of 'the

limited availability of expertise outside the

agency that is. independent of licensees.

Research is an important 'vehicle for

maintaining expertise in reactor fuel. NRC

research on reactor fuel during normal

operations ,and design basis accidents has

been concentrated in recent years on the

confirmation of regulatory decisions that allow

licensees to take light water reactor fuels to

bumups of nominally 62 GWd/t. This

research has largely resolved the issue of the

vulnerabilityof high-bumup fuel and cladding

to reactivity transients though some

confirmatory tests need to be completed.

Research results will allow regulatory changes

to better reflect the degraded capacity of highbumup fuel to sustain reactivity insertion

events.

The reactor fuel research has remained quite

productivewas examinations of high-bunup

fuel behavior under loss-of-coolant accidents

have been initiated. An important discovery

has been -,the synergistic effect on clad

ductility of hydrogen absorption during normal

operation and steam oxidation of the cladding

during an accident. Based on the research,

revised embrittlement criteria have. been

developed that could be incorporated into

10 CFR 50.46.

The research on high-burnup fuel is reaching

a substantial level of maturity. Some major

confirmatory experiments remain to be done -

notably experiments on reactivity insertion to

be done in a water loop at the CABRI reactor.

Plans Jor these experiments have -been

revised since our last report on reactor-fuels

research so the experiments which are part of,

an international collaborative effort now better:

meet agency needs. It is important that-this

work that is so well coordinated both with

agency needs and with international partners

be taken to completion. Still, major findings of

the research effort can be reduced to

regulatory practice now. This reduction to

regulatory practice needs to be initiated and

pursued aggressively.

It is evident that high-burnup fuel research will

soon achieve results that are adequate for

agency needs. The NRC has made clear that

it will expect the nuclear industry to provide

necessary safety analyses and experimental

data should the industry want to take fuel to

burnups that exceed the current regulatory

maximum. NRC needs to make these

expectations more explicit, particularly its

expectations for the experimental data

needed to support the analyses of highburnup fuel behavior under accident

conditions.

Completion of NRC's research on high-bumup

fuel raises the question of how NRC will

maintain expertise in fuel. Continued evolution

in fuel cladding alloys can'be anticipated.

Interest is developing within the industrŽ in

fuels with' enrichments exceeding 5% 25U.

These higher enrichment fuels may

necessitate NRC research. If use of MOX fuel

becomes more widespread than the planned disposal of excess weapons-grade plutonium,

additional research on MOX fuel with reactor

grade plutonium may be needed. Research

on both higher enrichment fuel and MOX fuel

can be done with substantial collaboration

with international partners, Such collaboration

will further the ideal of international safety

evaluations of nuclear power plants.

23 NUREG-1 635
Table 4. Reactor Fuel Research Activities

Job Code Title Comment

Y6586 Fuel Code Assessment for Improve FRAPCON and FRAPTRAN for

MOX calculating the behavior of MOX fuel rods; An important activity for licensing core loads for

excess weapons-grade plutonium disposal.

Y6580 Fuel Code Applications for Improve FRAPCON and FRAPTRAN for

High Burnup Fuel calculating the behavior of high bumup fuel rods;

an important activity as licensees press limits on ._ allowable fuel bumup.

Y6788 Halden Fuel Experiments Data on fuel behavior under operational transient

: Under Transient Conditions conditions for code development.

N6074 STUDSVIK Cladding Integrity Stress corrosion cracking, hydride embrittlement

Project and delayed hydride cracking study of ZIRLO

clad. Defueled clad segments provided for NRC

research.

VY6849 ZIRLO Cladding Performance Adequacy of criteria for ZIRLO cladding performance in a LOCA; an important study of

cladding used for high bumup fuel and the

synergism between hydriding and oxidation on

clad ductility.

Y6850 M5 Cladding Performance Adequacy of criteria for M5 cladding performance in a LOCA; an important study of cladding used

for high bumup fuel and the synergism between _______hydriding and oxidation on clad ductility.

G6923 Failure of HydridedZZircaloy Develop theoretical model of mechanical failure of

under Severe Loading hydrided Zircaloy cladding.

Conditions'

W6832 CABRI Water Loop NRC support for the CABRI water loop for RIA

testing of high bumup fuel; confirmatory testing of

high burnup clad and fuel vulnerability to reactivity

transient events.

Y6367 High Burnup Cladding LOCA testing of high bumup cladding behavior;

Performance important study of cladding used for high bumup fuel and the synergism between hydriding and

oxidation on clad ductility.

Y6723 International Agreement on Data report on BIGR pulse reactor tests.

Fuel Behavior and Materials

Science Research

Y6847 Clad Performance in ATWS Determine the adequacy of criteria and analysis of

clad performance in BWR power oscillations;

NRC needs to see if this problem can be solved

by analysis with minimal experimental

confirmation.

Y6195 Dry Cask Storage License for Develop criteria for dry-cask storage and

High Burnup Fuel transportation of spent high bumup fuel.

NUREG-1 635 24
7 NEUTRONICS AND CRITICALITY SAFETY

Neutronics and criticality safety are areas in

which NRC must maintain exceptional

capabilities through its research program. The

neutronics and criticality safety research

program is small but appears adequate to

ensure that the NRC has capabilities to meet

immediately foreseen regulatory needs. The

current NRC research activities in neutronics

analysis, core physics, and criticality safety

are listed in Table 5. Maintenance of the

SCALE suite of codes is essential for the

analysis of reactor core physics. These codes

are complemented by the PARCS code which

is part of the TRACE code and is discussed in

more programmatic detail in the Chapter 14 of

this report dealing with Thermal Hydraulics

Research. The availability of the NEWT lattice

code is important to licensees since it will be

essential for the use of more advanced

computer models in future regulatory

processes. Currently, this lattice code is being

used for the analysis of reactor cores fueled in

part with MOX fuel for the disposition of

excess weapons-grade plutonium. Several

other activities are under way to support the

licensing of MOX fuel core at the Catawba

reactor for this plutonium disposition activity.

These are appropriate programs at the

current time. It is noted that NRC is taking

advantage, to the extent feasible, of the

considerable European experience with MOX

fuel made with reactor-grade plutonium.

In the future, more innovative core designs for

advanced reactors may be submitted to the

NRC. Confirmatory analyses of reactor core

physics will be an essential part of the

regulatory process for these advanced

reactors. The capabilities now available to the

NRC in the area of core physics may well be

stretched. It will be useful to the agency to

understand these future needs. If long-term

development activities are identified, such as

those that might be needed for analysis of the

PBMR, additional research may be needed in

this area.

25 NUREG-1 635
Table 5. Research Activities In Neutronics Analysis,

Core Physics, and Criticality Safety

Job Code Title Comment

Y6846- SCALE Code Development for Essential code for neutronics analysis to

Reactor Physics audit licensee submittals and other

X__ XXregulatory needs.

Y6320 NEWT Lattice Code Generate lattice cross-sections for safety

analysis of MOX cores to support

licensing of cores for Pu disposal.

N6162 MOX Benchmark Confirmation of uncertainties in PARCS

code predictions of MOX core neutronics;

Also supports the licensing of Pu disposal

activities.

Y6403 Reactor Core Analysis Analysis to predict details of reactivity

transient in MOX core. Again, this

research supports regulatory activities

associated with the DOE program to

dispose of excess weapons-grade

-__ -plutonium.

Y6685 Experimental Data for High This project provides NRC with foreign

Burnup Spent Fuel Validation and domestic data on high bumup fuel

and MOX fuel for assessment of analytical

tools used to predict fuel inventories,

decay heating, and radiation shielding.

NUREG-1 635 26
8. HUMAN FACTORS AND HUMAN RELIABILITY RESEARCH

Human performance plays a critical role in the

safe operation of nuclear plower plants.

Human performance issues have been main

contributors to accidents and unsafe

conditions experienced by the current fleet of

operating reactors. They can be expected to

continue to have a major impact on nuclear

power plant safety. As licensees increasingly

rely.on risk-informed licensing applications

that require the quantification of human

reliability under accident conditions, the staff

needs to be able to evaluate the treatment of

operator actions in such applications. As new

reactor designs, likely dependent on a higher

degree of automation than the current fleet,

are introduced, the need for revised guidance

and tools for the NRC staff in human factors

and human reliability analysis will increase.

Therefore, it is very important that the NRC

maintain research programs in these areas.

The current NRC research activities in the

areas of human factors and human reliability

analysis are:

* Human Factors

(B7488, N6207, Y6843, N6137,

Y6529)

* Human Reliability Analysis

(Y6497, Y6496, N6248)

Current research in the human factors area

includes a continuing international

collaborative research program at the Halden

project (B7488). The ACRS is supportive of

this collaborative program and recommends

continued NRC participation.

The project "Development of -a Regulatory

Guide and Analytical technique for Assessing

NPP Staffing" (N6207) supports the

development of 'guidance. for staffing

exemption requests to 10 CFR 50.54(m). This

project is almost complete. Guidance is now

provided in the recently issued NUREG-1 791,

'Guidance for Assessing Exemption Requests

from the NPP Licensed Operator Staffing

,Requirements Specified in 10 CFR 50.54(m)."

Publication of this guidance is a significant

accomplishment that provides a more flexible

approach to staffing of current and future

reactors.

Human performance, issues, including

organizational issues are of great importance

to nuclear reactor safety. Inspectors at

nuclear power plants currently have limited

guidance or means with the Reactor

Oversight Process (ROP) to characterize

problems associated with human

performance. This issue has been highlighted

in a recent report from the Inspector General.

In response to a Commission request, the

project Y6843, "Develop Human Performance

Indicators,"has been initiated to study the

feasibility of establishing the technical bases

for indicators of' human performance that

would be used to supplement indicators

currently used in the ROP. This research is

appropriate and very important. It may lead to

significant improvements in the NRC

inspection program and the ROP.

There is evidence of degrading performance

of operations personnel, in the nuclear and

other industries due to operator overload. The

research project N6137, Impact of Operator

Workload on Human Performance," is a five-

,year effort to assess the impact of operator

overload on performance. -The plan is to

develop licensing requirements as well as

inspection guidance and techniques, for

reviewing the impact of workload on operator

performance and plant safety. This is an

important new project that deserves support

both for the currentileet of operating reactors

and for advanced reactor designs.

