NUREG-1635, Vol. 4

Review and Evaluation of the 

Nuclear Regulatory Commission 

Safety Research Program 

A Report to the 

U. S. Nuclear Regulatory Commission 

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 

U. S. Nuclear Regulatory Commission 

Washington, DC 20555-0001
AVAILABILITY OF REFERENCE MATERIALS 

IN NRC PUBLICATIONS

NRC Reference Material 

As of November 1999, you may electronically access 

NUREG-series publications and other NRC records at 

NRC's Public Electronic Reading Room at 

www.nrc.gov/NRC/ADAMS/index.html. 

Publicly released records include, to name a few, 

NUREG-series publications; Federal Register notices; 

applicant, licensee, and vendor documents and 

correspondence; NRC correspondence and internal 

memoranda; bulletins and information notices; 

inspection and investigative reports; licensee event 

reports; and Commission papers and their 

attachments. 

NRC publications in the NUREG series, NRC 

regulations, and Title 10, Energy, in the Code of 

Federal Regulations may also be purchased from one 

of these two sources. 

1. The Superintendent of Documents 

U.S. Government Printing Office 

Mail Stop SSOP 

Washington, DC 20402-0001 

Internet: bookstore.gpo.gov 

Telephone: 202-512-1800 

Fax: 202-512-2250 

2. The National Technical Information Service 

Springfield, VA 22161-0002 

www.ntis.gov 

1-800-553-6847 or, locally, 703-605-6000 

A single copy of each NRC draft report for comment is 

available free, to the extent of supply, upon written 

request as follows: 

Address: Office of the Chief Information Officer, 

Reproduction and Distribution 

Services Section 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

E-mail: DISTRIBUTION @ nrc.gov 

Facsimile: 301-415-2289 

Some publications in the NUREG series that are 

posted at NRC's Web site address 

www.nrc.gov/N RC/N U REGS/indexnum.html 

are updated periodically and may differ from the last 

printed version. Although references to material found 

on a Web site bear the date the material was 

accessed, the material available on the date cited may 

subsequently be removed from the site.

Non-NRC Reference Material 

Documents available from public and special technical 

libraries include all open literature items, such as 

books, journal articles, and transactions, Federal 

Register notices, Federal and State legislation, and 

congressional reports. Such documents as theses, 

dissertations, foreign reports and translations, and 

non-NRC conference proceedings may be purchased 

from their sponsoring organization. 

Copies of industry codes and standards used in a 

substantive manner in the NRC regulatory process are 

maintained at

The NRC Technical Library 

Two White Flint North 

11545 Rockville Pike 

Rockville, MD 20852-2738 

These standards are available in the library for 

reference use by the public. Codes and standards are 

usually copyrighted and may be purchased from the 

originating organization or, if they are American 

National Standards, from

American National Standards Institute 

11 West 4 2 nd Street 

New York, NY 10036-8002 

www.ansi.org 

212-642-4900

Legally binding regulatory requirements are stated 

only in laws; NRC regulations; licenses, including 

technical specifications; or orders, not in 

NUREG-series publications. The views expressed 

in contractor-prepared publications in this series 

are not necessarily those of the NRC. 

The NUREG series comprises (1) technical and 

administrative reports and books prepared by the 

staff (NUREG-XXXX) or agency contractors 

(NUREG/CR-XXXX), (2) proceedings of 

conferences (NUREG/CP-XXXX), (3) reports 

resulting from international agreements 

(NUREG/IA-XXXX), (4) brochures 

(NUREG/BR-XXXX), and (5) compilations of legal 

decisions and orders of the Commission and 

Atomic and Safety Licensing Boards and of 

Directors' decisions under Section 2.206 of NRC's 

regulations (NUREG-0750).
NUREG-1635, Vol. 4

Review and Evaluation of the 

Nuclear Regulatory Commission 

Safety Research Program 

A Report to the 

U. S. Nuclear Regulatory Commission 

Manuscript Completed: April 2001 

Date Published: May 2001 

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 

U. S. Nuclear Regulatory Commission 

Washington, DC 20555-0001
CONTENTS 

ABBREVIATIONS ......................................................... vi 

I. INTRODUCTION .......................................................... 1 

1. O bjectives ........................................................ 1 

2. Approach .................................................... 1 

3. Research Needs and Focus ............................................ 2 

4. Research Initiatives for the Future ...................................... 3 

4.a. New Power Plants and a Revised Regulatory Structure ............... 4 

4.b. Risk Implications of License Renewal and Power Uprates ............. 6 

4.c. Decisionmaking M ethod ....................................... 8 

II. ONGOING RESEARCH PROGRAMS ...................................... 10 

1. Probabilistic Risk Assessment Research and Applications .................. 10 

L.a. Risk Assessment Technologies in a Risk-Informed Regulatory Systems . 10 

1.b. Ongoing and Planned PRA Research Programs .................... 11 

2. Human Factors Research ............................................ 17 

a. History of Human Factors Research at the NRC .................... 17 

b. Ongoing and Planned Research Programs ......................... 18 

3. Fire Protection Research ............................................ 24 

4. Reactor Fuel Research .......................................... 30 

5. M aterials and M etallurgy ............................................ 34 

Embrittlement Estimation Methods .................................. 36 

Fracture Toughness Estimation Methods .............................. 36 

6. Thermal-Hydraulics Research ........................................ 43 

Future Directions in Thermal Hydraulics Research ...................... 45 

7. Severe Accident Research ........................................... 48 

Future Directions in Severe Accident Research ........................ 50 

Core Com petencies .............................................. 52 

8. Civil, Structural, and Seismic Research ................................. 55 

9. Research on Digital Safety Systems ................................... 60 

10. Spent Fuel Storage and Decommissioning Research ....................... 65 

11. Research in Criticality Safety ...... ............................... 69 

12. Radiation Health Effects ............................................ 72 

13. Analysis and Evaluation of Operational Data ............................ 74 

REFERENCES ............................................................. 77

o°i NUREG-1 635
LIST OF FIGURES 

Figure 1. Trend in the Occurrence Rate of Common Cause Failure 

E vents ......................................................... 16 

Figure 2. OERAB Programs and Activities ................................... 76

NUREG-1635 iv
LIST OF TABLES

Table Title Page 

1 Ongoing and Planned Probabilistic Risk Assessment Research ............... 15 

2 Ongoing and Planned Research Activities in the Area of Human 

Perform ance ....................................................... 23 

3 Ongoing and Planned Programs in the Area of Fire 

Protection ......................................................... 29 

4 Ongoing and Planned Research on Reactor Fuel ........................... 33 

5 Ongoing and Planned Research Programs in Materials and Metallurgy ......... 42 

6 Ongoing and Planned Thermal Hydraulics Research Programs ............... 47 

7 Ongoing and Planned Severe Accident Research Programs .................. 54 

8 Ongoing and Planned Civil, Structural and Seismic Research Programs ........ 59 

9 Ongoing and Planned Research on Digital Systems ........................ 64 

10 Ongoing and Planned Research Programs in Spent 

Fuel Storage and Plant Decommissioning ................................ 68 

11 Research Programs in Criticality Safety ................................. 71 

12 Ongoing and Planned Research on Radiation Health Effects ................. 73 

13 Past, Ongoing and Planned Activities in the 

Regulatory Effectiveness and Human Factors Branch ....................... 75

V NUREG-1635
ABBREVIATIONS

ACRS 

AEOD 

ANO-2 

ANSI 

APEX 

ASCE 

ASME 

ASP 

ASTM 

ATHEANA 

ATWS 

BEIR VII 

BWR 

BWRVIP 

CFD 

CDF 

CFAST 

CFR 

COMPBRN 

COTS 

CRSS 

CSNI 

DBA 

DPO 

EMI 

FIVE 

FRAPCON 

FRAPTRAN 

GSI 

GV 

GWD/t 

HSSI 

IAEA 

ICRP 

I&C 

IGR 

IPE 

IPEEE 

INPO

Advisory Committee on Reactor Safeguards 

Analysis and Evaluation of Operational Data 

Arkansas Nuclear One, Unit 2 

American National Standards Institute 

Advanced Plant Experiment 

American Society of Civil Engineers 

American Society of Mechanical Engineers 

Accident sequence precursor 

American Society for Testing and Materials 

A Technique for Human Event Analysis 

Anticipated transient without scram 

Biological Effects of Ionizing Radiation 

Boiling Water Reactor 

Boiling Water Reactor Vessel and Internals Project 

Computational fluid dynamics 

Core damage frequency 

Consolidated fire and smoke transport 

Code of Federal Regulations 

Computer Code for Modeling Compartment Fires 

Commercial Off-the-Shelf 

Center for Risk Studies and Safety 

Committee on the Safety of Nuclear Installations 

Design basis accident 

Differing Professional Opinion 

Electromagnetic Interference 

Five Induced Vulnerability Evaluation 

Fuel Rod Analysis Package-Gap Conductivity 

Fuel Rod Analysis Package-Transient 

Generic safety issue 

Garner Valley 

Gigawatt day/ton 

Heavy-section steel irradiation 

International Atomic Energy Agency 

International Committee on Radiation Protection 

Instrumentation and control 

Inert gas receiving 

Individual plant examination 

Individual plant examination of external events 

Institute of Nuclear Power Operations

vii NUREG-1635
JCCRER 

LERF 

LET 

LOCA 

LOFT 

LNT 

LWR 

MAAP 

MACCS 

MC 

MITI 

MOX 

MOV 

MRP 

NDE 

NEA 

NIST 

NMSS 

NOAA 

NPP 

NRC 

NRR 

NUPEC 

ODSCC 

OECD 

OERAB 

OSHA 

PRA 

PTS 

PUMA 

Pu 

PWR 

RELAP5 

RES 

RFI 

RHR 

RPV 

SAPHIRE 

SPAR 

SSC 

TRAC-BWR 

UCSB

Joint Coordinating Committee on Radiation Effects Research 

Large, early release frequency 

Linear energy transfer 

Loss of coolant accident 

Loss-of-fluid test 

Linear no-threshold 

Light water reactor 

Material access authorization program 

MELCOR Accident Consequence Code System 

Master curve 

Ministy of International Trade and Industry 

Mixed--oxide fuel 

Motor operated valves 

Modifications/Rework Package 

Non-destructive examination 

Nuclear Energy Agency 

National Institute of Standards and Technology 

Office of Nuclear Material Safety and Safeguards (NRC) 

National Oceanic and Atmospheric Administration 

Nuclear power plant 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Reactor Regulation (NRR) 

Nuclear Power Engineering Corporation 

Outside-diameter stress-corrosion cracking 

Organization for Economic Cooperation and Development 

Operating Experience Risk Analysis Branch 

Occupational Safety and Health Administration 

Probabilistic risk assessment 

Pressurized thermal shock 

Purdue University Multidimensional Integral Test Assembly 

Plutonium 

Pressurized water reactor 

Reacto:r Leak and Analysis Program 

Office of Nuclear Regulatory Research (NRC) 

Radio frequency interference 

Residual heat removal 

Reactor pressure vessel 

Systems Analysis Program for Hands-On Integral 

Simplified plant analysis risk 

Structure, system and component 

Transient Reactor Analysis Code-Boiling Water Reactor 

University of California, Santa Barbara

NUREG-1635 viii
I. INTRODUCTION

1.1 . Objectives 

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

recommendations that emerged from an assessment conducted from March 2000 through March 

2001 by the NRC's Advisory Committee on Reactor Safeguards (ACRS). 

In conducting this assessment, the ACRS sought to fulfill the following objectives: 

* Assess the ongoing and planned reactor safety research being carried out by the Office of 

Nuclear Regulatory Research (RES). 

* Recommend additional research that will be needed to promote the efficient and effective 

execution of the agency's mission in the future. 

The attentions of the report focuses on the research addressing issues of power reactor safety. 

Research dealing with the permanent disposal of radioactive waste in geological repositories will 

be discussed separately in a report by the NRC's Advisory Committee on Nuclear Waste (ACNW). 

1.2 Approach 

The ACRS assessment of the research programs focused on the following questions, rather than on 

the initial need for the research results: 

Has the research sufficiently progressed to a point at which its results are adequate to support 

regulatory decisionmaking? 

Does the research need to be expanded or otherwise modified to better meet the spectrum of 

agency needs? 

Does this research need to be performed independently by the NRC, rather than depending 

on information supplied by the licensee or nuclear energy institutions? 

In addressing these questions, the ACRS recognized that the agency's needs go beyond the 

immediate, identified regulatory decisions that might motivate research. Foremost among these are 

the need to maintain the technical competencies of the NRC staff, and to leverage the agency's 

participation in collaborative international research. The recent experience with safety issues 

associated with high burnup reactor fuels has demonstrated that certain technical competencies are 

crucial to the NRC, although they may not be continually used by the agency's line organizations 

and are difficult to acquire by contract. The difficulty can arise because there are few contractors 

who possess such episodically used expertise, or because the expertise that can be obtained is not 

sufficiently independent of licensees to meet the NRC's needs. When these circumstances arise, it 

is necessary for the NRC to maintain the competency perhaps by sponsoring research in the given

1 NUREG-1635
technical area. The ACRS bellieves that there are very few areas that require this type of resource 

commitment by the NRC to maintain competencies; however, competency in reactor fuels is 

certainly such one area. It would be useful if RES explicitly identified any others. Unfortunately, 

past efforts to identify such needs have used criteria that were too broad. The selection criteria 

should be limited to crucial imrportance to the NRC mission and unavailability outside the agency. The time required for outside experts to "spin up" on the peculiarities of regulatory processes and 

issues ought not be a criterion, since the agency's current staff is quite capable of providing this 

context for any technical issue. 

Reactor safety research is becoming a collaborative enterprise, with many countries joining together to carry out high-cost research initiatives, particularly in the area of experimental research. In order 

to maintain access to ongoing or planned collaborative projects that address the agency's critical 

research needs, it may well be necessary for RES to participate in other collaborative research 

projects that do not directly apply to the agency's needs. 

1.3. Research Needs and Focus 

The value of NRC-sponsored research in the past has been discussed at some length in previous communications from the ACRS to the Commission,"2 ,

3 and need not be reiterated here. It is 

noteworthy, however, that marty of the highest-impact results from NRC research have come from 

focused projects that utilized substantial fractions of the agency's research capabilities. Such major projects included the initial development of probabilistic risk assessment (PRA) techniques for 

reactor safety (WASH- 1400)., t:he assessment of risk at representative plants (NUREG- 1150), and 

the research on the severe accident source terms that led to the revised accident source term 

(NUREG-1465). 

The great majority of research activities now being sponsored by RES have been motivated by "user 

need" requests from the NRC's line organizations. The ACRS still has concerns about the user need 

process. As now constituted, it is not evident that this process provides a thorough indication of 

research that could benefit the agency's line organizations. The ACRS remains concerned that the 

current user need process leads to a nonuniformity in the research support for improving line 

organization capabilities. RES should have an understanding of the needs of line organizations and 

the opportunities to improve regulatory processes that is as complete as possible to begin the 

research prioritization effort. As part of its ongoing activities, RES should have its own mechanisms 

for identifying needs and opportunities that would benefit from research. The identifications that 

RES makes should enter into the hierarchy of possible research activities with equivalent standing 

to the identifications made in the user need requests. 

Nevertheless, motivation of the research by a user need request provides prima facie evidence that 

the information sought in the. :research is needed for the effective and efficient execution of the 

NRC's missions by its line organizations. 

The ACRS foresees continuing needs for NRC research, and disagrees with those claiming that 

nuclear power generation and its safety regulation have been sufficiently established such that no 

new safety issues requiring research will arise. Indeed, one has only to look at the substantial

NUREG-1635 2
changes taking place in the industry, as well as the changes within the NRC to forecast that research 

capabilities will still be needed. Within the industry, the important changes include power uprates, 

extended fuel burnup, "best estimate" safety analyses, licensing renewal, and shortened outages with 

greater use of online inspection and maintenance. Within the regulatory agency, the major change 

is the move to greater use of risk information in regulatory processes. This transformation of nuclear 

power reactor regulation has yielded processes for risk-informed changes to licensing bases, risk

informed oversight and monitoring, risk-informed enforcement of regulations, and new efforts to 

make the regulations themselves risk informed. The continuation of these changes in the regulatory 

processes, as well as the need to respond effectively and efficiently to changes in the nuclear 

industry, are very likely to require research to provide information and methods that are not currently 

available. 

The ACRS does see, however, qualitative changes in the focus of NRC research. In the past, much 

of the research has been on discovering, characterizing, and redressing vulnerabilities of nuclear 

power plants that might adversely affect the health and safety of the public. Certainly, the 

development of risk assessment methods was initiated explicitly for this purpose. Much of the 

research on severe reactor accidents, containment integrity, reactor vessel integrity, fire protection, 

and accident source terms was initiated in response to suspected vulnerabilities. The success of this 

research is largely demonstrated by the fact that we now believe that it is possible to quantitatively 

evaluate the risks that are posed by nuclear power plants in many circumstances. 

Although the identification and resolution of safety vulnerabilities should remain as the highest 

priority, the focus of research should be evolving more in the direction of developing improved 

methods and databases for the efficient and effective execution of the NRC mission. Because some 

research on safety vulnerabilities is not as complete as might be desired, there is today an expected 

tension and competition for resources between research with a focus on vulnerabilities and research 

directed toward improving and modernizing the capabilities of NRC line organizations. Such 

tension and competition becomes apparent in the discussions of both the ongoing research programs 

and the definitions of future research for the NRC. 

1.4 Research Initiatives for the Future 

RES seems to have a good appreciation of the trends and challenges that it will face in the future and 

the research that will have to be done to achieve the NRC's mission. These challenges include the 

following examples: 

*, aging of the existing fleet of nuclear power plants and the need to make license renewal 

increasingly risk informed 

* the increasing number of plants that are entering into decommissioning, and the need to 

explore alternatives to the current options for decommissioning 

* technological changes in digital electronics for instrumentation and control of power plants 

and new fuels and cladding

3 NUREG- 1635
technological change that will take the form of new types of power plants that may be quite 

different in design than the current fleet 

In the many research programs discussed in Part II of this report, one can see elements of research 

that are beginning to address these challenges for the future. It can also be anticipated that there will 

be continuing pressure on the NRC as a whole to carry out its mission more economically. As the 

current workforce retires, the NRC will be pursuing its mission with less-experienced staff. Thus, 

demands will arise for computational tools and information aids developed by the NRC's research 

to assist line organizations. 

The following sections offer additional suggestions regarding issues that should be considered by 

the research program in the future: 

1.4.a. New Power Plants and a Revised Regulatory Structure 

A substantial research effort is currently underway to examine the possibility of risk-informing 

individual regulations. As part of that effort, the staff is working on risk informing Title 10, Section 

50.46 of the Code of Federal Regulations (10 CFR 50.46) and the design-basis loss-of-coolant 

accident (LOCA). This laudable effort is challenging. Even if the piecemeal revision of regulations is finally successful, the results may still not be applicable to plants with designs that are 

significantly different from those of plants that are now licensed and in service. As a result, the 

Commission needs to determine, on a somewhat urgent basis, the requirements for a set of 

regulations that are not design dependent. 

The ACRS expects that the most viable option would be a major revision of the regulatory approach 

to be fully risk informed. For this to be a viable option, however, the staff must first explicitly define 

the full spectrum of regulatory objectives expressed in terms of risk acceptance criteria. These could 

include such things as prompt fatalities, latent fatalities, total deaths, injuries, land contamination, 

worker exposure, and releases of all magnitudes of radioactivity. The possible role in the new 

framework of the cornerstones that have been defined in the revised oversight process should also 

be explored. 

The safety goals (as re-interpreted in terms of the need for increased safety for new plants) could 

provide a starting point for this revision of the regulatory approach; however, the current safety goals 

are incomplete, and do not provide sufficient guidance with respect to uncertainties and defense-in

depth. To provide a proper perspective with respect to defense-in-depth and uncertainties, the risk

acceptance criteria need to be developed in terms of confidence limits. 

For such a regulatory system to work, there must be quality, peer-reviewed, acceptable PRAs 

available for the various new reactor concepts. This will require revisiting the concepts of 

neutronics, thermal-hydraulics, fuel behavior, severe accidents, and fission product source terms. 

New models and new data are very likely to be required. Consequently, the ACRS recommends that 

the staff begin now, and continue with early interactions with designers and developers to identify 

the needs in this area.

NUREG-1635 4
The needs for traditional defense-in-depth must be clarified in terms of the unacceptable uncertainty 

in meeting the risk acceptance criteria. Those events that are of high consequences, but with high 

uncertainty for their frequency, must be identified, and decisions must be made regarding how to 

deal with them in a traditional defense-in-depth manner. The current requirement for a containment 

must be re-examined for new reactors. The need for containment must be predicted on the level of 

confidence to be attached to the design's ability to meet the various risk acceptance criteria without 

a containment. 

