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NUREG-1635

Vol. 2

Review and Evaluation of the

Nuclear Regulatory Commission

Safety Research Program

A Report to the U.S. Nuclear Regulatory Commission

Manuscript Completed: June 1999

Date Published: July 1999

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

with contributions from the

ADVISORY COMMITTEE ON NUCLEAR WASTE

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

FOREWORD

In 1998, the Advisory Committee on Reactor Safeguards (ACRS) submitted to the Nuclear

Regulatory Commission (NRC) a comprehensive report of the NRC's Safety Research Program,

NUREG-1635, Vol. .1, "Review and Evaluation of the Nuclear Regulatory Commission Safety

Research Program," which documented the ACRS conclusions and recommendations. The ACRS

continues to support the conclusions and recommendations of that report. The present report is more

modest in scope and is intended to provide additional information concerning the research needed

to support Commission programs, especially risk-informed regulation. Not all research programs

are included in this report and our observations are often limited to certain aspects of a given

program. This does not imply in any way that these programs are less important to the

Commission's activities.

The primary reason the NRC carries out nuclear safety research is to provide the knowledge that

constitutes the technical bases for regulatory decisions. When the Commission receives an

application for licensing a new nuclear system or a major modification to an existing system, or is

confronted with an unexpected situation, it often must make decisions on the basis of available

knowledge. It is usually the responsibility of the NRC's Office of Nuclear Regulatory Research

(RES) to ensure that the necessary information is available or to initiate research to secure the needed

information. A properly structured research program can ensure that such information is available

when it is needed. For instance, the Commission has sponsored multi-year research programs into

the aging and deterioration of materials (e.g., heavy section steel) and components (e.g., steam

generator tubes). The resultant understanding is providing invaluable information in dealing with

deterioration of components in operating plants and with many of the issues associated with license

renewal.

This report addresses research in 11 areas:

IV.

II.

Vi.

VII.

VIII.

Ix.

XI.

Probabilistic Risk Assessment

Human Factors and Human-Machine Interface

High-Bumup Fuel Performance and Mixed Oxide Fuel

Instrumentation and Control Systems

Severe Accidents

Fire Protection Research

Thermal Hydraulics

Integrity of Steam Generators

Integrity of Pressure Vessels

Environmentally Assisted Stress Corrosion Cracking

Nuclear Waste-Related Research (Contribution by the Advisory

Nuclear Waste)

Committee on

iiU NUREG-1635
The scope and depth of our reviews of these 11 areas vary considerably and are related to changes

that have occurred in the past year or to our perception that additional attention is timely. The ACRS

is aware of RES's reorganization and self-assessment activities currently under way, as well as its

use of a systematic methodology for prioritizing research activities. The ACRS expects to review

those activities at a later date.

NUREG-1635 iiv
CONTENTS

Page

FOREW ORD ................................................................ iii

INTRODUCTION ............................................................. I

SUMMARY REVIEW OF RESEARCH ACTIVITIES ................................ 3

I. Probabilistic Risk Assessment ........................................ 3

II. Human Factors and Human-Machine Interface ........................... 5

III. High-Burnup Fuel Performance and Mixed Oxide Fuel .................... 6

IV. Instrumentation and Control Systems .................................. 8

V. Severe Accidents ................................................. 11

VI. Fire Protection Research ........................................ 12

VII. Thermal Hydraulics ............................................... 13

VIII. Integrity of Steam Generators ........................ .............. 17

IX. Integrity of Pressure Vessels ........................................ 18

X. Environmentally Assisted Stress Corrosion Cracking ..................... 19

XI. Nuclear Waste-Related Research ..................................... 20

XII. References ...................................................... 22

APPENDIX A: ACRONYMS .................................................. 25

v NUREG-1635

INTRODUCTION

The Nuclear Regulatory Commission (NRC) is undertaking a historic change in the philosophy of

regulating nuclear power plants with the introduction of risk-informed regulation. The potential

rewards of this approach are significant for the agency and for the licensees it regulates. It can, for

example, focus attention and resources on the most important threats to nuclear safety while at the

same time offering a potential reduction of the manpower and financial resources required for

regulation.

The underlying basis of risk-informed regulation is quantitative risk assessment. It is important that

risk-informed decisions be based on adequate knowledge, so that any change in the regulations does

not compromise the safety of the public. This means that the uncertainties that were originally

addressed by conservatism need clear understanding and quantification. To accomplish this,

research will be needed to upgrade the appropriate base of knowledge and methods associated with

probabilistic risk assessment (PRA).

The research program in PRA deals with risk assessment for nuclear power plants during normal

operation. This program is only minimally adequate for dealing with risk assessment for low-power

and shutdown operational modes, even though the perception is that these modes could contribute

significantly to the overall plant risk. The risk from fire-initiated accidents, which is applicable to

both operating and shutdown plants, is another area in which the PRA knowledge base is inadequate.

Additional work is needed on PRA model development. This report discusses research needed to

support such model development in the areas of severe accidents, human factors, fire protection, and

instrumentation and control systems. If the Commission is to have a comprehensive risk-informed

regulatory program, there must be a larger dissemination of PRA tools and results within the agency.

Nuclear power plants are subject to a variety of degradation processes. Some of these (e.g., fatigue

and uniform corrosion) were anticipated in the original design, but others (e.g., stress corrosion

cracking, thinning, denting, etc.) were not. The nuclear industry has expended large resources on

replacements, repairs, and inspection programs to manage this degradation. An area that is important

to the safe operation of plants during their current licensing periods and for license renewal is

assurance of the adequacy of aging management programs. For active systems, the maintenance rule

is intended to provide this assurance. Passive components, such as reactor coolant pressure

boundary, electrical cables, and reactor internals, are also subject to a variety of degradation

processes and are the focus of the current license renewal process. Fortunately, study of these

degradation processes has been a focus of NRC research for many years and has provided needed

insights into the regulatory issues arising from these degradation processes.

NUREG-1635
The ability to conduct more sophisticated and realistic analyses of the integrity of components such

as the reactor vessel and steam generator tubes offers the potential for reductions in conservatism.

The NRC must be able to independently verify that any changes to regulatory criteria will not

increase the probability of failure of these systems to unacceptable levels. The need for independent

verification of regulatory criteria and the technical complexity of the issues require a level of

technical expertise that can be maintained by the agency through the continued support of research

programs in these areas. The way in which several areas of current research respond to these needs

is described in more detail in the body of this report.

