ML19263B599
| ML19263B599 | |
| Person / Time | |
|---|---|
| Issue date: | 12/31/1978 |
| From: | Lawroski S Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| NUREG-0496, NUREG-496, NUDOCS 7901220150 | |
| Download: ML19263B599 (97) | |
Text
NUREG-0496 1978 REVIEW AND EVALUATION OF THE NUCLEAR REGULATORY COMMISSION SAFETY RESEARCH PROGRAM A Report to the Congress of the United States of America e,.
Advisory Committee on Reactor Safeguards U. S. Nuclear Regulatory Commission 790122oi9
Available from National Technical Information Service Springfield, Virginia 22161 Price: Princed Copy $6.00; Microfiche $3.00 The price of this document for requesters outside of the North American Continent can be obtained from the National Technical Information Service.
i NUREG-0496 1978 REVIEW AND EVALUATION OF THE NUCLEAR REGULATORY COMMISSION SAFETY RESEARCH PROGRAM A Report to the Congress of the United States of America Manuscript Completed: December 1978 Date Published: December 1978 Advisory Committee on Reactor Safeguards U. S. Nuclear Regulatory Commission Washington, D. C. 20555
-a p*
- E c o UNITED STATES y,
,g y ', '. -
l p, NUCLEAR REGULATORY COMMISSION 3
j ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g
WASHINGTON D. C. 20555 December 29, 1978 The Honorable Walter F. tbndale The President of the Senate t e Honorable Romas P. O'Neill, Jr.
The Speaker of the House Gentlemen:
I am pleased to transmit herewith the Advisory Committee on Reactor Safeguards' 1978 report to the Congress on the safety research program of the Nuclear Regulatory Commission.
%is report is required by Section 29 of the Atomic Energy Act of 1954 as amended by Section 5 of Public Law 95-209.
A copy of this report is being sent to the Chairman of the Nuclear Regulatory Corutission.
Respectfully submitted, Stephen Lawroski Chairman
TABLE OF C0tTPENTS Chapter Page S-1 Executive Summary.................................
1.
Introduction and Recommendations...................
1-1 2.
Ioss-of-Coolant Accident / Emergency Core Cooling....
2-1 Systems (LOCA/ECCS) 3.
Fuel Behavior......................................
3-1 Integrity..........................-
4-1 4.
Primary Syst s
5.
Operational Safety.................................
5-1 6.
Advanced Reacto r Sa fety............................
6-1 7.
Extreme External Phenomena.........................
7-1 8.
Radiological Effects...............................
8-1 9-1 9.
Waste Management...
10.
Sa fegua rds and Secur ity............................ 10-1 11.
Risk Assessment................................
11-1 12-1 12.
Improved Rewtor Safety Appendix A: Methodology o f ACRS Study..................
A-1 Appendix B: Glossary...................................
B-1 Appendix C: ACRS Charter and Membersnip................
C-1 v
7
EXEClJrIVE
SUMMARY
1.
Introduction and Recortynendations We first review by the Advisory Committee on Reactor Safeguards (ACRS) of the safety research program of the Nuclear Regulatory Com-mission (NRC) was conducted in 1977 and a report submitted to the Congress in December 1977.*
As in its 1977 Report to the Congress, the ACRS has interpreted the words " reactor safety research" as used in Section 5 of Public Law 95-209 to include safety-related research in all phases of the nu-clear fuel cycle.
We objective of the NRC safety research program has been broadened to include research on improved safety concepts in addition to con-firmatory research related strictly to the NRC's regulatory func-tion. The scope of the current F Wram is indicated by the estimated expenditures shown in Table 1.
Each program area listed in this table is the subject of a chapter in this report and of a section in the Executive Summary.
We folloving observations parallel and are similar to those made in the 1977 Report:
W e research program amounts to about one-half of the total NRC budget.
%is amount is not excessive in view of the importance and highly technical nature of the questions being addressed.
Expenditures for FY 79 are expected to be about 18 percent higher than for FY 78, but the major portion of this in-crease results from the transfer of operating expenses for the loss-of-Fluid-Test (LOFT) Reactor from the Department of Energy (DOE) to the NRC.
- Advisory Comr. ittee on Reactor Safeguards, Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program, NtM m-0392, December 1277.
