ML20009H321

Comments on NRC Safety Research Program Budget for Fiscal Year 1983

ML20009H321
Person / Time
Issue date:	07/31/1981
From:	Advisory Committee on Reactor Safeguards
To:
References
NUREG-0795, NUREG-795, NUDOCS 8108070261
Download: ML20009H321 (53)

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NUREG-0795 4

i Comments on the NRC Safety Research Program Budget

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NUREG-0795 Comments on the NRC Safety Research Program Budget for Fiscal Year 1983 atoYu ished uly 1

Advisory Committee on Reactr Safeguards U.S. Nuclear Regulatory Commission Washington, D.C. 20555 y-.,

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4 UNITED STATES 8

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NUCLEAR REGULATORY COMMISSION

_,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555

%,*****/g July 17,1981 The Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Commission Washington, DC 20555

Dear Dr. Palladino:

The Advisory Committee on Reactor Safeguards is pleased to transmit its comments on the Office of Nuclear Regulatory Research Budget for FY 1983.

Only that portion of the budget relating to Program Support has been considered. No attempt has been made to distinguish between Program Support Funds for research and for work related to standards devel-opment, since the latter are a relatively small proportion of the total.

The proposed funding levels considered are those recommended by the Executive Director for Operations for consideration by the Commission.

Sincerely, J. Carson Mark Chairman

t-TABLE OF CONTENTS i

t-l Page PAR 1 1: ' GENERAL C0MMENTS....................................,....

1 1.

I n tro d uc ti o n........................................

3 2.

Previous Rec ommendati on s............................

3 3.

De si g n-Rel a te d Re sea rch...........................

6 4.

Bu dge t Rec ommen da tion s.............................

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_ 5.

Speci fic Comments and Recommendations...............

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TABLE 1 - Office of Nuclear Regulatory Research Program Support Budget For FY 1983....................

9 PART II:

SPECIFIC C0MMENTS.......................................

11-1.

LOC A AN D TRAN S I ENT RESEARCH.................................

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l 1.1 I n troduc ti o n...........................................

13 1.2 Semisca1e..............................................

13 1.3 Separate Effects Experiments and Model Development.....

13 1.4 3-D Program...................

14 1.5 Code Improvement ar.d Maintenance.......................

15 1.6 Code As sessment and Appl ication........................

15 1.7 Fuel Behavior Under Operational Transien ts.............

16 1.8 Rec omme nd a ti o n s........................................

16 2.

L0FT........................................................

17 2.1 I n tre du c ti on...........................................

17 2.2 The LOFT Test Program..................................

17 2.3

' ac ommend a ti o n s........................................

17 3.

ACCIDENT EVALUATION AND MITIGATION..........................

18 3.1 I n tr o d uc t i o n...........................................

18 3.2 Behavior of Damaged Fue1...............................

19 3.3 Fu el Mel t B e h a v i o r.....................................

19 3.4 Fi ssion Product Rel ease and Transport..................

20 3.5 Ac c i den t Mi ti ga ti o n....................................

20 3.6 Recommendations........................................

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ADVANCED REACT 0RS...........................................

22 4.1 I n tr od uc t i o n.........................................

22 4.2 LMFBR Research.......................................

22 4.3 GC R R e s e a r c h.........................................

23 4.4 Rec omme n d a ti on s......................................

23 5.

RE ACTOR AND FACILITY ENGINEERING............................

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5.1 I n tr o d uc ti o n.........................................

24 5.2 Mechanical and Structural Engi neering................

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5.2.1 Mecha nic al Eng i nee ring...............................

24 5.2.2 Struc tu ral ' Engi nee ring...............................

25 5.2.3 Seismic Safety Margins Research Program (SSMRP)......

27 5.3 Prima ry Sy s tem I n teg ri ty.............................

28 5.3.1 Frac ture Mechani c s...................................

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5. 3.2 Operati ng Ef fects on Material s.......................

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5.3.3 Nondestruc tive Exami nation...........................

29 5.4 El ectrical Equi pment Qual i fication...................

29 5.5 Fuel Cycl e Faci l i ty Sa fe ty...........................

30 5.6 Ef fl uent Control and Chemical Systems................

30 5.7 Decommi s s i o ni ng......................................

30 5.8 Rec omme n d a ti o n s......................................

31 6.

FACILITY OPERATIONS AND SAFEGUARDS..........................

32 6.1 I r. tr o d u c t i o n.........................................

32 6.2 Human Engineering and Man-Machine Interface..........

32 6.3 Pl ant Instrumenta tion and Control....................

33 6.4 Oc c u pa ti o n al P ro tec ti o n..............................

33 6.5 Eme rgency P repa re d ne s s...............................

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6.6 S a fe g u a rd s...........................................

35 6.6.1 Material s Control and Accounting.....................

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6.6.2 P hy s i c a l Se c u r i ty....................................

35 6.6.3 Th re a t a nd St ra te gy..................................

35 6.7 Recommendations......................................

36 7.

WASTE MANAGEMENT............................................

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7.1 I n tro duc ti o n.........................................

37 7.2 High Level Waste.....................................

37 7.3 Low L ev el Wa s te......................................

38 7.4 Uranium Recovery.....................................

38 7.5 Recommendations......................................

39 vi

Page SITING AND ENVIRONMENT......................................

40 8.1 I n trod uc ti o n...........................................

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8.2 Earth Sciences.........................................

40 8.3 Siting.................................................

40 8.4 He a l th E f fe c t s.........................................

41

.8.5 Environmental Impacts..................................

41 8.6 Recommendations........................................

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9.

SYSTEMS AND REl.IASlLITY AilALYSIS............................

43 4

9.1 I n tro d uc ti o n...........................................

43 9.2 Ri sk Methods and Dat a Eval uation.......................

45 9.3 Reactor Risk and Reliabil i ty Analysi s.................

45 9.4 Transporta tion and Material s Ri sk......................

46 9.5 Re c omme n da ti o n s........................................

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1 APPENDIXES.......................................................

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AP PE NDI X A - REF ERE NCES.....................................

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l AP P E ND I X B - G L0 S S ARY.......................................

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2 PART I

GENERAL COMMENT

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' GENERAL C0ft4ENTS I

' 1.

Introduction i

Prior to our-last report to the. Congress on the ' Nuclear Regulatory Commission (NRC) Safety Research Program for _ FY 1982, NUREG-0751 (Ref. '1)*, we have taken the.

of view that all of the safety research being proposed by the Staff-was generally useful, and have recommended areas.which si J. be given priority orsincreases in: funding.

Hence, in our previous report to the Commission, NUREG-0699 (Ref. 2), although we recommended that the LCFT experi-mental program be terminated after FY 1982, we did not propose that LOFT funding be. cut in FY 1982 to permit the reccNnded increases

- in Severe Accident Phenomena and Mitigation Researci., and Systems:

and Reliability. Analysis, or to cover research on Fast Breeder.

Reactor 3.

However, in NUREG-0751, we took the point of view that budgetary constraints would exist, and recommended cuts in LCFT, in Loss-of-Coolant Accident (LOCA) and Transient Research, a'.(i in Waste Management to perinit our recommendation for-greater -emphasis and funding to be given to Plant.0perational Safety, Severe Accident Phenomena and Mitigation Research, and Systems and Reliability i

Analysis.

In this document we will again assume that budgetary constraints exist and that funding for the Advanced Reactors-program also must be accommodated within. a total funding level at or close to that proposed by the Executive ~ Director for Operations (ED0).

2.

Previous Recommendations In NUREG-0699, we outlined several steps that we believed needed to be taken by the NRC if those safety research areas which are judged to have - potentially the greater impact in protecting the public health and safety are to receive the necessary priority.

These steps are listed below, together with some brief comments on their current status.

'The Commission will have to provide policy guidance on the major open safety issues.

We believe that this basic task remains to be done.

• References appear in Appendix A 1

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'The NRC research user offices will have to reevaluate their approach to formulating requests for research and strive to consider these in some broad framework which takes into account the major issues confronting the agency.

Since December 1980, the Office of Nuclea'r Reactor Regulation (NRR) has begun to exhibit significantly more cohe.; ion and depth in its requests for research and in its comments on the safety research plan.

However, we believe that NRR will have to devote further attention tu developing priorities for its research requests and devote more effort to defining its longer-range research needs.

'The Office of Nuclear Regulatory-Research (RES) will have to reevaluate its current and proposed programs in terms of risk-reduction potential and major regulatory needs.

Although RES uses probabilistic risk analysis to assign priorities to unresolved safety issues and generic issues, to various alternatives proposed for the Degraded Core Cooling rulemaking, and to other major activities of NRR, it does not use such analysis to provide an important input into an evaluation of the efficacy of its own research program.

We recom-mend that it do so.

We also recommend that RES critically evaluate its research programs to see if its major experimental or theoretical programs are really capable of and likely to provide infomation of an importance commensurate with their expense.

'The NRC will have to judge whether some research, particular-ly that which involves large scale component testing or the application of existing methodology, should be the responsibil-ity of industry rather than the NRC.

1 There is a need for policy guidance on which research matters can and should be the responsibility of the industry.

'The NRC may have to reduce sharply some research which is merely confirmatory in nature where there is a good reason to believe that the current regulatory requirements provide adequate protection to the public.

The NRC Staff appears to have made a good beginning in this direction.