The ACRS believes that risk-informed regulation requires a strong research program. A strong

research program will also allow the United States to maintain its influence in nuclear safety

regulation throughout the world.

NUREG-1635 2
SUMMARY REVIEW OF RESEARCH ACTIVITIES

For each of the 11 research areas discussed in this report, we provide some background and our

views on program strengths and shortcomings, and make specific recommendations. The ACRS has

not attempted in this review to be all-inclusive of either individual areas or of the nuclear safety

research program. The Advisory Committee on Nuclear Waste has provided its comments and

recommendations on waste-related research and technical assistance programs.

I. Probabilistic Risk Assessment

Recommendations

* RES should expand its investigations of methods for addressing uncertainties beyond just

parameter uncertainty to include model uncertainty, especially in the areas of human

performance, fire protection, aging, and success criteria.

* Improved PRA models must be developed for the following items:

- Fires

- Software-based digital systems

- Aging of structures, systems, and components (SSCs)

- Human performance

- Safety culture

* PRAs should be performed for low-power and shutdown operations in an expedited manner.

These should be as rigorous as the PRAs for reactors at power.

Discussion

The cornerstone of risk-informed regulation is a PRA of acceptable quality. The release of

Regulatory Guides 1.174, 1.175, 1.176, 1.177, and 1.178 has focused attention on the estimation of

core damage frequency (CDF) and large, early release frequency (LERF), since these are the two

metrics that play a major role in risk-informed decisions.

Uncertainties due to model assumptions have been of concern in the severe accident arena for a long

time. NUREG-1 150 has identified one way of quantifying this uncertainty using expert judgments

dealing primarily with Level 2 phenomena. This work, of course, is very important in calculating

LERF. Model uncertainty has not been treated with equal rigor in the calculations of CDF (Level

1 PRA), although there are isolated instances in which this uncertainty has been addressed. An

example is the timing of reactor coolant pump seal loss of coolant accident (LOCA) associated with

loss of component cooling water.

3 NUREG-1635
There are two needs in this area at this time: (1) the major model uncertainties that exist in the

estimation of CDF must be identified, and (2) the methods for quantifying these model uncertainties

must be developed.

In addition, PRA models need to be improved to better treat fires, software-based digital systems,

aging phenomena, human performance, and safety culture. Further details on'these modeling needs

are discussed elsewhere in this report.

During our discussions with the staff on the ASME proposal for risk-informed in-service inspection,

it became evident that the model uncertainties associated with some aging mechanisms of SSCs

could be very large. Quantifying these uncertainties is important to the evaluation of strategies for

managing the risks from aging degradation of SSCs.

There is evidence from both the operating experience and the scientific literature that the

organizational structure of a nuclear power plant and the prevailing safety culture are major

determinants of human performance. The ACRS believes that RES should undertake an effort to

identify the contribution of these issues to human performance and to the estimation of CDF and

LERF. The ACRS realizes that there are sensitivities regarding the NRC's investigating

"management" issues, yet, it believes that there is a need to understand the ways in which these

elements may affect the CDF and LERF estimates before decisions are taken regarding regulatory

action, if any.

The ACRS notes that the current effort to make the reactor oversight process risk informed lists

"problem identification and corrective action processes" (clearly an organizational element) as an

issue of concern to the agency. Furthermore, it is decided that "no separate and distinct assessment

of licensee safety culture is needed because it is subsumed by either the Performance Indicators or

baseline inspection activities" (SECY-99-007, p. 15). The technical basis for these decisions and

conclusions is not evident to the ACRS. While they may be valid, the ACRS believes that a

systematic investigation of these issues would provide a more convincing argument as to what a riskinformed oversight process should include.

The risk-informed regulatory guides specify that the CDF and LERF values reflect all modes of

reactor operation. Plant configurations vary during low-power and shutdown operations, yet, only

a few have been investigated rigorously, e.g., mid-loop operations for PWRs and Plant Operational

State 5 during refueling for BWRs. Operating experience indicates that a large number of human

errors occurs during low-power and shutdown operations. In a risk-informed regulatory system, it

is important to have credible estimates of CDF and LERF for all plant configurations and modes of

operation. This should be achieved expeditiously.

NUREG-16354 4
H. Human Factors and Human-Machine Interface

Recommendation

The ACRS supports the RES strategy to develop a technically justified Human Performance Plan

(HPP) and urges RES to proceed aggressively in the development and implementation of the

proposed approach.

Discussion

In NUREG-1635, Vol. 1, the ACRS noted a lack of a systematic approach for achieving the three

goals of the mission statement of the human performance program; namely, (1) identifying human

performance issues important to public health and safety; (2) increasing understanding of the causes

and consequences of degraded human performance in such settings; and (3) implementing the

appropriate regulatory response to such issues. The ACRS is pleased to report that the strategy

presented to it on February 4, 1999 for developing a more technically defensible HPP is promising.

This strategy will begin with the identification of agency needs in the field of human performance

by reviewing the operating experience, contacting other organizations such as the Institute of Nuclear

Power Operations, performing sensitivity studies, and comparing the preliminary findings to error

classifications available in the literature. The list of human performance needs for NRC will then

be prioritized by a process now being developed within RES. Requirements and a closure strategy

for the priority activities will be defined, quantitatively where possible, using regulatory analysis

guidelines and risk criteria described in Regulatory Guide 1.174.

The ACRS recognizes that much work needs to be done to implement this strategy. The ACRS

continues to believe that a technically defensible HPP is essential to guide the agency's future

activities in this very important area. The ACRS, therefore, encourages the staff to continue its

efforts to develop such a HPP by following the proposed strategy. A technically defensible HPP will

establish the basis for realistically quantifying the risk associated with human performance.

5 5 NUREG-1635
HI. High-Burnup Fuel Performance and Mixed Oxide Fuel

Recommendations

The ACRS finds the confirmatory research program on high burnup fuel performance to be well

designed. The program can be improved by including considerations of accident source term

and information needs for fuel burnups in excess of current limits.

RES should conduct a phenomena identification and ranking process for mixed oxide fuel

(MOX) to ascertain potential research and regulatory needs.

Discussion

The staff has decided that the licensees will be responsible for providing all of the data and analyses

needed to approve requested extensions of fuel burnup beyond the current limit of 62 GWd/t. It is

very likely that the nuclear industry will make requests for burnups beyond this limit. Indeed, the

industry now has under way an extensive research program to support such requests. The ACRS

has agreed with this regulatory approach and suggested that the staff define the kinds of data and

analyses that will be needed to evaluate requests for extended fuel burnups.