S-1
TABLE 1 ESTIMATED RESEARCH EXPENDITURES FY 78 FY 79 (in millions)
PROGRAM SUPPORT IOCA/ECCS
$48.6
$67.2 Fuel Behavior 24.1 23.6 Primary System Integrity 8.0 10.1 Operational Safety 1.5 1.9 Advanced Reactors 15.7 15.0 Extreme External Phenomena 5.1 6.E Radiological Effects
- 7. 3 7.7 Waste Management 4.2 4.2 Safeguards and Security 6.6 6.2 Rick Assessment 3.1 3.4 Improved Reactor Safety 0.0 Subtotal
$124.2
$146.1 EQUIPMENT, PERSONNEL, ADMINISTRA-TIVE SUPPORT AND NON-SAFETY-RELATED RESEARCH**
14.3 17.4 Total
$138.5
$163.5 The NRC plans to reprogram approximately $0.8 million in FY 79 to initiate research in this area.
Includes non-safety-related environmental and socioeconomic re-search S-2
-i--,,--,------i.------.--i.-i-.i----i--
About 90 percent of the funds spent for research are for program support, that is, research contracted to outside organizations. We remaining funds are spent for persormel compensation and benefits, travel, equipnent and general administrative support.
We programs relating to the loss of coolant accident (LOCA) and emergency core cooling systems (ECCS) and to fuel behavior represent about 60 percent of the total ef-fort. %ese programs involve the use of large and expen-sive test facilities such as LOFT and Pcwer Burst Facility (PBF).
Specific comments regarding elements of each program are presented in th subsequent sections of this summary and in the body of the report.
We more general findings and recommendations are summarized below for each program area.
w'A/ECCS. A significant increase in funding will be re-quired in FY 80 to cover the planned operation of LOFT.
However, major reductions in expenditures are expected to result from completion of various projects within the next four to eight years and from the decommissionir.J of LOFT in the late 1980s.
%e ACRS concurs with the NRC decision that there is no need for the proposed ECC By-pass Test Facility, the large-scale multiple-purpose test facility, and a full-scale integral test facility.
An essential component of this research program is the in-dependent assessment of the best estimate codes and more definitive goals are needed for this effort.
Fuel Behavior.
The projects in this area are important and funding is at an appropriate level in view of the costs of operating PBF.
Since PBF is a unique and ex-tremely useful but expensive facility, the program of tests in PBF should continue to be closely monitored to assure that it meets clearly defined and justified regu-latory needs.
Primary System Integrity.
We Heavy Section Steel Tech-nology (HSST) project should be completed as planned.
We remainder of the program takes advant age of the large amount of research done by others.
Research on piping integrity, however, should receive a higher priority than other new wock being proposed.
S-3
Operational Safety.
his is a small program that could profitably be expanded to address the problems encoun-tered and the lessons learned from the many operating re-actors and to provide a broader and sounder data base for assessments of reliability and risk.
Advanced Reactor Safety.
%e projects in this area are important; the current levels of funding are marginal and modest increases for the next two or three years are ap-propriate.
Additional emphasis should be placed on iden-tifying and examining basic safety problems related to commercial-sized plants.
Extre;3 External Phenomena.
Earthquake-related research, which dominates this program, deserves a high priority.
The current level of funding is appropriate but will have to be increased in the years to come to support the seis-mic safety margins research program.
Radiological Effects.
his program is generally well fo-cused and progressing satisfactorily.
Increased attention is needed, however, to the development of methods for the compilation of data that can be utilized more effectively in evaluating possible relationships between occupational radiation exposures and various health affects.
Waste Management.
We problems in this area are important and will require significant amounts of research. W e cur-rent progran appears to be uncoordinated and unfocused and progress to date has been inadequate, at least in part be-cause of similar deficienciw in the national program. We immediate needs in this area are to establish a plan for the research and to improve internal management control.
Safeguards and Security.
We current program addresses both physical security of plants and transportation, and methods for control and accounting of special nuclear material.
We current levels of activity and funding are adequate.
Risk Assessment.
W e current research program is adequate for this stage in the development and use of probabilistic risk assessment.
%e ACRS recommends greater use of this methodology, but is concerned about the possible concentra-tion of that use in the Probabilistic Analysis Staff.
Growth in this area might better be srt ead throughout all segments of the research and licensing programs.
S-4
Improved Reactor Safety. W e ACRS considers this research on improved safe'.y concepts to be of high priority and recommends substantial funding ($1.5 million) in FY 79, by major reprogramming of other NRC funds if necessary, and at multimillion dollar levels in both NRC and DOE for complementary programs in FY 80.