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7 In E NUREG-0699, we' identified several areas as requiring emphasis, including the following:

' plant operational behavior as a function of design and control;

'the impact of control systems and other nominally non-

-safety systems on. safety;

' improve'd approaches to reduce the impact of design errors;

' reexamination of the' general design criteria.

In NUREG-0751, we recommended that several areas, including the above, be given higher priority and increased funding in FY 1982 at:

follows:

"(a) The role of control systems in safety; (b) Plant operational safety, including system behavior as a function of design; (c) Reliability analysis of existing plants, including emphasis on the more detailed understanding which might prevent many operating occurrences which have potentially serious implica-tions; (d)

Improved, more reliable shutdown heat removal systems, includ-ing dedicated and bunkered systems; (e) Studies of degraded core and core melt accidents with emphasis on the conceptual design and evaluation of features to miti-gate such accidents;

( f) Studies of the physical and chemical behavior of fission products in post-accident environments; (g) The early development of an approach to supplement or replace the single failure criterion; and (h) Other matters relevant to setting the principal design bases for plants to be constructed."

The NRC program for FY 1983 appears to be fairly responsive to our recommendations on i tems (c), (e), and (f) above.

Programs are being initiated on items (a), (b), and (g), but require substan-tially greater funding.

Items (d) and (h) seem not to be specifical-ly identified by the NRC Staff as important safety research matters 5

for FY 1983.

Although the Task Action Plan A-45 on Shutdown Decay Heat Removal Requirements (Ref. 3) deals with much of the content of item (d), its initial emphasis is on existing plants.

We believe RES should participate significantly in this work, and in addition, should carry on a more general investigation, the results of which can contribute to the design of future plants.

3.

Design-Related Research We wish to call attention to an aspect of safety research that has been weak or deficient in the past NRC program, namely, design-related safety research.

In order for the NRC to develop improved safety criteria, standards, and guides, it is important that the necessary knowledge of design possibilities, capabilities, tradeoffs, etc., be available.

This is the case whether the NRC is issuing

' design criter4a or performance criteria.

It is the case for the rulemakings on Degraded Core Cooling and Minimum Engineered Safety Features.

The NRC Staff has funded some reviews and introductory evaluations of design alternatives for improved shutdown heat removal systems, for containment approaches to mitigate degraded core accidents, and for design features to reduce the potential for successful sabotage.

These studies represent a potentially useful first step.

But their ultimate usefulness depends on the completion of a next step in which enough knowledge and information is developed to permit the promulgation of an appropriate NRC rul e.

A modest effort along these lines has been proposed for the Degraded Core Cooling and Minimum Engineered Safety Features rulemakings. However, the effort proposed is far too small to provide the needed information on a i

timely basis, and it does not include a program to develop design requirements with regard to sabotage or the specific objective of l

developing NRC requirements for future light-water reactors (LWRs),

particularly standard plants.

We believe that a sufficient emphasis and funding should be given to such design-related safety research to provide the needed informa-1 tion on a timely basis.

4.

Budget Recommendt.tions We make the following budget recommendations ft - the NRC safety research program in FY 1983, arranged by Decision Units and sum-marized in Table 1.

LOCA and Transient Research The proposed budget level is adequate.

However, $2 million should be reallocated within this Decision Unit from the program 6

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on Code Assessment and Application to the program on Code Improve-ment and Maintenance.

LOFT Only the funding level necessary to place the reactor in a standby status and to cover project closeout costs and a comple-tion of engineering and analysis tasks associated with the FY 1982 experimental program should be allocated for LOFT in FY 1983.

This has been roughly estimated by the NRC Staff to be about $14.5 million.

Accident Evaluation and Mitigation The proposed total funding of $50.4 million in FY 1983 appears to be adequate.

However, it should be reallocated internally to provide significantly greater effort on the program on Accident Mitigation, enough to provide the information needed for deci-donmaking in the Degraded Core Cooling rulemaking by the end of 1983.

This should be accomplished by a reevaluation of the merits of the currently proposed experimental programs.

Advanced Reactors l

A funding level of $20 million should be allocated for Liquid Metal Fast Breeder Reactor safety research and $2.5 million for Gas-Cooled Reactor safety research.

Reactor and Facility Engineering We recommend an increase of $2 million to the proposed funding i

of this Decision Unit to a level of $40.2 million; these funds are to initiato a more comprehensive safety research program on reactor aging effects and to permit augmentation of the work in probabilistic seismia safety,' including study of a BWR.

Facility Operations aid Safeguards We recommend an increase of $3 million to the proposed' total budget of this Decision Unit to a level of $19.8 million partly to enable a more appropriate emphasis on control systems, and on plant operational behavior.

Additional effort is also needed on the contribution of human error to risk. This should include not only operators and operating practices, but al so maintenance personnel and maintenance practices.

Studies on design steps to prevent sabotage should be accomplished within this Decision Unit, with appropriate input from the Division of Risk Analysis.

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Waste Management The proposed funding level is appropriate.

Siting and Environment The proposed funding level is appropriate.

Systems and Reliability Analysis We recommend an increase of ~ $4 million in the funding of this Decision Unit to a level of $25.7 million; t%ese additional funds should be used specifically for research c.9 the matters recom-mended in Chapter 9 of this report.

5.

Specific Comments and Recomendations Specific comments and recommendations regarding the scope, nature, 4

and funding levels of the various Decision Units of the research program are presented in Part II of this report.

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- TABLE 1 0FFICE OF NUCLEAR REGULATORY RESEARCH PROGRAM SUPPORT BUDGET FOR FY 1983

'(DOLLARS'IN MILLIONS) l l

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PROPOSED-RECOMENPATIONS 1.

LOCA AND TRANSIENT 31.0 31.C RESEARCH 2.

LOFT 40.6 14.5 3.

ACCIDENT EVALUATION 50.4 50.4 AND MITIGATION l

4.

ADVANCED REACTORS 2.5 (21.0)*

22.5 5.

REACTOR AND FACILITY 38.2 40.2-l ENGINEERING

6. -FACILITY OPERATIONS 16.8 19.8 AND SAFEGUARDS 7.

WASTE MANAGEMENT 19.6 19.6 I

8.

SITING AND ENVIRONMENTAL

'14.7 14.7 RESEARCH 9.

SYSTEMS AND RELIABILITY 21.7 25.7

' ANALYSIS

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TOTAL PROGRAM SUPPORT 235.5 238.4

• Set aside by ED0 for Conenission action.

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i PART II

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SPECIFIC COMMENTS a

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1.

LOCA~ AND TRANSIENT RESEARCH 1.1 Introduction This Decision Unit includes several programs which are directed primarily toward improved understanding of reactor behavior in loss-of-coolant accidents (LOCAs).

In previous reports, we have remarked on the extensive reorientation of these programs which gave increased attention to small-break problems. We continue to support this change of emphasis, and in addition suggest that the* research program be. broadened to consider wider classes of transients other than those associated only with LOCAs.

We shall reiterate this point in our discussion in the next chapter on LOFT.

Included in this Decision Unit is a program of improvement and assessment of codes which has as its objective an analytic descrip-tion and understanding of light-water-reactor (LWR) transients. The last group of programs in this Decision Unit is directed toward the understanding of core and fuel behavior under conditions in which the core is inadequately cooled. Comments on all of these programs follow.

I 1.2 Semiscale Deficiencies in size and configuration of Semiscale are compensated for, at least in part, by the economy of test operation and the relatively rapid turn around test time.

The NRC Staff has under-taken a serious study of the limitations of Semiscale and of the scaling questions which arise in translating obscevations in Semi-scale to full scale. We commend this effort and urge its continua-tion.

1 We support the program in Semiscale (Mod-2A) and also recommend another version of semiscale (Mod-5) directed toward the Babcock and Wilcox type pressurized water reactor (PWR).

This recommendation for the Mod-5 version of Semiscale was made in NUREG-0699 (Ref. 2)*,

and we hope that this program can be implemented in the near future.

1.3 Separate Effects Experiments and Model Development One of the programs in this item is the Two Loop Test Apparatus (TLTA) or the planned TLTA upgrade known as Full Integral Simulation Test (FIST).

Such a facility is the analogue of Semiscale for

• References appear in Appendix A.

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boiling water reactors (BWRs).

We have repeatedly; recommended. an improved program of this kind for BWRs.

Some questions also remain regarding the possible significance of core bypass and loss of l

emergency core cooling (ECC) spray water in a medium to. small break LOCA.. Tests to evaluate -the extent of this problem should be

' undertaken.

i Other programs in ' this area are 'FLECHT-SEASET 'at Westinghouse and.

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the. Thermal Hydraulic Test Facility (THTF) at Oak Ridge National Laboratory.(ORNL)..Useful data have been obtained in both programs.

The data from FLECHT-SEASET are of good accuracy but are limited to low pressure; ~THTF, on the other hand, does operate at high pressure.

i The.THTF program is scheduled to end in FY.1982.

As remarked :in NUREG-0699, we believe that the present effort on code assessment is~ inadequate.

For an improved program, additional experiments on. separate effects are needed.

Many of these experi-ments would not require large facilities since the experimentation should be directed toward establishing essential physical and engineering bases for the codes.

The Office of Nuclear Regulatory i.

l Research (RES) has increased its efforts in this direction, and we l-encourage further increases in these efforts.

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The model development program consists for the most om ' of rela-tively small projects in various research laborator%s.