Advanced reactor designs are likely to

introduce much greater automation than

exists in current reactors. Certainly, advanced

27 NUREG-1635
digital control and instrumentation methods as

well as new human-system interfaces can be

anticipated. These new features of plants are

likely to have some'' effects on human

performance. The NRC staff needs to prepare

itself to review new concepts and designs

proposed by licensees. The project "Human

Factors of Advanced Reactors" (Y6529) has

been initiated to address this issue and to

develop regulatory guidance and analytical

techniques to review human factors for

advanced nuclear power plants. The ACRS

views this five-year project 'essential for

preparing the staff in reviewing advanced

reactor designs.

-The :quantification of human reliability

continues to be a challenge; 'in risk

assessments. Many approaches to the

quantification of human reliability have been

proposed. However, the benchmark exercise

conducted by the Ispra Laboratory of the

European Union demonstrated that the choice

of model has a significant impact on the

results; obtained. Not much progress to

improve this'situation has been made since

that exercise was performed. The NRC staff

has recently completed an assessment of the

strengths and weaknesses of the various

methodologies now available for assessing

human reliability. The ACRS has been quite

impressed with this assessment and hope the

work leads to the identification of best

methods for the quantification of human

reliability in PRA.

Human reliability modeling introduces large

uncertainties in PRAs. The NRC staff needs

guidance in its review of the human reliability

models used by the industry in licensing

applications. The project Y6497, UHRA

Application and ATHEANA Maintenance," is

intended to improve NRC's ability 'to

independentlyrmodel human reliability and to

provide guidance concerning risk-informed

regulatory applications. Progress has been

made with the publication of NUREG-1 792,

"Good Practices for Implementing Human

Reliability Analysis (HRA)." Still, further

guidance is needed for reviewers of licensing

applications. The NRC has applied ATHEANA

model to the human performance issues

associated with its recent pressurized thermal

shock study. The NRC is also planning to

apply the ATHEANA model to a number of

ongoing risk assessments, including those for

fire and steam generator tube rupture'to

develop lessons learned on human reliability

analysis and to develop guidance for the staff.

If needed, mofifications to the Standard

Review Plan for licensee's applications will be

devised. The ACRS believes that this effort is

needed. ATHEANA is a state-of-the-art model

of human performance and is complicated to

use. Application of the tool will show whether

benefits derived from the analyses are

commensurate with the enhanced complexity.

Application may also show how the complexity

of ATHEANA can be reduced. Application of

ATHEANA is, however, very much behind

schedule. Resources and management

attention are needed to either accelerate the

efforts or to revise the scope of the application

efforts.

Both ATHEANA and SPAR-H (the HRA model

used in SPAR) quantify the probability that a

human unsafe act will be committed. This

probability depends on a number of

performance shaping factors (PSFs) that

determine the context within the crew

operates. The available time for action is one

of the PSFs estimated from thermal-hydraulic

considerations. The evaluated failure

probability ia understood to be the probability

that the required action will not be completed

within the available time.

An alternative approach to HRA is to

recognize the importance of time taken by the

crew to complete a task and to develop a

probability distribution for this time. The

failure probability, then, is calculated from this

distribution as the probability that this time will

exceed the available time.

NUREG-1635 28
Recent experiments performed at Halden,

Norway, have shown that there may be

significant variability in the time that crews

take to perform a given task. Such evidence

is very difficult to account for in iATHEANA

and SPAR-H. The alternative approach could

accommodate such evidence. In addition, the

staff is currently supporting research at Idaho

National Laboratory (INL) that develops 'time

lines" for past accidents. This evidence can

also be accommodated in the alternative

approach.

The staffshould evaluate the merits of an HRA

model that focuses on the time required for

action.

The project Y6496 is a continuing effort to

develop an event database called Human

Event Repository and Analyses. This

database and analysis capability should

significantly improve the treatment of human

reliability in nuclear reactors and provide a

realistic, performance-based database to

assess licensee's quantification of human

performance. This effort should be sustained

and made an ongoing part of the research

program.

The project N6248, "Advanced Reactor HRA

Development," is the first year of a proposed

five-year effort to develop HRA methods and

tools to support an independent staff review of

human reliability analyses submitted as part of

new reactor licensing applications. Given the

importance of human factors to reactor safety

and the likelihood that new reactor designs

may significantly alter the role of operators

and the human-system interface, this project

is valuable and should be continued to

completion.

I - I ; n I .

I j , I

29 NUREG-1 635
Table 6. Human Factors and Human Reliability Research Activities

*. .

Job Code Title Comment

Y6497 HRA Application and Apply ATHEANA to Fire Risk Requantification;

ATHEANA Maintenance, upgrade and improve ATHEANA. ATHEANA is

NRC's tool for analysis of human reliability.

Application of this tool will allow assessment of its

worth.

Y6496 Human Event Repository Develop a human event repository and analysis

and Analysis - tools. This program develops a useful data-base

for comparison to model predictions of human

events.

B7488 Halden Reactor Project International collaborative research project that

- addresses man-machine interaction and

verification and validation of software,

surveillance and support systems, advanced

control rooms and fuels and materials. This

international effort helps keep staff aware of

international developments in human factors and

human reliability.

N6207 Develop Reg. Guide and Support development of guidance for staffing

Analytical Technique for exemption requests to 1 0 CFR 50.54 (m). This is

assessing NPP staffing an important program as licensees look at

manpower costs associated with nuclear power

plant operations.

Y6843 Develop Human Determine availability and viability of human

Performance Indicators performance indicators for assessing

performance at nuclear power plants; This

program was undertaken in response to a

Commission SRM.

N6137 Impact of Operator An important new effort to assess the impact of

Workload on Human operator overload on operator performance and

Performance plant safety.

N6248 Advanced Reactor HRA The first year of a proposed five -year effort for

Development addressing human performance issues for new

reactors. This is a valuable project and should be

continued to completion.

Y6529 Human Factors of Develop regulatory guidance and analytical

Advanced Reactors techniques to review human factors for advanced

reactors. Essential work to prepare the staff in its

review of advanced reactor designs.

NUREG-1635 .30
9 MATERIALS AND METALLURGY

Research in the area of materials and

metallurgy is an important focus of the NRC

Safety Research Program. Current research

activities are concentrated in five areas:

* Environmentally Assisted Cracking in

Light Water Reactors

(Projects K6266, K6202, Y6270,

Y6388, N6007)

* Steam Generator Tube Integrity

(Projects Y6536, Y6588)

* Non-destructive Examinations

(Projects' Y6534, Y6604, Y6649,

Y6869, Y6867, Y6541, N6019)

* Proactive Materials Degradation Assessment

(Project Y6868)

* Reactor Pressure Vessel Integrity

(Projects W6953, Y6533, Y6378,

Y6638, Y6951, N6204, Y6870,

N6223, Y6485, Y6656)

These projects represent a significant

investment by the NRC to better understand

the issues of materials degradation in the

currently operating fleet of nuclear power

plants. Such investment is justified in view of

significant agency regulatory activities that

aging degradation research supports. As

plants age, known degradation mechanisms

will continue to affect components and new

degradation mechanisms may develop. The

current program is well focused on improving the agency's ability to independently evaluate

licensees' efforts to prevent, detect, and

mitigate environmentally assisted stress

corrosion cracking. The Proactive Materials

Degradation Assessment project is an effort

to identify potential material degradation

problems before they manifest in operating nuclear power plants. '

Unfortunately, the planning of NRC's research

in materials and metallurgy is not well

documented in the way planning for research

on digital instrumentation and control systems has been documented. It is, then, difficult to

explain the role and priority of each task within

each of the five project areas. In aggregate, the activities in the first four project areas

(Environmentally Assisted Cracking, Steam

Generator Tube Integrity, Non-destructive

Examinations, and Proactive Materials

Degradation Assessment) seem to be

appropriate. These are the very areas that

most challenge the industry and its ability to

detect component degradation. The agency

must develop the capabilities to assess the

acceptability of the industry's initiatives to deal

with these degradation challenges. The five

project areas are further discussed below.

Environmentally Assisted Cracking

Environmentally assisted cracking is a

complicated technical issue that continues to

afflict the industry as components age and

irradiation effect increases. In recent years,

the industry has experienced irradiation

assisted stress corrosion cracking (IASCC) of

components internal to the vessels of boiling

water reactors (BWRs) and stress corrosion

cracking of reactor vessel head penetration assemblies In pressurized water reactors

(PWRs).'Although the industry has responded

to these events with initiatives to prevent and

mitigate these types of degradation, the event

at Davis-Besse makes it readily apparent that

the NRC 'staff must be capable of

independently evaluating 'the adequacy of

licensees' initiatives. The research projects

now under way seem well designed to ensure

that the NRC has the needed technical

understanding of the stress corrosion cracking

Issues.

31 NUREG-1635
The project Y6388, "Environmentally Assisted

Cracking of LWRs," evaluates environmental

effects on fatigue of steels used in light water

reactors and provides the NRC with technical

data and analytical methods to assess

licensees' plans concerning mitigation. The

large effort includes tests of neutron-irradiated

specimens to improve the understanding of

IASCC initiation and stress relaxation. It also

provides data on the performance of probes

and monitoring techniques in radiation

environments. This work is essential and

should be continued. A new project,

"Investigation of Stress Corrosion Cracking in

Selected Materials" (N6007), will develop a

better understanding of stress corrosion

cracking in -PWRs.- Such cracking occurs

typically in the reactor coolant system

boundary. Understanding of such cracking in

this boundary is essential for maintaining the

defense-in-depth.

Environmentally assisted corrosion of reactor

materials is an international concern. The

CIR-I1 Cooperative Agreement (K6202) is a

collaboration with the international community

for studying the susceptibility of stainless steel

to IASCC. Certainly, this collaboration should

be continued.