As a step in the definition of the information needed, RES plans to use expert opinion to develop a

phenomena identification and ranking of issues that are likely to arise in the evaluation of an

application for extended fuel burnup. This expert judgment effort will attempt to define the

important physical and chemical phenomena, identify what is known about these phenomena for

high burnup fuel, and ascertain the nature of data and analyses concerning the phenomena that will

be needed to approve applications for extended fuel burnup. The ACRS believes that this is an

excellent way to approach the problem.

The current NRC research on high burnup fuel does not address some key technical issues that arise

concerning the effects of bumup. One such issue is the effect of high burnup on the releases of

fission products under accident conditions. The NRC has issued a revised source term for the

estimation of fission product releases during design basis accidents. This revised source term was

developed based on tests and analyses of lower burnup fuel. Correlations of the fission product

release when fuel cladding is breached, the so-called gap release, based on this data may seriously

underestimate the gap release from high burnup fuel. The ACRS has not heard of plans by the NRC

staff to conduct research to make the revised accident source term applicable to high burnup fuel.

Another issue that will eventually come to the NRC is the use of MOX in commercial nuclear power

plants. The U.S. Department of Energy (DOE) plans to utilize MOX in commercial reactors as a

way to dispose of excess weapons grade plutonium. Existing regulations for commercial nuclear

power plants have not anticipated the use of such fuel. There has been some experience worldwide

with the use of MOX, but little of this experience addresses conditions outside the range of normal

operations. Contentions are being made in some quarters that such use of MOX will pose

NUREG-1635 6
qualitatively different threats in the event of a reactor accident. The technical data to confirm or

refute these contentions are not now available. While not a priority issue now, such concerns may

well arise as the DOE efforts mature. The ACRS anticipates that experimental efforts of a fairly

long duration will be needed to address issues concerning the behavior of MOX under accident

conditions. The ACRS believes that RES should conduct a phenomena identification and ranking

process for MOX to ascertain potential research and regulatory needs for MOX. The phenomena

identification and ranking process will position the agency to respond promptly to an application by

a licensee to use MOX.

7 NUREG-1635
IV. Instrumentation and Control Systems

Recommendations

* The staff should complete research on the effects of lightning on instrumentation and control

(I&C) systems, specifically digital I&C systems, and continue research on the review of

software-based digital system specifications and requirements.

* The staff should assess whether "hot shorts" arising during a fire produce significant

electromagnetic interference (EMI) and, if so, whether current guidance on electromagnetic

interference/radio frequency interference (EMI/RFI) is adequate.

" The staff should continue its current research on methods for estimating failure probabilities of

systems that include embedded software digital components for use in PRAs.

" The staff should complete a review of advanced measurement methods and their potential to

improve measurement of safety-related variables in nuclear power plants.

Discussion

Digital I&C Systems:

The introduction of digital I&C systems in nuclear power plants offers many advantages; it also

presents new technical issues. These issues can be categorized as follows: (1) internal and external

stressors that may cause changes or failures and (2) quality and reliability of the software inherent

in most digital systems.

NUREG/CR-6579 provides a risk screening of external stressors on digital systems and indicates that

humidity, EMI from lightning, and smoke can be potentially risk significant. NUREG/CR-6406

reports that a combination of high temperature and high relative humidity is a condition that could

affect the performance of digital systems. These reports, as well as NUREG/CR-6543, conclude that,

by using accepted design methods, quality assurance, qualification, and/or commercial dedication,

along with surveillance and testing, the risk from these stressors can be reduced to levels acceptable

for safety-related systems. Completion of regulatory guidance for these external stressors along with

the recently issued regulatory guide on EMIIRFI, would provide guidance on all important external

stressors on digital systems except for the effects of lightning.

A risk-significant issue related to fire protection in nuclear plants is the potential impact of a fire on

I&C circuits, specifically "hot shorts" (shorts between two or more signals or control wires

connected to different circuits). If the hot short includes a high-voltage circuit, it could be a source

of EMI. The ACRS believes that the staff should assess whether there is EMI caused by "hot shorts"

and if there is, whether current guidance on EMI/RFI is adequate.

.NURIEG-1635 .8
Software, which is inherent in most microprocessor-based digital systems, is the basis for many of

the advantages of digital systems. Software-based systems, however, have attributes that require

them to be developed with a different and more thorough attention to detail and quality assurance

than is required to develop hardware.

Software is a set of instructions for the operation of hardware. If those instructions are written with

no errors and if they are designed to operate the hardware correctly for all anticipated conditions,

which include all foreseen and unforeseen external conditions and all circumstances inherent to the

hardware which the software is designed to operate, there will never be a failure. Software, however,

does fail, and failures are most often attributed to errors in system and/or software specifications and

requirements, and to a lesser extent to errors in writing software.

The accepted software engineering method for minimizing errors related to requirements and

specifications and to programming errors is to develop software in a highly structured and

disciplined process, which calls for internal testing to ensure the product of each stage in the

development process meets the requirements specified for that stage. The final test is an assessment

of the final product to see if it meets system design requirements. Unfortunately, this approach is

not able to ensure that errors in developing the requirements for complex systems are always reduced

to a level acceptable for some safety critical applications, such as safety-related applications in

nuclear power plants. Current regulatory guidance for digital system design provided in Chapter 7

of the Standard Review Plan is based on these software engineering methods.

In its June 1997 letter, the ACRS recommended that the Commission issue the update of Chapter

7 of the Standard Review Plan also recommended that "for systems that contain digital components,

the effect of software failure should be included in the probabilistic risk assessment (PRA)." This

recommendation was consistent with one of the National Research Council Study recommendations

that "The USNRC should develop methods for estimating failure probabilities of digital systems,

including commercial off-the-shelf (COTS) equipment for use in PRAs."

The staff currently has research in progress to improve the technical basis for review of softwarebased digital systems with emphasis on assessment of the adequacy of system and software

specifications and requirements and research to develop the technical basis for estimating failure

probabilities of digital systems for use in PRAs. The ACRS believes that this effort must be

continued and, if necessary for timely completion, be increased.