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this kind of program as being useful and produch ve.

Rt.. auld appreciate that this program provides a helpful interaction wich an f

important part of the - engineering and scientific community. and l

should therefore seek to broaden this contact by finding from time l

to time new contractors in this community.

The NRC still needs to improve its procedures 'ar awarding these contracts.

1.4 3-D Program l

This international program involves Japan, the Federal Republic of Germany (FRG), and the United States and was begun when LWR safety c

research was preoccupied with the problem.of large-break LOCAs in :

I PWRs.

Since the progra, as an' international obligation, cannot easily be cancelled, at the least it is in urgent need of redirec-tion. The primary need for this redirection comes from the expected decision by the FRG to proceed with the construction of the Upper l

Plenum Test Facility (UPTF).

The construction and test program for this facility can be expected to last over a decade and the cumula-

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tive cost to the United States can' be expected to be between $50 million and $100 million.

The mission of this program was deter-mined some time ago to provide information relating to special questions regarding large-break LOCAs. We believe that the informa-tion which will be obtained in UPTF. is not of primary importance for PWR safety, and that this information could be obtained at less cost and more quickly in other ways.

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I It must be emphasized :that there are other parts of this inter-national-program which have been very useful and have been connected.

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. Among these should be mentioned the

.with less c.;stly facilities.

PKL facility 'in FRG and. the Cylindrical Core Test Facility (CCTF) l' and the Slab Core Test Facility -(SCTF) both in Japan.

Further, there are plans in Japan ' to. construct-a large Scale, Test Facility 4

l (LSTF) which promises to be an outstanding facility for. the study of medium and small-break - LOCAs in PWRs.

The NRC ~ should. plan to.

participate in the program in this facility -together with the program in an associated facility, the Two-Phase Flow Test Facility.

j In the area of international programs, we recommend that the ' NRC take steps to get access to the experfuental data obtained in Japan from their test program for BWRs..In particular, it would be useful to have access both to the data.and to the facilities in which work j

has been done on LOCAs and ECC system performance for BWRs.

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understand that RES.is pursuing this possibility and wish to encour-l age them in this effort.

j 1.5 Code Improvement and Maintenance 1

i The tasks of developing best-estimate codes for PWR and BWR systems

.~ TRAC, an advanced, code, requires-and -'should j

are not complete.

j receive further work.

This work -includes both. development and j

assessment. The RELAP-5 Code also should be. supported; this code is j.

receiving general recognition as being useful.

l This program of Code Improvement 'and Maintenance can be of basic significance for reactor safety and should be given continued j

support.

The availability of very fest running codes for analysis of system i

behavior during transients for BWRs and PWRs'should be evaluated.

j If _ codes that have the necessary compromises on detail and on accuracy of modeling of. system behavior and physics are not avail-able i.o enable the running of transients very much faster than real.

1 time, such code development should be undertaken.

1.6 Code Assessment and Application f

It has already been remarked here that the code assessment program i

is. inadequate, particularly in the case of TRAC. We recommend, howeve'r, that both TRAC and RELAP-5 assessment be continued at a reduced level if necessary to permit the recomended code imorove-ment.

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1.7 Fuel Behavior Under Operational Transients This program includes pellet-clad interaction (PCI) investigations and fuel rod code maintenance and improvements.

These programs are being deemphasized in agreement with our previous recommendations.

Some PGI investigations, however, are still underway.

We believe that NRC should avail themselves of the considerable PCI work being performed by industry and ensure that work is not duplicated.

We have reservations about the validity and thus the value of the PROFIT Code as a guide.

1.8 Recommendations Aside from the 3-D international program which needs special consid-eration, the LOCA and Transient Research is budgeted at an economic and efficient level and we recommend that this budget l evel be approved.

We recommend that the budget level for the Code Improve-ment and Maintenance program be increased by about $2 million by reallocation within this Decision Unit; these funds should be taken from the program on Code Assessment and Application.

The 3-D Program is an international commitment to which the NRC must adhere.

There are many aspects of this program which contribute significantly to reactor safety in a cost effective w&y.

As our discussion above indicates, the only item of concern is the UPTF program in the FRG.

It appears that the FRG will proceed with the construction of this facility and the particular impact on United States expenditures will come from the long time span involved before completion. We recommend that the NRC continue participation, but make a serious effort to reduce the expenditures associated with UPTF.

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l' 2.

LOFT 2.1 Introduction The LOFT f acility has contributed to the understanding of the phenomena encountered in large-break LOCAs and has helped to demon-strate, within reasonable limits, the effectiveness of PWR ' emerge.ncy core cooling systems.

More recently, it has been useful in studies of the phenomena which may appear during transients and small-break LOCAs.

We reiterate, however, the view expressed in NUREG-0751 and l

NUREG-0699, that programs in LOFT do not make a cost effective contribution to reactor safety.

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2.2 The LOFT Test Program l

In NUREG-0751 and NUREG-0699, we recommended that - the LOFT test program be terminated by the end of FY 1982.

It was our strongly held view that there was an urgent need to transfer the funds thereby made available to other safety research programs-which would l

contribute much more effectively to reactor safety. 'This -is still our view.

2.3 Recommendations We recommend that a funding level of $14.5 million be provided for LOFT in FY 1983 to place the reactor in a standby status and to l

cover project closeout costs and a completion of engineering.and analysis tasks associated with the FY 1982 experimental program.

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3.

ACCIDENT EVALUATION AND MITIGATION 3.1 Introduction Most of the activities under this Decision Unit are meant to provide information that will be needed in the rulemakings that will deal l

with Siting, Minimum Engineered Safety Features, Hydrogen Control, and Degraded Core Cooling.

Because much of the content of these rules must deal' with territory-as yet only meagerly explored, a considerable amount of experimental and analytical work will be needed-for formulation of these rules.

Although the general areas of investigation presently'can be defined, it is necessary to raise and to answer a number of questions 'early in the process in order to focus the work properly.

Illustrative of the types of decisions needed are answers to the l

following:

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'Is the principal emphasis to be put on preventing core melt, or on dealm with the melted core after melting has occurred?

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'Should a final system design have to deal with 100% of the hydrogen that can be generated in any conceivable accident?

If not, what is the appropriate fraction?

'Is the degraded core problem to be dealt with primarily on a probabilistic or on a deterministic basis?

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'Is evacuation of the surrounding population to be seri-ously considered as an accident mitigation system?

If so, how much credit is to be taken for evacuation?

Answers to these, and similar questions that need to be raised, may be modified as the research develops, but they need to be formulated and some initial decisions are required to focus the research program.

In NUREG-0699 and NUREG-0751, we recommended that a high-level task force be established with the responsibility for recommending the research needs and for estimating the resources required to support these rulemaking proceedings.

A Degraded Core Cooling Steering Group was established and after some deliberation, recently completed a report.

The report does not respond to this need.

Guidance is still lacking.

We reiterate that decisions must be based on clear 18

l identification of information needs and that programs must be l

designed to respond to these needs. The Commissioners shvid assist l

by defining the safety philosophy and objectives to guide the work.

3.2 Behavior of Damaged Fuel As a bounding estimate, risk assessment studies have assumed that l

undercooling a core leads to melting the entire core.

This ve j conservative assumption has lead people to ignore the substantial difference between overheating and melting the oxide core.

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behavior, and margins to collapse, of an overhaated core is an essential yet largely unstudied question where answers are needed i

for decisions on accident evaluation and mitigation. The program is still in the formative stage but has made good progress.

l The examination of the Three Mile Island, Unit 2 (TMI-2) core is highly desirable and efforts should be continued to accelerate the cleanup and core examination at TMI-2.

The Severe Fuel Damage (SFD) test series in the Power Burst Facility (PBF) is intended to provide information cn the behavior of over-heated cores. It is potentially a much more valuable use for the PBF than the operational transient work now teminating and warrants priority. These experiments will be probing into a very complex area and special attention will be needed if they are to achieve their potential.

We believe that two new programs in this area should be deleted for l

l now and the funds thus made available be shifted to higher priority projects; these are:

(a) The Deformed Core Coolability (DECCA) program.

This is expen-i sive, duplicative and in our opinion unnecessary.

Better l

information is available from other sources.

t (b)

LWR Debris Coolability program.

This work is of questionable relevance to damaged core behas'or, is an expensive way to do l

heat transfer work, and tends to duplicate work being done on fuel melt behavior.

We would defer it until a clear need and relevance are shown.

3.3 Fuel Melt Behavior Research in this program aims at a better understanding of the behavior of molten fuel as it drops into the reactor cavity, melts through the vessel, drops on to the basemat below, and irtteracts 19

with the concrete basemat.

This investigation is of great impor-tance.

The experimental work is expensive, is difficult to carry out, and even more difficult to interpret.

What the NRC Staff described to us was primarily the. work planned for FY 1982.

Spe-cific plans for FY 1983 appear to be primarily an extension of what will be done in FY 1982. Th's is appropriate if the FY 1982 work is rc M ant.

However, in this area especially, policy guidance is needed on how much is to be attempted in the way of a detailed understanding of the processes being considered.

The Code Development work associated with these studies is extensive.

Here, especially, there should be considerable attention given to the question of how much detail is desirable or feasible. Otherwise, it is possible for the experimental program to beome a vehicle for code development while the code becomes primarily a tool for describ-ing the experiment.