Steam Generator Tube Integrity

Rupture of steam generator tubes in PWRs

can lead to accidents that allow radioactive

materials released from the core to bypass

the reactor containment and enter directly into

the environment. Severe accidents involving

containment bypass can be risk dominant at

some PWRs. Through the years, many

modes of corrosion of steam generator tubes

have been experienced. Regulations on the

corrosion were developed when erosion was

the dominant concern. Careful water

chemistry control by licensees has largely

eliminated erosion as a safety concern. But,

now, stress corrosion cracking has emerged

as the dominant threat to the integrity of

steam generator tubes. Incipient stress

corrosion cracking is much more difficult to

detect. NRC has two research projects to deal

with the degradation mechanisms in steam

generator tubes, "Steam Generator Tube

Integrity Program" (Y6588) and "PWR Primary

System Components Severe Accidents"

(Y6536). The first project, Y6588, deals with

potential tube degradation modes, their

resulting leak rates, and the effectiveness of

in-service inspections. The second project,

Y6536, seeks to improve methods and

models used to predict the behavior of

degraded steam generators and other PWR

components under severe accident loads.

Both of these research efforts are important

and should be continued.

Non-destructive Examinations

Non-destructive examinations are relied upon

to monitor the integrity of the reactor coolant

system. The reliability and effectiveness of

existing non-destructive examination

techniques remain open to question.

Certainly, a steam generator tube cracking

incident at the Indian Point reactor

emphasizes this point. Four projects are

under way to improve non-destructive

examination techniques (Y6534, Y6604,

Y6649, and Y6869) and this work should

continue. Two of these projects deal with the

effectiveness and reliability of non-destructive

examination of reactor vessel penetration

assemblies. As the ACRS noted in NUREG1635, Vol. 6, this is an area that needs

increased attention. A third project will provide

destructive examination data that should be of

tremendous value for the validation of nondestructive examination methods. The

project, N6019, will examine non-destructive

methods and leak monitoring techniques and

the requirements for light water reactor

components that have experienced

degradation or have been identified as being

susceptible to future degradation. The project

"Evaluate Reliability and Effectiveness of

Advanced NDE," Y6541, > will support

continued investigation of innovative methods

'NUREG-1635 32
to detect incipient amounts of wastage of

ferritic steel. All of these projects are

responsive to the NRC's needs and should be

continued.

Proactive Materials Degradation Assessment

The nuclear industry and the NRC have often

been surprised by unexpected material

degradation problems. As a result, they have

responded to such problems in a reactive

mode which has proven to be inefficient.

Reactive response does not enhance public

confidence in the safe operations of nuclear

power plants. The project "Proactive Material

Degradation Assessment" (Y6868) is an NRC

initiative to identify materials and locations in

light water reactors where degradation can

reasonably be expected in the future. The

goal of this project is to develop the technical

bases needed to implement regulatory actions

to proactively address materials degradation

problems. Current inspection and monitoring

programs at plants can be reviewed and

modified as needed to provide earlier

identification of incipient degradation before it

affects plant safety. The ACRS admires the

vision of this undertaking and supports its

continuation. The ACRS looks forward to

reviewing the initial results of this ongoing

effort soon and learning whether the

admirable goal of this project is, in fact,

feasible.

Reactor Pressure Vessel Integrity

The integrity of the reactor pressure vessels

has been studied for decades. Maintaining

the structural integrity of the reactor pressure

vessel in a nuclear power plant during both

routine operations and during postulated

upset conditions, including pressurized

thermal shock situations, is a longstanding

obligation of licensees. This obligation is

codified in three general design criteria (GDC

14, GDC 30 and GDC 31) as well as in 10

CFR 50.61 and the appendices G and H to 10

CFR Part 50. Technical bases for these

requirements were largely established in the

1 980s. NRC is continuing to devote

substantial resources to the study of pressure

vessel embrittlement though there does not

seem to be a comparable interest within the

industry who will have most of the research

benefits. Indeed, the number of projects in

this area seems to have grown since the

ACRS last reviewed the NRC research

program and questioned the need for

research in the area of reactor pressure

vessel integrity.

Some of the activities in this programmatic

area deal with the finalization of the NRC's

work on pressurized thermal shock which is

nearing completion. These activities will

contribute to the potential revisions of

Regulatory Guide 1.99 on radiation

embrittlement of reactor pressure vessel

materials and Appendices G and H to

10 CFR Part 50 on fracture toughness

requirements and reactor surveillance needed

to ensure low probability of reactor vessel

failure.

The project "International Pressure Vessel

Technical Cooperative Program" (Y6378) will

ensure NRC participation in the International

Atomic Energy Agency (IAEA) deliberation on

reactor pressure vessel integrity.

The NRC's comprehensive program on

reactor pressure vessel integrity has produced

significant results by providing better

understanding of the available margin in

reactor pressure vessel components.

Revisions to PTS screening criterion in the

PTS rule and the associated regulatory guides

and Appendices G and H to 10 CFR Part 50

are likely to provide great benefit to licensees

by relaxing current requirements and allowing

longer life of reactor pressure vessels. These

activities should be completed soon.

RES needs to reevaluate the need for

continued research into heavy section steel

components. This research may be justified if

there is a clear need for NRC to develop its

capabilities in the area of probabilistic fracture

33 NUREG-1 635
mechanics so that it can evaluate licensees'

applications. If this is the case, the research

needs to be clearly focused on this objective

and not the research that the industry should

perform to meet its responsibilities to ensure

reactor pressure vessel integrity. It appears

now, however, that it is NRC that is advancing

the state-of-the-art and making available

information that allows licensees to reduce

conservatism in their analyses.

- ..

I I I I

- I -

NUREG-1 635 34
Table 7. Research Activities in Materials and Metallurgy

Job Code 'Title Comment

Environmentally Assisted Cracking in LWRs

K6266 CIR-li Cooperative NRC contribution to international research on

Agreement irradiation assisted stress corrosion cracking.

K6202 Extension of CIR-il Assess the susceptibility of stainless steels to

Cooperative Agreement Irradiation Assisted Stress Corrosion Cracking.

This program allows NRC to stay abreast of

international developments.

Y6270 Environmentally Provide neutron irradiated specimens for NRC

Assisted Cracking research programs.

Y6388 Environmentally Develop data on irradiation assisted stress

Assisted Cracking of corrosion cracking in PWRs and BWRs. This

LWRs program provides NRC staff with the data and

analytical methods to review licensees' activities

and plans to limit corrosion.

N6007 investigation of Stress User need for a better understanding of stress

Corrosion Cracking In corrosion cracking in PWRs. This program

Selected Materials supports the regulatory process.

SteamGeneratorTubelntegrity

Y6536 PWR Primary System -Methods and models to predict PWR reactor

Components Severe coolant system component behaviors under

Accidents severe accident loads; This is an essential

~research program.

Y6588 Steam Generator Tube Wide-ranging program In support of the steam

integrity Program generator integrity action plan. ACRS supports this

action plan and regularly monitors its progress.

35 NUREG-1 635
Table 7. Research Activities In Materials and Metallurgy

(Continued)

Job Code -Title Comment

Non-destructive Examinations

Y6534 Piping NDE Reliability Program addresses Inconel cracking in weld metal

and base metal. This is an essential program to

ensure licensees adequately monitor nickel alloys

in plants.

Y6604 Evaluate Reliability of Addressing the inspection of cast stainless steel

NDE Techniques components and dissimilar metal welds; evaluation

of reliability and accuracy of in-service inspection.

This is an essential program to facilitate NRC

monitoring of licensee activities.

Y6649 Phase ii- Alloy 600 Independent assessment of industry analyses of

Cracking CRDM nozzle cracking. This is a classic NRC

program of confirmatory research.

Y6869 Barrier Integrity Evaluate RCS leakage experience and leak

Research Program detection capabilities. This is an essential program

to facilitate NRC monitoring of licensee activities.

Y6867 Cooperative Activities Complete non-destructive examinations of nozzles

Reactor Coolant, from vessel heads. Plan destructive tests. This is

System Pressure an important program to validate analyses NRC

Boundary Components uses in its regulation of licensee activities.

Y6541 Evaluate Reliability and Identify innovative NDE techniques in coordination

Effectiveness of with industry and international community. This

Advanced NDE program allows NRC staff to stay abreast of

international developments in NDE.

N6019 NDE & Leak Monitoring Assess adequacy of current inspection and

Requirements monitoring requirements. Assemble data on

probabilities of failure of passive components. This

is an essential program to facilitate NRC

monitoring of licensee activities.

-NUREG-1635 3

36
Table 7. Research Activities in Materials and Metallurgy

(Continued)

Comment

Proactive Materials Degradation Assessment

Y6868 Proactive Materials Identify materials and locations in LWRs where

Degradation degradation can reasonably be expected. This

Assessment program is intended to better equip NRC to

anticipate materials degradation problems at

nuclear power plants. This program should be

continued. The ACRS looks forward to reviewing

,_ _ _the initial results.

._ _ Reactor Pressure Vessel Integrity

Review and Revisions Support documentation of thermal hydraulics

N6204 of Pressurized Thermal analyses for pressurized thermal shock, and

Shock Reports NUREGs document Calvert Cliffs RELAP5 calculations to

1806 and 1809 support FAVOR calculations. This program should _____-_be completed.

Y6485 Technical Support - Support for the pressurized thermal shock

Pressurized Thermal rulemaking effort. This is essential support for the

Shock Rulemaking regulatory process.

W6953 Heavy-Section Steel Evaluation of Master Curve methodology for

Irradiation Program reactor pressure vessels. The ACRS questions the

need for the large investment in heavy section

steel research.

Y6870 Cooperative Program Development of a cooperative program with DOE

--on Irradiation to study reactor pressure vessel materials.

Y6378 International Pressure International cooperative effort to understand

Vessel Technical embrittlement of reactor pressure vessels and

Cooperative Program other components. This program will keep staff

aware of international developments in reactor

pressure vessel integrity.

Y6533 HSST-3 Development of fracture mechanics

(Heavy Section Steel methodologies; The ACRS questions the need for

Technology) the large investment in heavy section steel

research.