Sensors that measure the physical variables that are used by safety and control systems have not

changed significantly since the design of the first nuclear plants more than 40 years ago. New

measurement methods that integrate new sensor technology, such as optical-based sensors and

embedded microelectronics that perform on-line calibration, diagnostics, testing and compensation

for external effects, provide opportunities for increased measurement accuracy with lower

uncertainty. These new measurement methods represent the next I&C area where new technology

will have a significant impact on nuclear plant reliability, safety, and efficiency.

9 NUREG-1635
The NRC staff should have the information necessary to make prompt regulatory judgments

concerning licensee applications to use advanced sensors in nuclear power plants. The ACRS

believes that the staff should complete a review of advanced measurement methods and their

potential to improve measurement of key safety-related variables in nuclear power plants. This

review should include an assessment of NUREG/CR5501, which addresses a variety of advanced

sensors for making better measurements in nuclear power plants.

NUREG-1635 10
V. Severe Accidents

Recommendation

To meet its needs in the severe accident area, the agency should strengthen its participation in

international research activities that are already in existence (e.g., high bum-up fuel behavior,

aerosol behavior, fuel-coolant interactions, and thermal hydraulics). Such participation in

international research is necessary because of the very high cost of performing research in this area.

Discussion

The current NRC plans appear to be to essentially closeout research in the severe accident area while

providing only a minimum level of effort for maintaining the suite of NRC severe accident codes

and continuing with minimum (already committed) participation in international cooperative

programs. With this "minimum" strategy, the ACRS is concerned that information relating to severe

accidents needed to support the transition to a risk-informed regulatory system will not be available.

As the agency moves toward a risk-informed regulatory system, there will be increased tension

between the level of uncertainty in the risk determinations and the need for appropriate defense-indepth to ensure that there is sufficient margin with respect to undue risk. The ACRS believes that

the current level of uncertainty in the risk assessments caused by severe accident issues is so large

that the risk-informed regulations inevitably will have to remain highly conservative and include

substantial (and somewhat arbitrary) defense-in-depth provisions.

The determination of these uncertainties can be sharpened and their magnitudes substantially reduced

with additional severe accident research in the following selected areas:

- Fission product release from high bumup fuel

- Effects of mixed oxide fuel on core-melt behavior

- Release of fission products from molten fuel pools in the absence of sparging

- Hydrogen mixing and distribution in containment

11 NUREG-1635
VI. Fire Protection Research

Recommendation

RES should develop a definitive plan to enhance the agency's capabilities regarding fire risk

assessment methods.

Discussion

Fire protection requirements for nuclear power plants have been developed on a deterministic basis

in the past. Indeed, licensees have indicated a relative comfort with the deterministic fire protection

requirements, even though complying with these requirements has imposed significant costs on the

nuclear industry. Certainly, this comfort on the part of licensees has slowed the use of quantitative

methods to estimate risk to the public from fire-initiated accidents at nuclear power plants. Where

quantitative methods have been used in the individual plant examination of external events (IPEEE)

and in a few external events PRA, findings indicate that fire initiated accidents can be significant

contributors to the risk posed by a nuclear power plant. These findings suggest strongly that it will

be necessary to routinely estimate risks from fire in a risk-informed regulatory framework.

The fire risk assessment methods available to the NRC are not developed to a level comparable to

the methods for assessing risk during normal operations. The fire risk assessment methods and

insights from analyses using these methods have not been disseminated throughout the agency in a

way comparable to what has been done for risks from internal events. There is not within the NRC

an activity for assessing fire events comparable to the Accident Sequence Precursor program that is

used for evaluating other types of nuclear power plant events. Senior Reactor Analysts in the regions

do not have well-developed methods for assessing fire risks.

There is now a research activity in NRC to improve the fire risk assessment methods. This research

has begun to address a selected few of the many deficiencies previously identified in the fire risk

assessment technologies. There is not, however, a well-developed plan to show that these research

activities will, in fact, yield the kinds of tools that the agency will need in its move toward riskinformed regulation. There is not a demonstration that research to correct deficiencies not now being

studied can, indeed, be deferred to later years and still meet the aspirations the Commission has for

refining the regulatory process. Absence of a definitive plan for the development, application, and

dissemination of fire risk assessment methods within the agency can be attributed in no small part

to the slow process by which insights from the IPEEE submittals are being developed. The needs for

adequate capabilities to assess fire risks are becoming more painful to the agency as it proceeds

towards making NRC programs for enforcement, inspection, and plant assessment risk informed.

Difficulties have been encountered in the risk ranking of violation of fire protection requirements.

Decisions on changes to the fire protection core inspection as a result of the Fire Protection

Functional Inspection Pilot Studies will be greatly facilitated by good technologies for fire risk

assessment. One can easily imagine that decisions on applications for inservice testing and online

maintenance could be facilitated by reliable fire risk assessment methods.

NUREG-1635 12
VII. Thermal Hydraulics

Recommendations

The NRC needs to maintain a capability to run thermal-hydraulic codes and to evaluate code

predictions to support future regulatory decisions.

Code consolidation now under way should lead to efficient focus on one set of technical

enhancements and considerable savings in maintaining one code instead of three or four

different ones. Emphasis should be on improvement, rather than compromise that preserves

several poor features from each previous code. Figures of merit, such as gains in run time,

robustness, accuracy, and fidelity, need to be demonstrated by clear "before" and "after"

comparisons.

The NRC should identify and address those features of codes that need modification in order

to supply realistic, rather than conservative, evaluations of accident scenarios. Ways need

to be developed to quantify uncertainties in code predictions.

More specific justification of new experiments and the expected payoff from their results is

needed. New experiment efforts should be judged by how they complement past results and

how they improve the quality of specific capabilities in the thermal-hydraulic codes.

The NRC should continue to encourage new ideas that offer substantial benefits for

improving accuracy, coherence, and robustness of its thermal-hydraulic codes. Anticipatory

research that is planned is very modest for an area of such fundamental importance to the

realistic prediction of accident scenarios in which present methods of prediction have

significant limitations.

Discussion

As the nuclear industry is challenged to perform more competitively in the future, in response to

deregulation and low costs of alternative electrical generation, the NRC will be asked to approve

power uprates and relaxation of other technical constraints that have the potential to reduce real

margins of safety. The Commission will need assurance that these decisions can be made with

confidence. Since the consequences of accidents depend on thermal-hydraulic phenomena and there

is a sparse data base of experience with accidents, except for the Three Mile Island event and a

handful of lesser transients with no major consequences, reliance is placed on the predictions of

"best-estimate" thermal-hydraulic computer codes.