This Decision Unit includes also an evaluation of the MARCH Code.

Our discussions of the MARCH Code with the NRC Staff and others lead us to urge all who use the Code and its results to recognize its limitations.

For example, the behavior of a melted core as described by the MARCH Code is largely a matter of speculation. We 4

recommend further work on the MARCH Code to make it easier to use and to improve its capability.

This program includes also continuing work on hydrogen generation.

In light of the importance of the hydrogen problem, the work seems generally appropriate.

However, that part devoted to hydrogen generation by corrosion of zinc and ferrous alloys by coolant does not deserve a high priority.

3.4 Fission Product Release and Transport This work aims at an improved description of the radiological source term for severe accidents.

The NRC Staff recently completed and published NUREG-0772 (Ref. 4) which provides technical bases for estimating fission product behavior during LWR accidents.

This document points out what is known and identifies what is not known about fission product behavior.

Plans are to undertake further research to achieve better definition.

We believe this work and associated work on aerosol definition deserve high priority.

3.5 Accident Mitigation This program concentrates on systems which may be useful in mitigat-ing severe accidents.

Current activities include consideration of fil tered vented containment systems, core retention systems, an c' 20

m

[.

J general studies of severe accident sequences. We consider this work to oe of paramount importance because a thinking through of the many pessibilities and a detailed consideration of some of them is a necessary precursor for the work of the whole Decision Unit.

We believe additional emphasis should be given to this program.

y l-We are concerned that. information required for rulemaking proceed-ings' will not be available unless special attention is given to early definition of the needed research associated with this program..

We are indicating our evaluation of the importance of the work in this area by recommending an increase in the proposed budget.

We i

are not sure : that this is enough. We recommend that a thorough planning process. be carried out by representatives of appropriate units and that programs and needed fundt, be identified. If the funds currently proposed for this program are inadequate, we recommend that additional-funds be made available by transfer from other work within this Decision Unit.

3.6' Recommendations We endorse the requested funding level for this Decision Unit.

However, we recommend that the funding level for the A;cident.

Mitigation program be increased by $3 million by reallocation within this Decision Unit; these funds should be taken from the program t,n Behavior of Damaged Fuel.

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l 4.

ADVANCED REACTORS 4.1 Introduction This is a new Decision Unit, devoted solely to research on two types of advanced reactors:

Liquid Metal Fast Breeder Reactors (LMFBRs),

and Gas-Cooled Reactors (GCRs).

4.2 LMFBR Research We have bean informed by RES and NRR personnel that they expect NRC to be requested to reinitiate licensing of the Clinch River Breeder Reactor (CRBR) in the near future.

Further, they believe that the Department of Energy (00E) is moving clearly in the direction of a long-range LMFBR program and that development of a plant larger than CRBR will follcw the latter in a few years.

In view of the apparent inminence of CRBR licensing, RES has developed a phasad plan fcr expansion of the LMFBR safety research program during FY 1982 which will bring the rate of expenditure to a level of approximately $20 million per year by the start of FY 1983.

This plan involves maintaining generic work initially at a level of about $8 million per year, with the expansion of the program to be based upon de-tailed studies by RES and NRR jointly to identify NRR needs for licensing CRBR.

Some of these needs are expected to involve accel-eration of current work; however, we believe that the bulk of the expansion should involve new work such as core melt accident evalua-tion using risk analysis and other developing techniques, systems analyses, operational behavior, control system dynamics, relief valve behavior, and environmental effects on equipment.

We recommend a budget level of about $20 million for FY 1983 to carry out initial work under a plan to be developed jointly by NRR and RES defining safety research needs for CRBR licensing and L

follow-on licensing programs. We believe that at least half of the

$20 million expenditure should be devoted to work in new areas like those identified above.

l We believe also that much more attention should be devoted to long-range planning in LMFBR research than has been the case in the past. For example, there are no established and agreed-upon safety-related design criteria within NRC for LMFBRs and, without such criteria, licensing reviews are more ad hoc and it is difficult to define needed research programs, their costs, and their schedules.

Further, current research work is not dealing with all of the topics 1

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g y--

l we believe important; it may not be addressing the most orgent ones, and the level of expenditure may be quite unrealistic.

Therefore, we believe that NRC should move promptly to initiate design-criteria studies for demonstration-size LMFBRs, to define safety research needs for such units, and to define an appropriate safety research program.

Such a program should incorporate the views of NRR, RES, DOE, nuclear steam system suppliers, the utility ind::stry, and independent experts.

The program should be coordi-nated with ongoing safety research 'being performed by DOE as well as by foreign groups with which the NRC has cooperative agreements, and should include topics such as the following:

'What are the safety issues on which research is needed?

'How is each issue to be resolved?

'Who should do this work?

NRC?

DOE?

Industry?

Other?

'When is the result needed and why?

'What is the estimated cost to resolve the issue and is it practical to finance the effort?

Insofar as practical considerations and timing will permit, the results of this effort should be applied also to CRBR.

4.3 GCR Research The current safety research program on GCRs is devoted largely to support of safety evaluations related to the Fort St. Vrain high-temperature gas-cooled reactor (HTGR).

However, this pr'ogram will near completion in FY 1982, and the proposed program for FY 1983 is based primarily on the anticipation that further development of the HTGR may lead to a license application within the next several years.

On this basis, the proposed program and funding level is adequate, and we propose no changes for FY 1983.

4.4 Recommendations We recommend that the research programs in this Decision Unit be funded at a level of $22.5 million; $20 million for LMFBRs and $2.5 million for GCRs.

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5.

RaCTOR AND FACILITY ENGINEERING j

5.1 Introduction This is a new Decision Unit that has resulted from restructuring the research programs to reflect the consolidation of RES and the Office of Standards Development functions.

This Decision Unit deals with programs pertinent to plant and facility design engineering and component qualification.

Specific comments on these programs follows:

5.2 Mechanical and Structural Engineering 5.2.1 Mechanical Engineering The Mechanical Engineering Safety Research Program was organized in FY 1981 to develop a better technological basis for safety regula-tion as the aftermath of recent experience with less-than-satisfac-tory performance in safety-related equipment of a mechanical engi-neering nature.

The intent is to strengthen regulatory understand-ing of engineering practices and develop probabilistic approaches to the treatment of some engineering questions.

We support the general usefulness of this program.

The level of funding appears reasonable for the objectives but the portion invested in computational procedures seems excessive.

Experience with piping systems, especially seismic restraint devices, indicates that improvements in equipment by better design, maintenance, l

and inspection are needed if the credited safety capabilities are to be realized.

lhe requirements for equipment qualification are not well defined and considerable work is necessary to establish a basis for regulations.

Methods of checking computational procedures are needed in some areas.

The capacity and capability of safety and relief valves under the anticipated working conditions of gas, liquid and two-phase flow operation are particularly important as shown by recent discussions concerning ATWS regulatory requirements and experiences at TMI-2 and Crystal River.

The combination of dynamic loads effort appears to be concentrated on comp'ex computer analysis techniques that may not be usable because of insufficient data base and the need to use gross assump-tions of structural behavior in the analytical techniques.

This

~

~ effort would be better applied in determining how inherent capa-bilities of materials can be used to compensate for errors and to off-set overly conservative assumptions in combined load analysis, 24

e.g.,

factors that limit the size of double-ended pipe breaks, ductile response of piping, rate effects of fluid rel ease, and structural damping in seismic events.

We endorse the effort to participate in and capitalize on foreign research programs being pursued concurrently in these areas.

The interchange of information with experts in other countries having a different perspective will broaden the understanding of mechanical engineering applications and reduce the likelihood that important safety issues will be overlooked.

The RES Staff should examine some elements of the progrs to see whether its perception of need is consistent with that of the anticipated users of the information.

Notably, the project on Effects of Nonsafety Systems which is planned to start in later fiscal years, relates to Task Action Plan A-17 (Ref. 3) whose results would have been reported before the proposed program is ini tiated.

At the same time, much of mechanical component qualifi-cation requirements will have been instituted in NRC regulations long before the proposed research is completed.

This comment should not be interpreted as suggesting a lack of need for the proposed work, but rather a need to coordinate the research schedule with the application schedule.

The RES Staff should evaluate also whether it is appropriate for NRC or industry to finance these proposed work areas.

We support the value of some duplication of industry research to assure the availability of independent results and of consultinq expertise.

Work planned for the Safety / Relief Valve Test Program fits this criterion. We question the advisability of work done apart from the industry activity or as a substitute for work rightfully the respon-sibility of the industry. Snubber qualification, for example, might more appropriately be sponsored by the industrial suppliers and the industry users.

5.2.2 Structural Engineering We would assign relatively high priority to the programs on Load Combinations for Design of Structures and to the International Cooperation Program. The latter shows promise of providing knowledge from tests or from actual experience that will be valuable in assessing the accuracy and effectiveness of seismic design proce-dures.

We do not support the program on Effectiveness of Quality Assurance and Inspection Procedures with its present scope.

We believe that an analysis of construction errors or deficiencies to evaluate 25

their relative contributions to risk would serve to focus this program on those areas where improved QA or inspection would be of greatest benefit.

We do not support that portion of the program on Dynamic Testing and Damage Assessment that deals with the assessment of post-earthquake or post-accident structural damage.

It does not seem likely that the extent or type of damage that will actually occur can be fore-cast without an exceptionally comprehensive and expensive program, the cost of which cannot be justified by its relatively small potential contribution to risk reduction.