One likely constraint on the potential for licensing new reactors is early site permitting. The 

Commission may be called upon to create a bank of approved sites. As a result, the agency needs 

to develop criteria and guidelines for how this can be done, particularly in view of the fact that there 

may be intentions to place multiple modular units on a single site. Particular needs are to identify 

dominant accident sequences and associated source terms for the new reactor designs. These needs 

raise two main questions: 

How will the multiple units on a site affect the risk-acceptance criteria? 

* What will the NRC's position be with respect to multiple, automated, modular units that are 

managed by relatively few operators at a central facility? 

Another challenge to the agency with respect to licensing the new designs will be to determine the 

role to be played by the concept of "licensing by test" of a prototype. Again, this challenge raises 

several questions: 

* How many and what kinds of tests are necessary? 

* What data are required? 

* Who is to perform the tests, and who is to participate? 

What are the needs for validating the new computer codes for thermal-hydraulics and severe 

accidents for these reactor designs if the tests are not conclusive? 

Finally, some advance thought should be given to the needs for surveillance and inspection of new 

plant designs. Plants (such as the Pebble Bed Modular Reactor) that utilize continuous refueling will 

undoubtedly propose "online" continuous calibration of instruments, rather than periodic shutdowns 

for "hands- on" calibration. Automated surveillance and diagnostic systems, as well as artificial 

intelligence-based systems being developed at DOE national laboratories and universities, will 

undoubtedly be proposed. Even "noise" and vibration detecting systems monitoring core structures 

and mechanical equipment, traditionally operated off line, may have to be automated and integrated 

into an overall monitoring system.

5 NUREG-1635
I.4.b. Risk Implications of License Renewal and Power Uprates 

Although the NRC has set upon a course of making greater use of risk information in the regulatory 

process, two of the most important regulatory processes now underway, namely license renewal and 

power uprate, are being addressed by deterministic rules. Furthermore, the evaluations of plants for 

license renewal and power uprates, as well as evaluations for other significant regulatory actions 

such as extended fuel burnup, are essentially done independently of one another. Although the NRC 

can seek risk information on each of these types of regulatory actions for a plant, and can consider 

the synergisms that arise among these actions, this is not often done in detail because of the lack of 

data and computational tools that are well suited for evaluations of these regulatory actions or any 

synergisms among them. 

After the implementation of each of these changes, plants will still meet regulatory requirements. However, there is little question that the useful life of systems and components is being consumed 

by the higher demand imposed by these licensing actions. Vessels are more embrittled at 60 years of life than at 40, fuel cladding i:s more embrittled at higher burnups, mechanical components are 

closer to their fatigue limits, and the burst pressure of flawed steam generator tubes is decreased. 

Probabilistic risk assessments do not explicitly account for aging phenomena. A PRA that could 

account for aging of structures, systems, and components would provide measures of increases in 

risk metrics such as core damage frequency (CDF) and large early release frequency (LERF) as a 

plant ages from 40 years to 60 years of operation. These increases would result from a higher failure 

probability of long-lived components that are subjected to 20 additional years of service. These 

higher failure probabilities would be reflected in PRA results in several ways: 

* increased frequency of accidents caused by rupture of passive components 

* increased probability of cascading failures resulting from physical interaction of ruptured 

components with age-degraded components 

0 increased probability of failure of engineered safety systems 

* reduction in the structural capability of the reactor coolant system and the containment 

during severe accidents 

The magnitude of the increase in risk that would be calculated in such a hypothetical PRA cannot 

be estimated at this time. The risk increase may well be found to be small because the license 

renewal process is intended to provide assurance that aging management programs preserve regulatory margins in structures, systems, and components throughout the period of extended 

operation. Although licensees are not required to assess the changes in their risk profiles as a 

consequence of license renewal, the ACRS believes that the NRC should undertake an assessment 

of the impact of the rule on plant risk measures to demonstrate that the governing deterministic rule 

is effective in managing risk and. ensuring that any increases are small and consistent with the safety 

goals.

NUREG-1635 6
Assessments of the risk increases associated with license renewal and power uprates may need to 

consider risk metrics other than CDF and LERF. A study4 sponsored by the Swiss Federal Nuclear 

Safety Inspectorate assessed the risk associated with a 14.7% power increase of a Swiss BWR-6 with 

a Mark III containment. This study showed that the power upgrade would result in only minor 

increases in CDF and LERF, but the reactor would experience a 30% increase in the frequency of 

releases of radioactivity that were not sufficient to result in an increase in prompt fatalities, but could 

increase latent effects and land contamination. This risk increase is directly attributable to the 

increased radioactive inventory and the acceleration of events associated with the increased decay 

heat level at the uprated power conditions. This risk increase may be acceptable but, again, the NRC 

should undertake research to explicitly demonstrate the risks and their acceptability to ensure public 

confidence in upcoming power uprates. This work should also reconsider the criteria used to judge 

changes in risk, which are currently limited to CDF and LERF. 

Extended burnup of fuel, as noted in Part II of this report, has both societal and economic benefits. 

There seems to be little doubt that there will be interest in continuing the historical trend of ever 

increasing fuel burnups in power reactors. This trend has been arrested by recent regulatory fiat, but 

this interruption is likely to be temporary. The nuclear industry is marshaling the technical basis, 

independent of plant aging and power uprates, to justify burnups that go well beyond the current 

regulatory limit of 62 GWd/t. At the same time, events are being recorded of operational difficulties 

associated with high-burnup fuels, and research information is being accumulated regarding unusual 

fuel and cladding responses to upset conditions. 

Concurrent licensing actions (such as power uprates) also consume margins existing in the design 

of structures, systems, and components, and compete with aging for these margins. The proposals 

for power uprates are being considered without explicit consideration of aging. The ACRS believes 

that the NRC will have a growing need to be able to evaluate concurrent licensing actions such as 

license renewal, power uprates, realistic accident analyses, and extended fuel burnups. The 

exclusion of explicit consideration of aging risk from the evaluation of plants for license renewal 

does not mean that the aging effects that are allowed by license renewal should be excluded from 

consideration in assessing the risk posed by other licensing actions that may be affected by the aging 

of structures, systems, and components. For example, the risk associated with a power uprate could 

be significantly higher if the increased radioactive inventory and the time acceleration of events 

associated with the uprated power conditions were to occur in a containment with its ultimate 

capability degraded by aging. 

In principle, PRA could account for the individual and synergistic effects of concurrent licensing 

actions, such as license renewal, power uprates and extended fuel burnup. However, the supporting 

risk and phenomenological models are not yet sufficiently developed to do this. Model uncertainties 

are expected to be large, and the metrics that are commonly used in PRAs may not adequately 

represent the risks that are associated with these changes taking place in nuclear power plants. 

Therefore, the ACRS recommends that the NRC undertake, as part of the move toward risk-informed 

regulation, a research program to investigate the feasibility of improving PRA methods and the 

supporting phenomenological models to explicitly deal with aging, power uprates, extended fuel 

burnup, and other major changes taking place in nuclear power plants. This undertaking is likely

7 NUREG-1635
to be a multidisciplinary challenge akin to the effort that the agency is pursuing in connection with 

the Pressurized Thermal Shock Rule. Certainly, this undertaking will involve the agency's 

considerable expertise in risk analysis, materials and metallurgy, reactor fuel, and thermal 

hydraulics. 

1.4.c. Decisionmaking Method 

With the exception of some isolated instances such as the decisions regarding backfits and possibly 

the significance determination process used in the Revised Reactor Oversight Process, the 

decisionmaking processes that are used in the regulatory process have a subjective character with 

no assistance from formal decisionmaking methods. As an example, consider the integrated 

decisionmaking process discussed in Regulatory Guide (RG) 1.1745. Little guidance is provided 

regarding how to integrate the results of PRAs, defense in depth, and safety margins. Consider also 

the term "increased management attention," which is employed in RG 1.174 with very little 

explanation of what it means in practical terms. As another example, NRC decisions usually involve 

interactions among several affected parties both within and outside of the agency. Yet, these 

interactions lack structure. Simdilarly, the justification of the "action matrix" in the risk-informed 

reactor oversight process remains unclear. 

Some licensees have begun using methods to structure their decisionmaking deliberations. For 

example, the Expert Panel that categorized the structures, systems, and components according to 

their respective risk significances for the South Texas Project exemption request employed questions 

that were assigned weights of relative significance as well as rating scales. The staffs review of this 

exemption request could have been facilitated if the staff had been more familiar with the substantial 

body of literature regarding decisionmaking processes. 

The subjective nature of many regulatory decisions is sometimes justified by noting that the 

decisions have to be made in the face of significant uncertainties. To be sure, there is a body of 

literature on making decisions in the face of uncertainty. This literature covers a spectrum of 

approaches ranging from formal, -mathematical formulations to behavioral approaches. The common 

objective among all of these methods is to rank the desirability of decision options according to well

defined criteria. Many of the decisionmaking methods have matured to the point that relevant 

computer aids are marketed. 

The benefits of using more formldized methods for making decisions are similar to the advantages 

that have come from the use of FRA in safety analysis: 

Decision trees or influence diagrams help the decisionmaker structure the decision process, 

much as event trees and fault trees help structure safety analyses. 

Decision trees and influence diagrams serve as common communication tools for the 

disparate parties affected by the decision, much as event trees and fault trees communicate 

the nature of risk analyses.

NUREG-1635 8
Decision criteria are explicit, and this increases the probability that a consensus will be 

reached on a decision. 

It is possible to conduct understandable sensitivity studies to clearly see how the order of 

preferences of options changes as the inputs to the process are varied. 

The staff argues that it is engaged with other Federal agencies and foreign institutions in examining 

the ways that risk information is used in making decisions. This is too timid an approach. The NRC 

leads the world in the use of risk information, and must accept a leadership role. Therefore, the 

ACRS recommends that the staff initiate a program of research to investigate how best to use formal 

decisionmaking methods in regulatory decisions.

9 NUREG-1635
U. ONGOING RESEARCH PROGRAMS

The ongoing research programs have been divided into 13 technical areas. The programs in these 

areas are discussed in the subsections that follow. 

I1. 1 Probabilistic Risk Assessment Research and Applications 

II. L.a. Risk Assessment Technologies in a Risk-Informed Regulatory 

System 

The Commission has set upon a course of making greater use of risk information in the regulatory 

process and has made enormous strides toward achieving this goal. The Commission has set Safety 

Goals cast in the language of quantitative risk assessment. Risk information and changes in risk are 

used as input to decisions to change the licensing bases of nuclear power plants (Regulatory Guide 

1.174). Risk importance measures are used in categorizing the various structures, systems, and 

components for maintenance (10 CFR 50.65). Risk information is used in the development of 

inservice inspection programs and in specifying the allowed outage times for plant safety systems. Risk information has been used to design the Commission's new Reactor Oversight Program and 

the Commission has instituted a process of risk-informed enforcement of its regulations. The 

Commission has a risk-based process (Accident Sequence Precursor Program) for evaluating risk

significant operating plant events. The Commission is now looking at revising the regulations for 

nuclear power plants using risk information. 

Quantitative risk information that is at the heart of these evolutions in the regulatory structure comes 

from PRA. One would expect that the Commission and the line organizations of NRC that are 

responsible for implementing the risk-informed regulatory process would have available 

comprehensive, state-of-the-art risk-assessment tools capable of analyzing site-specific regulatory 

issues. In fact, the NRC has quite good risk assessment tools, but still it could be improved 

substantially. 

To a very real extent, the risk information available to the Commission and the staff that has been 

used to construct the technical bases in developing a risk-informed regulatory system has come from 

risk assessments done more than a decade ago for five representative nuclear power plants, and the 

individual plant examination (IE) submittals from licensees that are widely regarded to be of 

variable quality and, in many cases, out of date. The risk information available for the design and 

implementation of the risk-informed regulatory process focuses heavily on risk during normal plant operations and not shutdown operations during which so many events requiring special 

investigations by the NRC staff occur.

NUREG- 1635 10
1.1 .b Ongoing and Planned PRA Research Programs

There is a mortgage on the NRC's steps toward a risk-informed regulatory process. This mortgage 

must be paid in research to further develop the agency's capabilities to understand and characterize 

risk in order to carry out the Commission's mission. Table 1 lists the research programs that are 

currently under way and planned in the area of PRA. Comments on these research programs and 

others that should be under way in the agency are presented below. Human performance and human 

reliability are so important to the quantitative assessment of risk that a separate section of this report 

(Section 11.2) is used to comment on research in these areas. Fire protection risk that appears from 

the individual plant examination of external events (IPEEE) submittals to be a significant aspect of 

the risk profiles of nuclear power plants is also addressed in a separate section of this report (Section 

11.3). 

* Risk During Plant Transitions and Shutdown Operations 

Following directions from the Commission, the staff is not developing the capability to estimate risk 

during low-power and shutdown modes of operation. As a result, the staff must rely for risk 

information on its own scoping assessments of shutdown risk at two representative plants and a 

limited number of shutdown PRAs of varying scope and completeness. These scoping risk 

assessments are widely regarded as conservative in the sense that they overestimate risks during low

power and shutdown modes of operation. There is no doubt that the scoping assessments are out of 

date in the sense that they no longer reflect the nature of operations studied at the plants. 

The recent revisions to the Maintenance Rule make clear the regulatory expectation that licensees 

will manage risk during shutdown operations. The nuclear industry has made substantial efforts to 

improve operations during planned shutdowns and has greatly reduced the frequency of unplanned 

shutdowns. The industry has developed several quantitative and qualitative risk assessment tools to 

guide the planning of shutdown operations. In a workshop sponsored by the NRC, licensees 

acknowledged that there is still substantial risk associated with shutdown operations experience. 

Experience shows that the systematic search for accident sequences produces a far more complete 

picture of the way failures can occur in complex systems. That statement is true for normal 

operations; however, we do not have a similarly comprehensive and complete understanding of the 

risks during shutdown operations. Without this understanding, we cannot assess the effectiveness 

of current NRC regulations to deal with shutdown risk. The staff cannot independently evaluate the 

safety of plants for shutdown operations and interpret results of licensees' qualitative risk 

assessments. The staff relies on the licensees to evaluate the risk importance of the equipment that 

is available and not available during shutdown operations, the safety benefits of online inspection 

and maintenance versus inspection and maintenance during plant shutdown, and the risk significance 

of plant changes that will affect shutdown modes of plant operations. 

In addition, the categorization of the structures, systems and components that are needed in a risk

informed special treatment process (or Option 2) will be incomplete as long as the risks from plant 

transition processes and shutdown operations are not quantified. The risk management tools used

11 NUREG-1635
by the industry are not suitable for the shutdown risk determinations needed by the NRC. 

Specifically, the NRC needs the capability to realistically estimate the risk that arises from a set of 

plant configurations during shutdowns that are projected to occur over a plant's operating lifetime. 

The ACRS, therefore, continues to believe that the agency should undertake an effort to develop the 

capability to assess risk during low-power and shutdown modes of operation. We recommend that 

the Commission authorize the staff to undertake such an effort. 

& Standards for Probabilistic Risk Assessments 

As the NRC makes greater use of risk information in the regulatory process, it will be necessary for 

licensees to submit risk information in conjunction with regulatory actions. It will be impractical for the staff to repeatedly review site-specific PRAs associated with these processes. Consequently, the staff must have confidence that PRAs used by licensees in developing of their proposals meet 

some minimally acceptable standard. An effort, supported by the NRC, is under way to develop a 

consensus industry standard regarding the quality of PRAs for normal plant operations. This effort 

competes, in some sense, with an industry effort to peer review and certify for application PRAs 

done by individual licensees. The development of a standard for PRA quality and the approval of 

the industry's certification process have been a struggle. Difficulties arise in this process because 

the NRC has not attempted to definitively define the necessary risk assessment features that must 

be available to support regulatory decisions. The ACRS believes that there is a pressing need for the 

NRC to undertake a research process that will develop definitions of necessary PRA features. As 

the ACRS has noted many times, the sufficiency of information derived from PRAs for 

decisionmaking depends very much on the nature of the decision. It is probably not possible at this 

time to attempt to develop definitions of the sufficiency of PRA features for regulatory purposes. 

* The SAPHIRE Code 

The NRC risk-assessment code (SAPHIRE) continues to undergo development, and revision of this 

code appears at somewhat regular intervals. There is, however, no readily apparent strategy in these 

development efforts. An agency effort to define the necessary features of PRAs to support the 

regulatory process might well provide the agency with a more scrutable strategy for the development of this code. In any event, the ACRS believes that the SAPHIRE code has reached a stage of 

development that the public deserves to see a comprehensive peer review of this code that plays a 

substantial role in the risk-infoncr.ed regulatory system. 

* Common-Cause Failures 

Common-cause failures are of crucial importance in PRAs. The NRC has completed a 

comprehensive study of common-cause failures. Data presented to the ACRS shows that the 

frequency of events involving common-cause failures has been decreasing (See Figure 1). No 

developments in the methodologies for analyses of common-cause failures are planned. The current 

program is one of maintaining and updating databases. The ACRS believes that this program has 

served its regulatory purpose, and that the work can be brought to an orderly close.

NUREG-1635 12
Risk Importance Metrics

An important part of the effort to focus regulatory attention is a more 'risk informed' classification 

of structures, systems, and components. Processes have been developed to do this in the context of 

the Maintenance Rule, graded quality assurance, and currently reassessment of the need for 'special 

treatment.' Such classifications can be expected to be important if attempts are made to risk inform 

license renewal or the 10 CFR 50.59 process. Risk importance measures are critical tools for such 

classifications. Currently, the most widely used risk importance measures are the Fussell-Vesely 

and Risk Achievement Worth. The ACRS has commented in another report6 on the advantages and 

limitations of these metrics. Although the deficiencies in these metrics can probably be overcome 

in particular cases through the use of more qualitative assessment by an expert panel, such 

approaches require additional review by the staff and add to the burdens already placed on expert 

panels. Because of the importance of the classification process to risk-informed regulation, the 

ACRS believes that the NRC should have a research program that searches for metrics that do not 

suffer the disadvantages of those now in common use. The Top Event Prevention methodology and 

importance measures that are predicted on partial derivatives of CDF and LERF are examples of 

alternative approaches that should be investigated for the determination of risk significance. 

0 Improved PRA Capabilities 

The prominent role that quantitative risk assessment must play in a risk-informed regulatory system 

suggests that there should be continuing research to improve and expand PRA capabilities. The 

ACRS recommends three areas of such expansion, including plant aging, safety culture, and latent 

human errors. Inclusion of plant aging seems to be an inevitable need. Although the license renewal 

process is now highly deterministic, it will need to become more probabilistic and risk informed as 

plants requesting license renewal adopt risk-informed systems. There are some indications that it 

may be possible to quantify the effects of safety culture on human performance. Of even greater 

interest is the evidence that there could be an optimal level of regulatory involvement to encourage 

safety cultures. Latent errors (human errors that have no immediately observable impact, but later 

affect a plant) have been found in a study by the NRC staff to have been four times more common 

in risk-significant plant events than overt human errors. 

0 Quantification of Uncertainties in PRA Results 

Results obtained from PRAs are uncertain because of uncertainties in both the models used in the 

risk assessments and in the parameters adopted for these models. It is widely acknowledged that 

quantitative assessments of uncertainties are essential inputs to the regulatory decisionmaking 

process. Yet, carefully quantified uncertainties seldom appear in risk-informed regulatory 

discussions. When uncertainties are mentioned, they are typically the results of propagating 

parameter uncertainties through the analyses. Even with parameter uncertainties, there are questions 

especially with regard to plant-specific distributions, are typically produced by combining generic 

distributions with plant- specific data using Bayes' theorem. The issue of concern is that the 

Bayesian updating process will eliminate low-probability tails of the generic distribution, yet these 

tails may be applicable to real accidents. The ACRS believes that the staff should research the 

quantification of parameter and model uncertainties for plant-specific applications. (See also Section 

11.6.)

13 NUREG-1635
* Risk-Based Performance Indicator Development 

The new risk-informed oversight process is a major agency initiative. This process relies on a 

combination of performance indicators and baseline inspections. The existence of a good set of risk

based performance indicators will allow the agency to have a process that will be more performance based, will recognize plant-specific attributes, and will help define risk-significant trends. The 

ACRS is very supportive of this research.