It is evident from the above summary that the ACRS has deemed no major program area to be of low priority.
his is not to say, how-ever, that each program and its elements are being carried out in the most efficient manner and that all of the specific questions being addressed are of equal or even of high priority.
'Ihe ACRS believes that it would be possible in its 1979 review and report to develop a hierarchy of priorities within the various program areas and perhaps even across the entire safety research program.
In attempting to do this, the ACRS intends to evaluate each program element or subelement in terms of its relation to objectives such as:
To confirm that a proposed design may be licensed.
To understand and characterize complex phenomena in order that better licensing decisions can be made.
To establish the margins that exist in approved or pro-posed designs, and the sensitivity of perfonnance to specific system features.
To establish the reliability of important safety systems.
To disclose unexpected phenomena and interactions.
To examine approaches for potential improvements to re-actor safety.
To provide bases for the developnent of improved licens-ing criteria.
To maintain available a body of competent, knowledgeable experts to assist the NRC in 3 :s evaluation of complex technical proposals for licensing.
S-5
2.
Loss-of-Coolant Accident and Emergency Core Cooling Systems he loss-of-coolan;. accident (LOCA) and the emergency core cooling systems (ECCS) intended to mitigate its consequences cor:tinue to be important matters to the NRC, the nuclear industry, and to both the technical and lay public.
For licensing purposes, specific and con-servative rules for evaluating the adequacy of the ECCS were promul-gated after an extensive and exhaustive rulemaking hearing.
We research programs are directed at producing a reasonable quantitative evaluation of the safety margins embedded in the licensing processes.
Research on LOCA/ECCS was begun under the Atomic Energy Commission and is continuing at an increasing level under the NRC.
The NRC has proposed major reductions in funding within the next four to eight years as various projects in the program are completed.
The ACRS coxurs with the decision by the NRC that there is no need for the proposed ECC Bypass Test Facility, the large--scale multi-purpose test facility, and a full-scale integral test facility.
A cooperative program, involving large-scale test facilities, is being developed among tbc NRC and similar agencies in the Federal Republic of Germany and Japan.
We LOCA/ECCS program consists f two parts:
(1) experiments to study the physical phenomena invo Lved in LOCA/ECCS, and provide input for mathematical models, and additional experiments for independent assessments, and (2) the development of relatively complex computer codes to predict the course of a postulated LOCA and the realistic performance of the ECCS.
Although primary emphasis is on the large LOCA, the codes being developed provide a basis for investigating a range of postulated LOCAs and other off-normal transients.
Considering the progress being made and the status of the conserva-tive licensing processes, the ACRS believes that the LOCA/ECCS research should not continue to dominate the NRC research programs.
Nevertheless, the ACRS recommends that the large-scale tests of the international program and those planned for the next few years in LOFT be carried out.
We ACRS recommends further that the work on independent assessment of the best estimate codes be intensified.
The predictions of computer codes will never be perfectly accurate, and any attempt to make them so will require an unending expenditure of time and money. What is necessary now is for the NRC to determine what level of accuracy--or what level of uncertainty--will be ade-quate for its regulatory function.
The ACRS recommends that a quantifiable decision on this matter should be made within the next year in order to provide a yardstick against which the results of the S-6
LOCA/ECCS research program can be measured and to provide a basis for deciding when the program has been completed.
A clear-cut schedule for
'tination of the LOFT tests should be established.
Each major WFT t..st, where appropriate, should be preceded by blind predictions from best estimate codes; blind predictions should be made with the advanced code, TRAC, for the 3D-2D international tect series.
3.
Fuel Behavior his program is concerned with the behavior of the uranium oxide fuel and the zirconium alloy tubes (cladding) in which it is encased.
Wese tubes provide the first line of defense against the release of radioactive fission products.
%e research on fuel behavior is both analytical and experimental.
The analytical research is devoted to the develoirtent of computer codes to describe the behavior of the fuel under various abnormal conditions.
We experimental research involves tests on actual fuel rods (fuel plus clodding) in various test facilities, including the PBF in the U.S., the Halden Reactor in Norway, and the NRU Reactor in Canada.
The experimental program also includes studies of phenomena of potential importance to postu-lated core melt accidents including fuel-coolant inte raction, re-lease and transport of radioactive fission products from molten fuel, and the interaction between concrete and molten fuel.
All of these programs are important and producing valuable results.
We ACRS offers the following specific recommendations:
a) The high priority now assigned by the NRC Staff to experiments in PBF relating to reactivity-insertion accidents should be reassessed in view of the very low probability of such accidents.
b) The research program on phenomena significant to a core melt accident is important and shoul:1 continue.