That portion of the program concerned with the evaluation of analytical resul ts by comparison with the results from experiments has merit but should involve also comparisons with results from experience with actual structures.

Comparisons of analytical results produced by one computer code with those produced by another has little merit; this is being done in the Soil-Structure Interaction Program and has been proposed for others.

It is time to seek data from experiments or experience with which to compare the results from several of the complex computer codes now being used to analyze and design struc-tures for seismic and other loadings.

We understand that the Commissioners have disapproved the proposed contract for Benchmarking of Computer Codes Used in Struc tural Design as being too ambitious, among other reasons.

We understand also that NRR has indicated no strong desire for an in-house code to be used to validate those used by the industry.

We hav, no quarrel with these t'ecisions, but we do believe that there is a need for the licensing staff to have greater confidence in the ability of these codes to yield correct results and lead to satisfactory designs.

Correctness in this case means not only correctness of the algo-rithms and correctness of the mathematical models but correctness of the result in comparison with the best available physical evidence of how structures actually respond to loads.

We believe that this can be done by comparing the codes.with " standard problems" based to the extent possible on tests or measurements on real or model structures.

These comparisons should be made not by the NRC but by the architect engineers, vendors, licensees, or applicants, as appropriate.

NRC research should include determining whether such validation is possible and what its limits might be, selecting standard problems, and developing of instructions and requirements for validation and criteria for evaluating the results.

In other words, NRC research should be aimed at providing the NRC Staff or its consultants with the ability to ask the right questions and to evaluate the answers.

This approach should lead to a significant reduction in the scope and cost of the program.

Further reduction could be made by reducing the scope to include initially only one of the three task areas originally proposed; i.e., Seismic Analyses, Containments, or Category I Structures.

And finally, since it may 26

mt be possible to transmit all of the expertise gained from.this project to the NRC Staff, although this should be.done to the greatest extent practicable, it tuay be desirable to select a contrac-tor that would be available to serve as a consultant to the Staff in connection with subsequent uses of the code validation procedures.

The current consideration of accidents leading to degraded or melted cores requires the ability to predict the conditions of pressure, temperature, etc. for which the fission products will be released from the containment in amounts and at rates that will produce a threat to the public.

The program in Safety Margins for Contain-ments addresses this roblem but in a manner we consider wrong and unnecessarily expensive.

This program seems to be intended to provide the NRC Staff with the means to predict containment behavior, including presumably both " failure" and leak rate, for all of the large number of different types of containments that have been or are being built.

We believe that the responsibility for computing and justifying containment behavior and capacity should be placed on the applicant or licensee.

The NRC, of course, must have the knowl edge, background, and expertise to direct and evaluate the computations or demonstrations required, and the research program should be designed to provide those abilities. The NRC Staff should be able to define, on the basis of their consequences, those failure modes and limit states that must be calculated and probably should translate those limits into structural phenomena.

The NRC Staff should have the knowledge to request calculations appropriate to the various types of containments and to evaluate the results for each specific type.

The NRC Staff need not and should not attempt to develop methods of analysis appropriate to each of the many types of ccntainments, but it may t'e desirable to do so for a very limited number of cases in order to develop the expertise needed to direct and review the licensee or applicant program.

Similarly, the NRC Staff need not be responsible solely for tests to verify the various analyses, but might do some' testing to develop questions and to develop experience needed to evaluate the predictive capability of the analyses.

The NRC research in this area should have the active collaboration of personnel from the Divisier, of Risk Analysis (DRA),

since a probabilistic consideration of leakage rates and failure modes is important.

Those research programs not discussed here are considered of average priority and of generally acceptable scope.

5.2.3 Seismic Safety Margins Research Program (SSMRP)

It is clear that we need to know whether earthquakes are significant contributors to the risks from accidents in nuclear power plants.

It is our understanding that one objective of the SSMRP is to develop a methodology to answer this question.

27

c It is not yet clear how the methodology developed in the SSMRP can or will be used to evaluate seismic risks.

For one thing, it appears now that many of the data needed to describe the seismic input, structural response, component fragility, etc., are not only quite site-specific and plant-specific but are subject to consider-able uncertainties.

The importance of this cannot be evaluated until the planned sensitivity studies are completed, and an evalua-tion is made of other contributions to uncertainty or possible error.

A second problem relates to the likely usefulness of the current SSMRP methodology.

It appears to be very costly to implement and the accuracy of the results may be difficult to assess, aside from the large effects of uncertainties.

There is need for a practical method of assessing the seismic contribution to risk in the probabilistic risk analyses currently underway and those which are likely to be performed at an accele-rated rate in the future.

However, it is not clear that the current SSMRP will provide the NRC Staff with this capability.

We believe that a research program on probabilistic seismic safety is important.

However, we are not satisfied that the current SSMRP is properly structured to meet the needs.

We recommend the establishment of a task force that includes representation from NRR and the DRA to develop an appropriate program and scope of research in that area.

The work of this task force might result in redirec-tion or reorientation of the SSMRP or in recommendations for dif-ferent or additional research efforts on probabilistic seismic sa fety.

The current SSMRP effort dealing with specific reactors has been focused on a PWR.

In order to find out whether there are any special seismic safety questions for BWRs, we recommend that the SSMRP undertake a study of such a plant while the task force under-takes its review.

We recommend that additional funds be provided to support the BWR study and the task force study.

5.3 Primary System Integrity 5.3.1 Fracture Mechanics This is a good long-range program that is providing a sound basis far decisions on the integrity of pressure vessels and piping.

The question of thermal shock in pressurized systems represents an important uncertainty for the integri ty of the primary system, 28

i especially for older pressure vessels. Over the last several years, this program has been developing the techniques and information needed to. make informed regulatory decisions in this, as well as other areas.

In the piping area, RES should continue to work with NRR to define programs which will provide an acceptable basis for reducing the number of constraints or supports on piping systems, especially the primary loop, while maintaining adequate safety margins for all plant operating conditions.

5.3.2 Operating Effects on Materials The major contributor to uncertainties in assuring the integrity of the primary and secondary boundaries are the effects of operating environment, radiation, and water chemistry. This program addresses these issues in a sound, coherent manner.

One of its larger components is the examination of the Surry steam generator by -the Battelle Northwest Laboratory.

This provides a unique facility for needed NRC work on the verification of nonde-structive examination (NDE) techniques.

However, it could also be very useful to industry for activities sus.h as training of NDE operators, chemical cleaning studies, etc. We suggest that the NRC limit their exclusive control of the facility to a short period, say two more years, and look into then transferring the facility to private management so that a wider range of work of value to the industry can be performed with the steam generator.

5.3.3 Nondestructive Examination Periodic inspection of reactor components is regularly carried out to assure that no dangerous flaws are present in the primary coolant system pressure boundary.

NRC must be capable of judging how reliable these techniques are and must be able to develop criteria for the acceptability of new techniques.

The current programs address these questions at an adequate level.

5.4 Electrical Equipment Qualification Research in this area includes the qualification testing for LOCA and main steamline break environments, the development of accele-rated aging methods, the determination of representative LOCA source terms, and fire protection evaluations.

The proposed program for FY 1983 contains funding for the perfor-mance of fire replication tests.

In NUREG-0699 and NUREG-0751, we stated that the NRC funding of these tests cannot be justified by 29

the information that will be obtained.

We have no new information to change. this _ view.

We do r.ot support this part of the program.

Selected TMI-2_ instrumentatior and electrical equipment will be removed, examined, and tested to obtain a better understanding of the performance of equipment which had been qualified under existing standards.

This program will be carried out in cooperation with DOE and industry.

We support this work, but believe its value is diminished significantly unless it can be done in the near future.

5.5 Fuel Cycle Facility Safety This program includes research to provide the licensing offices the competence to evaluate source terms during nomal operation of fuel cycle facilities, to develop models for realistic analyses of accidents within such facilities, and to assess the effects of dry storage of spent fuel.

Specific attention is dire &J to analyses of accident ;cenario', leading to _ the generation transport, and release of aerosols from fuel cycle facilities.

We agree with the importance of this effort and we support the plans for it.

5.6 Effluent Control. and Chemical Systems This program includes research on improving the accuracy for evaluat-ing the performance of effluent control systems in LWRs and fuel cycle facilities. Included are studies of hydrogen control and mitigation systems.

In order to achieve these objectives, a major effort is being directed to obtaining more accurate radionuclide source term data.

We endorse these efforts and we are encouraged to note the newly planned associated studies for developing QA/QC procedures for effluent monitoring systems.

Inasmuch, however, as 4

studies of Licensee Event Reports (LERs) from operating nuclear power plants have shown a relatively high frequency of failures in the equipment installed to monitor the performance of effluent i

control systems, we recommend that a portion of the funding for this program be directed to research on the evaluation and correction of these failures.

Data show this to be a continuing problem with no signs of improved perfomance in recent jears.

1 5.7 Decommissioning This program supports research designed to establish and validate the criteria by which the licensing offices will ddarmine whether nuclear power plants and other facilities have been satisfactorily decommissioned and their sites can be released for unrestricted public use.

We reiterate our recommendation in NUREG-0657 (Ref. 5) that features of design that will facilitate later decommissioning of nuclear facilities should receive priority within this unit. The 30

current ' program does.not reflect this emphasis.