37 NUREG-1 635
Table 7. Research Activities In Materials and Metallurgy

_________(Continued)

Job Code Title Comment

Y6951 Fracture Mechanics Fracture mechanics of heavy section steel. The

Technology for LWR ACRS questions the need for the large investment

in heavy section steel research.

Y6638 Statistical Analysis of Assist NRC staff in developing a revision to

RPVSteels Regulatory Guide 1.99, 'Radiation Embrittlement

of Reactor Vessel Materials." This research

directly supports the regulatory process.

N6223 FAVOR 4.1 Sampling Validation of new features of the FAVOR computer

Validation code for fracture analysis of vessels. FAVOR is

NRC's computer code for fracture mechanics

analysis and is used extensively.

Y6656 Risk Inform Appendices Develop a risk-informed revision to 10 CFR 50,

G & H Appendix G on Fracture Toughness Requirements

and Appendix H on Reactor Vessel Material

Surveillance Program.

N6227 SMIRT-18 Conference Costs associated with presentation of papers on

Registration NRC research projects at the Structural Mechanics

in Reactor Technology meeting.-

N6097 SMIRT 18 Financial support to publish proceedings of the

1 International SMIRT conference.

NUREG-1635 38
10 OPERATIONAL EXPERIENCE

The analysis of operating data is a

cornerstone in the NRC's increased use of

risk information in regulatory processes. Such

analysis provides current information on

initiating events, component failure data, and

the risk profiles of licensees. Comparison of

these results to goals in the agency's Strategic Plan provides a measure of

regulatory effectiveness and inputs for the

agency's annual report to Congress on

significant operating events.

The NRC research activities associated with

operational experience are listed in Table 8.

The Accident Sequence Precursor (ASP) Program, Y6815, and the Industry Trends

Program, Y6546, alertthe staff and industryto

component failures as old or replacement

components age or operations change. Data

derived from operating experience will validate

or refute the assumption that aging management programs are sufficient to

ensure the operability of both active and

passive components. The operating experience programs provide data that can

be the bases for regulatory decisions to

improve safety. These programs also support the Reactor Oversight Process, including the

determination of the safety significance of

inspection findings and the development of

industry performance indicators.

Two tasks in the research of operational

events, "Method to assess Effect of Design and Operations Margins," N6082, and

"Procedure Development for External Events,"

Y6814, are important efforts to extend the use

of quantitative risk assessment into external

events, including fire, and low power and

shutdown operations.

ACRS is supportive of the research activities

in the area of operational experience and

recommends that these activities be

continued. In light of the limited resources

Uses of Operational Data and Analyses in

Regulatory Activities

U

U

U

U

U

U

U

U

U

U

U

U

U

U

U

U

allocated to these tasks, RES has done a

commendable job in producing outputs in

well-documented and thorough fashion. Tasks

that are currently in the 2005 Research Plan

related to Operational Experience should

remain funded and should be continued for

the foreseeable future.

Staff engaged in the collection and analysis of

operating experience data might also be able

to improve the state-of-the-art in PRA

modeling. Specifically, they might be able to

use operating experience data to derive

higher resolution models of system and

component operability. Currently, PRAs use

success criteria models. A system or

component that meets the success criteria is

deemed operable This "go/no go" model is

not entirely realistic. There is no assessment

of margins, equipment aging, changing plant conditions, etc. Success criteria models may not provide adequate answers for some

applications such as power uprates, containment overpressure credit, license

39 NUREG-1635

t

m m
renewal, sump screen clogging, or any set of

plant conditions that are in some way offnormal or even outside 'the design

specifications of the equipment. There have

been several events that were surprises

because the phenomena that caused or

contributed to the failure mode had not been

realistically modeled. Certainly, the recent

Davis-Besse event involving corrosion of the

reactor pressure vessel head penetrations

comes to mind. Staff granted a small

extension to ordered shutdown date for

reactor pressure vessel penetration

inspections. They did so, in part, because the

calculated risk was small. Unfortunately, the

phenomenological modeling of the head

penetrations and their corrosion was

incorrectly used in the risk assessment.

Development of improved models of system

and component operability models will require

that choices be made concerning areas where

improved modeling will yield useful

improvements in the risk predictions. The

issues of interest may themselves dictate

where choices for improved modeling should

be made. Some modeling improvements are

being made now on an adhoc basis. There is

no need to continue to 'do so if a more

structured approach could result in better

models with wider applications..,

e , I .

I 11t

; i

I

NUREG-1 635 40
Table 8. Research Activities in Operational Experience

Job Code Title Comment

N6082 Method to Assess Effect Provides a methodology to assess the effects of

of Design and Operations changes to design and operation on plant safety

Margins margins. This program provides direct support

for the regulatory process.

Y6468 Reactor Operating Collect operational data for reactor systems, Experience Data for Risk components, initiating events, common-cause

Applications failures and fire events. Data collected in this

program is of use for validation of PRA models.

Y6546 Industry Trends Program Includes grid concerns. This is an essential

program for NRC.

Y6864 Operating Event Support for technical expertise in operating

Technical Support events.

Y6816 SDP/ASP Standardization Develop analysis guidelines for operating events

during low power/shutdown conditions. This

program will extend the ASP program to include

events during shutdown operations.

Y6815 Accident Sequence Systematically screen, review and evaluate

Precursor Analysis operating events. This is a flagship program at

NRC.

Y6987 Expert Elicitation Process Develop guidelines for obtaining and using -Accident Sequence expert opinion in ASP analyses. The useful

Precursor Program elicitation of expert opinion is of growing importance in the risk-informed regulatory

system.

Y6814 Procedure Development Expand the scope of ASP analyses to include

for External Events the calculation of risk from external events and

from low power and shutdown modes of

operation. This program will help extend the

scope of the ASP program.

41 NUREG-1 635

11 PROBABILISTIC RISK ASSESSMENT

Probabilistic risk assessment has become an

essential technology for NRC as it evolves the

regulatory system to make greater use of risk

information. The NRC- research activities in

probabilistic risk assessment are shown in

Table 9. Probabilistic risk assessment has

become pervasive within the research

program. Other activities nominally part of the

development of PRA are addressed in other

Chapters of this report. See especially the

discussions of Digital Instrumentation and

Control Systems (Chapter 4), Fire Safety

Research (Chapter 5), Human Factors and

Human Reliability Research (Chapter 8), and

Operational Experience (Chapter 10). The

staff involved in PRA research has been

extraordinarily productive since the last ACRS

report the NRC research program. A major

focus of the current PRA research is to

support the ROP, which uses risk information

-for monitoring the operations of nuclear power

plants and acting on inspection findings and

deviation of performance indicators from

established thresholds.

The ROP makes heavy use of the SAPHIRE

computer code and the SPAR models of

specific plants. The SPAR model development

program has become an essential element of

the ROP. The ability to develop a SPAR model

for each nuclear power plant has only been

feasible because of the existence of Level I,

internal events, PRAs for each plant. Each

SPAR model begins with a basic model of a

plant system for a generic category of plants

(e.g., a BWR4 reactor with a Mark I

containment). The SPAR model is then made

plant specific through upgrades based on

discussions with the licensee. NRC has found

it essential to develop its own risk-assessment

model for each plant as a matter of

practicality. It would be difficult for the NRC

staff to take a variety of plant PRAs, which use

different platforms and approaches, make

them operational at NRC, and have

knowledgeable staff available to'execute and

update each plant model. NRC development

of SPAR models for individual plants has also

enhanced the plants' risk assessments.

A major Issue that confronts the use of risk

information in nuclear power plant regulation

is the question of incompleteness of individual

plant risk assessments. The Individual Plant

Examination (IPE) program and subsequent

evolutions at the nuclear power plants led to

development of Level I, internal events, PRA

models of all of the operating. These PRAs

meet (or with modest effort can meet) the

requirements of industry standards for internal

events PRAs. The same is not true for the

assessment of risk from fires, floods, seismic

events and for plant modes of operation that

differ from full power operations.

Furthermore, the capabilities to assess risk at

Level II, radionuclide release and source

terms, lag far behind the Level I capabilities.

The NRC staff has plans to expand the scope

of the SPAR models' to include treatment of

risks from fire-initiated events, seismic events

and shutdown modes of operations. These

plans are, however,'not well developed. There

is furthermore the question of availability of

resources needed to undertake these efforts.

The expansions of the scope of SPAR models

will be challenging because all licensees do

not-have 'sophisticated risk assessments in

--these areas for comparison and validation of

NRC's SPAR models with expanded scope.

The NRC staff could develop generic models

accounting for the major features of the plant

designs, but the staff would not be able to

upgrade the generic models to become plantspecific models as was done for the

treatments of risk from internal events. In

addition, fire and seismic risk assessments

differ qualitatively from internal events risk

assessments since the events occur in

"areasr of a plant and affect multiple systems

43 NUREG-1 635
ratherthan just specific components in specific

systems. Fire and seismic risk assessments

require detailed knowledge of spatial

relationships in addition to functional

relationships. Spatial relationships, of course,

vary substantially even among plants of the

same generic type. Despite these challenges,

the regulatory oversight value of full-scope

SPAR models is very high. Over the next year,

the staff should develop its approach and

plans for the expansion of the scope of the

SPAR models to treat external events,

shutdown modes of operation and even to go

to Level II analyses that include, accident

progression and the release of radionuclides

to the environment. Even if it is not possible to

have plant-specific models in the near term,

the generic shells should be available and can

be adapted to be plant specific in the future or

can be upgraded in particular areas to address

specific regulatory issues.

Another barrier to the greater use of risk

assessment in the regulatory process is the

question of uncertainty in the risk predictions.

There are, of course, parametric uncertainties

and the agency has active programs to better

understand the important parametric

uncertainties (See especially Chapter 10,

Operational Experience). There are also

issues of uncertainty in the models adopted in

PRA. Uncertainties in the models of human

reliability and passive system reliability are

significant examples. It has become common

now for the NRC and the licensee to agree

upon a model appropriate for particular

regulatory activities. This agreement can often

be based on familiarity or expedience. The

disturbing trend is for the staff to conclude,

then, that there are no longer uncertainties

associated with the results predicted by the

agreed upon models. Staff needs to ensure

that it treats uncertainty in risk assessments in

a more defensible manner. Research needs to

provide the tools and understanding so that

this can be done.