NUREG-1635 14
Table 1. Ongoing and Planned Probabilistic Risk Assessment 

Research

Task Description 

J8263 Risk-Based Performance Indicator Development 

Y6211 Development of Risk Based Performance Indicators 

L1429 SAPHIRE Maintenance and User Support 

W6224 Risk-informing 10 CFR Part 50 

W6241 Plant Database Development for SAPHIRE 

W6528 Root Cause Investigation Improvements 

W6970, Support in Development of Consensus PRA Standards 

W6971 

Y6036 Modify 10 CFR Part 50 

Y6194 Common Cause Failure Database 

Y6184 Revised Reactor Oversight Program Support 

planned Develop standards for the application of risk-informed and performance 

based regulation in conjunction with national standards committees 

planned Develop improved methods for calculating risk 

planned Develop and maintain analytical tools for staff risk applications 

planned Risk inform NRC regulations 

planned Develop guidance on PRA applications to ensure uniform 

comprehensive application of PRA models 

planned Improve risk-informed decisions associated with natural hazards 

planned Preliminary results of a study using finite element modeling of coastal 

flooding to resolve concerns with the NOAA model predicting beyond 

design basis storm surges 

planned Develop PRA methods for NMSS regulated facilities and devices

15 NUREG- 1635
80 81 82 83 84 85 86 87 88 

Year

I I 9 I 9 I 9 89 90 91 92 93 94 95

C Estimated - itd- 5% Bound -- 95% Bound )

Figure 1. Trend in the Occurrence Rate of Common Cause Failure Events* 

* U.S. Nuclear Regulatory Commission NUREG/CR-6268, "Common-Cause Failure 

Database and Analysis System," June 1998.

NUREG-1635

0.35

0.3 

20.25 

Lu 

0. 

0.15 

0.1 

0.05 

0 

0

I

16
11.2. Human Factors Research

Human performance and human factors are currently viewed as important, al beit unquantified (in 

some cases), contributors to the risks posed by nuclear power plants. It is thought that the importance 

of human factors may well increase as competitive pressures grow within the nuclear power 

production industry. The issues of interest are not solely confined to the plant operators; rather, they 

include the entire workforce involved in the design, operation, and maintenance of a power plant. 

Issues that are cited in this regard include changes in staffing, management, and organization. The 

issues of interest include overt failures in human performance, as well as latent errors that have no 

immediately observable impact, but later contribute to events at a plant. The issues of human factors 

and human performance may well be of even greater interest if generation of nuclear power at future 

plants is achieved by multiple, automated, modular units that are managed by relatively few 

operators at a central facility. 

In the discussions of human factors and human performance presented here, human reliability 

analysis is included along with those projects that the NRC has typically included in this research 

area. This is done because the ACRS believes that there needs to be a closer integration between 

human reliability analysis and other human factors research at the NRC. On the other hand, 

discussion of human factors associated with the introduction of digital instrumentation and control 

(I&C) systems is deferred to the discussion of these digital systems later in this report. 

a. History of Human Factors Research at the NRC 

The NRC established its first Human Factors Program following the accident at Three Mile Island

Unit 2. The NRC commissioned the Human Factors Society to develop a plan to assist in 

establishing this first human factors program. The agency adopted some of the recommendations 

of the Society, and published its first Human Factors Program Plan in 1983. 

In 1985, the human factors research function was discontinued within RES. The function was 

reestablished in 1987 on the basis of a recommendation in the National Research Council's report, 

"Revitalizing Nuclear Safety Research." That report recommended a research agenda that included 

human-system interface design, personnel subsystem (training, licensing, work scheduling), human 

performance (measurement, prediction, and modeling of human error), management and 

organization, and research on the regulatory environment. In 1989, published a report entitled 

'Human Factors Regulatory Research Program Plan" was published, in part, to respond to the 

recommendations of the National Research Council's report. The plan addressed all of the 

recommended areas of research, with the exception of research on the regulatory environment. 

In 1994, a Human Factors Coordinating Committee was formed to coordinate activities within the 

various NRC offices that deal with human factors. The Human Factors Coordinating Committee 

published the "Human Performance Program Plan" in 1996. The ACRS7 was critical of this plan, 

since it was more an inventory of activities than a plan. A revision to the plan, published in 1998, 

included in its mission statement:

17 NUREG-1635
* Identify human performnance issues important to public health and safety. 

Increase understanding of the causes and consequences of inadequate performance by 

humans. 

Implement appropriate regulatory responses to such issues. 

The ACRS agreed with the mission statement of the revised plan, but found that the staff did not 

have a systematic approach for achieving the three goals of the mission statement. 8 A revision was 

published in 1999 and the ACRS9 commented, "...the staff described a disciplined strategy for future 

development of a more technically justified [human performance plan]. We believe the following 

two elements of this strategy are valuable: review of the Accident Sequence Precursor (ASP) data 

to identify the contribution of human performance to significant events and interaction with other 

organizations, such as the Institute of Nuclear Power Operations (INPO), that have a strong focus 

on human performance. The NRC conducted a workshop in April 2001, and to utilize input obtained 

in this workshop to produce a detailed, 5-year, human performance implementation plan. 

The NRC's human performance plans include the acquisition of experimental data to validate human 

reliability analyses. This is being done largely through the NRC's participation in the Halden 

Program. Human reliability analysis research done at the NRC is not a part of the plan. This 

research has focused on the issue of evaluating human errors of commission in PRA and has been 

developing A Technique for Human Event Analysis (ATHEANA) methodology in what appears to 

be substantial isolation from the rest of the human performance activities at the NRC. 

The evolution to a Revised Reactor Oversight Process at the NRC has included an acknowledgment 

of the importance of human performance. The staff has not identified specific indicators for human 

performance in the operation of a nuclear power plant. Instead, it has asserted that any degradation in human perform- ance will be revealed by degradations in other areas that are addressed by performance indicators. The ACRS2 has termed assertions concerning the ability to detect 

degradation of human performance from performance indicators associated mostly with plant hardware an "untested assumption." The ACRS 10 has noted that results of the staffs analyses of ASP 

data have shown that latent human errors with no immediately observable impact were four times 

as common as overt errors. 

b. Ongoing and Planned Research Programs 

Table 2 lists the ongoing and phmned research programs in the area of human performance and 

human factors. Some comments on these programs are provided below: 

* Corrective Action 

The intention of this program is to develop a guidance document to assist in the review of human 

performance aspects of a licensee's corrective action program. This will be an important aid to 

inspectors in the Revised Reactor Oversight Process. This work should be completed, and

NUREG-1635 18
allowances should be made in future research programs to evaluate the effectiveness of the guidance 

and any lessons learned from its use. 

0 Halden Experiments 

The NRC has long been a partner in the international cooperative effort known as Halden project. 

A much proclaimed feature of the Halden project is the study of operator performance in an 

instrumented simulator. Unfortunately, there is not a close tie between the NRC's need to develop 

methods of human reliability analysis and the experiments being done with the Halden simulators. 

Despite several attempts, the ACRS was unable to identify instances where information obtained in 

the Halden simulator experiments was ever used in the human reliability analysis program. This is 

not to say that the information could not be used. There must, however, be some greater coordination 

of the activities. Furthermore, there must be some demonstration that data obtained in a program 

conducted by the Committee on the Safety of Nuclear Installations (CSNI) using a simulator for a 

VVER Russian reactor and Finnish operators will yield results that are transferable to U.S. nuclear 

power plants. 

* Task 20: Determining the Impact of Human Performance on Risk 

* Task 21: Evaluation of Human Performance in the Revised Reactor Oversight Process 

This activity has involved the examination of ASP results, and has yielded a determination that in 

20 events with conditional core damage probabilities between 1xl0 5 and 5x10 3 , "... the average 

contribution of human performance to the event importance was over 90%."01 The study also found 

that latent errors were four times more numerous than overt or active errors. This work is being 

continued to examine the issues of human performance in support of the Revised Reactor Oversight 

Process, and should test assumptions that have been made in the design of the process related to the 

influence of human performance on plant safety. The ACRS believes that this type of work which 

attempts to quantify the impact of human performance on the basis of operational data is of critical 

importance to the design of a useful NRC research program in the area of human factors. The ACRS 

believes that the work should be continued-and even expanded-answers to the following questions: 

What is the risk significance of the human element in the operation of a nuclear 

power plant? 

* When is human performance at a nuclear power plant "good enough"? 

* What are the significant latent errors, and why do they occur? 

* Are current regulatory staffing requirements having a meaningful impact on safety?

19 NUREG-1635
• Root Cause Investigations

The Revised Reactor Oversight Process has altered the role of inspectors with respect to 

investigating of human performance events. Specifically, root causes of the events will be examined 

by the licensees, and for findings of very low safety significance (green findings) NRC inspectors will no longer be involved in the-se detailed examinations of the events. Instead, they will examine 

the licensee's corrective action program to ensure that its methods and executions are adequate. This 

work provides the necessary rev;isions to the human performance protocol. As such, it is a necessary 

activity to support the agency's evolution to the Revised Reactor Oversight Process. 

0 Develop Review Guidance (NUREG-0700) 

The detailed design review guidance for control rooms provided by NUREG-0700 has become 

anachronistic in the face of the NRC's move toward more risk-informed and performance-based 

reactor oversight. In the current regulatory environment, the prescriptive guidance provided in 

NUREG-0700 is more likely to inhibit innovation in the design of control rooms than it is to 

promote safety. Furthermore, the performance of hardware and operators in the control room, rather 

than the details of design, are to be the focus of regulatory attention within the Revised Reactor 

Oversight Process. This will be even more true for the designs of control rooms for future reactors. 

Consequently, the work in this project no longer appears to be consonant with agency needs and 

should be brought to a close. The NRC should be looking for performance-based guidance in this 

area. 

* Develop a Human Performance and Reliability Plan 

This program involves the development of a more detailed, 5-year plan for human performance activities and-research at the NRC, as mentioned above. The ACRS is supportive of the development of this plan, since the directions cf the efforts now seem to be along systematic avenues consistent 

with the objectives in the mission statement for the current plan and supportive of the NRC's 

evolution toward risk-informed and performance-based regulation and the Revised Reactor 

Oversight Process. The ACRS notes that work such as that discussed above in connection with 

Tasks 20 and 21 is more likely to provide defensible technical bases for planning than the NRC will 

obtain by continuing to collect urtquantified opinions at workshops. 

The ACRS reiterates its advice given in regard to a previous incarnation of the human performance 

program plan, the development of a plan for research on human factors is certainly not a simple task. 

This task would be made easier and the recommendations more convincing if the task were guided by a high-level model that identifies the important elements that influence the likelihood of unsafe 

human acts. The need remains for this high-level model as the basis for planning. Human 

performance will not receive the attention that its importance merits in the regulatory scheme if it 

is treated at the margin as seems to be the case in the Revised Regulatory Oversight Process. There 

must be an overarching strategy that properly places human factors in the overall approach to safety. There is no regulatory model that establishes the place of human factors issues in the regulatory 

structure. Are human factors moire important than, less important than, or of equal importance to

NUREG-1635 20
equipment reliability? Without agreement on the importance of human performance in the regulatory 

scheme, there will be no agreement on the research needs, no basis for allocating resources to meet 

these needs relative to other needs, and no agreement on when these needs have been met. 

0 ATHEANA Development and Application 

The ATHEANA effort was initiated with the laudable, if ambitious, goal of expanding the 

capabilities of PRA to include human errors of commission, as well as human errors of omission that 

are currently considered. ATHEANA has attempted to use the concept of error-forcing context in 

its development and the ACRS has applauded this effort. At the same time, the ACRS has identified 

important elements that are missing from the identification of error-forcing contexts, such as the 

safety-conscious work environment. Applications of ATHEANA have failed to demonstrate that 

the contextual approach has provided any benefits when compared to traditional methods. A major 

problem with ATHEANA is its failure to model the relationship between error-forcing contexts and 

the probability of unsafe acts. The staff has acknowledged the need to initiate quantification in 

ATHEANA. After several years of development aimed at expanding the capabilities to quantify 

risk, ATHEANA is only now beginning to address the issue of quantification. The development team 

is now investigating the applicability of other approaches that are not necessarily consistent with the 

much-publicized idea of error-forcing context. ATHEANA has not been evident in any of the NRC's 

programs dealing with human performance, especially mission-critical programs such as the Revised 

Reactor Oversight Process. The ACRS has been told that ATHEANA will be applied to the human 

performance aspects of the Pressurized Thermal Shock program now under way at the agency (see 

Materials and Metallurgy, later in this report). The ACRS concludes that the greatest achievement 

of the work to date is to sensitize people to the concept of error-forcing contexts. The ACRS, 

therefore, recommends terminating the ATHEANA effort, and developing a new plan to quantify 

the probability of unsafe human acts. 

* Develop Human Factors Regulatory Guidance 

This program of future research activities in the area of human performance will be affected by the 

planning efforts that the staff now anticipates. Current plans are to focus on four areas: 

0 effects of deregulation and organization changes, including consolidation of the 

nuclear utilities 

0 quantification of experience and data to improve human reliability 

* analysis for events such as steam generator tube ruptures or anticipated transients 

without scram 

* addressing the issue of latent errors in probabilistic risk assessments

21 NUREG-1635
I

Although the individual elements of these future plans may have some merit, the ACRS is concerned 

with the approach that is being adopted to develop these future plans. The planning process for the 

human performance program seems to be driven by the identification of possible or expected changes in the regulatory environment, examining the human dimension of the regulatory issues 

involved and then defining a project to address specific human performance concerns. This 

approach can produce a list of plausible research activities, but it treats human performance at the 

margins of the regulatory process. There is no articulation of the place of human performance in the 

regulatory structure relative to all of the other elements in the structure that compete for support. Thus, there is no context for judging either the importance of individual issues or the completeness 

of the program.

NUREG- 1635 22
Table 2. Ongoing and Planned Research Activities in the Area of Human Performance 

Task Description 

Y6315 Crew Systems Ergonomics Research Information Analysis 

Center* Support 

Y6315 Corrective Action 

B7488 Halden Experiments 

E8238 Task 20: Determine the impact of human performance on risk 

Task 21: Evaluation of human performance in the NRC Revised 

Reactor Oversight Process 

W6528 Root Cause Investigations 

W6546 Develop Review Guidance (NUREG-0700) 

W6994 ATHEANA 

planned Develop human performance and reliability plan 

planned Develop human factor regulatory guidance 

planned Develop the technical bases for review guidelines that will be used to 

evaluate human performance

*Now called "Human Systems Information Analysis Center."

23 NUREG-1635
11.3. Fire Protection Research

In the aftermath of the fire al: the Browns Ferry Unit, the Commission established additional fire 

protection requirements for nuclear power plants. These highly prescriptive requirements seemingly have been effective. Although significant fires continue to occur at nuclear power plants, none has 

posed a threat to the public health and safety comparable to the fire at Browns Ferry. 

Because the fire protection requirements were "backfits" to the safety profiles of many nuclear power plants, a complicated regulatory quilt of Appendix R plants and license-condition plants exists. All 

these plants were reviewed against the guidance contained in the Branch Technical Position (BTP) applicable to its vintage. There are many hundreds of variances, inaccurately called exemptions, to 

specific provisions of the requirements. As a result, monitoring and inspecting fire protection 

programs at nuclear power plants is a technically complicated process for the NRC. Furthermore, 

the protection achieved by the prescriptive requirements has been obtained at substantial costs to the 

licensees. As commonly found with prescriptive safety requirements, the current condition probably does not take optimal advantage of the engineering and managerial creativity available to the 

licensee, and does not encourage the taking of fire protection measures beyond those that are 

mandated by the regulations. 

The NRC has asked licensees to estimate in the IPEEE effort the residual risk of core damage associated with accidents initiated by fires as part of its vulnerability assessment. Although insights have been slow to emerge from this effort, anecdotal accounts suggest that it has often been found 

that the core damage frequency associated with fire initiators can be commensurate with the core 

damage frequency from all other operational initiators. Such high estimates of the core damage frequency despite the current, comprehensive fire protection requirements may be accurate and not 

a reflection of excessive conservatism in the fire risk assessment methods. Fire protection systems 

are not safety related, and do not have the diversity and fault tolerance requirements of other safety related plant protection systems. If, indeed, anecdotal accounts concerning the risks of fire at nuclear 

power plants are accurate, although the NRC has invested resources in fire-related issues, it may be 

argued that fire has not received the attention (relative to other safety issues) that is merited by its 

risk significance. 

Today, the NRC finds itself with substantial limitations in its abilities to apply risk concepts to the 

regulation of fire protection at nuclear power plants: 

*, Fire growth models used by the NRC do not have the technical sophistication or capabilities of models developed at the National Institute of Standards and Technology (NIST) and in 

other industries. 

Inspections of licensees' safe shutdown programs in response to fires are specialized activities that cannot be totally done by the NRC's regional staff, and require attentions by 

specialists from the headquarters staff.

NUREG-1635 24
* Neither the onsite inspectors nor the Senior Reactor Analysts have fire risk assessment tools 

that allow them to independently evaluate fire hazards and qualitative fire hazard 

assessments done by licensees. 

The staff disagrees with licensees regarding the assessment of electrical circuit failures 

during fires. Computational methods should be developed to facilitate the circuit analyses 

and risk inform the entire process. 

These burdens on the staff and licensees could be relieved by research in the area of fire protection. 

The ACRS has benefitted from an excellent document prepared by the staff entitled, "NRC Fire Risk 

Research Plan: Fiscal Years 2001-2002." This plan is an example of the type of document that is 

needed to define research objectives, identify programs, estimate costs, set milestones, and identify 

future goals. 

* Fire Risk Assessment Tool Development 

This task is intended to develop many of the technical tools and databases that are needed to do a 

risk analysis: 

* improved estimates of challenging fire frequency 

* fire effects database including characteristic heat release rates 

* guidance for identifying scenarios for which smoke effects may be 

important 

* improved estimates of the probability of fire and fire containment 

• configuration and condition dependent fire protection system 

reliability estimates 

0 improved estimates of the probability of spurious actuation of 

equipment during fires 

* adequacy of current methods for analyzing fires in turbine buildings 

• risk methods for prioritizing fire barrier penetration seals 

The ACRS understands that these needs for improvements to fire risk assessment methods have been 

selected from a more extensive list of risk assessment weaknesses, and the ACRS is supportive of 

the proposed research. The ACRS believes that the plans for this research could be improved in the 

following ways: 

* The proposed improvements of fire risk analysis tools should be part of a systematic 

assessment of the scope and quality of fire risk assessment technologies that the NRC needs 

(and will continue to need) to execute its mission now and in the future. 

The proposed improvements should be predicated on an explicit understanding of whether 

fire risk assessment at the NRC will remain the function of headquarters, technology used 

by Senior Reactor Analysts, or technology that is more generally available for use by line 

organizations.

25 NUREG-1635
The ACRS is concerned about an interminable code development effort with episodic revisions that 

are not targeted to meet regulatory needs. Augmentation of the effort with a careful examination of 

the fire risk assessment information that is needed for the regulatory process would allow better 

planning of the needed research. Similarly, quantitative goals for improvements in the technology 

are useful, as long as they are treated as goals and not requirements, to aid researchers in the 

identification of productive paths for their work. 

The ACRS notes that the plans include efforts to characterize realistic heat release rates and other 

features of prototypic fires. Successful completion of this part of the research will be most useful for 

many of the technical issues concerning fire that have arisen in the regulatory process such as fire 

barriers and electrical circuit damage. This research could allow the NRC to use realistic estimates 

of fire characteristics that are appropriate for nuclear power plants, rather than bounding fire 

characteristics that are intended for generic situations found in standards such as those advanced by the American Society for Testing and Materials (ASTM). 

* Fire Risk Requantification Study 

This program will attempt to develop the kinds of insights concerning fire risk that were derived by the NUREG-1 150 study for plants during normal operations. The ACRS is enthusiastic about the 

concept behind this research, but details of the undertaking are not currently available. It is important that the level of effort needed for this work, including the detailed analysis of uncertainties not be 

underestimated. The insight derived from this study could lead to changes in the priority, or even 

the need for other research in ihe plan. 

* Fire Model Benchmarking and Validation 

This task is part of a collaborative program that will compare the capabilities of codes like the 

Computer Code for Modeling Compartment Fires (COMPBRN) and the Fire-Induced Vulnerability Evaluation (FIVE) method with more sophisticated fire models such as the CFAST model developed by the National Institute of Standards and Technology (NIST) and with experimental data. The 

ACRS feels that this is a step in the process of defining the types and technical sophistication of fire 

growth models that the NRC needs tocarry out its regulatory mission. The collaborative effort 

cannot be the-entire scope of work for defining improvements to fire effects and growth models. 

There are aspects of fires important to the safety analysis of nuclear power plants that are not treated 

by the NRC's current model, and these need to be taken into account in the definition of fire 

modeling needs. Eventually, the NRC will need models of fire growth and behavior that are in the 

public domain. This work could be expanded by planning for the development of such models. 

* Fire Risk Assessment Guidance Development 

This task is to develop guidance from the results of the fire risk requantification study mentioned 

above. This may be a first step in defining a standard for fire risk assessments, especially if it builds

NUREG-1635 26
upon a bottom-up assessment of the fire risk information needs of line organizations to carry out the 

NRC mission. 