In addition, we suggest that there is a need for a greater effort, on the part of-the licensing staff, to incorporate into their rules and regulations the benefits of the research that has been conducted in this subject area in the past.

5.8 Recommendations We recommend that the FY 1983 funding for this Decision Unit be increased by $2 million, from $38.2 million to. $40.2 million, to help fund a more comprehensive safety research program on reactor aging effects and to permit augmentation of the work in probabilis-tic seismic safety, including study of a BWR.

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6.

FACILITY OPERATIONS AND SAFEGUARDS 6.1 Introduction This Decision Unit represents a new grouping of programs as a result of the reorganization within RES; it currently includes programs on:

Human Engineering and Man-Machine Interface, Plant Instrumentation and Control, Occupational Protection, Emergency Preparedness, and Safeguards.

This Decision Unit encompasses programs related to many eases of facility operation and control which we believe require r imme-diate attention.

Particular new emphasis is recommendec or the following:

' Plant operational behavior as a function of design and control.

'The impact of control systems and other nonsafety systems on safety.

'The need for improved radiological protection.

' Design features to prevent sabotage.

' Ability to respond to sabotage attempts.

Specific comments on these programs follow.

6.2 Human Engineering and Man-Machine Interface The currently outlined program for research in the area of human factors includes studies of human error rate, reviewing control room design from a human factors perspective and enhancing operator sel ection, training, and performance.

The research in the man-machine interface area will include information control and display design and evaluation; human reliability and human performance standards; procedures development and analysis; and computer based aids to human performance.

T1e goal of the research in this area is to improve NRC's basic understanding of the impact of humans on reactor safety and of factors that affect the performance of the man-machine system.

We have previously identified this program as a high priority area and suggested areas for additional research in NUREG-0751. The ultimate objective of this research is to reduce the human contribution 32

to risk to an acceptably low level. The current program seems to be a reasonable attempt to address the concerns regarding human factors which resulted from the TMI-2 accident.

The fomation of a' program plan by the Human Factors Society for RES appears to be an acceptable method of assuring that those areas of

- research, deemed to have the most risk reduction potential, are addressed first. We continue to believe that this should be an area of expanding research over the next few years and believe that the funding levels proposed represent the minimum amount appropriate for this area.

It is still our belief that additional areas of research i

as outlined in NUREG-0751 should be pursued and that commensurate funding be made available.

6.3 Plant Instrumentation and Control A program on improved plant instrumentation was begun in FY 1980 in response to various recommendations that improved information on plant conditions during abnormal occurrences was needed by operators.

Work is being conducted on instrumentation systems for the detection of inadequate core cooling. Current emphasis is on that part of the system that indicates liquid level in the pressure vessel.. Assess-ment will be made of instrumentation and of qualification programr proposed in response to Regulatory Guide 1.97, Revision 2 (Ret.

6) requirements.

Work 'on noise diagnostics is scheduled to be i

carried into FY 1983.

We believe that this work is directed toward a useful objective.

Work is also being planned to examine the problems associated with electromagnetic pulse and lightning protec-tion.

Evaluations will be performed to develop an improved understanding of and solutions for problems encountered with operating safety-4 related instrumentation and electrical equipment.

Existing operat-ing experience will be used to identify problems associated with equipreent of this type, and to develop regulatory recommendations.

We believe that this work can reduce the risk associated with plant operation and continue to endorse an expar sion of this effort.

t In addition, there are plans to evaluate the safety implications of control systems. We endorse this activity and recommend that its funding level be augmented and that it include active participation by members of the DRA in order to accomplish an appropriate interac-tion between the deterministic and probabilistic approaches.

b.4 Occupational Protection Recent data and projections for the future show a continuing in-1 crease in the collective occupational doses associated with the 33 f

n sw

i operation of commercial nuclear power plants.

Whereas a few years ago, the generally accepted value for a single power plant was about 500 person-rem per year, the latest tabulation published *(Ref. 7) by the NRC Staff showed that the average collective dose per operating unit increased by 19% between 1976 and 1979 and now approximates 600 L

person-rem per year.

The collective dose per megawatt-year in-creased by 30%.

Projections are that some plants will have total-collective doses of as much as 5,000 person-rem for 1982, and that average values will continue to increase dramatically over the next few years.

According to the Long' Range Research Plan (Ref. 8) there is a need for development of a better understanding of the " corrosion, erosion, transport, and deposition phenomena within the primary coolant system and the effects on. these of changes in design, raterials of construction, quality control, housekeeping practices, and opera-tions."

The Plan states also that "there are few methods or data for analyzing the performance and reliability of LWR decontamination i

systems or for evaluating the net contribution (or reduction) they might make to occupational exposure."

l The continued increase in collective occupational doses at operating nuclear. power plants makes research in these areas of progressively greater interest.

We believe that it should be given priority.

l In terms of assessing current occupational exposures, we are encour-aged. by. the research activities planned for improving the QA/QC l

aspects of personnel monitoring systems. This inc% des attention to l

performance testing of survey instruments and personnel monitoring systems as well as bioassay procedures.

We endorse these efforts.

In terms of nuclear power plant personnel, consideration also needs l

to be given to studies to determine the optimum staffing from the standpoint of maintenance, operation, and minimization of individual and collective doses.

In view of the need for greater emphasis on the control of collec-tive occupational doses and on the associated newly proposed QA/QC programs, we recommend that funding for this program be increased.

6.5 Emergency Preparedness Included in this program area are evaluation and testing of emer-gency instrumentation for assessing nuclear power plant radiation levels and releases under accident conditions; evaluation of public l

warning systems; study of human factors affecting the respons'e of I

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the plant staff and the general public during an emergency; the effect of site selection on emergency preparedness; and the efficacy l

of countermeasures, recovery, and mitigation actioris to be applied in the event of an emergency.

l Although research in this program he been considerably expanded in recent years, several areas remain tnat need to be addressed in a l

more vigorous manner.

These include the assessment of groundwater contamination in the event of a serious breach of containment l

(currently proposed at an FY 1983 funding level of $0.1 million),

l and the development of methods for the restoration of lands contami-nated in the event of accidental radionuclide releases.

Both of these subjects have implications relative to the siting of nuclear power plants.

I Although we accept the proposed reduction in the requested level of NRC funding for this program for FY 1983, we do so only with the understanding that much of the research previously conducted by NRC in this subject area will now be handled by the Federal Emergency Management Agency (FEMA).

We urge the NRC to support FEMA in its request for funds to assume these activities.

l 6.6 Safeguards l

l This program is composed of three segments:

(1) Materials Control and Accounting, (2) Physical Security, and (3) Threat and Strategy.

i l

6.6.1 Materials Control and Accounting Funding for this program will be drastically reduced.

Emphasis l

will continue to be placed on improved material management capabil-ities including increased attention to determining the amount of material held up in processing equipment.

6.6.2 Physical Security l

The emphasis of the effort for physical security will be devoted to applying te:hniques already developed for use in the licensing and regulatory process.

Emphasis will be placed on special studies l

which will help to provide answers to safeguard problems and to l

resolve safeguard issues.

6.6.3 Threat and Strategy The work in this area is to develop appropriate responses to threats or appropriate actions in the event of successful sabotage or theft.

l Human factors aspects of the interactions between the adversary and safeguards systems will be addressed.

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l We recommend that the work in this Decision Unit include those studies necessary for the NRC to be able to develop requirements with regard to design to protect against sabotage by an insider.

This program should, if possible, be completed by the end of FY 1983.

6.7 Recomendations The funding level for this Decision Unit should be increased by

$3.0 million, from $16.8 million to $19.8 million, with emphasis as obtlined above.

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7.

WASTE MANAGEMENT 7.1 Introduction This Decision. Unit includes research on the safety problems of handling and ultimate disposal of high and low levei radioactive wastes and uraninun mill tail ings.

The safe disposal of all of these types of wastes has been and continues to represent a major public concern in the application of nuclear energy for large-scale power generation.

7.2 High Level Waste The objectives of this program are to identify failure mechanisms that affect long-term waste isolation capability; to identify technical requirements that may be needed to mitigate the conse-quences of accidental or unplanned movement of radionuclides; and to l

define the uncertainties or confidence levels in the data, analyti-l cal methods, and predictions for each of these areas of concern.

l One of the major controversies in determining criteria for assuring the retention of deposited radioactive wastes over a period of thousands of years is the lack of definitive data to verify pro-jected release and migration rates.

We are encouraged by plans of the NRC Staff to gain insight into these problems through studies of the migration of naturally occurring radionuclides in various soil structures.

Full advantage should be taken of this approach.

l Other studies deserving emphasis include the development of techni-ques for monitoring conditions w' thin a repository in which waste containers have been placed, evaluations of the potential impacts of l

short-term climatic changes on waste isolation factors such as the direction and rate of groundwater flow, and an assessment of the requirements for the disposal of transuranic wastes (these need not necessarily require the same approach as proposed for application to high level wastes).

We also urge that increased attention be i

directed to research to provide basic technical data, in addition to research of an applied or confirmatory nature.

In the past we have supported the high priority given to this program by the NRC Staff.

We believe that research work on high level waste handling and disposal should be vigorously pursued so that the necessary technical information is made available on a timely basis for decisions regarding licensing and regulatory activities.