The staff has also been revising 10 CFR 50.46

to account better for risk information. This is

challenging and important work. Even more

challenging is the effort to develop a

"technology-neutral" alternative to the current

regulatory framework. The ACRS views such

a technology-neutral regulatory framework as

essential in the future and feels that it needs

more attention.

Altogether the scope and the number of PRA

research activities are quite impressive. The

ACRS cautions, however, that NRC should

not allow its work in such a crucial technology

as risk assessments become totally devoted

to the support of line activities. Methods

development is still important. As an

example, the ACRS notes that considerable

research is being reported in the literature

regarding Binary Decision Diagrams'as tools

for solving large fault tress without resort to

cutoff frequencies as is now done. Some

researchers report that the unavailability of

highly redundant systems could be

underestimated significantly when cutoff

frequencies are used for the analysis.

Although no definitive evidence has yet been

produced to show that methods used in the

NRC's SAPHIRE code are inadequate, the

staff needs to review the literature concerning

Binary Decision Diagrams and evaluate the

need to adopt this technology. The growing

importance of the SAPHIRE code and the

-SPAR models in the regulatory process

warrants such an investigation.

. . . e

NUREG-1 635 44
Table 9. Probabilistic Risk Assessment Research Activities

Job Code Title Comment

N6027 PRA for Dry Cask Storage A variety of tasks including uncertainty

Follow Up analysis and extension to multiple casks.

This program supports licensing and

inspection oversight of cask vendors.

N6105 Guidelines for the Complete the technical basis for the internal

Communication of Risk risk communication guidelines. This task

Information completes the technical basis for internal risk

communication guidelines. The ACRS

remains concerned that publically available

information on risk analyses may not be

sufficient to ensure public confidence in a

risk-informed regulatory process.

Y6842 Guidance for the Quantitatively assess the importance of

Development of Latent latent errors and the treatment of latent

Errors errors in PRAs. This project has been

deferred until FY2007. The ACRS cautions

that operating experience shows that latent

errors may be four times more common than

active errors in important reactor events. The

work should not be deferred further.

J8263 Reactor Oversight Process Development of performance indicators to be

Support incorporated into the ROP.

Y6370 Development of Risk-based Support for the Mitigating Systems Performance Indicators Performance Index.

These programs support the ROP.

Y6626 Access to INPO's EPIX Data-base on equipment performance and

C System reliability.:

J8258 International Common Cause Sharing of data on common-cause failures

Exchange Project with the international reactor safety

community. This program keeps staff abreast

of international findings concerning commoncause failures.

N6008 Passive Components This program should prioritize passive

Conditional Core Damage components for consideration in the

Probability proactive materials degradation assessment

(Project Y6868, Materials and Metallurgy,

Chapter 9).

45 NUREG-1 635
Table 9. Probabilistic Risk Assessment Research Activities

(Continued)

Job Code Title Comment

Y6153 SPAR Model Development: Develop SPAR models for evaluation of large

Level2/LERF early release frequencies.

N6090 SPAR Model Development: Develop logic models for analyzing low Shutdown Models power and shutdown internal events.

SPAR Model Development: Identify methods to characterize risk during W6355 Low Power Shutdown low power or shutdown operations.

SPAR Model Development: Revision of Level 1 SPAR models to better

W6467 Level 1 Rev. 3 Models reflect as built and operated plants.

.Y6595 SPARModel Development: Development models of external events for

External Events Analysis the SPAR codes

SPAR Model Development: These are important programs to support the N6075 Enhanced Level 1, Revision expanded scope of the SPAR models.

3 Models

Y6394 Maintain and Support Testing to ensure that SAPHIRE is a stateSAPHIRE Code and Library of-the-art PRA code.

-_ of PRA

N6172 Participate in the MERIT Base program supports risk informing 10

Program CFR 50.46 and includes development of a

(Maximizing Enhancements probabilistic LOCA code, non-piping

In Risk Informed component degradation, and pressurized

Technology) water stress corrosion cracking. This

international program supports one of the

important NRC initiatives.

NUREG-1635 46
Table 9. Probabilistic Risk Assessment Research Activities

(Continued)

Job Code Title Comment

N61M11 Technical Support for 10 Quantification of the effect of break size

CFR 50.46 Task Order 3 reduction and alternative break locations on

margin to existing alternate acceptance

criteria

Y6538 Technical Development of LOCA Frequency Provide LOCA frequency estimates for use in

Distributions revision of 10 CFR 50.46.

These programs are needed to support risk

___________ informed revisions to 10 CFR 50.46.

K6081 PRA Techniques In Risk- Develop methods for uncertainty analysis for

informed and Performance- risk-informed purposes.

based Regulation This is a cooperative agreement with a broad

scope. In addition to potential methodological

contributions it has an educational value.

N6107 10 CFR 50.48c related In collaboration with EPRI, develop a

Technical Activities comprehensive set of risk methods, tools and

data to understand and evaluate risks from

fires.

W6224 Risk-informing Part 50 Develop recommendation on changes to 10

CFR Part 50 to make it risk-informed.

Y6492 Assess Possible Part 50 Develop recommendations to specific Risk-informed Changes requirements in 10 CFR Part 50 to make

them risk-informed.

These program support the initiative to risk

inform 10 CFR Part 50.

W6970 Support to Develop Provide guidance on the use of industry

Consensus PRA Standards standards for PRA.

Revise Regulatory Guide 1.200 based on

W6971 Support In Development of industry pilots and Revision 1 to ASME PRA Consensus PRA Standards standard.

These program support the Commission's

phased approach to PRA quality.

47 NUREG-1 635
Table 9. Probabilistic Risk Assessment Research Activities

(Continued)

Job Code Title Comment

Y6103 Low Power and Shutdown Program to extend the scope of SPAR

Risk Study - Level 2 models to include accident progression for

accidents initiated during shutdown

operations. Premature at this point.

N6133 Development of Consensus Support for staff in development of ANS Low

on PRA Power and Shutdown operations PRA

___ _ Standard.

N6134 Low Power/Shutdown Level Project provides support for staff involvement

I and Fire Risk Standard in the development of ANS standards on

PRA for low power/shutdown operations and

. fire-initiated events.

Y6371 Risk Associated with Cable Addresses the inclusion of aging effects intoAging PRA.

NUREG-1 635 48
12 SEISMIC RESEARCH

As the design of nuclear power plants

improves, the seismic-hazard and seismic

response of the plants can make an

increasingly important contribution to risk.

Seismic hazard analysis and structural

response are not areas where NRC must

maintain state-of-the-art expertise. Such

expertise is available to the NRC on a

contractual basis. As noted in our previous

report, seismic research activities at NRC can

be confined to support needed updates to

regulatory guides and collaborative work with

the international community to stay abreast of

developments in other Countries. The current

research program is, indeed, largely focused

on needs of the regulatory process and a few

important international collaborations.

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49 NUREG-1 635
Table 10. Seismic Research Activities

Job Code Title Comment

N6020 Seismic-induced Passive Review of work by national laboratories and

Component LOCA industry on piping degradation and failure under

Frequencies earthquake loads; Work being done to upgrade Frequencies Regulatory Guides.

Y6481 SSHAC Method 10-year update of the Probabilistic Seismic

Hazard Assessment used in evaluation of early

site permits, work to support update required by

regulations.

Y6718 Sofl-structure Interaction for Review adequacy of current NRC guidelines

Buried Structures concerning soil-structure interactions; work to

update Regulatory Guides.

N6112 Evaluation of Seismic Siting Review of ASCE Standard 43-05, "Seismic

Design Criteria for Structures, Systems and

Components in Nuclear Facilities."

N6076 Japanese Collaboration on Collaboration with Japan on seismic tests and

Seismic Issues analyses; Collaborative work give NRC access to extensive work underway in Japan.

W6081 Japanese Collaboration on Supports work in U.S. in connection with

Seismic Issues collaboration.

N6102 Reg. Guide 1. 165 Update Review of technical advances in the development

Technical Basis of seismic response spectra; prepare draft

revision to Regulatory Guide 1.165.

N6103 Enhancement of the CARES The CARES computer code is used to predict the

Code free field and structural response to seismic input.

(Computer Analyses for Rapid

Evaluation of Structures)

N6219 Resolve Regulatory Guide Regulatory Guide provides up-to-date guidance 1.92 Public Comments for using the response spectrum and time history methods for estimating seismic response of

power plants.

N6104 Ground Motion Seismic Collection and review of new data on the

Hazard Studies propagation of earthquake motion in the Central

and Eastern U.S.; work to support required update

in regulations.

Y6796 IAEA Coordinated RES NRC contribution to international effort to

Project on Seismic Ground understand earthquake effects on nuclear power

Motion plants. Collaborative effort keeps NRC staff

abreast of any international developments.

Y6757 Containment Capacity Confirmatory analyses of structural response and

Studies failure modes of containments under extreme

loading including seismic loads.

NUREG-1 635 50
13 SEVERE ACCIDENT RESEARCH

In the past, NRC invested heaviiy in the

experimental and analytical characterization of

severe reactor accidents. A substantial

technology has been established to

understand the progression of severe reactor

accidents and the radiological consequences of such accidents. Once its immediate needs

were met to understand severe reactor

accidents sufficiently well to estimate risks to

the level of confidence needed to provide

assurance of adequate protection, the NRC

substantially curtailed its investments in

severe reactor accident research. The current d

NRC research activities in the severe accident

area are listed in Table 11.