* Fire Protection for Nuclear Power Plants 

The focus of this effort is to model fire detection methods. The ACRS agrees that there must be 

confidence in the modeling of fire detection systems, but it is not apparent why this work is not part 

of the Fire Risk Assessment Tool Development effort. The need for this work would be better 

appreciated if it were placed in the context of a complete description of the scope and precision of 

fire risk assessment capabilities that the NRC needs for its mission. Again, any plans for the 

improvements in modeling would be helped by the establishment of the goals for the improvements, 

preferably in quantitative terms. 

0 Fire Protection for Gaseous Diffusion Plants 

Fire is very likely to be a risk-dominant accident in gaseous diffusion plants. This task is to develop 

analytical tools to support the placement of sprinklers for fire suppression, fire properties of 

combustible liquids, and unspecified fire protection and risk information needed for the regulation 

of gaseous diffusion plants. The plans do not seem to address issues of smoke and the transport of 

toxic, corrosive, and radioactive materials within plants during fires. Again, the ACRS believes that 

the planning in this activity could be improved by defining the complete scope of fire risk 

assessment needs for regulation of gaseous diffusion plants. The ACRS assumes that much of the 

work done for the gaseous diffusion plants will be applicable to the facility that will be licensed by 

NRC for the fabrication of mixed-oxide (MOX) fuel elements, as further discussed elsewhere in this 

report. Fire is very likely to be the risk-dominant accident at this facility, and transport of 

radioactivity in smoke will be the mechanism by which fire events have significant consequences. 

0 Fire Significance Determination Process Support 

The ACRS believes that the fire-significance determination process must be corrected to have 

technically defensible bases (which it does not currently have) and, to minimize the dependence that 

now exists on subjective evaluations. This research should be directed toward these ends. 

0 Fire Risk Assessment Tools for Precursor Analysis 

The ACRS is very supportive of this element of the proposed fire research. If, indeed, fire poses the 

risks suggested in the IPEEE results, then it will be important to have a process for evaluating fire 

precursors similar to that available for examining precursors during normal plant operations (ASP). 

An area of continuing controversy is the analysis of safe shutdown circuits. It is known that the 

industry is trying to develop a probabilistic model to address this issue and to provide the data 

needed to evaluate the model. The NRC needs to have in hand technical information to define what 

would be prototypic fires that could yield meaningful probabilistic data for the resolution of this 

issue. The NRC may even need to carry out its own confirmatory experimental studies.

27 NUREG-1635
Because manual fire suppression is often an important element of a nuclear power plant's fire 

protection plan, there are human factors that need to be considered in the assessment of fire risks. 

The NRC needs a systematic method to assess how smoke will affect the performance of fire 

fighters, and how the Occupational Safety and Health Association (OSHA) two-in and two-out rule 

will affect the effectiveness of manual fire suppression strategies. It is our expectation that eventually 

experimental data will be needed to resolve these issues. 

A challenge to the staff that is now on the horizon is the licensing of a facility for the fabrication of 

MOX fuel assemblies for use in light-water reactors as a means to dispose of some of the world's 

excess weapons-grade plutonium. Fire will be a major safety concern at this facility. Although most 

of the requirements for the facility can be drawn from industrial fire protection standards, the 

peculiar element of plutonium aerosol dispersal in fires will require specialized attention by the staff. 

Indeed, there are issues concerning smoke aerosol dispersal in nuclear power plants, since corrosive 

components in smoke can affect electrical and electronic circuits long after the fire has been 

extinguished. The ACRS believes that the staff needs to have the technical capability to predict the 

dispersal of aerosols associated with smoke and fires both in the MOX fuel fabrication facility and 

in nuclear power plants. 

In summary, the ACRS is supportive of the current research activities in fire protection being undertaken by the staff. The ACRS feels, however, that in light of the failure of the attempt by the 

National Fire Protection Association to produce a useful, performance-based, risk-informed fire 

protection standard, and in light of results that seem to be coming from the IPEEE effort, it is time 

for a comprehensive re-examination of the agency's needs with respect to fire protection and the 

resources it takes to monitor licensees' fire protection programs.

NUREG-1635 28
Table 3. Ongoing and Planned Programs in the Area of Fire 

Protection

Task Description 

W6733 Screening Review of Internal Fires-IPEEE 

Y6037 Fire Risk Assessment 

W6593 Effects of Aging and Emerging Issues on MOV Performance 

Y6041 Assess Debris Accumulation on PWR Sump Performance 

planned Fire Protection for Gaseous Diffusion Plants 

planned Fire Significance Determination Process Support

29 NUREG-1635
II.4. Reactor Fuel Research 

An important development within the nuclear industry today is the use of reactor fuels to extended 

levels of burnup. Until recently, 12- to 18-month fuel cycles were common. At discharge from the 

reactor, fuel burnups were typically about 40 GWd/t. Today, 24-month cycles are becoming common 

and the burnups of reactor fuels at discharge from the reactor are approaching 60 GWd/t. Some in 

the nuclear industry expect that competitive pressures will lead to industry interest in fuel duty 

cycles of up to 36 months and fuel bumups of up to 75 GWd/t. 

The NRC had been approving ever higher bumups for reactor core loads. The Current agency policy 

is to limit the peak rod burnups to 62 GWD/t. This policy stems, in part, from the irradiation 

experience gained from lead fuel rod testing programs conducted in the 1980s, which involved the 

successful irradiation of several thousand fuel rods to burnups approaching this regulatory limit. It 

must be noted, however, that fuel rods in the high burnup demonstration programs of the 1980s were 

usually placed in nonaggressive, low-power core locations that do not reflect the more aggressive 

environments that are typical of present day reactors (higher power levels, higher core temperatures, 

and higher core flow rates for boiling-water reactors). It is not surprising, therefore, that some 

problems have been revealed for high bumup fuel in recent years. Although the lead rod test 

experience has demonstrated the feasibility of achieving high-burnup levels under normal operating 

conditions, there is not a satisfactory demonstration of performance of high- burnup fuel for design 

basis or severe accidents. For design-basis accidents (DBAs), the available database extends to 

certainly no more than about 35 GWd/t and is predominantly in the range of only about 17 GWd/t. 

This regulatory strategy was based on a confidence that there was a thorough understanding of fuel 

behavior after many years of research sponsored by NRC and that the database developed in the past 

experimental programs could be extrapolated to higher burnups and higher operating power densities 

that have become common because no new phenomena would emerge. 

Of course, new phenomena did emerge. The discovery by French and Japanese experimentalists that 

high burnup fuels were vulnerable to modest reactivity insertion events is now well known. Clad 

embrittlement by hydride formation and the development of a high-porosity, low-thermal 

conductivity rim on fuel pellets are also well known. Less appreciated are the operational difficulties 

that have been encountered with high burnup fuel, including axial offset anomalies in the core 

neutron flux attributed to boron absorption on higher temperature portions of fuel rods and control 

rod insertion difficulties. 

The ACRS regularly reviews the research program on reactor fuel. The confirmatory research 

program developed by RES is exceptionally well conceived, focused, and executed. It has taken 

maximum advantage of its limited resources by developing partnerships with the Electric Power 

Research Institute for the experimental studies of cladding material properties and behavior during 

LOCAs and other types of accidents, and partnerships with researchers abroad who still have the 

capabilities to conduct in-pile fuel safety tests. The program is extending the NRC's empirical fuel 

behavior codes, FRAPCON and FRAPTRAN, to the higher burnups that are allowed today.

NUREG-1635 30
The program designed by the staff is very economical. The limitations on resources has meant some 

approximation have been made. Among these was the assumption that properties of the various 

types of zirconium cladding now in use could be inferred by investigation of only a subset of these 

claddings. Recent events have suggested that the embrittlement of cladding may be more sensitive 

to the alloy composition and clad fabrication methods than supposed when this approximation was 

adopted. As a result, it may be prudent to consider expanding the testing program to include all the 

major types of cladding, and even multiple examples of these clads. The ACRS remains concerned 

that current resource constraints on the program may limit the ability to demonstrate the repeatability 

and to quantify the experimental errors of the various tests. 

The program is constrained by the user need process to its confirmatory role to examine fuel only 

up to (or close to) currently permitted fuel bumups. We know, however, that the nuclear industry 

is interested in fuel burnups that are much higher than 62 GWd/t. There are both societal and 

economic advantages associated with use of fuel to higher burnups. It is the NRC's position that 

licensees will have to supply all the information needed to justify use of fuel to very high bumups. 

The ACRS challenged the research staff to find ways within their limited research program to 

identify definitively the type and quality of data that licensees should make available to support 

applications for yet higher fuel burnups. Especially important is to identify where experimental data 

must be used to substantiate predictions of analyses. The research staff responded with a brilliantly 

conceived program of phenomena identification and ranking. They assembled a prestigious panel 

of world experts in fuel behavior to carry out this effort. The results of the effort are providing 

continuing guidance to NRC research and the efforts to develop regulatory guidance concerning 

higher burnup fuels. 

Still, the constraints on the program are significant. A broader range of investigations is merited in 

light of the societal and economic incentives to use fuel to higher burnups. Furthermore, the NRC 

should be very wary of again abandoning core competency in the area of reactor fuel. An active, 

exploratory research program could go a long way toward sustaining this core competency. The 

following areas could merit investigation: 

Develop predictive models of fuel behavior as a function of burnup capable of predicting rim 

effects and other new phenomena that become apparent only in reactor fuel that is taken to 

higher burnups. 

, Investigate the degradation of high burnup fuel, which may involve fuel swelling and 

slumping processes that are not currently considered in the NRC's severe core damage 

models. 

Investigate fission product release from high burnup fuels which is thought to differ from the 

release rates from lower burnup fuels that were the basis for the NRC's Revised Accident 

Source Term as well as the NRC's severe accident models. 

Investigation of boron interactions with cladding, and the propensity of boron adsorption to 

distort the neutron flux profiles in operating reactors.

31 NUREG-1635
The new element in the NRC's fuel research program involves mixed plutonia-urania fuel that the 

Department of Energy proposes for use in selected light-water reactors as a means for disposing of 

some of the world's excess weapons-grade plutonium. Existing U.S. fuel regulations do not 

specifically address the use of MOX fuel, although prior to the U.S. moratorium on MOX, several 

plants loaded partial MOX cores and LTAs so there has been some U.S. experience using MOX fuel. 

Following the moratorium in the 1970s, other countries decided to continue to pursue MOX fuel; 

therefore, experience exists in Europe and Japan on the use of MOX fuel. The fuel used in these 

countries originates from reprocessed spent fuel and therefore, the isotopic mixture of the plutonium 

and uranium differs from the MOX fuel proposed for the U.S. mission reactors. The European and 

Japanese experience consists of operational data and testing of the fuel. This experience and testing 

has shown that the fuel composition does create some operational and accident behavior differences 

between uranium and plutonium based fuels. Because the industry in Europe and Japan has not 

provided data to support a higher burnup limit, the current limit remains at the initial licensing limit. 

There are differences between MOX fuel produced from recycled fuel and MOX fuel that is 

fabricated for burning weapons-grade plutonium, Weapons- grade plutonium is almost entirely Pu2 3 9

, 

while reactor-grade plutonium contains a mixture of the isotopes Pu239, Pu240 , and Pu241. There are 

also morphological differences. Plutonium in MOX fuel made from recycled fuel is distributed in 

a uranium oxide lattice at the atomic scale. MOX fuel made from weapons-grade plutonium will be 

fabricated by blending particles of plutonium dioxide and uranium dioxide so that there are 

macroscopic islands of plutonia distributed within a matrix of urania. Because plutonium fissions 

preferentially, islands of burnup and porosity become distributed in the fuel as it is consumed. The 

limited testing that has been done suggests that there are substantial differences in the retention of 

fission products in the matrix of MOX fuel. There are also differences in the mechanical and thermal 

properties between MOX fuel and conventional uranium based fuels. The biggest differences 

between MOX fuel and conventional fuels and, indeed, differences between MOX from weapons

grade plutonium and MOX from reactor-grade plutonium, are likely to be neutronic differences. The 

NRC neutronic models are not currently able to model these differences, and substantial upgrades 

are very likely to be needed. The NRR MOX user need letter to RES identifies the changes needed 

to the codes. 

It is evident that there are technical issues to be resolved in the safety regulation of MOX fuel for 

use in light-water reactors. The burden of resolving these issues will fall primarily on the applicant. If experience with high-burnup fuel is applicable, the licensee will provide persuasive evidence of 

acceptable fuel behavior under normal operating conditions. Paper analyses substantiated with little 

experimental data will be provided to argue for adequate fuel behavior under off-normal conditions 

of design basis accidents. Severe accident behavior will not be addressed at all. The NRC will need 

a strong technical foundation to critically review the licensees submittals to ensure that analyses are 

both plausible and correct. The NRC research program will have to provide this strong technical 

foundation.

NUREG-1635 32
The research program that the staff will pursue in connection with MOX fuel is only now being 

defined. Early indications are that it also will make use of the Phenomena Identification and 

Ranking process that has been successfully applied in the research on high burnup fuel. The program 

will have to investigate how the fission products differ from conventional fuels. The information 

will provide the technical bases for specifying accident source terms for the safety evaluation of 

plants that will use MOX fuel. Additionally, developing the neutronics capabilities identified in the 

NRR user need memo on MOX fuel to the Office of Research will provide the necessary tools for 

analyzing a heterogeneous mixed uranium and MOX fueled core. These analytical tools in addition 

to the collaborative programs that the Agency participates in and the knowledge gained from other 

regulatory partners who have experience with MOX, will provide the resources and information 

necessary to make a safety determination on the use of MOX fuel in current LWRs. 

Table 4. Ongoing and Planned Research on Reactor Fuel 

Task Description 

W6200 Code Development and Analysis for High-Burnup Fuel 

W6500 IGR High Burnup Fuel Tests 

Y6367 High Burnup Cladding Performance 

Y6403 Reactor Core Analysis 

planned Provide reports that serve as the basis for confirming or revising criteria 

for high burnup fuel in Regulatory Guide 1.77 and Standard Review Plan 

Chapter 4.2 

planned Develop Agency plan on MOX fuel 

planned Provide updates to the FRAPCON code for predicting fuel performance

33 NUREG-1635
II

11.5. Materials and Metallurgy 

The major ongoing and planned research programs in the field of materials and metallurgy are 

summarized in Table 5. These programs fall into four important topical areas: 

* improved evaluation of the risk of pressurized thermal shock 

* reactor vessel integrity 

0 steam generator tube integrity 

• environmentally assisted cracking 

These programs are discussed in the subsections that follow: 

* Pressurized Thermal Shock 

The NRC has investigated the susceptibility of reactor pressure vessels to pressurized thermal shock 

(PTS). On the basis of this research, the agency established a rather conservative screening process that imposes constraints on the startup and shutdown of nuclear power plants. As the plants age and 

the reactor vessels accumulate more radiation damage, the constraints imposed on the heatup and 

cooling of the plant become more onerous. It can be anticipated, especially in light of the enthusiasm 

for license extension, that a point will be reached at which licensees will need to have less 

conservative pressurized thermal shock constraints on operations. Perhaps in anticipation of licensee 

requests, the NRC has undertaken a remarkable, multidisciplinary program to re-examine the 

phenomena and risk associated with pressurized thermal shock. This program brings together the 

agency's rather considerable capabilities in: 

* accident frequency assessment 

* thermal hydraulics 

* probabilistic fracture mechanics 

To prepare best estimate assessments of PTS, special attention to the development of defensible 

uncertainty analyses of both phenomenological and probabilistic aspects of the research. 

The PTS program has been examined several times over the past year by the ACRS and its 

subcommittees. Although, at first blush, it might appear that the research NRC is doing in this 

program ought to be done by licensees seeking relief from current requirements and guidance, the 

NRC research program can be justified as needed to independently evaluate licensee requests for 

such relief. There is a strong basis for believing that the program will be successful in light of the 

substantial advances that have occurred in the fields of probabilistic fracture mechanics and risk 

assessment in recent years. The program is well organized and well-managed, and appears to be 

progressing toward its goals in a technically defensible way. It may well set a standard by which 

future multidisciplinary research programs at the NRC are judged especially if the promised detailed 

quantifications of uncertainties are done successfully. In any event, the program will be of generic interest, since it is the first major application of the risk-informed technologies that the NRC is 

developing to a rule that has its genesis rooted in adequate protection.

NUREG-1635 34
0 Reactor Vessel Integrity

Nearly all PRAs that the frequency of spontaneous reactor pressure vessel failure is negligibly low. 

Regulatory oversight of the reactor pressure vessels is intended to ensure that this low probability 

is maintained despite the aging of the vessels. The PTS rule, 10 CFR 50.61, provides a conservative 

screening criterion that effectively limits the level of embrittlement of the reactor pressure vessel. 

This screening criterion was intended to ensure that the probability of failure due to pressurized 

thermal shock is less than 5X10 6 per reactor year. Research on the integrity of reactor pressure 

vessels has been supported by the NRC since its formation and prior to that by its predecessor 

agency, the Atomic Energy Commission. Despite this long history of effort, NRR issued a series of 

user requests in 1991, 1995 and 1997 for additional information to address the issues of pressurized 

thermal shock, pressure and temperature limits for startup, low Charpy upper-shelf energy, setpoints 

for low-temperature overpressure protection, characterization of flaw distributions, thermal 

annealing, fatigue and stress corrosion crack growth, and extended operation under plant license 

renewal. 

The fundamental degradation mechanism in the reactor pressure vessel is embrittlement under 

irradiation. The technical issues that must be addressed to meet the overall objectives of the aging 

research program and provide the technical bases for NRR assessments include the availability of 

adequate fracture mechanics analysis methods, reactor dosimetry methodologies, understanding of 

the effects of embrittlement on material properties, and the capability to adequately characterize the 

flaw distribution in reactor pressure vessels. 

The NRC research efforts have provided great insight into the regulatory questions arising from 

reactor pressure vessel embrittlement. The reactor surveillance program, together with Regulatory 

Guide 1.99, Revision 2, provide conservative estimates of upper-shelf toughness loss and transition 

temperature shifts in most cases. Fracture mechanics tools have been developed that can be used to 

demonstrate that embrittled vessels still have adequate upper-shelf toughness. The generic analyses 

in NUREG/CR-6023 and the submittals from the owners groups demonstrate that all the pressurized 

water reactors and boiling water reactors licensed by the NRC should have adequate upper-shelf 

toughness at least through the end of their current licensing periods. If a plant were faced with a PTS 

limit problem and wished to consider annealing as a solution, models for embrittlement recovery 

during annealing and reembrittlement rates are available. The technical expertise provided by the 

program was used successfully to address the recent request by the Boiling Water Reactor Vessel 

and Internals Project (BWRVIP) for inspection relief for boiling water reactor vessels based on 

fracture mechanics arguments for extremely low probabilities of failure of circumferential welds. 

These results represent a substantial accomplishment in meeting regulatory needs. Today, the reactor 

vessel integrity research program is addressing regulatory needs in two key areas: embrittlement 

estimation methods and fracture toughness estimation methods for embrittled materials. While 

related, these areas address distinctly different approaches to estimating fracture toughness for use 

in pressure vessel integrity analyses. Such analyses affect continued operation of the power plant in 

terms of PTS, as already discussed, allowable pressure and temperature limits that govern plant 

startup and shutdown operations, and the acceptability of flaw indications detected during periodic

35 NUREG-1635
inservice inspections. These operational considerations have proven to be limiting for some plants 

during the current license period and will be more so during license renewal periods. 

Embrittlement Estimation Methods 

The current management of vessel embrittlement relies heavily on materials surveillance programs. 

When the surveillance program includes appropriate samples of the limiting pressure vessel material, 

the use of surveillance materials reduces the uncertainty in estimates of vessel embrittlement. Even 

if new, currently unknown embrittlement phenomena should occur, the surveillance materials would 

still provide good estimates of the embrittlement. Unfortunately, we do not have sufficient materials 

in all cases to continue surveillance programs through the license renewal period. Even now, in 

some cases, we do not have surveillance materials (e.g., sample coupons) of the actual belt line weld 

material. We must rely, therefore, on surrogate materials and adjust the results through the use of 

models, which account for the differences in composition between the actual material and the 

surrogate material. Addressing this issue requires models that include the effects of all pertinent 

variables and assessment of potential distributions of these variables. Development and validation 

of these models are important elements of the research program. Although tremendous progress has 

been made in the development of these models over the past decade, statistical analysis of the current 

data shows that there are long-term irradiation/aging phenomena occurring that are not accounted 

for in the current models. This could indicate that, at longer times, other mechanisms of 

embrittlement in addition to the familiar nucleation of nanoscale copper-rich precipitates are 

occurring. Anticipatory research efforts indicate that as irradiation time increases, other 

microstructural changes such as the nucleation of copper-catalyzed, manganese-nickel-rich 

precipitates can produce additional embrittlement. 