Nuclear accidents involve multiphase flows of steam, water, and noncondensable gases. In extreme

cases, other materials may melt, vaporize, or be entrained, contributing to the complex mixture to

be analyzed.

13 NUREG-1635
The thermal hydraulics of multiphase flow is far from being a precise science. It is not based on

exact equations. Rather it uses "engineering" methods that evolve as more complete formulations,

empirical results, and theoretical insights are achieved. Present codes contain many assumptions and

coefficients that do not have an extensive validation for reactor conditions and may even be "wrong"

in an academic sense if they yield an incorrect answer to a simple, well-defined problem. The codes

must be carefully compared with representative data because they may not provide accurate

predictions for situations not considered when the codes were developed.

The original confirmation of the adequacy of codes for Appendix K calculations was a long process,

relying on extensive experiments (e.g., Loss of Fluid Test) which NRC is unlikely to repeat. When

doubt existed, conservative assumptions were adopted in the codes. To assess the risk associated

with the changes in plant configuration or operating procedures that burden-reduction measures

require, code predictions.that are closer to reality are needed. In addition to the "best estimate" of

what is likely to happen, the uncertainty in these code predictions emanating from model

assumptions and parameter values must be quantified.

The validation of present codes for best-estimate use is problematical because their many

assumptions are not "realistic." The assumptions were made as expedients when codes were

developed decades ago. A mature level of technical knowledge, sophistication, and tenacity is

required in order to discern how far the codes can be trusted, when they need improvement, and

when recourse must be made to experiments. Code review and validation is a major task that must

be done thoroughly, otherwise faults may remain hidden until misleading predictions are revealed

in a regulatory context.

The ACRS perceives the need for "best estimate" code review and validation as being insufficiently

appreciated, and even discouraged, by the NRC management. The eventual check is a thorough

time-consuming investigation by the ACRS -- far too late in the process to ensure efficient use of

agency resources. Our experience with reviewing codes for AP600 and preliminary experience with

codes presently under review suggests that the ACRS and its consultants will find many deficiencies

related to physical modeling and code documentation. These deficiencies should be discovered by

the NRC staff reviewers who are in a much better position to delve into the coding details.

Key technical issues may require resolution by RES's thermal-hydraulic research group which has

lost several FTEs during the past year.

Recently, the NRC has taken steps to improve the effectiveness and logic of its regulatory processes.

Real improvements in public safety may be elusive if the technical bases for modeling accident

phenomena are neglected. Vendors may be expected to restrain the enthusiasm with which they

probe the limitations in their codes unless the NRC leads the way. The NRC has an obligation to

public safety to develop and exercise its own capability to check prediction methods and results.

Consultants and contractors may help with specific expertise, but the staff must be knowledgeable

and insistent enough to ensure critical evaluation of sufficient quality. To do this, the NRC staff

needs experience with running its own "best-estimate" thermal-hydraulic code.

NUREG-1635 14
Formerly, the NRC supported the development of a number of light water reactor thermal-hydraulics

codes (e.g., TRAC-PWR, TRAC-BWR, RELAP), with most of the development and applications

work being done at the national laboratories. The thermal-hydraulics program is now focused on the

formulation of a single thermal-hydraulics and neutron transport code. Although some outside

contractor development help will be needed, the goal which the ACRS supports is to achieve the

in-house code capability needed for the evaluation of future license applications.

This research program is more than a code consolidation effort. It will also correct some of the

identified weaknesses in the previous thermal-hydraulics codes. For example, in the course of the

AP600 design certification review the NRC staff and the ACRS uncovered many deficiencies in the

existing suite of thermal-hydraulics codes and databases for the application of these codes to new

designs.

While the ACRS endorses this short-term strategy of code consolidation and improvement, it is

concerned that the NRC may be too easily satisfied with the latest evolution of what it has called

"archaic codes" that are "becoming obsolete and are difficult to use and maintain." Only recently

has RES begun to thoroughly investigate the details of the TRAC codes, discovering unrealistic

physical models, unexplained coefficients and undocumented changes that have been patched

together over many years. These codes may require a quantum leap in quality improvement if they

are to form a reliable basis for "realistic" or "best estimate" calculations that should provide

opportunity for relaxing earlier conservative assumptions (they may also reveal if the assumptions

were really conservative when many phenomena interact during a long scenario). Moreover, future

codes must provide a quantitative measure of uncertainty, which is not now possible. Additionally,

risk assessment will require that sensitivity studies be made of the effects of assumptions or

empirical coefficients. This will require codes that run much faster than they do currently.

When existing codes have been successfully consolidated, they will still be based on decades-old

methods while the state of the art has not remained static. NRC should plan a strategy for identifying

computational needs of the future and incorporating improved numerical methods and physical

modeling into its expertise in order to achieve more confidence and precision in predicting transients

and accidents. In the commercial arena, computational fluid dynamics (CFD), including multiphase

flow, is a very active and competitive area of development with new features continually emerging.

While the small NRC-sponsored program at Rensselaer Polytechnic Institute provides a window on

a particular academic development effort in multi phase CFD, it should be supplemented by a

broader assessment of possible benefits from adopting capabilities from other sources.

The work on interfacial area modeling, centered on Purdue University, is a rare example of

long-range research. The objective is to replace the present computationally awkward set of flow

regimes and correlations with a unified predictive approach. While this is an ambitious and

unproven endeavor, there are large potential benefits in terms of code simplicity, stability, speed, and

structural cohesion. The ACRS regrets that this is the only promising new idea currently being

pursued in response to the need, already mentioned, for considerable improvements in code quality,

which should still be possible in this relatively immature branch of science.

15 NUREG-1635
Since practical two-phase thermal-hydraulic phenomena cannot now, and probably never will, be

predicted from first principles, there is an ongoing need for an experimental reality check on all

analyses and assumptions used for predicting the course of nuclear accidents. Separate-effects tests

are needed in order to improve modules in the codes and, eventually, to reach a stage where

complete systems perhaps need not be tested every time a design is changed, as is now necessary

because of uncertainty in combined uncertainties. NRC should better articulate the specific needs

for new tests and the payoff in terms of improved predictions and regulatory consequences. The

primary goal should be to support NRC's own code development, with the expectation that industry

will bear the burden of providing new data for evaluating its codes.