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I 7.3 Low Level Waste In NUREG-0657, NUREG-0699, and NUREG-0151, we emphasized the need for sufficient research work to expedite the licensing and regula-tion of the handling and disposal of low level radioactive wastes.

We reiterate tnat position for FY 1983.

The existing situation mandates the selection of new disposal sites within the near future.

Research related to the development of criteria for judging accept-ability of such sites should be expedited.

Acquisition of the technical data to support development of a proposed rule on Low Level Waste Disposal is extremely important.

This should be given priority over efforts directed to solving problems at existing low level waste disposal sites.

Some research effort should also be directed to exploration of the possible disposal of low level wastes in sites at an intermediate depth below the earth's surface, as contrasted to shallow land burial.

A problem area not apparently being addressed in the research program currently proposed is the treatment and disposal of wastes resul ting from the decontamination of systems within operating reactors and of facilities, such as THI-2, that become contaminated as a result of an accident. Because of the possible presence of chelating agents, such wastes may require special handling and disposal procedures. We recommend that these problems be addressed.

The volumes of low level radioactive wastes currently being gen-erated at commercial nuclear power plants are far larger than they need to be.

Inis is because such wastes contain large amounts of s

nonradioactive wastes as well as void spaces.

Al though one ap-proach to solving this problem is to locate additional waste dis-posal sites, we believe that more attention needs to be addressed to solving this problem at its source.

7.4 Uranium Recovery The disposal of uranium mill tailings which resul t from uranium recovery and concentration operations has long been a public concern.

We agree that the work to develop criteria for dealing satisfac-torily with the large number of existing uranium mill tailings and to provide early guidance for the licensing and regulation of new mills warrants the amount of funding requested.

In reviewing specific research projects in this subject area, we were disappointed in the lack of studies directed toward solving the basic problem as contrasted to those designed merely to ameliorate releases from existing tailing piles.

38

7.5 Recommendations l

The proposed budget for research on Waste Management for FY 1983 is

.less than that proposed for FY 1982, with a major reduction being 4

made in the area of H10h Level Waste.

We believe that with proper rearrangement and consolidation, coupled with some shifting of funds among subel emen ts, the proposed reduction can be accommodated.

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I t

39

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8.

SITING AND ENVIROMEllT 8.1 Introduction Research in this program area includes studies of the Earth Sciences (i.e.,

seismology, geology, meteorology, and hydrology), Si ting, Health Effects, and Environmental Impacts.

8.2 Earth Sciences This. program is devoted to research in the areas of seismology, geology, meteorology, and hydrology.

We believe that these studies are of considerable importance to the development of a new rule on Siting (10 CFR 100), to the improved seismic design basis for future nuclear facilities, to an improved assessment of the seismic safety of existing facilities, to improved techniques for the evaluation of the potential effects of flooding an nuclear power plants, to the improvement of the models used in the CRAC Code, and to several other imnortant problems facing the NRC.

A major portion of this work.13 devoted to developing a better understanding of the seismic and geologic behavior of several important regions of the United States.

We continue to support this work and believe it is respon-sive to our previous recommendations.

8.3 Siting This program pertains to research on improving methods for evaluat-ing alternate sites vor nuclear f acilities, on obtaining additional information on the relationships of population density, land-use patterns, and siting alternatives for incorporation into the nuclear facility site-selection process, and on providing validated techni-cal bases for standards and siting criteria. This work will include studies to provide input for developing criteria for assessing single versus multi-unit sites, for evaluating changes in land-value and -use patterns and population density, and for providing the wide ranging types of information necessary for support of the rulemaking on Siting.

As would be expected, work in several other Decision Units will provide data applicable to nuclear facility siting problems.

This includes improvement in the CRAC Code, advances in the development and application of risk assessment techniques, the development of minimum requirements for engineered safety features, and the estab-lishment of safety goals.

40 f

1

i 4

Because of the fundamental importance of siting in assuring adequate protection of the health and safety of the public', we believe this program fully justifies the' increased funding proposed - for FY 1983.

8.4 Health Effects The objectives of this program are to improve the-understanding of the relationship between exposure to radiation and the magnitude of the ' biological effects produced, to provide information on the metabolism of inhaled and ingested compounds containing radio-nuclides not previously investigated, to improve the methodology for predicting deaths and illnesses as a result of radiation exposures, and to evaluate the effectiveness of protective actions and of instruments to detect and measure radiation. Some of this work will-help support the current e.ffort *o evaluate, for possible incorpora-tion into NRC regulations, the ommendations of the International Commission on ~ Radiological Protection (ICRP) as expressed in its Publications 26, 29, and 30 (Refs. 9-11).

The proposed research on the biological effects of neutrons is needed to evaluate the signif-icance of neutron exposures at nuclear power plants.

Although, in the past, studies have been directed to the development of techniques for increasing the removal of radionuclides ingested or inhaled by radiation workers, we note that such efforts are no longer included in this program.

We recommend that consideration be given to correcting this deficiency.

4 In NUREG-0751, we recommended that the work in this program be reviewed, evaluated, and coordinated with the National Institutes of Health Research Committees, the dational Council on Radiation Protection and Measurements (NCRP), the National Academy of Sciences, and other federal agencies (e.g., DOE and EPA).

We are encouraged to see progress in this area, especially the close cooperation being developed with the NCRP.

8.5 Environ 3 ntal Impacts This program includes laboratory and field studies to provide a better understanding of the movement of radioactive releases through aquatic environments, including their transport in rivers and in coastal zones, and their ultimate passage through ecosystems and food chains to people.

Associated studies include the development of mathematical models to simulate such transport, including the dispersion and diffusion of radionuclide effluents and the asso-ciated effects of sediment deposition and resuspension.

Al so included is research to provide more quantitative methods for predicting the socioeconomic impacts of such releases, both-under normal conditions and accidents.

Such work is required to provide 41 m

~.

l l

l the NRC with the necessary tools for assessing the environmental impacts of nuclear power plants as required by NEPA and to evaluate and incorporate into NRC regulations newer recommendations as published by the ICRP and the NCRP.

+

8.6 Recommendations Overall, we believe that the proposed level of funding for this Decision Unit is appropriate.

0 42

9.

SYSTEMS AND RELIABILITY ANALYSIS 9.1 Introduction This Decision Unit now has a somewhat changed scope from that included in the Decision Unit, Systems and Reliability Analysis (SARf), as discussed in NUREG-0751.

Human error data analysis is now fomally included in the new Decision Unit, Facility Operations and Safeguards, rather than in SARA, and the program on Transporta-tion and Materials Risk has been moved into SARA.

In NUREG-0751, we recommended a change in emphasis and a consider-ably expanded research program in this Decision Unit for FY 1982, as follows:

"We recommend that this Decision Unit be allocated at least $23.9 million for FY 1982, an increase of $9.0 million over the proposed budget, and that this additional funding be allocated approximately as follows:

(a) A large increase in emphasis and resources for the task on alternate decay heat removal systems, including consideration of sabotage, enabling enough effort to provide a basis for regulatory decisionmaking no later than the end of FY 1982

($2.0 million).

(b) A very considerable acceleration in the development of informa-tion needed to estimate the likely effect on risk of various potential design changes intended to mitigate accidents leading to severe core damage or core melt in LWRs ($1.25 million).

(c) The development of accident precursor screening techniques and their extensive application to the existing operating plants

($1.0 million).

(d) The early development of.a focused, cohesive program to provide the information needed.to determine the appropriate regulatory approach to control systems and to information needs of the reactor operator ($1.0 million).

(e) Critical review and evaluation of probabilisitic analyses and risk studies perfomed by licensees and construction permit holders ($1.0 million).

(f) An examination of possible weaknesses in the current applica-tion of the single failure criterion, and the early development of an improved approach ($0.5 million).

43 i

e

(g) The development of a basis for an improved approach to minimiz-ing significant design errors ($0.5 million).

(b) A systematic. approach to possible design steps to reduce the potential for serious accidents which might be caused by sabotage by an insider ($0.5 million).

(1) A program to better define property damage from accidents involving large releases of radioactive materirls, including the effect on societal resources ($0.5 million).

(j) A critical evaluation of the merits of LWR regulatory require-ments in other countries which differ significantly from those of the NRC ($0.5 million).

(k) The early development of quality assurance (QA) criteria for probabilistic analyses to be used in the regulatory process

($0.25 million).

We recommend that the matters listed above be given priority, even if it means reducing the funding for other programs, ongoing and proposed, in this Decision Unit."

The NRC Staff has described programs which indicate that an in-creased emphasis will be given in SARA or in other Decision Units to many of the items listed above, particularly items (b), (c), (e),

(f), (g), and (k).

Ilowever, on most of these items, the level of support in FY 1983 is much less than that recommended by us in FY 1982.

The FY 1983 program provides a negligible or far from ade-quate effort on items (d), (h), and (j) and for a reasonable amount of effort on item (1).

With regard to item (a), a substantial effort on decay heat removal systems is contemplated as part of the Task Action Plan A-45; however, this effort is aimed primarily at existing plants.

We believe that the DRA should participate significantly in this work and in addition should carry on a more general investigation, the results of which can contribute to setting design requirements for future plants.

NRR has stated its plans to build up its capability to apply prob-abilistic methodology to iicensing matters, a step we strongly support.

However, we do not believe that the buildup of this capability in NRR will be swif t enough or large enough by FY 1983 that there will not remain a continued need for DRA to apply probabi-listic risk assessment (PRA) methodology during this time period, especially in a peer review mode.