The surveillance program only provides estimates of the embrittlement at the inner wall of the 

vessel. For reactor pressure vessel (RPV) integrity analyses, such as PTS, in which a flaw initiates 

and propagates deeper into the wall, the attenuation of embrittlement is of importance for 

determining whether cracks that initiate will arrest before penetrating the wall and whether failures 

will be small leaks or catastrophic ruptures. Calculations of the variation in embrittlement through 

the RPV wall arising from attenuation of the neutron flux have substantial uncertainties, and more 

accurate estimates could reduce the conservatisms in estimates of the consequences of crack 

initiation during PTS events. 

Changes in core loading could potentially lead to a significantly harder neutron spectrum at the 

vessel wall, even if shielded assemblies are used to reduce the total flux. Such a shift in spectrum 

would require reevaluation of the embrittlement correlations and the throughwall attenuation of 

embrittlement damage. 

Fracture Toughness Estimation Methods 

The current methods for estimating fracture toughness use the knowledge of the material chemistry, 

unirradiated properties, neutron fluence, and the embrittlement estimation methods to adjust generic 

fracture toughness curves for both crack initiation and crack arrest. Because of the considerable

NUREG-1635 36
uncertainties in each step in this process, relatively large margin terms are included, making the 

overall estimate of fracture toughness quite conservative. An alternative approach to determining 

fracture toughness, the so-called Master Curve (MC) has been evaluated under the NRC's research 

program and currently has gained significant licensee interest. 

This approach has a better physical basis than the current empirical approach of relating the shift in 

the fracture toughness to the shift in 30 ft. lb. Charpy transition temperature, and could reduce the 

conservatism that was introduced to address the larger uncertainties associated with the empirical 

approach. However, a number issues must be addressed in implementing of the MC approach. One 

is the surrogate material problem noted in the discussion of the surveillance program. Others include 

the shape of the MC, its applicability to the determination of crack arrest toughness, the use of small 

specimen tests to integrate MC with the existing database, confirmation that the shape of the MC 

does not change with irradiation, and a predictive method so that changes in fracture toughness due 

to through-wall fluence attenuation can be estimated. 

Current regulatory requirements for pressure vessels do provide assurance of the vessel integrity. 

However, the conservatisms in these requirements impose significant regulatory burdens, operational 

constraints, and large costs on the industry. These conservatisms are detrimental to safety because 

they may distort the allocation of resources, and because they directly increase exposures to 

personnel and restrict the operational flexibility of the plant during heatup and shutdown. Although 

reductions in the conservatisms in the assessment of vessel integrity can and should be made, there 

is danger in removing some of the conservatisms without an overall understanding of all the sources 

of conservatism and uncertainty. Excessive conservatism in one aspect may be compensating for 

nonconservatism in another aspect. The ongoing research program is addressing important issues 

in our understanding of embrittlement. The understanding of the technical issues associated with 

vessel embrittlement has reached a level of maturity that justifies a reduction in the effort that has 

historically been associated with this area. It can be argued that the elimination of any excess 

conservatism should be the responsibility of industry. However, because of the fundamental 

importance of vessel integrity for reactor safety and public confidence, the NRC must be able to 

independently verify that any changes to regulatory criteria will not increase the probability of 

failure to unacceptable levels. 

The ACRS feels that the need for independent verification of regulatory criteria and the technical 

complexity of the issues require additional information and a level of technical expertise that the 

agency can maintain only through the continued support of a research program in this area. RES 

should work with NRR to review the projected needs and ensure that the program is directed to meet 

those needs. 

0 Steam Generator Tube Integrity Program 

Steam generators have been the most troublesome of the major components in pressurized water 

reactors around the world. Industry efforts have been successful in managing the degradation of 

steam generator tubes due to wastage, pitting, and denting. Fretting, stress corrosion cracking, and 

intergranular attack have proven to be more difficult to control. Even for those plants that have not

37 NUREG-1635
yet had to replace their steam generators, inspection, monitoring, and repair are very expensive. As 

a result, there is substantial industry interest in operating pressurized water reactors with modestly 

degraded steam generators without compromising safety. 

To ensure the structural and leak integrity of degraded steam generators, it is necessary to be able 

to evaluate and characterize degraded tubes, evaluate structural integrity and leakage associated with 

degraded tubes, and characterize the progress of degradation over future inspection cycles. 

The activities of the ongoing research are intended to address these technical issues. The industry 

has developed inspection technologies, performance demonstration and qualification programs that 

have improved the effectiveness and reliability of steam generator inspection programs. The research 

program is carrying out an independent assessment of steam generator inspection reliability through 

an industry round-robin on a ste am generator mockup developed as part of the research program. The 

mockup contains hundreds of cracks and simulations of artifacts such as corrosion deposits, tube 

support plates and the like that make detection and characterization of cracks more difficult in 

operating steam generators than in most laboratory situations. An expert group has reviewed the 

signals from the laboratory grown cracks used in the mockup to ensure that they provide reasonable 

simulations of those obtained from real cracks. The number of tubes inspected and the number of 

teams in the round-robin are intended to provide better statistical data on the probability of detection 

and the crack characterization accuracy than is currently available from industry performance 

demonstration programs. 

In addition to the round-robin activities, there is also a task on advanced technology for 

nondestructive examination (NDE) of steam generator tubes. There are two motivations for this 

work. First, the work is intended to maintain expertise in area so that industry efforts to improve the 

accuracy and reliability of inservice inspection can be independently examined. Second, the work 

on advance NDE techniques is needed for the round-robin testing with the mockup which raises the 

problem of determining the true state of the flaws. Although this can be done by metallographic 

sectioning and burst and leak testing, such methods are too expensive and time consuming to be used 

for hundreds of cracks, and a nondestructive assessment of the true state of the flaws is needed. 

A tube failure and leak measurement facility has been constructed that can provide prototypical 

conditions for both normal operating and main steam line break conditions. It is being used to assess 

industry models for tube integrity and leak rate. The entire progression of the failure from the 

initiation of a small leak, stable crack growth under increasing pressure, to unstable crack growth 

and gross rupture can be simulated in the facility. A high temperature tube failure facility has been 

used to study the behavior of f [awed tubes under severe accident conditions. 

In most cases, degradation of steam generator tubing by stress corrosion cracking is currently 

managed by plug or repair on detection, because current NDE techniques for characterization of 

flaws are not accurate enough to permit continued operation. This is very conservative in many 

cases, since flaws less than 40% throughwall or even deeper short flaws have very little impact on 

tube integrity. On the other hand, current inspection technologies and procedures can miss flaws that 

will lead to steam generator tube ruptures.

NUREG-1635 38
The development and qualification of inspection and degradation management strategies are the 

responsibility of industry. Current regulatory guidance is properly more performance oriented and 

less prescriptive that it has been in the past. Because steam generator tubes account for well over 

50 percent of the primary coolant pressure boundary and failure can lead to containment bypass, it 

is important to maintain the capability and technical expertise to independent assess industry 

developments. The program has already made important contributions to the assessment of issues 

not within the original proposed scope of the program such as thermally induced failure of flawed 

tubes, behavior of the tubes with ElectrosleeveTM repairs, and the potential for tube failure 

propagation under design basis and severe accident conditions. It is currently addressing the issue 

of the effect of pressurization rate on tube failure pressures that has arisen from consideration of 

performance demonstration issues at the Arkansas Nuclear One Unit 2 plant (ANO-2). The NRC 

must be prepared to expect other similar issues to arise during continued operation with degraded 

steam generators. Even in the case of generators that have been replaced, laboratory studies show 

that Alloy 690 is susceptible to cracking in environments that can occur on the secondary side of 

steam generators. 

The recent examination by the ACRS of technical issues arising out of the differing professional 

opinion (DPO) on alternative repair criteria identified a number of technical issues that need to be 

further addressed and which should be considered by RES in the formulation of the steam generator 

research program: 

the effects of the loads and vibration due to the large blowdown forces that occur during 

rapid depressurization such as would occur following a main steam line rupture on tube 

leakage and integrity 

the effect of tube support plate movement during depressurization on the tubes 

* a better understanding of the heatup of the steam generator tubes associated with the 

countercurrent flows produced by natural convection under some severe accident situations 

• a better understanding of the radionuclide releases associated with iodine spiking 

the DPO issues focused on the use of alternative repair criteria for outside diameter stress 

corrosion cracking at drilled-hole tube support plates. A more critical issue is whether current 

inspection techniques in areas of potentially high noise are adequate to identify significant 

tube degradation. This issue is of particular concern at locations like U-bends and the roll 

transition region where the tubes are not constrained by the tube support plate and where the 

additional loads due to blowdown could be even more significant than in the case 

degradation in the tube support plate region 

Although the problems associated with degraded steam generators should greatly diminish as more 

licensees replace the current models with designs much less susceptible to degradation, assurance 

of steam generator integrity will remain a significant problem for some time yet.

39 NUREG-1635
• Environmentally Assisted Cracking

Since 1967, the NRC (and its predecessor the Atomic Energy Commission) have conducted research 

that addresses aging of reactor components. Environmentally assisted cracking of reactor structural 

materials has been a recurrent problem in operating reactors since the mid 1970's. The NRC research 

in this area was originally initiated to address boiling water reactor pipe cracking problems. Since 

that time, the focus of the rese;rch has shifted to address other problems in environmental cracking such as irradiation assisted stress corrosion cracking. This research has been used to evaluate and 

establish regulatory guidelines to ensure acceptable levels of reliability for light water reactor 

components. The products of this program have been technical reports, methodologies for evaluating licensee submittals, and other inputs to the regulatory process. These results have led to the 

resolution of regulatory issues as well as the development, validation, and improvement of 

regulations and regulatory guides. 

Fatigue and stress corrosion cracking continue to be significant degradation mechanisms in piping 

systems of nuclear plants. Because pipe failures are included in the DBAs and effective mitigating systems are available, such failures typically do not rank high in terms of contributions to the risks 

posed by a nuclear power plant. The potentials for such failures have a significant impact on plant 

operability, worker exposure, and public confidence. 

High radiation levels in a reactor core can increase the susceptibility of the core structural materials 

to stress corrosion cracking because of changes in the water chemistry due to the radiolytic decomposition of water and degradation of the materials themselves. Although many of the affected 

components can be replaced, replacement is frequently difficult or impractical. Catastrophic failure 

of some internal components That are needed to maintain core geometry such as the top guide in 

boiling water reactors could have serious safety consequences, but internal structures tend to be 

highly redundant and tolerant of cracking. But again, failure in such components has a large impact 

on plant operability, worker exposure and public confidence. Currently, NRR has accepted crack 

growth rate curves only for materials subject to fluence levels less than 5x10 20 n/cm 2. For materials 

that have sustained higher fluences, they are forced to assume crack growth rates thought to be very 

conservative. 

To ensure the integrity of the reactor coolant system boundary and the reactor internals, susceptible materials and conditions must be identified, the effectiveness of mitigating measures demonstrated, and crack growth rates must be determined to ensure that selected inspection intervals are adequate 

to ensure integrity. 

The industry has a very substantial research effort in this area to address these issues. Much of the 

work is currently done through broad industry cooperative efforts such as the BWRVIP and the 

modifications/rework package (MRP), but individual owners groups and vendors sponsor additional 

work to address problems that are unique to a smaller class of systems or are proprietary such as the 

Noble Chem process to reduce susceptibility to irradiation assisted stress corrosion cracking in 

boiling water reactors.

NUREG- 1635 40
The objective of the NRC research is to provide sufficient data and physical understanding to enable 

the NRC staff to assess the validity of industry analyses of the likelihood of core internal component 

degradation and failure as a result of irradiation assisted stress corrosion cracking, to evaluate 

licensee submittals concerning inspection and remediation and the industry models used for 

establishing inspection intervals and repair criteria. The results support licensing decisions related 

to operating plants and license renewal reviews. Currently, three aspects of environmentally assisted 

cracking are being investigated: 

* fatigue of reactor structural materials 

* irradiation assisted stress corrosion cracking 

* environmentally assisted cracking of nickel-based alloys and weld metals 

The NRC effort is leveraged through participation in the Cooperative Irradiation Assisted Stress 

Corrosion Cracking Research Program managed by the Electric Power Research Institute which 

includes vendors, utilities, and regulatory authorities from the U.S., Europe, and Japan. Participation 

in the program also provides access to industry irradiation programs in the BOR-60 reactor in 

Russia. Each member contributes funds, which are used for joint research work as well as in-kind 

results from their own research programs. The NRC contractors and staff have also worked with the 

Pressure vessel Research Council in their efforts to develop a consensus position on the importance 

of environmental effects on the fatigue life of reactor materials. The experimental studies on 

irradiation assisted-stress corrosion cracking depend on tests being performed at the Halden reactor 

as part of the NRC participation in the Halden Project. 

Maintenance of the integrity of the reactor coolant boundary system is a fundamental aspect of 

defense in depth for reactor safety. It is clear that the primary responsibility to demonstrate that 

degradation mechanisms can be managed adequately belongs to the industry and they do appear to 

making the necessary investments to provide this information. The NRC research programs in this 

area are substantially leveraged through cooperative international programs and provide an 

independent capability that is important to maintaining public confidence. Dealing with degraded 

components is not a prospective problem. Staff must make judgments now on the operability of 

degraded components. The ongoing research programs help to maintain a technical capability to 

ensure that such judgments are valid and defensible. The additional information provided by the 

programs and the industry efforts will lead to more effective and efficient management of the 

degradation that occurs.

41 NUREG-1635
Table 5. Ongoing and Planned Research Programs in Materials and Metallurgy 

Task Description 

Y6095 Underwater Welding of Irradiated Stainless Steel 

W6275 Assessment of Reliability of Ultrasonic Testing and Nondestructuve 

Evaluation Methods 

W6953 Heavy-section Steel Irradiation (HSSI) Program 

W6878 Statistical analysis of Fracture Toughness Behavior of RPV's 

Y6127 PTS Risk Assessment 

Y6193 Probabilistic: Fracture Mechanics for Revision of PTS Rule 

Y6249 Analysis of BWR Reactor Vessel Metallic Samples 

W6610 Environmentally assisted cracking of LWRs 

W6631 International. Pressure Vessel Technical Cooperation Program 

W6212 Radiation Embrittlement Damage Analysis and Predictions II 

W6986 Fracture mechanics technology for LWR Materials 

W6487 Steam Generator Tube Integrity Program 

planned Develop Regulatory Guide for Implementing Leak Before Break Concept for 

Pipes 

planned Develop Thermal Hydraulic Conditions to assess Steam Generator Integrity 

planned International Cooperation on Environmentally Assisted Cracking 

planned Generic Flaw Density and Size Distribution for Reactor Vessel Welds

NUREG- 1635 42
11.6. Thermal-Hydraulics Research

The ongoing and planned research programs in the area of thermal hydraulics are summarized in 

Table 6. The centerpiece of the research on thermal hydraulics at the NRC is the consolidation of 

the various thermal hydraulic codes into a single NRC code. Through its Subcommittee on 

Thermal-Hydraulics Phenomena, the ACRS reviews this code consolidation effort regularly. The 

ACRS believes that this code consolidation is an essential activity that has been well-planned 

and is progressing very well despite the difficulties that the NRC has had in staffing and funding 

this work. Should this effort not fully meet its goals in the planned time frame for the work, the 

ACRS believes that this simply reflects the inexact nature of research planning and not the 

exceptional diligence and skills of those conducting this effort to provide a consolidated code. 

The ACRS strongly supports the NRC having its own thermal-hydraulic code. Because of the 

evolving state-of-the-art in this technical field, independent calculations must be made for reactor 

transients using a code that is familiar to the analysts. The availability of a familiar thermal

hydraulics model makes it possible for the. staff to independently assess applicant predictions, 

investigate the effects of assumptions and approximate physical modeling, discover sensitivities 

and weaknesses, incorporate improvements in the analytical and solution methods, and provide 

predictions in interpretations of plant transients. 

A second important component of the ongoing thermal hydraulics research is the contribution to 

the staff initiative in reassessing pressurized thermal shock. This initiative is discussed further in 

this report in the Section 11-5 dealing with Materials and Metallurgy. Suffice here to say that 

thermal hydraulics is an important element of the comprehensive reexamination of pressurized 

thermal shock being undertaken by the staff and that the ACRS fully supports this work. 

Several of the ongoing research programs involve maintenance and user support for thermal

hydraulic codes that will eventually be replaced by NRC's consolidated code. In general, the 

ACRS believes that maintenance of mature codes should be done by the NRC staff and not be a 

contracted activity. In the particular cases of the thermal hydraulic codes, the ACRS understands 

the maintenance of other thermal hydraulic codes such as RELAP5 and TRAC-BWR to be 

essential interim measures taken to meet the needs of the NRC line organization until the 

consolidated code is available for routine use by these organizations. 

The following paragraph present the ACRS comments regarding other ongoing thermal

hydraulic research programs: 

0 PUMA Integral Test Facility 

RES has a strong desire to keep at least some of the ever decreasing number of thermal-hydraulic 

test facilities viable. The Purdue University Multidimensional Integral Test Assembly (PUMA) 

facility is the only BWR test facility available to the NRC that can model the interaction between 

the containment and the reactor coolant system, which has some importance in the analysis of 

event scenarios involving an anticipated transient without scram (ATWS) and other BWR

43 NUREG-1635
instabilities. Understanding of such issues will become more important as the industry moves to 

use fuel with higher burnups, increases the power of plant operation, and adopts best estimate 

safety analyses. It is essential that the NRC better define the role that the PUMA facility can play in improving the understanding that the NRC has of these issues, and the independent validation 

of licensee's contentions on the basis of proprietary data. It is important to develop and evaluate 

the tradeoffs between continuing to use the PUMA facility and developing a better facility that is 

more suited to the agency's research. 

* Oregon State University Integral Test Facility 

This facility is being used to address particular, current issues in thermal hydraulics. However, 

the future maintenance of this, facility needs to be better justified in terms of specific needs to 

improve the NRC's understanding of thermal-hydraulic issues and to validate and improve its 

consolidated thermal- hydraulics code. 

* Rod Bundle Heat Transfer 

Research at the Pennsylvania State University into rod bundle heat transfer is intended to provide 

a mechanistic model of core reflood. Uncertainties in current, empirical models are too high to 

support the NRC efforts to mal:e requirements in 10 CFR 50.46 risk informed. Uncertainties in 

the current reflood models jeopardize the ability of the NRC to reduce conservatisms in the 

decay heat model licensees are required to use and this may have collateral effects on regulatory 

activities associated with spent fuel storage both in water pools and in dry casks. The ACRS fully 

supports this work and has been quite impressed by its technical quality. It is imperative to 

coordinate the ambitious data collection in this program with the development of models for the 

thermal-hydraulic computer codes. Both data collection and model development may have to be 

modified as computational uncertainties are quantified. 

* Two-phase Flow and Heat Transfer for CFD 

An emphasis in this work is on the tracking of liquid-vapor interfaces. The ACRS is surprised 

that such research is needed. Vendors of commercial computational fluid dynamics (CFD) codes 

have done quite a lot of work on developing means to track such interfaces. 

* CFD Code Models 

This work is an example of longer term research sponsored by the NRC. The work is undertaken 

in anticipation of the day when. thermal hydraulics codes used for reactor safety assessments will 

bear a closer resemblance to the commercial CFD codes that are now available. Even now, the 

agency is finding advantages in the use of CFD models for the analysis and resolution of local 

thermal hydraulic issues that arise in the regulatory process. The ACRS believes that it is, 

indeed, important for NRC to mi.aintain a foothold in this important direction in the development 

of thermal hydraulic modeling. '[t is, however, also important to ensure that the work being done

NUREG-1635 44
in this program does not digress into more academic realms rather than remaining focused on the 

issues of regulatory significance. 

0 Steam Generator Tube Rupture 

This activity is directed at a current regulatory concern. The ACRS supports this work, but 

wonders if the work is sufficiently extensive enough to finally resolve the issues. The following 

examples illustrate the particular areas where additional research may be needed: 

* shock and sympathetic vibrations of the reactor coolant system accompanying the system 

depressurization in a main steamline break or steam generator tube rupture 

"* identification of parts of the reactor coolant system (other than reactor nozzles, surge lines, 

and steam generator tubes) that are vulnerable to thermally induced failure as a result of 

countercurrent natural convection under severe accident conditions 

"• 3-Dimensional Neutronics for MOX Fuel 

Neutronic analysis codes that are currently available to the NRC have been tuned to data on 

urania-based fuel. A number of differences complicate the use of these neutronic codes for the 

analysis of fuels composed of mixtures of plutonia and urania. Not the least of these is the 

approximate treatment of the delayed neutron fraction. Consequently, this work is viewed by the 

ACRS as essential and in anticipation of applications by the Department of Energy and current 

licensees to burn MOX fuel in power reactors. 