Research on severe accidents has been

continuing in other countries. Substantial

programs are under way in both Europe and

Japan. NRC has developed an effective

strategy to maintain the technology for severe

accident analysis and to update this

technology with research results from

international programs. The body of

knowledge coming the NRC's past work and

the ongoing international work are

systematized in the useable form in the

MELCOR accident analysis code. At the same

time, the NRC is entering into international

cooperative research programs to obtain data Aeros5 If ap.i.g in a Steam for validating the MELCOR code and___ improving its accuracy and realism. NRC

provides the Cooperative Severe Accident

Research Program (CSARP) as a forum for is pa g in a tton the exchange of severe reactor accident c iv e pr ogan information among Countries. One outcome e m stu in P of this focus of the NRC's research into S easus on severe reactor accidents is that many acciet tP~ htbps eco Countries and institutions have adopted the .ts.a MELCOR code as the preferred tool for the isomntfrWsThhghiksassocited severe accident analysis. itsuhacdnsmyStefrmcneats

i tarhe .aerosolt decontamination assumsed -in 

A new version of the MELCOR code has been acid a gets released to users. NRC is collaborating with Test rest aex dtoproidthe basis-fo researchers in Russia to modernize MELCOR more realistcanalyss of accidents. to use FORTRAN 95 coding. MELCOR is

51 NUREG-1 635
being used for licensing actions. The

capabilities developed to perform detailed

parametric uncertainty analyses with the code

are especially attractive.

RES is also maintaining the MACCS code for

the analysis of consequences of accidents at

nuclear facilities. This code is widely accepted

in the U.S. as a tool for consequence

analysis. Its maintenance at near the state-ofthe-art is important to the agency and the

ACRS is supportive of the current research

programs.

Collaborative severe reactor accident

research programs that NRC has joined are

making good technical progress and there

have been notable accomplishments in the

last 2 years.

* PHtBUS-FP

The Phebus-FP program consists of largescale prototypic experiments involving the

degradation of irradiated reactor fuel,

release of fission products as vapors and

aerosols, and transport of these fission

products through a model of a reactor

coolant system into a model of a reactor

containment. These are the most prototypic

and most comprehensive severe accident

experiments that have ever been performed.

The last of these tests was completed

recently. The experiments have proved to

be invaluable for the validation and

improvement of the MELCOR code and the

validation of the alternative source term

used for a large number of licensing actions.

The program has revealed a number of

unanticipated phenomena and refined

understanding of other phenomena. NRC

has joined a second-generation program

that will involve about 15 Nations to conduct

separate effects tests to further understand

the important accident phenomena revealed

in the PHEBUS-FP test program. This

follow-on program addresses the

containment chemistry of radioactive iodine,

fission product chemistry in the reactor

coolant system, the effects of boron carbide

control rods on core degradation and fission

'product chemistry, and the release of fission

products from high-burnup fuel and MOX

fuel.

* ARTIST

TheARTISTtest program is an international

collaborative effort undertaken in

Switzerland to ascertain the amount of

decontamination that can occur in the

secondary side of steam generators in PWR

accidents initiated by steam generator tube

ruptures or initiated by other means but

involving steam generator tube ruptures.

Such accidents have been found to be risk

dominantforsome PWRs. During lastyear, the scoping test program -ihas been

completed. Results of the tests show that

-decontamination is modestly larger than

what had been anticipated in accident

analyses. Plans are being formulated now to

conduct integral system tests and additional

tests to support modeling of secondary side

decontamination.

* MASCA

-The MASCA test program and its

predecessor the RASPLAV program were

undertaken to understand the technical

feasibility of retaining core debris within

reactor 'pressure vessels, especially' with

water flooding the outside of the vessel.

These programs were conducted in Russia

and involved the development of technology

to produce large scale melts of prototypic

core debris involving U02, ZrO2 and Zr.

The major tests in the program have now

been completed. Efforts are under way to

identify and maintain' the experimental

capabilities that have been developed for

the- MASCA program since these

capabilities may be essential for the

investigation of severe accidents in reactors

that do not use light water technology.

NUREG-1 635 52
OECD-MCCI

This is an international collaborative

experimental study being conducted at the

Argonne National Laboratory to investigate

the viability of using an overlying layer of

water to cool core debris interacting with

structural concrete. This program is nearing

completion.

Planned modifications of the MELCOR code

to address the ACR-700 have been curtailed

since the application for certification of this

reactor has not been submitted. There still

may a need to upgrade the modeling of iodine

chemistry in reactor containments to respond

to recent findings concerning the effects of trisodium phosphate buffer in reactor sumps on

sump pump screen blockage.

The ACRS is very supportive of the strategy

NRC has developed to maintain and update

its capabilities for severe accident analyses.

The leveraging of resources through

international collaborative experimental

research is especially important. The planned

extensions and continuations of current

collaborations are well worth the investment.

This type of collaboration in experimental

research could be emulated in other NRC

research areas such as fire safety research

and thermal-hydraulics research.

53 NUREG-1 635
Table 11. Severe Accident Research Activities

Job Code Title Comment

Y6321 Benchmark, MOX Fuel International Collaborative follow-on to the

Release, Source Term PHEBUS-FP experiments.

Experiments

Y6328 Assessment and Analysis of In-kind support for the follow on to the

PHEBUS-ST PHEBUS-FP experiments. This work is

providing data on fission product behavior

during reactor accidents for use in MELCOR

development.

Y6628 Consequence Models and Uncertainty analysis of the MACCS code for

UncertaintyAssessment computing reactor accident consequences.

Y6313 OECD-MCCI Program International collaborative research on the

interactions of core debris with concrete. This

program should be completed next year

Y6690 Analysis Support for OECD- In-kind and financial support for the

MCCI Program international collaborative research on exvessel core debris interactions with concrete.

Y6312 MASCA Program International collaborative research on the

behavior of molten core debris in the lower

plenum of a reactor vessel. This program has.

resolved safety issues with respect to invessel

retention of core debris. The program has

developed the capability to produce and test

large-scale melts of uranium dioxide that may

be of use in advanced reactor safety model

development and validation.

Y6802 MELCOR Severe Accident Computer model for the analysis of severe

Code Development and reactor accident and repository for severe

Assessment accident research results. This is the agency

tool for Level 2 PRA including source term

characterization; MELCOR is the repository for

severe accident research results obtained by

the agency.

Y6721 AGT WABRAE-RAS on Support for Russian investigators in the

Nuclear Safety Analysis development of a FORTRAN-95 version of

Codes MELCOR. This program is modernizing the

coding in MELCOR by cost-effective use of

expertise in Russia.

NUREG-1635 54
Table 11. Severe Accident Research Activities

(Continued)

Job Code Title Comment

Y6848 High Burnup Fission Refine release models in MELCOR for the

Product Release Data effects of high fuel bumup; code analyses will

be used to create a licensing source term

applicable to high-burnup fuel and reflecting

improved modeling of severe accidents.

Y6517 High Burnup Source Term Establish the technical basis for the extension

for Storage of regulatory guide on spent fuel heat

generation in a spent fuel storage facility to

include high-burnup fuel

Y6504 Steam Generator Fission International collaborative research on the

Product Retention retention of aerosols on the secondary sides of

steam generators in containment bypass

accidents (ARTIST program). This program

provides an experimental resolution of a longstanding issue of source terms from accidents

that bypass containments.

Y6607 Support ARTIST Tests In-kind support for the ARTIST program - see

Y6504 above.

Y6486 Severe Accident Initiated Investigation of the potential for induced steam

Steam Generator Tube generator tube failure during severe accidents

Rupture Sequences leading to containment bypass. This is an

important part of the Steam Generator Action

plan and the analysis of plant behavior under

accident conditions.

Research Programs to Maintain the MACCS Code for Consequence Analysis

Y6785 Plume Model Adequacy Test the assumption that simple plume

Evaluation treatments in MACCS code are adequate by

comparing with the state-of-the-art dispersion

model. This activity is important to show MACCS

is adequate for regulatory needs.

Y6628 MACCS Uncertainty Support for emergency planning.

Assessment for

Consequence Models

Y6469 Evaluation of Radionuclide Upgrade information on uptake pathways.

Pathways and Uptakes This project upgrades the code to take

advantage of more recent information.

_

55 NUREG-1 635

14 THERMAL-HYDRAULICS RESEARCH

Thermal hydraulics, especially the dynamics

of two-phase flow, have always been essential

elements of the regulatory evaluation of

design basis accidents. NRC confirmatory

evaluation of licensees' submittals in the area

of thermal hydraulics has long been a major

element of many licensing actions. Thermalhydraulic analyses have grown ever more

sophisticated. This trend is likely to continue

for existing plants as licensees seek power

uprates and take advantage of NRC's

willingness to allow best-estimate analyses (with scrupulous attention to uncertainties) in

the place of deliberately bounding,

conservative analyses. To evaluate the

adequacy of the licensees' analyses, NRC

must have state-of-the-art thermal-hydraulic

computational tools and equally sophisticated

understanding of both thermal-hydraulic

phenomena and the limitations of computer codes. NRC attempts to maintain its

competence in the thermal-hydraulic field

through its research program.

Major elements of the current NRC thermalhydraulics research program can be grouped

into three general areas:

* PWR sump screen blockage issues

* TRACE computer code development

* Experimental studies of thermalhydraulic phenomena

These major features of the current thermalhydraulics research program are discussed

below.

PWR Sump Screen Blockage

The sump screen blockage issue for PWRs is

the analog of a previous issue identified for

BWRs. Debris from coatings and insulation

can be generated during the high-pressure

blowdown of the reactor coolant system

Chemical Effects/Head-Loss Tests in a

Simulated PWR Sump Pool Environment

GSI-191 addresses the potential for debris

accumulation on PWR sump screens to affect

emergency core cooling system (ECCS) pump net

positive suction head margin. In response to a

concern expressed bytheACRS, RES has initiated

a program to investigate the potential for chemical

reactions that can occur in the containment pool to

produce chemical products that can increase the

head losses over those due to the physical debris

alone.

NRC and the nuclear utility industry jointly developed an Integrated Chemical Effects Tests

(ICE7) program to determine if chemical reaction

products can form in representative PWR post- LOCA containmentsump environment. These tests

were conducted by Los Alamos Natonal

Laboratory (LANL) at the University of New Mexico

(UNM). Chemical products were observed in all

five test series.

A head-loss loop was set up at Argonne National

Laboratory (ANL) to investigate the potential head

loss associated with the chemical products observed in the ICET tests.

These recent research results indicate that a

simulated pool environment containing phosphate and dissolved calcium can rapidly produce a

calcium phosphate precipitate that, if transported to a fiber bed covered screen, produces significant head loss.