Furthermore, our recommendations for expanded research programs in the SARA Decision Unit generally do not fall in the category of licensing applications of PRA methodo-logy, but rather involve the use of PRA methodology to develop 44

r l

information whicn can form a part of the basis for generic regula-tory decisionmi kin).

Hence we believe that funding greater than that recommended by the E00 should be provided for SARA in FY 1983.

We recommend also that positive steps be taken by RES management to I

ensure that the needed close collaboration between DRA and the other research divisions occurs.

I 9.2 Risk Methods,and Data Evaluation l

We support the overall level of funding proposed for this program but recommend ome reorientation of the work.

We believe that the effort which relates to common cause and external events should be substantially increased.

We recommend also that efforts be insti-tuted on methods to enable inclusion in PRAs of the effects of l

design errors, sabotage and externally and internally produced l

flooding.

We recommend that the DRA should be asked to review and critique the SSMRP program which is funded under Decision Unit 5,

" Reactor and Facility Engineering."

We also recommend that work be i

initiated on improved mechods for analyzing reliability requirements for control systems and on additional screening techniques for accident precursors.

9.3 Reactor Risk and Reliability Analysis We endorse and recommend much increased support for the proposed i

l research in support of a Degraded Core Cooling Rule and on a much broadened interpretation of minimum engineered safety features. We also support the proposed research on regulatory analysis, on IREP/NREP, on accident sequence analysis, and on consequence analy-sis.

As discussed in Section 9.1, we believe that there are several items that are missing from the program or are much too weak.

We are recommending additional funding for this program specifically to deal with those items and to provide expanded support for the major rulemakings referred to above.

l One of the key items in the codes used for assessment of accident l

consequences is the model used to estimate the impacts on the l

neighboring population.

Currently, the mainstay of the analyses l

used for this purpose is the CRAC Code.

Although ef forts are i

being made to improve this Code, we are concerned about the pace of progress.

Since this Code is important to emergency preparedness, probabilistic risk assessment, and the support of the rulemaking on l

Siting, we believe that efforts for its improvement should be expedited.

At the present time, the Code is deficient in terms of 45

4 l

the range, depth, and flexibility of countermeasures that can be l

considered, and in terms of the assessment of the health effects of various exposure rates and levels.

There is also a need to develop an ancillary code for evaluating the impacts of releases via the liquid pathway.

We recommend that this work be given the needed support.

9.4 Transportation and Materials Risk This program includes research on risks in the fuel cycle, in transportation of irradiated fuel, and from nuclear materials.

We believe that, while work in this area is useful, it is generally of lower priority than the proposed research in the program on " Reactor Risk and Reliability Analysis" (Section 9.3), as well as many of the recommendations presented in Section 9.1 of this report.

9.5 Recommendations We recommend that the FY 1983 funding for the SARA Decision Unit be increased by $4 million, from $21.; million to $25.7 million, and that these adoitional resources be used to implement our recommenda-tions.

4 v

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i APPENDIXES i

i 47

9

.i L

~

APPENDIX.A L

l REFERENCES i

1. ~ Advisory Comittee on Reactor Safeguards. U.S. Nuclear Regula-tory ' Commission,' " Review and Evaluation of the Nuclear Regular tory Commission Safety Research Program for Fiscal Year 1982 -

A-Report to the-Congress of the United States.of America,"

NUREG-0751, - February 1981.*

2.

Advisory Committee on Reactor Safeguards, U.S. Nuclear Regula-tory Commission, " Comments on the NRC Safety Research Program Budget. for Fiscal Year 1982," NUREG-0699, July 1980.*

3..U.S.

Nuclear Regulatory Commission, " Unresolved Safety-Issues Summary," NUREG-0606, Vol. 3, No. 2, May 15,1981.*

4.

U.S.. Nuclear Regulatory Commission, " Technical Bases for-Estimating Fission Product Behavior During LWR Accidents,"

l NUREG-0772, March 1981.*

5.

Advisory Committee on Reactor Safeguards, U.S. Nuclear Regula-tory Commission, " Review and Evaluation of the Nuclear Regula '

i tory Commission Safety Research Program for Fiscal Year 1981 -

A ' Report to the Congress of the United States of America,"

NUREG-0657, February 1980.*

6.

U.S.

Nuclear Regulatory Commission, Regul atory Guide

1. 9 7,.

' Revision 2,

" Instrumentation for Light-Water-Cooled Nuclear

. Power Plants to Assess Plant and Environs Conditions During and Following an 1.ccident."

7.

U.S.

Nuclear Regulatory Comission, " Occupational Radiation.

Exposure at Comercial Nuclear Power Reactors,1979,"

NUREG-0713, Vol. 1, March 1981.*

8.

U.S. Nuclear Regulatory Commission, "Long Range Research Plan -

Fiscal Years 1983 - 1987,"' NUREG-0740, March 1981.*

9.

International Commission on Radiological Protection,."Recom-mendations 'of the ICRP," Publication 26, Annals of the ICRP Vol. 1, No. 3, 1977.

10.

International Comission on Radiological Protection, " Radio-nuclide Release into the Environment:

Assessment of' Doses to Man," Publication 29, Annals of the ICRP Vol. 2, No.1,1979 11.

International Comission on Radiological Protection " Limits for Intakes of Radionuclides by Workers," Publication 30, Annals of I

i the ICRP, Vol. 2, No. 3/4, Part 1, 1979.

j

• Available for purchase from the NRC/GP0 Sales Program, U.S. Nuclear I

Regulatory Comission Washington DC 20555, and/or the National Technical Information, Service, Spr,ingfield, VA 22161.

49 L

h APPLNDIX B GLOSSARY l-L ATWS Anticipated Transients Without Scram t

BWR Bo!1tng Water Reactor CCTF Cylindrical Core Test Facility CRAC Calculation of Reactor Accident Consequences CRBR Clinch River Breeder Reactor L2CCA Deformed Core Coolability l

DOE Department of Energy l

DRA Division of Risk Analysis, RES ECC Emergency Core Cooling EDO Executive Director for Operations EPA Environmental Protection Agency FEMA Federal Emergency Management Agency FIST Full Integral' Simulation Test FRG Federul Republic of Germany FY Fiscal Year I

'GCR Gas-Cooled Reactor I

HTGR High Temperature Gas Cooled Reactor ICRP International Commission on Radiological Protection IREP Interim Reliability Evaluation Program LER Licensee Event Report LMFBR Liquid Metal Fast Breeder Reactor 51

LOCA Loss-of-Coolant Accident LOFT Loss of Fluid Test LSlf Large Scale Test Facility LWR Light-Water Reactor MARCH Meltdown Accident Response Characteristics Code NCRP National Council on Radiation Protection NDE Nondestructive Examination l

NEPA National Environmental Policy Act NRC Nuclear Regulatory Commission NREP National Reliability Evaluation Program NRR Office of Nuclear Reactor Regulation ORNL Oak Ridge -National Laboratory PBF Power Burst Facility l

PCI Pellet-Clad Interaction PKL Test Facility In Germany designed to model plant systems behavior during Loss-of-Coolant Accidents and Transients PRA Probabilistic Risk Assessment l

PWR Pressurized Water Reactor l

l QA/QC Quality Assurance / Quality Control RELAP-5 Advanced System Code used to model Loss-of-Coolant Accidents RES Office of Nuclear Regulatory Research SARA Systems and Reliability Analysis SCTF Slab Core Test Facility 52

SFD Severe Fuel Damage SSMRP Seismic Safety Margins Research Program THTF' Thermal Hydraulic Test Facility TLTA Two Loop Test Apparatus TMI-2 Three Mile Islands Unit 2 TRAC Transient Reactor Analysis Code UPTF

. Upper Plenum Test Facility 6

53 i

U.S. NUCLE AR R500LATORY COMMISSION y

BIBLIOGRAPHIC DATA SHEET NUREG-0795

4. TITLE AND SUSTITLE (Ader 1/odunne Na, // apprapnes; 2.(Leavee/,,e) l Comments;on the NRC Safety Research Program Budget for Fiscal Year 1983 -

3. REC 6PIENT'S ACCESSION NO.

7. AUTHORISI

5. DATE REPORT COMPLETED Advisory Committee on Reactor Safeguards July

$81

9. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS (tac /ude 2,p Codel DATE REPORT ISSUED

~ MONTH l YEAR

(

Advisory Committee on Reactor Safeguards July 1981 U.S. Nuclear Regulatory Commission (t,,,,,,,,i; j

Washington, DC, 20555 I

a. (Leave Nenki 12, SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (include les Codel

10. PROJECTITASK/ WORK UNIT NO.

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.Same as 9,, above*

i t. CONTR ACT NO.

13. TYPE OF REPORT PE Rico cove RE D (inclusere darrs)

15. SUPPLEMENTARY NOTES

14. (Leave etash/

16. ABSTRACT 000 words or sess)

Recommendations of the Advisory Committee on Reactor Safeguards are presented to the Commissioners for their consideration for FY 83 budget for the NRC safety research program.

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17. KEY WORDS ANO DOCUMENT ANALYSIS 17a DESCRIPTORS l

17b. IDENTIFsE R$iOPEN ENDED TERMS

16. AVAILABILITY STATEMENT

19. SECURITY CLASS (Thas re00rr)

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