* Graphical User Interface for Thermal-Hydraulics codes 

The graphical user interface for the RELAP5 code has been completed. Adaptation of the 

interface to the TRAC-M code will be completed by the end of FY 2001. The program will 

become, then, a maintenance and improvement activity. 

Future Directions in Thermal Hydraulics Research 

It is evident that in the future more realistic thermal-hydraulics analyses rather than highly 

conservative, bounding analyses will be used in the safety assessment of power reactors. In order 

to support this evolution and the evolution to a more risk-informed regulatory process, the NRC 

staff needs to address two questions: 

* When is the realistic estimate obtained from a code good enough? 

* How uncertain is this estimate? More specifically, could the probability of exceeding 

some limit be unacceptable? That is, what are the quantitative risk implications in the 

uncertainties of the predictions obtained with the code?

45 NUREG-1635
At present, there are no loguical quantitative approaches to answer these questions. They are, instead, answered by the use of someone's judgment. Already there have been examples of 

different staff members having quite different judgments on these questions and there may even 

be differences in the judgments of the NRC staff and its management. 

The staff needs a more logical and defensible approach to answering these two questions. 

Answers to thermal-hydraulic questions will be acceptable when the bias in the predictions of 

key parameters is small enough and when the uncertainty of the combined bias and deviation of 

the predictions is small enou~gh. What is enough in this regard depends on the particular regulatory decision that is to be made. Both the measures of uncertainty and the criteria for 

acceptance depend on the risk significance of the answer. They also depend on how closely some 

limit is to be approached. The smaller the margin, the smaller is the acceptable uncertainty. Therefore, the modeling uncertainty needs to be determined and must be quantitatively related to 

the risk predicted in probabilistic risk assessments. 

The present approach to uncertainty analysis is often just the variation of some selected 

coefficients in the code and descriptions of how answers are changed as a result of these 

variations. This depends on the selection of suitable sets of coefficients for variation and suitable 

ranges for the variations. The process does not address the effects of the code itself, the form of 

the equations solved by the code or the solution method adopted in the code. 

In any case, it is not enough to estimate uncertainty entirely by theoretical means. There must be 

some comparisons to meaningful data. Often, there is a very small set of data chosen for code 

assessment. For example, code assessment may involve one comparison with a single loss-of

fluid test (LOFT) facility and a comparison with a BETHSY test. The predictions are seen to be 

not far from the data and it is guessed that this is an adequate demonstration of code viability for 

predictions that may not be closely related to the tests chosen for the comparison. 

The ACRS does not believe that thermal-hydraulic aspects of risk-informed regulation can 

proceed very far until the questions posed above can be satisfactorily addressed. It will take 

research to develop these answers, and the main thrusts of this research should include the 

following: 

"* Define the need for evatuating of the adequacy of the outputs of realistic thermal hydraulic 

codes. Specify what needs to be determined about the realism, accuracy, and other 

qualities of the code predictions. Develop quantitative measures of the performances of 

thermal-hydraulic codes. 

"* Develop tools for calculation of these quantitative measures for use by the NRC line 

organizations. 

* Develop means to assess these quantitative measure from data obtained in separate effects 

tests and integral response tests. Find ways to reduce arbitrariness and excessive reliance 

on judgment in the assessment of code predictions of experimental results.

NUREG- 1635 46
"* Use the consolidated NRC thermal-hydraulics code to develop worked examples 

demonstrating that the above have been accomplished. This may involve the restructuring 

of the codes, perhaps to incorporate uncertainty evaluations into the code itself at 

fundamental levels rather than as responses to user inputs to the code. 

"* Ensure that outputs of the work is in forms best suited to the users in the NRC line 

organizations. 

"* Explicitly address model and parameter uncertainties. 

Table 6. Ongoing and Planned Thermal Hydraulics Research Programs

NUREG-1635

Task Description

W6240 Boron Mixing Experiments 

W6245 TRAC-P Maintenance and Consolidation 

W6698 PUMA Integral Test Facility 

W6699 Oregon State University Test Facility 

W6855 Rod Bundle Heat Transfer 

W6995 Two-phase Flow and Heat Transfer for Computational Fluid 

Dynamics 

W6996 Development of Models for NRC's Computational Fluid 

Dynamics Code 

planned Continue development and upgrading of NRC's 

thermal/hydraulic code 

planned Develop measures of uncertainty in code predictions 

planned Develop quantitative measures of code assessment 

planned Perform experiments in support of development of models 

which can significantly improve code capabilities 

planned Define the functional requirements for realistic codes in 

response to anticipated regulatory needs

47
11.7. Severe Accident Research

The risk to the public by nuclear power generation arises if accidents progress to a point that 

results in fuel degradation and the release of large quantities of radioactive material into a plant's 

environment. The NRC has invested heavily in the investigation of severe reactor accidents and 

has developed an integrated, systems-level computer code for the analysis of severe reactor 

accident progression. Research on severe reactor accidents has been curtailed substantially in 

recent years. The remaining severe accident research programs are listed in Table 7. Summary 

comments concerning these programs are provided immediately below. These summary 

comments are followed by a discussion of the continuing needs for severe accident research. 

The following subsections summarize the ACRS comments regarding the ongoing severe 

accident research programs: 

* MELCOR Code Development and Assessment 

The NRC has done a good job consolidating the diverse results of its research into an integrated, 

systems level analysis tool called the MELCOR code. The continued development of this code is 

needed as research results from worldwide programs on severe accidents, such as the PHEBUS

FP program and the planned ARTIST program, become available. The ACRS believes that the 

following topical areas (among others) should be included in the continued development of this 

code (as further discussed below): 

degradation of high-burnup and MOX fuels with hydrided cladding, and the effects of fuel 

swelling on damage progression 

* aerosol transport into and through steam generators during bypass accidents 

improved modeling of natural convection within pressurized reactor coolant systems 

during severe accidents 

* modeling of loop seal clearing and reformation during severe accidents 

effects of air ingression on residual fuel behavior following vessel rupture in a severe 

reactor accident 

There are many opportunities being made available worldwide for the testing and evaluation of 

the MELCOR code. These opportunities consist of International Standard Problems that involve 

the comparisons of code predictions to results of tests. The NRC participation in these exercises 

has been limited by the availability of funds. The NRC could benefit through participation in 

international programs as a way of validating its codes. The ACRS believes that resources 

should be made available for more participation in these international collaborative efforts.

NUREG-1635 48
. MACCS Maintenance and User Support

The MELCOR Accident Consequence Code System (MACCS) is a computer model of the 

dispersal of radioactive material in the environment and the health consequences of such 

dispersal following a severe reactor accident. This is clearly a calculational capability that the 

NRC needs to have. The ACRS believes that as a general rule, the maintenance and user support 

for a mature code like MACCS should be done by the NRC staff and not as a contracted activity. 

The current program may be needed as an interim measure since the MACCS code is becoming 

outdated relative to accident consequence models developed elsewhere in the world. Additional 

research to improve MACCS may, then, be needed to support the agency's move toward a risk

informed regulatory system. 

* Probabilistic Consequence Model Development 

This program is intended as a contribution to a collaborative European program on the use of 

expert elicitation to develop distributions for uncertain inputs to models of the dispersion of 

radioactive materials and the consequences of this dispersion. This is a crucial element in 

determining the uncertainty in estimates of risk associated with the use of nuclear power 

generation. 

The ACRS believes there to be a strong justification for this research program. The research will 

augment the study of uncertainties in risk estimates begun with the NUREG-1 150 study of five 

representative plants. 

It will also assist in the identification of areas where the MACCS code needs major improvement 

in order to reduce the uncertainties in the calculations of accident consequences. This program 

also fits nicely into the need for NRC to maintain a level of participation in pertinent, 

international cooperative research programs. 

0 Analytical Support for CONTAIN Code Development 

This program is intended to provide maintenance and user support for a mature code dealing with 

containment phenomena under accident conditions. Line organizations at the NRC make some 

use of this computer code. Again, the ACRS believes that maintenance and user support for a 

mature computer code should be done by the NRC staff and not as a contracted service. 

Furthermore, the CONTAIN code has become anachronistic since the maturation of the 

MELCOR code. The agency should have a strategy to wean line organizations away from the use 

of the CONTAIN code, encourage the use of the more modern code MELCOR, and promptly 

eliminate the need to provide further user support for the CONTAIN code. 

* OECD Lower Head Failure Program 

The objective of this program is to characterize the timing, size and location of the failure of the 

reactor vessel lower head under severe accident conditions when this head is exposed to molten

49 NUREG-1635
core debris. The program was particularly pertinent to the NRC when examining the safety of the 

proposed advanced light water reactor AP600 when in-vessel retention of core debris was part of 

the safety strategy for this reactor. This safety strategy was abandoned and the program is now 

justified as supporting the assessment of severe accident management strategies related to ex

vessel fuel coolant interactions and direct containment heating. 

Although this program is part of an international collaborative research agreement, the ACRS 

believes the research is not needed by the NRC. There are no safety strategies being proposed 

now that involve invessel retention of core debris. The ACRS believes that severe accident 

management strategies shoukld. not be dependent on the precise rupture behavior of reactor 

pressure vessels. The ACRS recommends that this research be brought to a prompt and orderly 

conclusion. 

0 Technical Support for ]V[olten Fuel Coolant Interactions Research 

This program provides the technical support for the development of instrumentation capable of 

mapping fuel-coolant mixing regions in order to characterize coarse breakup and fine 

fragmentation of core debris suddenly immersed in water. This breakup and fragmentation is the 

first step in a process that leads to explosive interactions of molten core debris and water. The 

current understanding of energetic interactions of molten core debris with water is not sufficient 

to predict the energetics of interactions expected to occur during severe reactor accidents. 

Research would be useful to provide a fuller predictive capability. The research needs for molten 

core debris interactions with water include large-scale tests with prototypical materials. 

Instrumentation being developed in this program would not likely be useful in such tests. 

Worldwide, the experimental investigation of fuel-coolant interactions has been abandoned. The 

need for this program should be reevaluated. 

Future Directions in Severe Accident Research 

With its risk-informed initiatives, the NRC is engaging in a major shift in the way that it 

determines the regulations and conducts its regulatory activities. This shift requires a good 

capability to determine risk and its associated uncertainties. Risk is dominated by severe 

accidents involving core degradation and the massive release of radioactive materials. It is then 

expected that severe accident analyses should play a prominent role in the agency's risk-informed 

initiatives. Currently, however, the agency relies on CDF and LERF as risk metrics. It is thought 

that these metrics and their uncertainties can be readily determined using available PRA 

technology and the current understanding of severe accidents. Because of this, there is a belief 

that sufficient severe accident research has been done. This belief persists because current PRAs 

require very little in the way of severe accident input to determine CDF and require only 

knowledge of those severe accident phenomena that lead to early containment failure to 

determine LERF. The need is to determine the containment loads due to blowdown dynamics, 

hydrogen production and combustion, direct containment heating, and to evaluate the mechanical 

response of containment to these processes. It is believed that there is sufficient severe accident 

knowledge to do this.

NUREG-1635 50
The ACRS does not share this belief regarding the current adequacy of knowledge concerning 

severe accident processes. First, the determination of CDF requires input on the success criteria 

for the emergency core cooling system to successfully terminate accidents before they can 

progress to extensive fuel melting. Current success criteria are based on information developed 

for fuel with burnups less than about 38 GWd/t. The trend in the nuclear industry is to push for 

much higher burnups certainly to 62 GWd/t and perhaps as high as 75 GWd/t. There is evidence 

that the flow blockage behavior of such high-burnup fuel under accident conditions differs 

significantly from the present understanding, which is founded on studies of lower-bumup fuel. 

This raises questions concerning the beliefs regarding what constitutes a success path during an 

accident. Research on the behavior of high-burnup fuel under accident conditions is needed. 

With respect to LERF, the severe accident issue is not so much the determination of the risk 

metric but what constitutes an acceptable value of this metric. The LERF acceptance criterion 

discussed in Regulatory Guide 1.174 is a surrogate for the prompt fatality safety goal. This LERF 

surrogate value has been determined from level III PRA consequence analyses that have been 

based on the fission product releases in steam expected from modest-burnup fuels. 

There are a number of issues associated with the LERF risk metric that should be addressed with 

continued severe accident research. First, the current LERF acceptance values are based on the 

release of fission products in steam and do not consider the changes in the source term that might 

accompany air ingression following rupture of the reactor vessel. Air ingression is expected in 

many reactor accidents and is expected to alter the fission product release from residual fuel 

within the reactor vessel. The alterations caused by air ingression may make the releases more 

hazardous to the public than what was considered in the development of the current LERF 

acceptance criterion. 

Second, there is evidence that the release of fission products increases with increased burnup and 

that reactivity insertion accidents of lower energetics than currently evaluated in PRAs will make 

significantly higher contributions to risk. Neither of these is accounted for in the current LERF 

acceptance criterion. 

Finally, as margins are eroded and the LERF acceptance criterion is more closely approached, it 

is likely that there will be a need to determine site specific values of risk acceptance criteria. 

Such site-specific determination will require that greatly improved atmospheric dispersion 

models be available. 

0 Countercurrent Natural Circulation in the Reactor Coolant System 

Several important severe accident phenomena such as direct containment heating, steam 

generator tube rupture, and invessel fuel coolant interaction have been relegated to low frequency 

based partly on the expectation that natural convection flow will fail parts of the reactor coolant 

system and the system will depressurize. The thermal hydraulic calculations that are the basis for 

this expectation are inappropriate for two-phase natural convection flow and do not include any

51 NUREG-1635
determination of uncertainty so that probabilities can be attached to the speculated system failure 

that leads to depressurization. 

0 Electrostatic Charging and Aerosol Physics 

The consequences of severe reactor accidents are mitigated substantially by the tendency of 

aerosols of radioactive materials to agglomerate and deposit along flow pathways and in the 

reactor containment. The NRC has invested heavily in the development of technology to predict 

agglomeration and deposition of aerosols. The models that have been developed do not account 

for the real possibility that aerosols in a radiation field will be electrostatically charged. Such 

electrostatic charging will create interparticle forces that are far stronger than those that are now 

modeled and could significantly alter the tendencies for aerosols in reactor accidents to 

agglomerate and deposit. The effects of aerosol charging are not likely to affect estimates of the 

LERF risk metric. Charging could mean that fission product aerosols will remain airborne in 

containment much longer than is currently predicted. If so, this will impact the calculated 

consequences of latent fatalities and land contamination for late containment failures and may 

even make these the limiting consequences. Research is needed to better understand the potential 

for electrostatic charging of aerosols under accident conditions and the consequences of 

charging. 

0 Bypass Accident Consequences 

High risks are associated with severe reactor accidents that involve bypass of the reactor 

containment such as accidents involving the rupture of steam generator tubes. High risks are 

ascribed to these accidents because of the limited source term mitigation thought to occur along 

flow pathways in bypass accidents such as flows through the secondary sides of steam generators 

in pressurized water reactors. Models that have been used to predict such low mitigation are not 

especially sophisticated. A cooperative research program to experimentally investigate the 

potential mitigation of radioactive material releases through steam generators is being proposed 

by the Swiss. The ACRS believes that NRC should join in this collaborative program called 

ARTIST. Based on the results obtained in this program, the NRC could decide if improved 

models of source term mitigation along containment bypass flow pathways are needed in its 

accident analysis codes. 

Core Competencies 

The ACRS believes that it is essential that the NRC maintain some core competencies in the area 

of severe reactor accidents. In particular, the NRC needs competencies in the following areas: 

"* the behavior of aerosol and the chemistry of fission products under reactor accident 

conditions 

"* reactor fuel degradation under accident conditions.

NUREG- 1635 52
Competencies in these areas will be increasingly difficult for the agency to obtain by contract. 

Maintenance of the competencies will not be achieved by reliance on user needs. Continued 

participation in the international collaborative PHEBUS-FP program and its successors will 

maintain these competencies. This participation has become passive in the sense that NRC 

attends meetings and receives results of the work. It is not now actively using its models and 

codes for the analysis and planning of the tests in the PHEBUS FP program. The ACRS believes 

it important for the maintenance of core competencies that resources be made available for more 

active participation by the NRC in this program.

53 NUREG-1635
Table 7. Ongoing and Planned Severe Accident Research Programs

T -

NUREG- 1635

Task Description

K6987 Cooperative Agreement with UCSB 

W6203 MELCOR Code Development and Assessment 

W6231 MACCS Maintenance and User Support 

W6352 Probabilistic Consequence Model Methods Development 

W6758 Analytical Support for CONTAIN Code Assessment 

Y6058 OECD Lower Head Failure Program 

Y6073 Scientific Software and Data Distribution 

Y6232 Technical Support for Molten Fuel Coolant Interactions Research 

W6199 OECD RASPLAV 

planned Provide NRR with consolidated severe accidents codes 

planned Provide NRR with technical assistance in the review of the MAAP Code 

planned Provide improved severe accidents codes

54
11.8 Civil, Structural, and Seismic Research

The NRC has heavily invested in testing the capabilities of containment designs and 

understanding the seismic threats to containment and piping integrity. Much of this work has 

involved collaborative efforts with partners from abroad. Like severe accident research, many of 

the technical areas addressed by this work are not essential for the agency's current efforts to 

develop risk-informed regulatory practices, using CDF and LERF as metrics. There is not now an 

active effort to construct new plants that would require state-of-the-art capabilities to review 

structures and design capabilities though there are at least plans to examine the issues of the 

AP1000 nuclear plant containment design and more innovative containment designs that are 

being proposed. Over the next several years, however, it is likely that the pertinent areas of most 

interest for the NRC mission will involve the aging of structures, systems, and components 

(SSCs) of the existing fleet of power plants. There are needs to better understand the capabilities 

of partially degraded SSCs to withstand accident loads. Such an understanding will be important 

in the evaluation of licensee programs of aging management and to assess the efficacy of 

inspections. 

Ongoing and planned civil, structural, and seismic research programs are listed in Table 8. 

Several programs in this table allow the NRC to achieve the following objectives: 

"* leverage the benefits of a much larger investment by partners in collaborative projects 

"* keep receiving seismic data from organizations such as the US Geological Survey, and 

maintain core competencies and participate in industry activities 

These considerations are significant enough to justify some level of continued investment in this 

area over the next few years. The programs have been grouped into categories for comment here: 

Work to assess the structural capability of containments and piping components 

experiencing varying degrees of age-related degradation and to validate the analytical 

tools to resolve related regulatory issues and concerns: 

- W6042 Capacity of Aged/Degraded Containments 

- W6043 Inspection Aged/Degraded Containment 

- W6164 Validation of Degraded Methods 

- W6775 Integrity of Nuclear Piping - Structural/Material Issues 

- W6081 Japanese Collaboration on Seismic Issues 

- W6684 Assessment of Aged & Degraded Structures & Components 

This work is needed to address the immediate and future regulatory needs generated by the aging 

of the current fleet of nuclear power plants. In many cases the projects in this category leverage 

the work of international and national partners that are investing significant resources. Of 

particular current interest is the international effort (W6775) in response to cracks identified at 

Summer nuclear plant, Ringhals 4, and Biblis. The ACRS supports these research efforts, but

55 NUREG-1635
feels a comprehensive plan is needed on research concerning the aging of structures and their 

abilities to sustain accident load:s. 

"* Work needed to complete major efforts on which significant resources have already been 

expended: 

W6959 Degradation and Failure Characteristics of Nuclear Power 

Plant Containment Protective Coatings 

W6233 GV Downhole Seismic Array OP/Analysis of Data 

W6412 Garner Valley Strong Motion Study 

These programs have had value to the NRC and the work should be completed. 

"* Participation in progra-nts to test the ultimate capability of pre-stressed concrete 

containments and many seismic projects. 

Key motivations for the research in this category include: 

"* leveraging limited investnents to garner the benefits of a more significant effort being 

carried out by partners 

"* determining if results of research conducted outside of the nuclear industry can be utilized 

to improve seismic related nuclear regulation 

"* maintaining core competencies in the seismic area 

These motivations may be significant enough to justify the projects. It must be borne in mind 

that the uncertainties in both accident loads and seismic loads on containments and structures 

will be dominated by the uncertainties in the loads. We comment on the various programs in this 

category of work individually be low: 

0 Y6131 Joint Containment Model Test 

This work is part of a collaborative agreement with Japan. The objective is to test to failure the 

pre-stressed concrete containment built for this purpose by the Nuclear Power Engineering 

Corporation (NUPEC). The NRC's involvement is to evaluate the predictive capabilities of 

models used to identify failure modes and margins. It is not apparent how this work will affect 

the safety regulation of the current fleet of U.S. nuclear power plants. 