4.
Primary System Integrity We primary system, comprised of the reactor pressure vessel, other vessels, and piping, contains the cooling water under high pressure.
Wis system constitutes the second line of defense against release of radioactivity to the environment.
Failure of the reactor pressure vessel is considered by the NRC to be of such low probability that it need not be postulated as a design basis accident.
Failure of the primary system piping, how-is considered to be more probable and is in fact assumed as
- ever, the design basis loss-of-coolant accident.
S-7
%e objective of research in this area is to assure that the proba-bility of failure in the primary system is suitably low.
In prac-tice, the desired high quality of the primary system is sought by utilizing appropriately conservative procedures in design, fabrica-tion, inspection, and operation.
A large amount of research in all of these areas is being carried out by organizations other than the NRC.
Except for its sponsorship of the Heavy Section Steel Tech-nology (HSST) project, the chief requirements of the NRC are to maintain a high level of cognizance of what is being done by others in this country and elsewhere, and to sponsor a moderate amount of research of its own to provide confirmation of and confidence in the results obtained by others.
%is program is important and deserves an overall 'i h priority.
In g
general, it is being conducted in a manner suitabu to the needs of NRC as mentioned above.
However, the following recommendations are made regardina relative priorities and emphasis within the program:
a) The HSST program should be completed as planned.
b) Research on quantification of piping reliability de-serves a high priority.
c) An expanded effort is needed to study the effects of coolant chemistry on the formation and growth of cracks in primary system components, d) Additional effort is needed to evaluate the findings from operating experience relating to possible satura-tion effects in radiation embrittlement.
e) The research program on steam generator tube integrity should be reviewed to assure that it is appropriately confirmatory in nature and does not unnecessarily duplicate the industry programs.
5.
Operational Safety This is a small program, carried out chiefly by the Research support Branch to address safety questions related to the operation of nu-clear power plants.
We current program includes research on quali-fication testing of equipnent that n,ight be exposed to hostile environments in the event of a fire or an accident, fire protection, noise diagnostic procedures, and the role of human errors in reactor safety, including the question of control-room design and man-machine interaction.
S-8
All of these problems, and others like them that will arise from time to time as more operating experience is obtained, are important but will vary in priority on an ad hoc basis.
The ACRS believes that the question of man-machine interface deserves a high priority.
Both the adverse consequences of errors made by people and the fa-vorable aspects of man's adaptability and ingenuity need further examination with regard to reactor safety.
The advantages and disadvantages of a greater degree of computer-controlled automation should be explored, as should the potential for computer-aided guidance to operators during anticipated events.
In addition, a more systematic review and evaluation of operational experience and operational incidents in U.S. plants and in similar plants in other countries should be undertaken in order to provide a broader and sounder data base for assessments of reliability and risk.
6.
Advanced Reactor Safety Advanced reactors include breeder reactors, high-temperature 7as-cooled reactors, and advanced converters such as the spectral-shic and heavy-water reactors.
Research on advanced reactor safety is unique in that no licen-4 actions relating to such reactors are now in progress.
Howev.
some form of advanced reactor is considered a viable source of power for the future.
Accordingly, it is prudent for the NRC to assune that such reactors will require licensing in the future and to be prepared to take such action when called upon.
This could involve the Clinch River Breeder Reactor at an early date, depending on decisions made at the national level, or could involve different types of reactors approaching commercial size, on a longer time scale.
Particularly for the latter case, it is important that basic safety research be pursued on a sound, logical, long-range basis which is devoid of sharp funding perturbations and which will lead to the necessary information in an efficient manner and on a timely basis.
'Ihe ACRS believes that a modest increase in funding of this program for each of the next two or three years is needed in order to permit implementation of recommendations such as:
a)
A broad-scale study should be made of the liquid-metal fast-breeder reactor (LMFER) to determine which types of accidents are the greatest contributors to risk in pl a-t.c of commercial size.
Depending on DOE inten-tions in the high-temperature gas-cooled reactor and gas-cooled fast reactor areas, similar studies may be
- ~
needed for gas-cooled reactors.
S-9
b) Greater consideration should be given to measures ifor preventing EMEBR accidents and to means for mitigat-ing their consequences.
c) Greater attention should be placed on keeping abreast of foreign safety research and foreign advanced-reactor licensing activities.
'Ihe study of the core-disruptive accident and associated problers should be continued.