NRC has maintained medium-scale test facilities at Oregon State University (APEX) and Purdue

University (PUMA), primarily to support evaluation of new reactor concepts. They are being

adapted to answer a few unresolved technical questions, such as phase separation at tees. A

thorough and comprehensive evaluation needs to be made about how many such questions remain

and a decision made whether these facilities are the best place to answer them. However, we have

qualms about the NRC giving up too many experimental capabilities out of short-term

considerations, since We anticipate that future issues will emerge that require experimental

resolution. Rebuilding facilities from scratch will involve delays and costs.

On a related issue, the ACRS is concerned lest the wealth of data generated in the heyday of

thermal-hydraulic research decades ago become lost, inaccessible or unappreciated, leading to

unnecessary duplication of effort. For example, a major program has recently been initiated at

Pennsylvania State University to reinvestigate PWR core reflood, as a result of deficiencies found

in existing codes. It appears to repeat the objectives of older FLECHT tests. The ACRS would like

to see what new information is being generated by this new program and how the anticipated results

will quantitatively improve predictions relevant to safety decisions.

NUREG-1635 16
VIII. Integrity of Steam Generators

Recommendation

RES should provide additional data, methods, and models to evaluate the technical credibility of

licensee programs on steam generator integrity.

Discussion

Steam generators have been historically the most troublesome components in pressurized water

reactors. Industry efforts have been largely successful in managing degradation due to wastage,

pitting, and denting; however, fretting, stress corrosion cracking (SCC), and intergranular attack have

proved more difficult to manage. The replacement cost for steam generators is on the order of

$100M plus the cost of replacement power. Even for those plants which have not yet had to replace

their steam generators; inspection, monitoring, and repair are expensive. Thus, there is substantial

industry interest in operating with degraded steam generators, and a significant fraction of the plants

will probably not replace their steam generators during their operational life.

Highly prescriptive requirements have been used by NRC to regulate the integrity of steam

generators. This regulatory process ensures safety because a "conservative" approach is taken when

sufficient information is not available. A performance-based approach, better focused on the types

of the degradation actually occurring in steam generators, has the potential for reduced burden in

steam generator tube plugging requirements with minimal changes in plant risk. The staff and the

industry are developing performance-based regulatory guidance for the assessment of the integrity

of steam generator tubes. Steam generator tubing is unique in that it has an important containment

function as well as being by far the largest component of the reactor pressure boundary. The NRC

must have the capability to independently evaluate the technical credibility of licensee programs to

meet the performance criteria being developed.

The objectives of the NRC research are to provide the data, correlations, and models needed to

permit the NRC to assess the validity of the programs proposed by industry to manage the

degradation of steam generators. The focus of the program is on: (1) assessment of personnel,

procedures, and equipment used for inservice inspection (ISI); (2) validation and assessment of

correlations and models for evaluating the integrity of and leakage from degraded steam generator

tubes under design basis and severe accident conditions; and (3) validation and assessment of

correlations and models for predicting degradation generation and progression in steam generator

tubes.

17 17 NUREG-1635
IX. Integrity of Pressure Vessels

Recommendation

RES should continue its efforts at verification of current regulatory criteria and maintain technical

expertise in this area to respond to anticipated industry initiatives.

Discussion

The Reactor Vessel Integrity Program has already provided great insight into the regulatory questions

arising from vessel embrittlement. The generic analyses in NUREG/CR-6023 and the submittals

from the owners groups demonstrate that all PWRs and BWRs in the U.S. should have adequate

upper shelf toughness at least through the end of their current licensing period. The recent change

from "crack arrest" toughness to "crack initiation" toughness for the calculation of pressure

temperature limits for reactor vessels will significantly decrease heatup constraints for operating

plants. A screening limit for pressurized thermal shock (PTS) of PWR vessels has been developed.

Almost all U.S. plants will be well below this screening limit through their current licensing period,

although PTS concerns could pose difficulties for a significant number of plants considering license

extension.

Issues related to the embrittlement of reactor pressure vessels (RPVs) will continue to arise both for

operating reactors during their current licensing period and in the context of license extension.

Current regulatory requirements for pressure vessels are believed to ensure vessel integrity. The

conservatisms in these requirements, however, impose significant regulatory burdens, operational

constraints, and large economic costs on the industry. These conservatisms could be detrimental

to safety because they may distort the allocation of resources. They may also increase the exposure

of personnel performing unnecessary inspections and may unnecessarily restrict operational

flexibility during plant heatup and cooldown operations.

Although reductions in conservatism in the assessment of vessel integrity can and should be made,

the NRC must be able to independently verify that any changes to regulatory criteria will not increase

the probability of vessel failure to unacceptable levels. The ACRS feels that this need for

independent verification of regulatory criteria and the technical complexity of the issues require a

level of technical expertise that can be maintained by the agency through the continued support of

a significant research program in this area. The budget for the program has been significantly

reduced from FY 98 to FY 99. Further reductions may be forced by competition for resources by

other high priority research needs, but a core program will need to be maintained. Work should

continue on the "regulatory applications" of this research in the development of a revised PTS rule

and the updating of Regulatory Guide 1.99, Revision 2. Work should continue on the basic

processes of embrittlement and fracture to assess potential nonconservatism in the present

understanding.

NUREG-1635 118
X. Environmentally Assisted Stress Corrosion Cracking

Recommendation

A cooperative research program by industry and NRC on environmentally assisted cracking (EAC)

is needed to ensure the integrity of the reactor pressure boundary and reactor internals.

Discussion

Environmentally assisted cracking of reactor structural materials has been a recurrent problem in

operating reactors since the mid-1970's. The initial focus was on Boiling Water Reactor (BWR)

pipe cracking problems. Since that time, the emphasis has shifted to other problems in EAC such

as Irradiation Assisted Stress Corrosion Cracking (IASCC). The industry, largely through the BWR

Vessel and Internal Project Program, is developing inspection, mitigation, and repair/replacement

programs to address these problems. Because cracking depends on a complex interaction between

materials, fabrication processes, irradiation conditions, and water chemistry, there are still substantial

uncertainties in crack growth rates which must be reflected in inspection requirements and in

integrity assessments. NRC and industry research in this area have recently provided the basis for

the staff to accept a reduction of a factor of two in the crack growth rates used for integrity

assessments of defected stainless steel components, which reduces the need for mid-cycle

inspections in many cases.