, Y6251 Collaboration Seissmic Proving Tests of Concrete Containment 

This work is part of the collaborative agreement with Japan's NUPEC/Ministry of International 

Trade and Industry (MITI). NUPEC will provide results of a large scale seismic test of a large 

prestressed concrete containment and a large reinforced concrete containment with liners. The 

NRC will verify the accuracy of current modeling techniques in design and margins to ensure 

integrity of containments for seismic loadings. The NRC investment in this program leverages a

NUREG- 1635 56
much larger investment by NUPEC/MITI. It is not apparent what current NRC need(s) the 

research addresses. The results of this work will help improve the techniques for modeling 

containments and determining their seismic margins. Certainly, the results and models will not 

affect the current fleet of plants. Technical issues are sufficiently specific to the peculiarities of 

containment designs that it is not apparent that the results of the tests and comparisons to model 

predictions will be of great value in the analysis of containment designs proposed for future 

reactors. 

0 6828 Risk/Performance-Based Response to External Events 

This program will review previous efforts by the NRC, American Society of Mechanical 

Engineers (ASME), American Society of Civil Engineers (ASCE), and the Department of Energy 

to develop risk-informed and performance-based approaches to evaluate the performance of 

piping and other structures under the loads imposed by seismic events. The objective is to risk

inform regulatory practices in external event evaluations. The work could support revisions to 

regulatory guides and sections of NRC's Standard Review Plan. These changes are intended to 

lead to more uniform margins and to provide a graded approach to design, commensurate with 

the potential risks. The ACRS views this work as appropriate to support the NRC's move to risk

informing the reactor regulations and helps the staff stay abreast of developments made by 

industry. 

0 W6691 Displacement-based Seismic Design 

Changes in seismic engineering practices outside of the nuclear industry raise questions of 

whether NRC seismic requirements in nuclear power plant design or modifications also warrant 

similar changes. For Example, are displacement criteria more appropriate than site peak ground 

acceleration for seismic design? The ACRS views this as valuable research that maintains NRC 

core compe-tencies and anticipates possible changes in design criteria that might be adopted in 

the future. 

* W6166 Cooperative Geoscience Research 

This project leverages work by the U.S. Geological Survey to improve the definition of seismic 

hazard in the central and eastern U.S. This has been an area of controversy and there are two sets 

of seismic hazard curves being used for risk analysis. Although the ACRS finds this research to 

be valuable, it is disappointed that this work is not part of a larger effort to resolve the 

differences between seismic hazard curves developed by the Lawrence Livermore National 

Laboratory and by the Electric Power Research Institute. Data generated in this program and 

work the NRC has completed on the methodology for resolving these differences could eliminate 

an area of continuing concern and discussion. 

Future work in this program will maintain cooperation with the U.S. Geological Survey. The 

work is intended to reduce uncertainties associated with propagation of earthquake ground 

motion. The ACRS view this involvement as maintaining NRC core competencies in the seismic

57 NUREG-1635
area and preventing isolation of the regulatory process from developments within the seismic 

community. 

0 Tadotsu Shaker Table 

The modest investment by the NRC yields a wealth of information from the unique experimental 

facility that the Japanese have developed. One presumes that this involvement also assures access 

by the NRC to the facility and the expertise in the event of important regulatory decisions that 

can be addressed by this experimental capability. 

* CSNI International Standard Problem 

This program allows the NRC participation in an international standard problem exercise on soil

structure interaction modeling. The ACRS is generally supportive of this type of activity for 

assessing and improving calculational capabilities including capabilities in the seismic area. The 

ACRS would be more enthused by the participation if it were part of a bigger effort to ensure 

inhouse NRC capabilities in the use of seismic analysis tools.

NUREG- 1635 58
Table 8. Ongoing and Planned Civil, Structural and Seismic Research Programs

Task Description

W6081 Japanese Collaboration on Seismic Issues 

W6251 Collaboration Seismic Proving Tests of Concrete Containment 

W6691 Displacement Based Seismic Design 

W6775 Integrity of Nuclear Piping - Structural/Material Issues 

W6959 Degradation and Failure Characteristics of NPP Containment Protective 

Coatings 

Y6131 Joint Containment Model Test 

Y6166 Cooperative Geoscience Research 

Y6233 GV Downhole Seismic Array Operation / Analysis of Data 

Y6063 Update and Maintenance of Seismic Data Analysis System 

W6684 Assessment of Aged & Degraded Structures & Components 

J6043 Inspection Aged/Degraded Containment 

Y6164 Validation of Degraded Methods 

Y6167 CSNI PWG-3 Subgroup Support 

W6828 Risk/Performance-Based Response to External Events 

planned Validate analytic tools to resolve concerns about containments that have 

experienced degradation 

planned Validate tools to resolve issues and concerns about structures that have 

experienced degradation 

planned Provide confirmatory data on the structural capacity and failure modes for a 

variety of containment designs up to and including severe accident pressures 

planned Conduct tests of degraded piping in cooperation with international partners 

planned Provide data on concrete containment and other seismic category 1 structures from the Tadotsu 

Shaker Table 

planned Collaborate with NUPEC on seismic structure engineering 

planned Reduce uncertainties associated with propagation of earthquake ground motion 

planned Examine risk-informed and performance bases approaches implemented in recent ASCE and DOE 

standards 

planned Develop the technical basis for deciding if changes in earthquake engineering practices warrant 

changes in NRC seismic requirements for NPP design or modifications 

planned Evaluate available soil-structure interactions analysis techniques

59 NUREG-1635
11.9. Research on Dizital Safety Systems

The nuclear industry is being forced to change from the past practice of using analog systems for 

instrumentation and control of nuclear power plants to the use of digital systems simply because 

replacement parts for analog systems are becoming increasingly difficult to obtain. Digital 

systems and especially software-controlled digital systems are widely regarded as providing the 

ability to operate power plants more optimally and to provide enhanced safety of power plants. 

Digital systems have greatly different operational and reliability characteristics than the analog 

systems. Whereas analog systems can process inputs and safety functions in parallel, digital 

systems must process these inputs and functions in a sequential fashion so much engineering 

attention must be paid to the timing and scheduling of algorithms used to control the digital 

systems. Digital systems do not fail typically by 'wearing out'. The unreliability of software

controlled digital systems is; dominated by system design errors. In particular, inadequate 

specification of the system requirements leads to failure of systems to function properly in 

response to all appropriate inputs. Testing does not provide high confidence that these 

vulnerabilities do not exist because, it is the unexpected inputs that will not be included in the 

testing that could lead to the failures. 

Because of the peculiarities of reliability of digital safety systems, it is not readily apparent how 

these systems are to be addressed in quantitative risk assessment methods that are so important 

for the move toward risk informed regulation, the implementation of the Maintenance Rule and 

the Revised Reactor Oversight Process. 

The NRC staff has emphasized control of the design and development processes rather than 

testing of the products of the processes to ensure reliability of digital systems. Unfortunately, the 

standards for controlling the design and development process that are available to the nuclear 

industry have been developed for the larger digital electronic community and are not focused on 

the peculiar and usually simpler albeit higher reliability requirements of the nuclear safety 

community. Furthermore, the nuclear community is not so large that it can greatly influence the 

digital systems community with respect to hardware and software development. The nuclear 

industry is often forced to use: hardware and software developed for other markets (so-called 

"Commercial Off-the-Shelf" or COTS systems) for which design and development have not been 

strictly controlled to the specifications of the standards. Consequently, review of digital safety 

systems for nuclear power plants is a time consuming process. It is for this reason NRR has 

asked RES to conduct research that will lead to: 

* Improved methods and tDols for the review of digital systems 

* Risk and reliability models for digital instrumentation and control systems 

* Regulatory guidance for emerging technology in the digital field

NUREG-1635 60
RES has produced a rather comprehensive plan for research on digital systems through FY2004. 

The major elements of this program fall in the following four categories (Table 9 shows more 

detailed breakdown of the research programs). 

* Systems Aspects of Digital Technology 

This program element is to address 1) environmental stressors such as tempera-ture, humidity, 

smoke, lightning, and electromagnetic interference to digital systems, 2) digital requirement 

specifications on the functions of digital instrumentation and control systems and the interfaces 

of these systems with other plant systems, 3) diagnostics and fault tolerance features that enable 

digital systems to detect and report internal problems and either avoid or handle these problems, 

but greatly complicate the systems, and 4) operating systems that control the basic functions 

of digital instrumentation and control systems. 

0 Software Quality Assurance 

This program element is to develop objective software engineering criteria that provide a 

measurable acceptance level for software quality for use in the regulatory process. Criteria for 

software testing are to be defined as part of the software quality assessment. 

* Risk Assessment of Digital Instrumentation and Control Systems 

This program element involves 1) analysis of digital instrumentation and control system failure 

data, 2) examination of digital failure assessment methods used by the defense and aerospace 

industries, 3) identification of the risk importance of digital instrumentation and control systems, 

and 4) development or adaptation of digital reliability assessment methods. 

0 Emerging Instrumentation and Control Technology and Applications 

This program element involves preparing the staff to consider technological improvements that 

the nuclear industry may adopt, including 1) online equipment monitoring systems to determine 

what types of maintenance is needed and when, 2) advance instrumentation for measuring flow, 

temperature, pressure, and neutron flux that might serve as support for power uprates, 3) "smart" 

transmitters that can compensate for measurement error or alter sensor functionality, 4) wireless 

transmission of data from sensors to operators, and 5) "firewalls" to bar accessing computer 

systems and corrupting or degrading the performance of computer systems. 

RES is to be congratulated for preparing such a detailed plan of its activities over the next several 

years. The plan makes transparent the precise products and uses that are to come from the 

research. The ACRS offers the following comments on the proposed research programs: 

Overall, the planned work appears to be complete with regard to the digital safety systems 

issues of concern to the regulatory system.

61 NUREG-1635
The ACRS also supports those aspects of the plan intended to provide improvements to 

the efficiency and effectiveness of digital safety system reviews by line organizations. If, 

indeed, the prognostication of improved safety from digital systems are to be had, it is 

imperative that the regulatory process and the reviews it involves not constitute a 

significant barrier to the adoption of these systems. 

The examinations of environmental stressors on digital electronic systems focuses on the 

issues of electromagnetic interferences and gives scant attention to other stressors cited in 

the plan as possibly affecting the performance of digital systems. Most notably, the plan 

does not address the issue of deposits of smoke from fires on digital system performance 

especially long after the ýfire has occurred. 

"* The ACRS endorses the staff's questioning of the efficacy of fault tolerant methods and 

self diagnostics (Program Element 3.2.5) that can be incorporated into digital safety 

systems. The ACRS is concerned that complexity may make digital systems more 

susceptible to errors caused by rare events. The increased complexity brought on by the 

proliferation of features in digital systems is usually done: 1) to improve the utility's plant 

operations, and 2) to continuously assess the validity of signals being sent to safety 

systems. The staff should find means to assure that such performance enhancing systems 

are isolated from digital safety systems to eliminate interactions that might degrade safety. 

"* The ACRS is pleased by the attention given in the plan to the possibility of removing 

conservatisms that inhibit the effective use of COTS systems in digital safety systems. 

"* It is apparent from the plan that the staff is attempting to garner advantage by examining 

broadly data on digital systems failures experienced within and outside the nuclear 

community. 

"* The plan expresses a willingness to develop techniques for qualitative rather than 

quantitative risk assessment of digital systems and their controlling software. The ACRS 

believes that to be most useful to the regulatory process the risk analyses must be 

quantitative. Techniques the staff develops or adapts must be able to be incorporated into 

the existing PRA technology. It is evident that the staff is struggling in this area. They are 

certainly -searching for existing methods that can be applied to the particular issues 

confronting NRC. This might be an area where the NRC should make a broader 

solicitation of ideas, even to the extent of modestly funding multiple institutions and 

researchers for the competitive development of ideas and approaches, rather than using the 

approach toward contracting research. 

"* The ACRS agrees that it: is of some importance for the research staff to stay abreast of 

technological developments in the field of digital instrumentation and control that are 

likely to be applied to aspects of nuclear power systems that are regulated by the NRC. 

The staff cannot be expected to stay abreast of all aspects of so dynamic a field as digital 

systems. This activity should be kept specific to advances that are known to be pertinent

NUREG-1635 62
to the nuclear community. The topics listed in the plan, predictive maintenance (3.5.2), 

advance instrumentation (3.5.3), smart transmitters (3.5.4), and wireless communications 

(3.5.5) do appear to be such pertinent technologies that the NRC research office should 

stay aware of. 

The plan needs to specify in greater detail what unique technical information the staff will 

derive from its participation in the Halden project and how this information contributes to 

the goals of its research and the missions of the NRC. 

During its review of Generic Safety Issue (GSI)-168, the ACRS was shown several low

voltage cables that had been artificially aged to the equivalent of 40 years of operation and 

then subjected to an environment expected to exist inside containment following a LOCA. 

The deterioration of the insulation on some cables was quite serious whereas on other 

cables the deterioration was only modest. Some experts attribute the differences in cable 

deterioration to differences in the insulation. Others question the use of Arrhenius 

activation energy as the basis for artificial aging of the cables. Since the behavior of 

electrical cables in the age range of 40 to 60 years is an important consideration in the 

license renewal process, better understanding of the issues of cable qualification would 

seem important. The ACRS is pleased to hear that the NRC is participating in an 

Interagency working group on cable insulation. The Department of Energy, in cooperation 

with the Electric Power Research Institute, has undertaken a research program dealing 

with condition monitoring of both low- and medium-voltage cables. The NRC should 

establish a connection with this research. Furthermore, the issues of cable aging are of 

sufficient interest throughout the worldwide nuclear power reactor community that it may 

be possible to establish international cooperative research activities in this area.

63 NUREG-1635
Table 9. Ongoing and Plainned Research on Digital Systems*

"* Systems Aspects of Digital Technology 

3.2.1 Verify EMI/RF[ Qualification Levels 

3.2.2 Complete Environmental Qualification Guidelines 

3.2.3 Lightning Protection Guidelines 

3.2.4 Investigate Requirements Specification Assessment Methods 

3.2.5 Diagnostics ard. Fault Tolerant Techniques 

3.2.6 Operating Systems 

"* Software Quality Assurance 

3.3.1 Investigate Software Engineering Practices and Measures 

3.3.2 Investigate Criteria for Software Testing 

"* Risk Assessment of Digital Instrumentation and Control Systems 

3.4.1 Perform Data Analysis on Digital Instrumentation and Control 

Failures 

3.4.2 Investigate for Digital Failure Assessment Methods 

3.4.3 Identify the Risk Importance of Digital Instrumentation and 

Control Systems 

3.4.4 Investigate Digital Reliability Assessment Methods 

"• Emerging Instrumentation and Control Technology and Applications 

3.5.1 Review of Future Applications for Digital Instrumentation and 

Control and Research Infrastructure 

3.5.2 Predictive Majintenance/Online Monitoring 

3.5.3 Advanced Instrumentation 

3.5.4 Smart Transmitters 

3.5.5 Wireless Communications 

3.5.6 Firewalls 

*The project numbering in this table comes from the excellent plan developed by the staff for the research 

entitled "NRC Research plan - D]igital Instrumentation and Control".

NUREG-1635 64
11.10. Spent Fuel Storage and Decommissioning Research

It is evident that it will be some time before a permanent repository for spent nuclear fuel is 

available to relieve licensees of the continuing burden of storing spent fuel on individual reactor 

sites. The NRC and the industry have been actively addressing the limited capacity of spent fuel 

storage pools and have developed the capabilities to store fuel on site in dry casks. These dry 

cask storage systems have been designed and are regulated based on the best available data. The 

available database, however, is not extensive. The NRC staff has examined the issues of 

hydrogen generation, corrosion, and fuel decrepitation, but further work is still needed. The 

prolonged duration of spent fuel storage on individual reactor sites that can now be anticipated 

means that these spent fuel systems can pose safety issues for a very long time. 

Spent fuel storage pools themselves have become issues especially for plants that are undergoing 

decommissioning. Although there is now great enthusiasm within the nuclear community over 

the successful license extension process that the NRC has developed, there is potential that, in 

the near future, several plants may elect to undergo decommissioning and, of course, all plants 

eventually will have to be decommissioned. Episodically, there have been recommendations that 

the NRC's regulations for decommissioning be reexamined and made more coherent. Today, 

such a reexamination should be done with an eye toward making the regulations more risk 

informed. Until the spent reactor fuel can be removed from the site, the dominant risk during 

plant decommissioning comes from accidents involving the spent fuel storage pool. The exact 

magnitude of this risk is being assessed by the NRC. 

Clearly, there are areas within the general topics of dry cask storage of spent fuel and the 

decommissioning of reactors where research would be of significant aid to the regulatory 

process. The planned and ongoing research programs in these areas are listed in Table 10. 

Comments about the individual programs are presented below: 

* Spent Fuel Storage Fees 

This is a legacy requirement imposed on the NRC involving the continued storage of fuel used in 

reactor safety research tests conducted at the Idaho National Engineering and Environmental 

Laboratory. There is, of course, no current options available for relief from this continuing 

charge against research resources. 

"* Seismic Behavior of Spent Fuel Storage Cask Systems 

"* Seismic Criteria for independent spent fuel storage installation 

The objective of this program is to provide the technical basis for the evaluation of seismic safety 

of a dry cask storage system. It is a cooperative program with the Electric Power Research 

Institute and there is the potential for collaboration with Taiwan and Japan. 

The seismic issues of spent fuel casks ought to be fairly easily resolved with existing analytical 

techniques. Experimental verification of the analyses will be more challenging. The ACRS

65 NUREG-1635
would be more enthusiastic about this program if it were a part of an overall effort to establish 

the risk associated with dry cask fuel storage on reactor sites. Such an overall effort would 

require not only the understanding of the seismic stability of the fuel cask systems, but also the 

understanding of the consequences of rupture of the fuel casks. Such ruptures could well involve 

the release of radionuclides in manners different from those made familiar by the analysis of 

reactor accidents. Of particular concern would be the ignition of cladding as a result of air 

interactions with zirconium hydrides in clad that had been broken and fragmented in a seismic 

event. There is great similarity between these concerns and the risk concerns associated with 

transportation accidents involving high-burnup fuel. There are clear opportunities for 

coordination of research efforts, as further discussed below. 

"* Dry Cask License Renewal Criteria 

"* Characterize condition and material behavior of spent fuel and safety related components 

in dry cask storage systems 

"* Support for resolution. of fuel integrity issues with dry cask 

storage systems 

"* Demonstrate the performance of a cask loaded with spent, high 

burnup fuel 

"* Examinations of CASTOR-V/21 Fuel 

These seem to be important programs to establish data on the performance of spent fuel storage 

systems. In the tasks, the experience and the material behavior of spent fuel storage systems at 

Department of Energy sites will be gathered and evaluated for its implications concerning spent 

fuel storage systems at comlnuercial nuclear power plant sites. Also, in a joint program with the 

Electric Power Research Institute, and the Department of Energy, the NRC will examine 

specimens of fuel that have been housed in the CASTOR-V/21 fuel for about 15 years. 

Information derived from these examinations will provide technical foundations for license 

renewal of dry cask storage systems. One of the most exciting elements of these programs is the 

instrumentation and monitoring of an actual cask loaded with high-bumup fuel. This could be 

exceptionally useful source of real data on the behavior of the system. The staff faces two 

challenges in conducting this activity. The first is defining the instrumentation that will be 

applied to detect phenomena and processes that may not be readily anticipated today. The second 

is, of course, maintaining the level of interest in the effort for the many years that it will have to 

be continued -to gather the kind of data needed for the regulatory process. The ACRS is 

supportive of these initiatives that the staff has undertaken to establish a good technical 

foundation for licensing dry cask storage systems: 

"* Performance of spent fael transport casks during severe 

accidents 

"* Confirmatory testing for spent fuel transport casks under 

severe accident conditions 

Evaluate source terms; i:or risk assessment of dry cask storage 

and transportation accidents

NUREG- 1635 66
The NRC methods for evaluating risks associated with transportation accidents involving fuel 

casks are now rather old. They were developed at a time that quantitative assessment of risk was 

not a pervasive philosophy at the NRC. The assessments of fuel damage during an accident were 

for fuel that had not experienced the levels of burnup that are now becoming common so that 

cladding on the fuel was not as embrittled as it is now. Source terms were estimated in quite 

conservative and bounding fashions. Studies that the NRC has completed have indicated that 

there are substantial opportunities to improve these risk assessments substantially. The ACRS 

finds that there is merit in these proposed research activities, but cautions that substantial 

resources will be required to make improvements in area where the risks to the public health and 

safety are probably not at all large. The cost-benefit tradeoff of these studies should be examined 

with some care before undertaking what could be a substantial effort. 