However, related code-develognent work should not carry a priority higher than that of other work such as in (a) and (b) above.
7.
Extreme External Phenomena Extreme external phenomena include natural phenomena such as earth-quakes, tornadoes, hurricanes, floods, tsunamis, and lightning, and man-made phenomena such as aircraft crashes, turbine-generated mis-siles, and explosions. Research in this area is devoted chiefly to earthquakes, with a much smaller effort devoted to other natural phenomena; none of the current effort is devoted to the effects of man-made phenomena.
Because research on extreme external phenomena, and particularly earthquakes, addresses questions relating to the siting of all types of reactors and fuel-cycle facilities, it should be assigned a high priority in the NRC safety research program, and should be funded at increasing levels over the next few years 8.
Radiological Effects Research projects in this program include those related to the health effects of 1(;w-level radiation exposure, movement of radionu-clides through the environment, the control of accidental radio-nuclide releases, and the development of procedures for decontamina-tion and reentry.
'Ihe ACRS believes that the NRC research effort in this program generally is well focused and progressing satisfactority.
- However, the ACRS believes that increased attention should be directed to:
a)
Examination of the NRC data bank on human occupa-tional radiation exposures by qualified epidemiolog-ists, with a view toward developing data that can be utilized more effectively in evaluating possible relationshipo betwen radiation exposures and various health effects.
S-10
b) Research to develop a better underscanding of the basic factors that govern the buildup and control of radionuclides in reactor cooling systems.
c) Research to develop improved methods for, and the data base supporting, the calculation of radi-ation doses to population groups residing in the vicinity of nuclear facilities.
9.
Waste Management We primary cbjective of the radioactive waste management program should be to control and minimize, to the extent reasonably achiev-able, both the maximum individual doses and the collective population doses resulting from all aspects of the handling and ultimate dis-posal of radioactive wastes.
As a regulatory agency, the NRC has the responsibility to establish criteria to assure that the handling and disposal of radioactive wastes will be conducted in a safe manner.
The ACRS review of the NRC waste management program has shown that its ef'.ts toward achieving the above objectives are uncoordinated and unfocused.
This is due, at least in part, to similar deficien-cies in the national program. Within the NRC program, there appears to be a lack of systematic processes for identifying research needs in this field and for assigning priorities for their accomplishment.
There also appears to be a lack of adequate interaction and communi-cation among the several NRC groups involved.
Wis is particularly true in the communication of waste management research needs to the Office of Nuclear Regulatory Research.
There is a need also for better interaction and communication between the NRC and other governmental agencies having responsibilities in this area.
The ACRS recommends that the research on waste management emphasize the following areas:
a)
Identification of the dominant contributors to risk in radioactive waste nanagement operations, and quan-tification of the uncertainties in the risk esti-mates.
b)
Continued development of criteria necessery for licensing the design and operation of radioactive waste facilities, c) Continued development of licensing criteria to facilitate the decontamination and decommissioning of nuclear facilities.
S-11
Increased funding above current levels will be required in order for this program to meet its current objectives on a timely basis.
Inasmuch as related research on this subject is being conducted in several foreign countries, the ACRS recommends that the NRC Staff vigorously pursue cooperative programs with these groups.
10.
Safeguards and Security
'Ihe term safeguards refers chiefly to means to prevent the theft of special nuclear material (SNM) from fixed sites or during trans-portation. The term security refers principally to the protection of nuclear facilities from sabotage or takeover with a subsequent release or threat of relecse of radioactivity.
Most of the current problems related to safeguards and security are not new, and much work has been done on them by the Departnent of Energy and the Department of Defense, as well as by the NRC.
'ihe NRC's need for rt. search in these areas is to provide the bases for evaluating the effectiveness of procedures proposed or employed by licensees for preventing theft of SNM and for protecting materials in transit or facilities against sabotage.
Current research addresses both physical security systems and material control and accounting methods.
The ACRS believes that this program is appropriate and adequate, and makes the following specific recommendations:
a)
Plan fo r work on safeguards and security should provide for a program at about the present level of effort for at least the next several years, with some allowance for the possibility that it may be necessary to increase the level should the national policy call for early adoption of new fuel cycles or new reactor types.
b)
In one aspect of the current program, an attempt has been made to determine the minimum number of essential components in a nuclear power plant which, if fully protected, could enable the plant to be shut down safely even if all other components were sabotaged.