57 NUREG-1635
following a major pipe break. This debris can

clog the screens protecting the intake pumps

forthe emergency cooling system and prevent

adequate coolant flow. Blockage issues have

been exasperated by the discovery of

mechanical and chemical effects that magnify

the blocking effects of debris trapped on the

sump screens. As a result, it is difficult to

design screens that are of sufficient size to

ensure emergency core cooling. The industry

is looking to the NRC for guidance on

acceptable methods for sizing screens to

protect the sump intakes of the cooling

pumps.

The NRC is still in the exploratory phase of

research on sump screen blockage. It is still

identifying phenomena that affect blockage. It

is far from developing tools and methods that

can -be used 'with confidence for making

predictions. NRC staff is now analyzing the

licensees' responses to Generic Letter 2004-

02, 'Potential Impact of Debris Blockage on

Emergency Recirculation During Design Basis

Accidents at Pressurized Water Reactors."

These responses should reveal the licensees'

views of current predictive techniques and

their applicability, as well as indicate what'

methods they expect to use to assess the

adequacy of their current and modified screen

systems. The NRC- staff needs to have

sufficient technical knowledge to evaluate

these methods. Current NRC research is

focused on significant gaps in knowledge,

establishing what phenomena play significant

roles, and on developing general awareness

of what analytical steps are needed to

describe the phenomena adequately. The

ACRS would expect that many of the details

of predictive methods, such asthecoefficients

in correlations, computational schemes, and

methods for developing suitable conservatism

to account for uncertainty, could be left to the

licensees or to industry-sponsored

organizations such as EPRI. This is possible,

however, only when the phenomena are well

understood and a technical basis has been

established for their prediction. When this is

not the case, the NRC may need to develop

sufficient* predictive ability of its own to

achieve authoritative competence to evaluate

licensees' submittals.

For example, the NRC-sponsored research

has revealed -the "thin bed effect'. This

appears to involve a dense agglomeration of

fine particles that fill the pores in a layer of

debris, such as fiberglass, but the mechanism

by which it occurs and how' it influences the

pressure drop are not understood. Previous

NRC acceptance of pertinent Nuclear Energy

Institute (NEI) guidance' now appears

premature in light of confirmatory research

that has revealed much larger influence of the

bed structure (e.g. up to a factor of about 100

on pressure drop for the same mix of fibers

and 'particles) than had previously been

thought to be possible. Research in this area

sh6uld be continued and expanded as needed

in order to reduce the very large uncertainties

surrounding these effects and to determine if

a predictive capability is feasible.

Other important phenomena, such as

chemical and downstream (of the screen)

effects are now being investigated by RES.

These are essentially exploratory studies that

have uncovered some significant effects, but

have yet to reveal their-scope and magnitude.

Predictive capability remains to be

demonstrated. The NRC needs to evaluate

the results of these studies and determine

how much it can rely on the nuclear industry

to develop reliable predictive tools'and how

much independent predictive capability it

requires. Development of a predictive

capability may require, investment of

substantial resources and time.

TRACE Computer Code Development

Several years ago, the NRC recognized that

it could 'not sustain the continued

maintenance of several thermal-hydraulic

NUREG-1 635 58
codes for each general type of nuclear power

plant. It elected to consolidate its existing

codes for the confirmatory analysis of licensee

submittals on design basis thermal-hydraulic

issues into a single code now called TRACE.

The consolidation is now largely completed.

The TRACE computer code is viewed by the

NRC research staff as "as good as anything

else that is out there." The long-term

validation and improvement phase of code

development is at hand. Current research is

devoted to improving features of the TRACE

code, making It easier to use and validating it

against' available data. Some of the data

already exist -and other data 'are being

generated. In addition, the integration of the

TRACE code, coupled with the CONTAIN

code to model containment response and the

PARKS code for neutronic analyses into the

regulatory processes of the agency has

begun.

The TRACE code is reputed to-now be able to

serve as the "workhorse" thermal-hydraulic

analysis code for the agency. In the course of

its work to consolidate thermal-hydraulics

codes into TRACE, the research staff has

found many ways to improve the code. Such

improvements should be done. Now, however,

it is far more important that the integration of

TRACE into the regulatory process be

completed in an expeditious manner. The

research staff working on the development

needs to have input from users of the code on

needed features and capability of the code.

Inevitably, the introduction of a- new

computational tool will slow and detract the

regulatory process for some transient period.

There is no way to counter this difficulty

associated -with -the introduction of a'new

computer code. It must be endured and the

sooner this is done, the sooner the challenges

associated with the use of a new code in the

regulatory process can be overcome. Once

TRACE is integrated into the regulatory

process, the developers will receive valuable

advice on how their efforts to improve the

code should be directed to enhance the

regulatory process.

Highest priority should be given to the

integration of TRACE code into the regulatory

process. As this integration progresses, the

research t staff can continue its efforts to

improve and further develop TRACE on a

"time available" basis. The ACRS is

concerned that efforts to improve TRACE

lack prioritization and defensible organization.

Placing the TRACE code in the hands of

users will also identify a host of needed

improvements. Prioritization of technical

improvements might be aided substantially by

commissioning a detailed peer review of

TRACE. To do this, the staff will have to have

available code documentation of outstanding

scope 'and quality. Such high quality code

documentation will also be needed if the code

is to become part of the regulatory process.

Code documentation, then, is a task that

ought to take precedence in the thermal

hydraulic research effort.

Experimental Studies of Thermal-Hydraulic

Phenomena

Thermal-hydraulic phenomena involving the

flow of two-phase mixtures of steam and

water are very complicated especially those

involving .blowdown from high pressure

systems. Thermal-hydraulic phenomena that

arise in advanced light water reactor designs

that emphasize passive response to accidents

are driven by subtle forces- that require

sophisticated understanding to ensure plant

safety. As a consequence, NRC has long felt

that it cannot rely solely on computer code

projections of thermal-hydraulic phenomena

to ensure adequate protection of the public

'health and safety. Experimental confirmation

is also required. As the computer models

used to analyze thermal-hydraulic phenomena

-have- become more sophisticated, the

experiments needed to validate model

predictions have become progressively more

59 NUREG-1 635
integral in nature. Experimental facilities have

become larger and more complex. RES has

an interest in maintaining these facilities for

use in addressing future as well as current

regulatory issues. Maintenance of large,

complex experimental facilities has become a

significant expense in this research area. The

major experimental facilities used by NRC in

the U.S. are the APEX and PUMA facilities as

well as RBHT facility at Penn State University.

Abroad, NRC is conducting tests at the PKL

facility, the SETH tests and tests at the ROSA

facility. Additional experimental needs may

arise in connection with the design

certification of the ESBWR.

APEX is a medium-size, scaled, integral test

facility that proved useful for the certification

of the AP600 and API 000 reactor designs. It

has been modified to provide data crucial to

the analysis of thermal shock to reactor

-vessels. It is proposed now that the APEX

facility be used for confirmatory analyses for

AP1000 and for some "thermal hydraulic

integral experiments." These proposed

applications would benefit from review to

assess their focus and applicability.

PUMA is a medium size, scaled facility

especially suited for evaluating passive

emergency core cooling systems. It is being

modified to be applicable to testing the

emergency core cooling systems for the

ESBWR.

The RBHT test program has been under way

for a number of years with the purpose of

improving core reflood models that are a key

part of evaluating the adequacy of pressurized

water- reactor emergency core cooling

systems. The reflood models may become

critical if applications are submitted for large

power uprates in PWRs. The proposed

research program at the RBHT facility needs

evaluation to see if the quality, scope and

detail of the data are properly matched to the

proposed uses of these data.

NRC has wisely not sought to duplicate large

test facilities available overseas. Use of these

facilities is possible through international

programs. The SETH program was useful for

resolving Generic Safety Issue (GSI) 185 and

assessing, the emergency heat removal

systems in the ESBWR. Future work under

this program at the ROSA and the PKL

facilities in support of the TRACE code needs

to be more clearly focused.

It is essential for NRC to maintain an ability to

assess thermal-hydraulic phenomena that

occur both in existing reactors and in future

reactors. It is evident that the development of

computer codes to predict thermal hydraulic

phenomena and the experimental validation of

these predictions will grow more burdensome

with time. Major development efforts can be

anticipated if very innovative designs using

coolants other than water are brought forward

for certification. It is not likely that the nuclear

institutions of any one country will be able to

develop adequate codes and conduct

sufficient validation of these codes alone.

International cooperative development of

codes and conduct of experiments appear

essential as NRC research moves beyond

TRACE with its current capabilities and

especially if analyses are needed for coolants

other than water. NRC already takes

substantial advantage of international

experimental capabilities. Extending this

intemational flavor in thermal-hydraulics

research to include the development -of

computer codes will contribute to current

ideas of multi-national design approval

process. It may slow code development. It

also may ensure that sufficient resources for

code development are available so that it is

feasible to meet the more exacting standards

that are likely to be demanded in the future.

NUREG-1 635 60
Table 12.- Thermal-Hydraulics Research Activities

Job Code Title Comment

N6106 Confirmatory Head Loss Testing Experiments to measure head loss

across sump pump strainers in PWRs.

Y6871- PWR Sump Screen Penetration Experiments to determine the type and

and Throttle Valve Testing quantity of debris that can pass through

:____-_-_;_-_typical PWR sump screens.

N6100 Head Loss Testing Assess the susceptibility of recirculation

screens to debris blockage during

design basis accidents.

Y6999 Integrated Chemical Effects Tests Five tests to determine representative

chemical and material environments in

PWRs that can contribute to sump

blockage.

N6121 GSI-191 Chemical Effects Experiments to determine chemical

Simulations effects that can contribute to sump

screen blockage.

N6198 Transportability of Coatings Parametric study to ascertain if coatings

can be transported to sumps under

accident conditions.

N6083 BWR ECCS Suction Concerns Technical assessment of Generic Issue

193"BWR Suction Concems."

Y6769 PUMA Test Facility Facility for the conduct of thermal

hydraulics tests. This facility can

produce data for natural circulation

systems for use in ESBWR design

certification.