Data and models for assessment of public exposure to environmental releases of 

radioactive materials from site decommissioning. Develop tools for estimates of 

radionuclide transport for decommissioning scenarios 

The ACRS' enthusiasm for these research programs is tempered by the fact that the programs do 

not seem to be part of a larger effort to define the risks associated with decommissioning and the 

fact that with the possible exception of accident scenarios involving the spent fuel pool, source 

terms from decommissioning events will be small and are not likely to pose significant risks to 

the public. By themselves, the programs do not seem to hold the promise of benefits 

commensurate with their costs. If the programs could be viewed within the context of a larger 

effort to determine the risks of decommissioning and make the NRC regulations on 

decommissioning risk informed, they might be more readily justified. In any event, these 

activities would seem to have some features in common with ongoing efforts to address 

uncertainties and improve the MACCS code (see the discussion of Severe Accident Research in 

Section 11-7).

67 NUREG-1635
Table 10. Ongoing and Planned Research Programs in Spent 

Fuel Storage and Plant Decommissioning

Task Description

A6893 Spent Fuel Storage Fees 

W6829 Seismic Behavior of Spent Fuel Storage Cask Systems 

Y6038 Dry Cask License Renewal Criteria 

Y6248 Examinations of CASTOR-V/21 Fuel 

Y6301 Human Reliability Analysis of Dry Cask Storage Activities 

planned Data and models for assessment of public exposure to 

environmental releases of radioactive materials from site 

decommissioning 

planned Characterize condition and material behavior of spent fuel 

and safety related components in Dry Cask Storage Systems 

planned Demonstrate performance of a cask loaded with spent, high 

burnup fuel 

planned Evaluate source terms for risk assessment of dry cask 

storage and ixansportation accidents 

planned Develop tools to support license termination decisions 

planned Develop tools for estimates of radionuclide transport for 

decommissioning scenarios 

planned Seismic criteria for independent spent fuel storage installation 

planned Support for resolution of fuel integrity issues with dry cask 

storage systems 

planned Performance of spent fuel transport casks during severe 

accidents 

planned Confirmatory testing for spent fuel transport casks under 

severe accident conditions

NUREG- 1635 68
II.11. Research in Criticality Safety

Criticality safety is the name given to the arcane aspect of nuclear safety concerned with 

criticality operation with nuclear materials outside the configuration of a reactor assembly. 

Activities in refining, enriching, blending, and fabricating fuels for reactors; shipping and storing 

these fuels prior to exposure in a reactor and the handling; storing, and disposing of spent fuels 

after use all fall in the realm of criticality safety. In assessing the criticality safety of these 

operations, heavy reliance is placed on standards issued under the aegis of the American National 

Standards Institute (ANSI). These standards rely on safe limits on fissile material concentration, 

total material inventory, and container configuration to guard against criticality. Supplemental 

use of neutron absorbers and restriction on the availability of moderating materials also plays a 

role in defining a criticality safe material configuration. The use of these standards along with the 

so-called double contingency principle (the requirement that an acceptable design is judged to be 

safe base on independently satisfying two of the safe limits) has proved to be effective. 

There is an important point here. Fatal exposures to radiation from fissile materials have 

occurred in operation with nuclear materials in the USA. These took place at Department of 

Energy facilities and involved operation with higher enrichment uranium or plutonium. An 

important qualifier may be placed on these events after careful examination of the circumstances 

under which they took place. In every case where a fatality occurred, an unusual activity outside 

the scope of the design of the facility was being attempted. The most common activity of this 

type was the recovery of high value scrap from fabrication processes. The ad hoc use of 

equipment to conduct operation without design review has been forbidden form many years now, 

and there have been no further incidences involving fatalities. Furthermore, a review of more 

recent criticality events in other countries supports this assessment. When fatalities have 

occurred, the ANSI Standards were not followed and the requirement for independent safe 

assessment of new operation was not met. 

It must be recognized that for all of the success in using the standards

based approach to criticality safety, the resulting products (designs, procedures, etc.) are quite 

conservative, and this has meant that they are expensive. Process equipment, storage facilities, 

and criticality alarm systems designs have all had substantial margin. Shipping rules have also 

been conservative. 

The current criticality safety assessment methods do not have the precision necessary to allow a 

defensible comparative evaluation of risk to be made to support a choice between competing 

alternative approaches to safety. They also do not meet the needs of a risk-informed regulatory 

system. It may well be that the approach to criticality safety should remain deterministic. 

Unfortunately, there are occasions where a risk-informed assessment could be useful. An 

example is in the resolving a dispute between NRC requirements and requirements of the 

International Atomic Energy Agency (IAEA) on spent fuel burn-up credit. The NRC regulations 

impose significantly less burden on the licensees than do the IAEA requirements. Resolution of 

this dispute could be a worthwhile goal of NRC research efforts.

69 NUREG-1635
A development that will affict the NRC involvement in criticality safety is the interest in. using weapons-grade plutonium blended with uranium to make a MOX fuel for use in light-water 

reactors. Initial efforts are now under way to acquire a batch of the weapons grade plutonium from the former Soviet Union for demonstration of the concept using a light-water reactor 

operated by Duke Power. A review of the division of safety responsibilities between the NRC 

and the Department of Energy in conducting this demonstration has led to the decision to have 

material blending and fabrication done under license from the NRC and compliance with NRC 

regulations. The blending operation can pose a range of criticality problems, as a batch of 

plutonium is fissionable wit:h fast neutrons without a moderator present. Indeed, a subcritical 

mass of unmoderated plutonium can become highly supercritical if it is moderated by accident. 

The sensitivity of a mixture of uranium and plutonium to the energies of the neutrons will vary as 

blending progresses. A competent criticality safety assessment of the blending process is needed. 

The NRC must assess this analysis and may require some experimental verifications with 

intermediate mixtures that arise in the blending process, While there is no doubt that pressures will be imposed on the staff to review this assessment expeditiously, it is imperative that the 

review be done correctly. NRC experience with plutonium is limited. 

With this background, comments are -presented below on the specific tasks in the NRC's 

criticality research program (See Table 11). 

0 Development of Criticality Safety Data for Licensing Review 

As the availability of experimental facilities for criticality studies is reduced due to concerns 

about operating costs, it is prudent to assess the available data for completeness in addressing the 

parameter ranges of current and anticipated interests. It is also useful to test the abilities of 

computer codes to predict experimental results. Some latitude in defining the experimental data 

sets that might be needed for these assessments must be accepted. 

Modernization of the codes to use the most recent ENDFB cross section files is an important part of this task. This is Version 6 of these cross-sections, the same set that the NRC has found to be 

most reliable in predicting neutron attenuation of pressure vessel walls. It is reassuring to see the 

emergence of demonstrated generality in the application of a cross section set to a range of 

different problem areas. 

The task includes an assessment of U.S. data and analysis methods in the resolution of the 

differences in the NRC and IAEA regulations on allowable bum-up credit. This effort is central 

to maintaining a capability to address power reactor related concerns in criticality safety. This 

effort also provides NRC access to the capabilities at Los Alamos National Laboratory to address 

plutonium criticality safety assessment that will be so important to the resolution of fuel blending 

concerns discussed above. 

The ACRS is supportive of this proposed activity. Some greater detail in specifying the 

individual tasks and the need for the expected results of the activities would be appropriate. A 

strategy by which maintenance of criticality codes becomes an in-house activity of NRC should

NUREG-1635 70
be considered. The value of these activities in future years rests heavily on decisions concerning 

use of standards-based regulation versus risk-informed regulation in the area of criticality safety. 

* Preservation of Criticality Benchmark Data 

This task is to collect and archive the documentation of criticality experiments conducted at a 

number of national laboratories. The data are to be stored at the Criticality Safety Information 

Resource Center at Los Alamos National Laboratory. Many of the experimenters are now retired 

and there would be great value in upgrading some of the less expansive entries in the old log 

books. The ACRS is supportive of the modest contribution NRC makes to this overall effort. 

* REBUS Experimental Program 

This program is an international effort involving reactivity tests of selected light water fuel 

bundles for direct evaluation of the burn-up credit. The program thus directly addresses the 

issues concerning bum-up credit between the NRC and the IAEA. The ACRS supports this 

program. 

Table 11. Research Programs in Criticality Safety

NUREG-1635

Task Description

W6479 Development of Criticality Safety Data for Licensing Review 

Y6035 Preservation of Criticality Benchmark Data 

Y6225 REBUS Experimental Program 

planned Conduct studies on materials criticality safety to preserve criticality 

benchmark data and provide a methodology for use of codes and update 

codes

71
11. 12. Radiation Health Effects

Ongoing and planned research activities in the area of health effects of exposure to radiation are 

listed in Table 12. The first cf these, the BEIR VII program, is a continuation of the National 

Research Council's examination of the stochastic health effects of low doses of ionizing 

radiation. For some time now, there have been scientists who argued against the linear, no 

threshold (LNT) hypothesis for extrapolating the stochastic effects of radiation from higher doses 

where the effects are measurable, but still stochastic, to low doses where there are not reliable 

data. They have based these arguments on epidemiological studies and analogies to the toxic 

effects of chemicals. Unfortunately, the epidemiological studies lack statistical power. C. Lamb 

of the National Cancer Institute has presented persuasive statistical arguments that the sample 

sizes for meaningful epidemiological studies to resolve controversies concerning the LNT 

hypothesis are prohibitively large. 

Phase 11 of the BEIR VII study will reexamine the database available to support the LNT 

hypothesis for low linear energy transfer (LET) radiation (gamma and beta rays) and, perhaps, 

refine our understanding of this hypothesis. (Data supporting the LNT hypothesis for high LET 

radiation and neutrons is more persuasive.) It will not end the controversy. 

The technical basis of the LNT hypothesis is a biological model concerning the reliability of 

repair mechanisms for damage caused by radiation. Resolution of the debate concerning the LNT 

hypothesis will be possible only after those opposed to the LNT hypothesis formulate their 

concepts in terms of a biological model that can be tested with cells and animals. The 

considerable advances in instrumentation being sponsored by the Department of Energy offer the 

hope that definitive testing of alternative models will be possible in the coming years. 

Several of the programs deal with the collection of occupational exposure data. It seems to be an 

obligation of NRC to stay abreast of these data. 

A most interesting program (Y6112) is the adaption of the RESRAD and RESRAD-BUILD 

codes by the NRC for site-specific probabilistic dose assessments. RESRAD and RESRAD

BUILD are widely respected computer codes developed by the Department of Energy. The NRC 

use of these codes seems to be a meaningful step toward the greater use of risk information in 

dose assessments. Results and insights obtained from the work with these codes may well have 

some importance to the further development of NRC's MACCS code for accident consequence 

analysis (See Section 11.5). 

One of the programs (G6251) provides support to the work of the International Committee on 

Radiation Protection (ICRP). The ACRS is surprised that there are not efforts focused on 

determining if NRC should upgrade its radiation protection standards from ICRP30 to ICRP60. 

The differences between these standards could well be a point of contention in the licensing of a 

facility for the fabrication of mixed plutonium oxide - uranium dioxide fuel (MOX) assemblies 

for disposal of some of the world's excess supply of weapons-grade plutonium dioxide. The dose 

estimates for Pu 239 in ICRP30 are more conservative than the estimates that come from the use of

NUREG-1635 72
methods described in ICRP60. Indeed, ACRS expects that the entire accident analysis for the 

MOX fuel fabrication facility will require the research delineated in Table 12 to put consequence 

analyses for inhalation of Pu239 on a firmer technical foundation especially when the Pu 23 9 is 

mixed with uranium dioxide. 

Table 12. Ongoing and Planned Research on Radiation Health Effects

NUREG-1635

Task Description

G6071 BEIR-VII 

G6143 NEA Information System on Occupational Exposure 

W6801 JCCRER Project 2.3 - Deterministic Effects of Occupational Exposure to 

Radiation 

Y6112 Default Parameters for RESRAD and RESRAD-BUILT Computer Codes 

Y6133 Collection and Analysis of Occupational Radiation Exposure Data 

G6251 Radiation Protection Standards Development 

planned Develop technical basis to support rulemaking in the area of occupational 

radiation protection 

planned Validate current health effects models 

planned Develop and maintain analytic tools for health effects applications 

planned Develop dose assessment for clearance of materials 

planned Develop radiation protection standards and guidance for use of byproduct 

materials

73
11.13. Analysis and Evaluation of Operational Data

From the start of operation of licensed power reactors, value was ascribed to the review of 

accumulated experience and assessing the lessons to be learned over the broad range of designs and operations, including human performance. These assessments are conducted by the NRC and 

the industry, and have proven to be very useful both in the design of the regulatory system and 

the evaluation of its effectiveness. 

The Data Collection and Analysis staff reviews the experience from nuclear power plant operations as an essential feature of the effort to protect the health and safety of the public. Plants 

from the entire fleet of operat;ing plants in the country are monitored to identify vulnerabilities as 

they develop with time. Using- consistent statistical criteria, potentially significant events can be 

detected rapidly and common problems identified. The assessment of the root causes of events 

can identify plant system performance, human performance, and externally induced events that 

warrant special attention by the regulators. Indeed, operating events sometimes disclose 

unanticipated interactions, phenomena, and behavior. The objective is to develop strategies to 

avoid event initiation and to mitigate event consequences. An important activity that has 

impressed the ACRS, recently, has been the effort to evaluate the effectiveness of rules such as 

the Station Blackout rule and the rule concerning anticipated transients without scram. 

Until recently, the activities in this area were the responsibility of the NRC's Office for Analysis and Evaluation of Operational Data (AEOD). The Office's responsibilities were reassigned to the 

RES. At the time of this reassignment, the ACRS was concerned for the long-term commitment 

to continue these activities, aand the possibility that those doing the data evaluation may lose 

some of the independence that they enjoyed in AEOD. The ACRS concerns have not been 

realized, and the work in this; area continues at the high standards that it had in the past. 

Activities in this program area are carried out in two branches. The Operating Experience Risk 

Analysis Branch (OERAB) conducts activities involving the assessment of operational data, analysis of data on an industry-wide basis, plant -specific event analyses, and identification and 

evaluation of risk-based performance indicators that may eventually replace those now in use in 

the NRC's Revised Reactor Oversight Process. The elements of this range of activities are shown 

in Figure 1. The Branch has planning that extends well beyond the current budgeting cycle that 

reflects a comprehensive view of its mission. 

The Regulatory Effectiveness and Human Factors Branch examines accumulated experience to 

address specific questions on plant behavior. Some of the recent and ongoing activities of this 

branch are listed in Table 13. The Branch is now considering reviews of human performance during steam generator tube rupture events and a review of unnecessary paperwork oriented rules 

and requirements. The later of these has the potential of providing the basis for ameliorating substantially burdens on both licensees and the NRC staff. The ACRS regularly reviews the 

products of the two branches involved in the analysis and evaluation of operational data. The 

ACRS is enthusiastic about the programs now under way in these branches and their plans for the 

future.

NUREG-1635 74
Table 13. Past, Ongoing and Planned Activities in the 

Regulatory Effectiveness and Human Factors Branch 

Routine Events Assessment Activities: 

Diablo Canyon electrical bus fire and Loss of Onsite Power event 

Arkansas Nuclear RHR pump common mode bearing failure 

Hatch scram and failure to open of safety relief valves 

Indian Point steam generator tube rupture 

Long-Term Studies: 

Evaluation of Air-operated Valves 

Scram occurring during surveillance and maintenance activities 

Regulatory effectiveness of the ATWS rule 

Regulatory effectiveness of the station blackout rule 

Causes and significance of design basis errors 

Ongoing and Planned Studies: 

Audit of the plant specific backfit process 

Regulatory effectiveness of Appendix J - Containment Leakage 

Electrical system performance following a reactor trip 

Latent failures in human performance 

Topics Under Consideration: 

Human performance during steam generator tube rupture events 

Unnecessary paperwork oriented rules and requirements

75 NUREG-1635
1490 

ieuogwendo

" ""WOleo~pfl 

L *Xwu~d*mwdinbaI 

1.T

IJOW2IPUI SO~muqnw 

"B's) ewoimp~

-1

L ------------ T -

sOsAIBuV 4uaA3 I L

lubUhdOASP JQPunl sessooAwd pus soumqqu a........-------..... 

1 1 .p~~oolpqeeaa*ýu$W 

Xl~al SUhI6$:euo lildPS IO) .uww *obs 

I- ---- - -

4 

*" spom moum IEYm nw

WoawdolaAaa 

Id9GI

cq

L ----------------------------- - - -----------------

S3UWALOV UNV SYVUDJO~1d 8VU30 - z 3unowL
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, Advisory Committee on Reactor Safeguards, 

NUREG-1635, Vol. 2, Review and Evaluation of the Nuclear Regulatory Commission 

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

1999. 

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

NUREG-1635, Vol. 3, "Review and Evaluation of the Nuclear Regulatory Commission 

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

2000. 

4. U. Schmocker, M. Khatib-Rahbar, E.G. Cazzoli, and A. Kuritzky, "An assessment of the 

risk impact of reactor power upgrade for a BWR-6 Mark -11I plant," in: Probabilistic 

Safety Assessment and Management, P.C. Cacciabue and I.A. Papazoglou, Eds., pp. 

1842-1847, Springer, London, 1996, (Proceedings of ESREL '96 and PSAM III). 

5. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, "An Approach for Using 

Probabilistic Risk Assessment in Risk-informed Decisions on Plant-Specific Changes to 

the Licensing Basis," July 1998. 

6. Report from D. A. Powers, Chairman, ACRS, to Richard A. Meserve, Chairman, NRC, 

Subject: "Importance Measures Derived from Probabilistic Risk Assessments," February 

11,2000. 

7. Report from R. L. Seale, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC, 

Subject: Human Performance Program Plan, February 13, 1997. 

8. Report from R. L. Seale, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC, 

Subject: Proposed Final Draft of the NRC's Human Performance Plan, June 12, 1998. 

9. Letter from D. A. Powers, Chairman, ACRS, to William D. Travers, EDO, NRC, Subject: 

SECY-98-244, NRC Human Performance Plan, February 19, 1999. 

10. "Summary of INEEL Findings on Human Performance During Operating Events," Report 

No. CCN 00-5421, transmitter by letter dated February 29, 2000.

77 NUREG- 1635
11. Memorandum dated March 28, 1997, from John C. Hoyle, Secretary of the Commission, 

to L. Joseph Callan, Executive Director for Operations, NRC, Subject: Staff 

Requirements-COMSECY-96-066 Research (Direction Setting Issue 22). 

12. Staff Requirements Memorandum dated April 15, 1997, from Annette L. Vietti-Cook, 

Acting Secretary of NRC, to L. Joseph Callan, Executive Director for Operations, NRC, 

Subject: COMSECY-96-061 - Risk-informed, Performance-Based Regulation, April 15, 

1997. 

13. National Fire Protection Association (NFPA) 805, "PerformanceBased Standard for Fire 

Protection for Light Water Reactor Electric Generating Plants," 2001 Edition. 

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

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

15. Committee on Nuclear Safety Research, Revitalizing Nuclear Safety Research," National 

Academy Press, Washington, DC, 1986.

NUREG-1635 78
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-'9) (Assigned by NRC, Add Vol., Supp., Rev., 

NRCM 1102. and Addendum Numbers, if any.) 

3201,3202 BIBLIOGRAPHIC DATA SHEET 

(See instructions on the reverse) 

2. TITLE AND SUBTITLE NUREG-1 635, Volume 4 

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 Commission May 2001 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

Annual 

7. PERIOD COVERED (Inclusive Dates) 

Fiscal Year 2001 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. 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 ADDREISSL (If NRC, type 'Same as above'; ircontractor, provide vR u Lo r'eg, u.O. '17 rr,=1.4UlOiU0_y n, 

and mailing address.)

Same as above

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less) 

This report to the U.S. Nuclear Regulatory Commission (NRC) presents the findings and recommendations that emerged from 

an assessment conducted from March 2000 through March 2001 by the NRC's Advisory Committee on Reactor Safeguards 

(ACRS). 

In conducting this assessment, the ACRS sought to fulfill the following objectives: 

Assess the ongoing and planned reactor safety research being carried out by the Office of 

Nuclear Regulatory Research (RES). 

Recommend additional research that will be needed to promote the efficient and effective 

execution of the agency's mission in the future. 

The attentions of the report focuses on the research addressing issues of power reactor safety. Research dealing with the 

permanent disposal of radioactive waste in geological repositories will be discussed separately in a report by the NRC's 

Advisory Committee on Nuclear Waste (ACNW).

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT 

Unlimited 

14. SECURITY CLASSIFICATION 

Nuclear Reactors Safety Engineering (This Page) 

Nuclear Reactor Safety Safety Research Unclassified 

Reactor Operations (This Report) 

Unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (2-89)
Federal Recycling Program
NUREG-1635, Vol. 4 REVIEW AND EVALUATION OF THE NUCLEAR REGULATORY COMMISSION 

SAFETY RESEARCH PROGRAM

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, DC 20555-0001 

OFFICIAL BUSINESS 

PENALTY FOR PRIVATE USE, $300

MAY #0