The ACRS reviewed the current research on the EAC which deals with three issues of EAC: fatigue

of reactor structural materials, IASCC, and EAC of nickel based alloys and weld metals. This

includes work done by NRC contractors and participation in the Cooperative IASCC Research (CIR)

Program, which is managed by the Electric Power Research Institute (EPRI) and which includes

contributions from vendors, utilities, the NRC, and regulatory authorities in Europe and Japan. The

overall objective of the program is to provide data and physical models to assess environmentally

assisted degradation of primary pressure boundary components in light water reactors. In addition,

crack growth models developed by industry and submitted for evaluation and approval by the NRC

are being studied. The findings from this research program will enable the staff to make

independent evaluations of industry assessments of the operability of degraded components. The

ongoing work may lead to additional inspection relief. The work is also strongly leveraged through

the access it provides to the information being developed in programs in Europe and Japan.

19 NUREG-1635
XI. Nuclear Waste-Related Research

1999 UPDATE ON THE REVIEW OF THE NRC SAFETY RESEARCH PROGRAM

CONCERNING WASTE-RELATED WORK

As reported last year (NUREG-1635, Vol. 1), the Advisory Committee on Nuclear Waste (ACNW)

is charged with review of safety research in waste management. There is a relatively small ($2.3

million a year in Radionuclide Transport and $1.3 million a year in Radiation Protection and Health

Effects) research program in the Office of Nuclear Regulatory Research (RES) in this area. The

Office of Nuclear Material Safety and Safeguards (NMSS) contracts with the Center for Nuclear

Waste Regulatory Analyses (CNWRA) for technical assistance ($13.1 million a year), much of

which is, in essence, research related to the Yucca Mountain program. Thus, the ACNW also

considers work at the CNWRA to be included in its review.

The RES staff made a presentation to the ACNW at its 106th meeting in February 1999. Although

we submitted a series of questions to NMSS, we received no formal answers to them, nor were any

members of the staff available to make presentations at our meetings primarily because the viability

assessment review consumed most of the staff's time during this period. Our comments on the

program at the CNWRA are based on reading the Technical Program Description for the CNWRA.

The ACNW staff did meet with several scientists from NMSS and the CNWRA to discuss some of

the questions that we posed. We are not currently in a position to comment on the adequacy of the

CNWRA to provide all of the technical support necessary for review of a license application from

the Department of Energy (DOE). We plan to visit the CNWRA in the near future.

Office of Nuclear Regulatory Research

Our main recommendation last year was that RES adopt a way to prioritize the research that is

sponsored. With only modest funds available for waste-related research, the only way to have an

effective program is to focus on critical issues. We understand that RES is considering the use of

an Analytical Hierarchy Process to aid it in establishing research priorities. We also understand that

RES is engaged in a self-assessment that may affect future directions with respect to prioritization

of research programs. We look forward to hearing about the revisions to the current research

program that result from these efforts.

We have become concerned that thewaste-related research program within RES may be too small

to accomplish what the NRC mayneed. This is not to say that specific projects being supported are

not valuable; indeed, the program is supporting some excellent individuals and important work. We

do recommend, however, that definitive goals for the program be set and that funding commensurate

with the goals be provided. Perhaps this will be a product of the self-assessment process. A

program that is too small to be effective can be continued for the short term, but it should not be

allowed to consume resources with little likelihood of producing important results for the regulatory

program in the long term.

NUREG-1635 .20
Office of Nuclear Material Safety and Safeguards - CNWRA

In the continuing interactions that the ACNW has with the staff on the Yucca Mountain project, we

have gained considerable insight on the nature and quality of the work being performed in high-level

waste. Our conclusion is that much important work is being accomplished by the staff on

performance assessment, corrosion, geochemistry, and so forth. We are generally favorably

impressed with specific projects of the ongoing work. Nevertheless, we have concerns about the

process for prioritizing work.

The ACNW was told last year that results from the performance assessments of Yucca Mountain

were the major drivers for prioritizing work in the CNWRA program. On the basis of this

information, the ACNW recommended last year that NMSS continue to use Total Performance

Assessment (TPA) code results to guide its work. Our review of the program at the CNWRA for

this year leads us to question whether NMSS truly is doing what it says. We list two pieces of

evidence below that lead to our concern that the overall program is not being driven by high-level

strategic concerns.

1. The CNWRA work for the coming year includes substantial effort related to Igneous Activity.

We have seen no convincing argument that such work should receive significant support. In

fact, the TPA results that we have seen led us to suggest more than a year ago that the program

in Igneous Activity be phased out.

2. The ACNW wrote a letter on the Engineered Barrier System (EBS) on September 9, 1998. In

it we recommended that consideration be given to enhanced research (i.e., technical assistance)

in a few areas related to near-field chemistry and corrosion. (Increasing emphasis by DOE on

the EBS, along with TPA results, suggests that these areas should have high priority.) The

EDO response said that there was limited flexibility in the program and that the ACNW should

indicate what items might be cut to make room for new, high-priority research. Again, this

response suggests that the status quo is the baseline for the program.

We recommend that NMSS adopt a procedure for prioritizing work at the CNWRA using results

from the TPA code as a primary driver and that decisions be based on the prioritization. In

particular, we reiterate our recommendation that work on Igneous Activity be ended unless a credible

case can be made for its continuation.

We also reiterate our recommendation of last year that the CNWRA make use of external experts

as it takes on work. This advice may be especially important in the area of engineering, where we

have not seen clear indication that either NMSS or the CNWRA has sufficient depth in design

engineering to allow full evaluation of DOE's repository system for Yucca Mountain.

21 NUREG-1635
XII. References

1. U. S. Nuclear Regulatory Commission, Advisory Committee on Reactor Safeguards,

NUREG- 1635, Vol. 1, "Review and Evaluation of the Nuclear Regulatory Commission

Safety Research Program, A Report to the U. S. Nuclear Regulatory Commission," June

1998.

2. U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research,

Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in RiskInformed Decisions on Plant-Specific Changes to the Licensing Basis," July 1998.

3. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.175, "An Approach for

Plant-Specific, Risk-Informed Decisionmaking: Inservice Testing," August 1998.

4. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.176, "An Approach for

Plant-Specific, Risk-Informed Decisionmaking: Graded Quality Assurance," August

1998.

5. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.177, "An Approach for

Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," August 1998.

6. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.178, "An Approach for

Plant-Specific, Risk-Informed Decisionmaking: Inservice Inspection of Piping,"

September 1998.

7. U. S. Nuclear Regulatory Commission, NUREG-1 150, "Severe Accident Risks: An

Assessment for Five U.S. Nuclear Power Plants," December 1990.