Y6852 PWR Thermal-Hydraulics Integral Tests at the APEX facility at Oregon

Experiments State University.

N6042 OECDIROSA Program International collaborative tests of

reactor accident thermal hydraulic

-_______iphenomena.

Y6945 Rod Bundle Heat Transfer Test Experiments at Penn State University in

Program ,-. Phase 3 support of TRACE code analyses of

small and large break loss of coolant

accidents. To date, there is little

evidence that data from this facility can

be of value for TRACE code

development. Further work in this facility

should be scrutinized carefully to assure

that it meets agency needs.

61 NUREG-1 635
Table 12. Thermal-Hydraulics Research Activities

(Continued)

Job Code Title Comment

Y6589 Thermal-Hydraulic Research Perform analytical and small-scale

experimental work in support of the

TRACE code. Neutronic work in this

program in nearly complete. Long-range

thermal hydraulic work needs to be

shown necessary for agency needs.

N6043 Thermal-Hydraulic Sub-channel Analysis for international standard

International Standard problem for a BWR subchannel

benchmark.

Y6571 SETH Program - Test Facilities Thermal-hydraulics tests in two

international efforts: PKL on boron

dilution and PANDA in support of

ESBWR certification.

Y6974 OECD-PKL Program and Test International collaborative research on

Facility boron dilution accidents including'rmidloop operation.

N6213 TRACE Verification and Verification and validation of the TRACE

Validation thermal-hydraulics analysis code. This

work is viewed as vital to the verification

and validation of TRACE.

Y6673 TRAC-M Development and Simulate separate effects tests with the

Assessment - Small LOCA TRACE code and show acceptable

Processes agreement with predecessor codes.

(in the past, the TRACE code was Good progress has been made in this

called TRAC-M) important work.

Y6666 Advanced Numerical Methods In Advanced numerical methods for the

TRAC-M TRACE code. This work is not essential

(In the past, the TRACE code was for the current range of efforts to make

called TRAC-M) TRACE useful to the agency.

N6147 TRACEDevelopmentand Use TRACE code to evaluate level swell

Assessment Against Specified tests done at several facilities. This is a

Tests small part of the TRACE validation and

verification effort.

N6201 Gravity Reflood and SBLOCA Use the TRACE code to assess PUMA

TRACE Assessment facility tests. This work necessary to

lend credibility to TRACE for ESBWR

analysis.

: ...

NUREG-1 635 62
Table 12. Thermal-Hydraulics Research Activities

(Continued)

Job Code Title Comment

Y6525 TRAC-M Code Maintenance Maintenance of the TRACE code. This

(In the past, the TRACE code was is an essential activity.

called TRAC-M)

N6040 Data Acquisition Recover old input decks for the TRACPWR model.

N6072 Implementation of ACR-700 This work is no longer necessary.

(Misleading title, Project deals

with PUMA Input deck)

Y6198 Continuation of Support for Support for the SCDAP/RELAP5

System Code Analysis computer code and the analysis of

steam generator tube rupture accidents.

Y6392 Maintenance, Application, Consolidation of RELAP5 capabilities

Assessment and Development of into TRACE. This work appears to

NRC Computer Codes overlap most of the TRACE

development tasks. Incorporation of

RELAP capabilities into TRACE has

proven difficult because of code

philosophy differences.

Y6667 SNAP Implementation Graphical user interface for TRACE and

other NRC computer codes. This work

is important because of poor direct input

methods inherited in TRACE from the

underlying TRAC models.

Y6662 AP1000 Confirmatory Confirmatory thermal hydraulic

Thermalhydraulics Analysis analyses of a wide range of design

basis accidents hypothesized to occur

in AP1 000. This work is complete.

Y6526 Administer CAMP Meeting Meeting of users of NRC thermalhydraulics codes. This program will

assist in the international acceptance of

TRACE.

N6030 Flow-induced Vibrations and Analysis of component vibration that

Effects on BWR components can lead to fatigue failure in BWRs.

I ..

63 NUREG-1 635

15 REFERENCES

1. U.S. Nuclear Regulatory Commission,

"Review and Evaluation of the Nuclear

regulatory Commission Safety Research

Program," Advisory Committee on

Reactor Safeguards (ACRS), NUREG1635, Vol. 6, March 2004.

2. Code of Federal Regulations, Title 10,

Part 50 (10 CFR 50), "Domestic

Licensing of Production and Utilization

Facilities," U.S. Government Printing

Office, Washington D.C., 2005.

3. 10 CFR 50.46, "Acceptance Criteria for

Emergency Core Cooling Systems for

Light-Water Nuclear Power Reactors,"

U.S. Government Printing Office,

Washington D.C., 2005.

4. 10 CFR 50.48, "Fire Protection," U.S.

Government Printing Office, Washington

D.C., 2005.

5. 10 CFR 50.54, "Conditions of Licenses,"

U.S. Government Printing Office,

Washington D.C., 2005.

6. 10 CFR 50.61, "Fracture Toughness

Requirements for Protection Against

Pressurized Thermal Shock Events,"

U.S. Government Printing Office,

Washington D.C., 2005.

7. Appendix G to 10 CFR Part 50, "Fracture

Toughness Requirements," U.S.

Government Printing Office, Washington

D.C., 2005.

8. Appendix H to 10 CFR Part 50, "Reactor

Vessel Material Surveillance Program

Requirements," U.S. Government

Printing Office, Washington D.C., 2005.

9. 10 CFR Part 52, "Early Site Permits;

Standard Design Certifications; and

Combined Licenses for Nuclear Power

Plants," U.S. Government Printing Office,

Washington D.C., 2005.

10. U.S. Nuclear Regulatory Commission,

Supplement 4 to Generic Letter (GL) 88-

20, "Individual Plant Examination of

External Events (IPEEE) for Severe

Accident Vulnerabilities - 10 CFR

50.54(f)," June 28, 1991.

i1. U.S. Nuclear Regulatory Commission,

"Procedure and Submittals Guidance for

the Individual Plant Examination of

External Events (IPEEE) for Severe

Accident Vulnerabilities," NUREG-1407,

June 1991.

12. U.S. Nuclear Regulatory Commission,

"Staff Requirements Memorandum

(SRM), SECY-4-01 11, Recommended

Staff Actions Regarding Agency

Guidance in the Areas of Safety

Conscious Work Environment and Safety

Culture," August 30, 2004.

13. Pouchet, A., "The European Benchmark

Exercise on Human reliability Analysis,

'Proceedings of American nuclear

society International Topical Meeting on

probability, reliability, and safety

Assessment (PSA '89), pp. 103-110,

Pittsburgh, PA, April 2-7, 1989.

14. U.S. Nuclear Regulatory Commission,

Regulatory Guide 1.99, "Radiation

Embrittlement of Reactor Vessel

Materials," Rev. 2, May 1988.

65 NUREG-1 635
15. U.S. Nuclear Regulatory Commission,

Regulatory Guide 1.200, "An Approach

for Determining the Technical Adequacy

of Probabilistic Risk Assessment Results

for Risk-Informed Activities," February

2004.

A16. U.S. Nuclear Regulatory Commission,

Regulatory Guide 1.165, "Identification

and Characterization of Seismic Sources

and Determination of Safe Shutdown

Earthquake Ground Motion," March

1997.

17. Generic Safety Issue 191,:"Assessment

of Debris Accumulation on PWR Sump

Performance."

18. Generic Safety Issue 185, "Control of

Recriticality Following Small-break Locas

in PWR."

NUREG-1 635 66
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(9.2004) (Assigned by NRC, Add VoL., Supp., Rev.,

NRCMD 3.7 and Addendum Numbers, it any.)

BIBLIOGRAPHIC DATA SHEET

(See kistucons on the reverse) NUREG-1635, Vol. 7

2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED

Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program - A MONTH YEAR

Report to the U.S. Nuclear Regulatory Commission May 2006

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

Hossein Nourbakhsh Technical Report

7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (IfNRC, provide Dvision Otfice or Region U.S. Nuclear Regulatory CommissIon, and mailing address; if contracor.

pmovide name and matng address.)

Advisory Committee on Reactor Safeguards

U. S. Nuclear Regulatory Commission

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION -NAME AND ADDRESS (ffNRC, type 'Same as abovey Ifcontractor, provide NRC Dvision, Office orRegion, U.S. NuclearRegulatory commission,

and malinig address.)

Same as above

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

This report presents the observations and recommendations of the Advisory Committee on Reactor Safeguards (ACRS)

concerning the NRC Safety Research Program being carried out by the Office of Nuclear Regulatory Research (RES). This

report focuses on that portion of the NRC research program dealing with the safety of existing nuclear reactors and advanced

light water reactor designs, Economic Simplified Boiling Water Reactor (ESBWR), submitted for certification. The research

strategy for more advanced reactors that are not based on water reactor technology such as the Generation IV reactors being

studied by the Department of Energy is also discussed. The ACRS considered the programmatic justification for the research

as well as the technical approaches and progress of the work. The evaluation identifies research crucial to the NRC missions.

The ACRS also attempts to identify research that had progressed sufficiently to meet current and anticipated regulatory needs

so that It could be curtailed in favor of more Important activities. This report does not address research on the security of

nuclear power plants. Comments on such research will be reported separately. Also, the ACRS does not comment on research

activities dealing with nuclear waste issues. The Advisory Committee on Nuclear Waste (ACNW) will report on these research

activities.

12. KEY WORDS/DESCRIPTORS (List words orpirases fht lasstrsearchers inlocaing thereport) 13. AVAILABILITY STATEMENT

advanced reactor research probabilistic risk assessment unlimited

digital instrumentation and control systems reactor fuel research 14. SECURITY CLASSIFICATION

fire safety research seismic research (is Page)

human factors and human reliability research severe accident research unclassified

materials and metallurgy thermal-hydraulics research (Ths Rep)

neutronics and criticality safety unclassified

observations and recommendations

operational experience 15. NUMBER OF PAGES

16. PRICE

NRC FORMU 335 (9-2004) PRINTED ON RECYCLED PAPER
Federal Recycling Program

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