8. Memorandum dated January 8, 1999, SECY-99-007, from William D. Travers,

Executive Director for Operations, NRC, . for the Commissioners, Subject:

Recommendations for Reactor Oversight Process Improvements.

9. U. S. Nuclear Regulatory Commission, NUREG/CR-6579, Brookhaven National

Laboratory, BNL-NUREG-52536, "Digital I&C Systems in Nuclear Power Plants -RiskScreening of Environmental Stressors and a Comparison of Hardware Unavailability

With an Existing Analog System," January 1998.

10. U. S. Nuclear Regulatory Commission, NUREG/CR-6406, Oak Ridge National

Laboratory/Sandia National Laboratories, ORNIJrM-13122, "Environmental Testing

of an Experimental Digital Safety Channel," September 1996.

NUREG-1635 22
11. Letter dated June 23, 1997, from Robert L. Seale, Chairman, ACRS, to L. Joseph Callan,

Executive Director for Operations, NRC, Subject: Regulatory Guidance for

Implementation of Digital Instrumentation and Control Systems.

12. U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research,

NUREG/CR-6543, Sandia National Laboratories, SAND97-2544, "Effects of Smoke on

Functional Circuits," October 1997.

13. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research,

NUREG/CR-5501, "Advanced Instrumentation and Maintenance Technologies for

Nuclear Power Plants," August 1998.

14. U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research,

NUREG/CR-6023, Oak Ridge National Laboratory, ORNILTM-12340, "Generic

Analyses for Evaluation of Low Charpy Upper-Shelf Energy Effects on Safety Margins

Against Fracture of Reactor Pressure Vessel Materials," July 1993.

15. U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research,

Regulatory Guide 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Materials,"

May 1998.

16. U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research,

NUREG/CR-6327, Modeling and Computing Services, MCS 950302, "Models for

Embrittlement Recovery Due to Annealing of Reactor Pressure Vessel Steels," May

1995.

17. Report dated September 9, 1998, from B. John Garrick, Chairman, ACNW, to Shirley

Ann Jackson, Chairman, NRC, Subject: Issues and Recommendations Concerning the

Near-Field Environment and the Performance of Engineered Barriers at Yucca Mountain

23 NUREG-1635

APPENDIX A: ACRONYMS

ACNW Advisory Committee on Nuclear Waste

ACRS Advisory Committee on Reactor Safeguards

APEX Advanced Plant Experiment

ASME American Society of Mechanical Engineers

BWR Boiling Water Reactor

CDF Core Damage Frequency

CFD Computational Fluid Dynamics

CIR Cooperative IASCC Research

CNWRA Center for Nuclear Waste Regulatory Analyses

COTS Commercial Off The Shelf

DOE Department of Energy

EAC Environmentally Assisted Cracking

EBS Engineered Barrier System

EDO Executive Director for Operations

EMI Electromagnetic Interference

EPRI Electric Power Research Institute

FLECHT Full-Length Emergency Cooling Heat Transfer

FTE Full-Time Equivalent

GWd/t Giga Watt Days/Ton

HPP Human Performance Plan

IASCC Irradiation Assisted Stress Corrosion Cracking

I&C Instrumentation and Control

IPEEE Individual Plant Examination of External Events

ISI Inservice Inspection

LERF Large, Early Release Frequency

LOCA Loss of Coolant Accident

LOFT Loss of Fluid Test

LP/SD Low Power/Shutdown

LWR Light Water Reactor

MOX Mixed Oxide Fuel

NMSS Office of Nuclear Material Safety and Safeguards

NRC Nuclear Regulatory Commission

PRA Probabilistic Risk Assessment

PTS Pressurized Thermal Shock

PUMA Purdue University Multidimensional Integral Test Assembly

PWR Pressurized Water Reactor

RELAP Reactor Leak and Analysis Program

RES Office of Nuclear Regulatory Research

RFI Radio Frequency Interference

RPV Reactor Pressure Vessel

25 NUREG-1635
SCC Stress Corrosion Cracking

SECY Office of the Secretary of the Commission

SSCs Systems, Structures, and Components

TPA Total Performance Assessment

TRAC Transient Reactor Code Analysis

VIP Vessel Integrity Program

NUREG-1635 26
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(2-89) (Assigned by NRC, Add VoL., Supp., Rev..

NRCM 1102, and Addendum Numbers, If any.)

3201,3202 BIBLIOGRAPHIC DATA SHEET

(See instructans on the reverse)

2. TITLE AND SUBTITLE NUREG-1635, Volume 2

Review and Evaluation of the Nuclear Regulatory Commission 3. DATE REPORT PUBLISHED

Safety Research Program MONTH YEAR

A Report to the U. S. Nuclear Regulatory Commissioh July 1999

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

Annual

7. PERIOD COVERED (Inclusive Dates)

Fiscal Year 1999

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC. provide Division, Office or Region, US. Nuclear Regulatory Commission, and mailing address; if contractor,

provide name and mailing address.)

Advisory Committee on Reactor Safeguards

U. S. Nuclear Regulatory Commission

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type 'Same as above*. if contractor, provide NRC Diwsion, Oftfce or Region, U.S. NudearRegulatory Commission,

and mailing address.)

Same as above

10. ,SUPPLMENTIARY NUOTES

11. ABSTRACT (200 words orless)

In 1998, the Advisory Committee on Reactor Safeguards (ACRS) submitted to the Nuclear Regulatory Commission (NRC) a

comprehensive report of the NRC's Safety Research Program, NUREG-1 635, Vol. 1, "Review and Evaluation of the Nuclear

Regulatory Commission Safety Research Program," which documented the ACRS conclusions and recommendations. The

ACRS continues to support the conclusions and recommendations of that report. The present report is more modest in scope

and is intended to provide additional information concerning the research needed to support Commission programs, especially

risk-informed regulation. Not all research programs are included in this report and our observations are often limited to certain

aspects of a given program.

12. KEY WORDSIDESCRIPTORS (List words or phrases that vwlt assist researchers in ocating the report) 13. AVAILABILITY STATEMENT

Unlimited

14. SECURITY CLASSIFICATION

Nuclear Reactors Safety Engineering (his Pa"e)

Nuclear Reactor Safety Safety Research Unclassified

Reactor Operations (s Report)

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2-89) This form was electronically produced by Elhe Federal Forms, Inc.


onrecycld Lpanpe

Federal Recycling Program

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