The ACRS believes that an extensive study of this matter is appropriate since, if successful, it could lead to greater assurance of safety, at less cost, for both old and new plants.
c)
It is recommended that studies be made to determine whether the use of alternative fuel cycles would change significantly the nature or importance of the types of safeguards measures now being studied.
S-12
%ese studies should include also an estimate of how soon new questions night arise and how long would be required to solve them.
d)
In connection with the developnent of computer codes directed at security problems, the NRC staff should give careful prior attention to the type of question for which the code might provide answers, the use to which such answers would be put, and the amount of effort likely to be needed to obtain them.
- 11. Risk Assessment Research on probabilistic risk assessment methodology and the use of that methodology as a research tool is the task of the Probabilistic Analysis Staff (PAS).
In addition, the PAS provides guidance, as-sistance, and instruction to other members of the NRC Staff in the uses of this methodology.
Because many of its activities are conducted in-house, the PAS is relatively large.
We amount of research being done by outside o rganizations is increasing but is limited to some extent by the number of people having the necessary skills.
W e ACRS offers the following findings and recommendations:
a) The current and proposed research prog rams are adequate for this stage in the developnent and use of probabilistic risk assessment.
b) The ACRS considers it especially important that the methodology of probabilistic risk assessment, and the insights to be gained from its use, be utilized to the greatest extent feasible in the planning of the NRC safety research program.
This was suggested by the Risk Assessment Review Group *, it is being done now in some cases, and its expanded use is en-couraged.
We ACRS is concerned about the possible concentration of risk-assessment activity in the PAS, and believes that a strong effort should be made to spread the capability and activity in this area to a greater extent throughout all seqnents of the re-search and licensing programs.
- H.W. Lewis, et al., Risk Assessment Review Group Report to the U.S. Nuclear Regulatory Comission (NUREIi/CR-0400), September 1978.
S-13
c)
Effort to increase the use of this approach should have a high priority because of its potential in-fluence on a large range of researc' activities.
d)
The ACRS considers the investigation of acceptable risk important. However, the ACRS reconnends that, in addition to the work being done within NRC, the many other Federal agencies with responsibility for setting safety standards also participate in a broadly based study of the question of acceptable risk.
12.
Improved Reactor Safety The basic purpose of this research is to investigate concepts that have the potential for improving safety of light-water reactors.
In response to a request in the FY 78 Budget Authorization Act, the NRC sufr:itted to the Congress on April 12, 1978 its Plan for Research to Improve the Safety of Light-Water Nuclear Power Plants (NUREG-0438).
This plan recommended five projects for immediate or early initiation, identified several other topics for scoping studies, and proposed to develop improved methodolcgy for evaluating re-search topics.
The ACRS endorsed this proposed research program in a letter dated March 13, 1978 to NRC Chairman Joseph M. Hendrie.
In its 1977 report, the ACRS pointed out the distinction between the research on improved safety concepts that should be done by the NRC and the DOE, as follows:
"It is both desirable and appropriate for the NRC to conduct research on new safety concepts, but their develognent and implementatic.i should be carried out by the nuclear industry or the Department of Energy."
The ACRS believes that this research program should be given high priority.
The ACRS considers it unfortunate that this NRC prcgram could not be initiated in FY 78 and recommends substantial funding
($1.5 million) in FY 79, by reprogramming of other NRC funds if necessary.
The ACRS recommends that in subsequent years this pro-gram be funded at the level needed to permit effective pursuit of all of the research projects and scoping studies proposed in NUREG-0438.
The ACRS recommends further that emphasis be given to the work on alternate containment concepts, on bunkered dedi-cated shutdown heat removal systems, on improved in-plant response to accidents or potential accidents, improved methodology for on evaluating research topics; and to scoping studies on the topics relating tn prevention or mitigation of offsite consequences re-sulting fr om postulated core melt accidents via liquid pathways and to possible design measures for protection against sabotage.
S-14
'1he ACRS believes that there are complementary roles for both NRC and DOE in research to improve light-water reactor cafety, and that aggressive programs at multimillion dollar funding levels should be pursued by each agency with appropriate coordination.
S-15
1.
INTRODUCTION AND RECOMMENIRTIONS 1.1 NRC Research Program he objective of the Nuclear Regulatory Commission (NRC) safety research program has been broadened to include research on improved safety concepts in addition to confirmatory research related strictly to the NRC's regulatory function.
The scope of the current program is indicated by the estimated expenditures shown in Table 1.
Each program area listed in this table is the subject of a chapter in this report.
It should be noted that the listing of program areas in Table 1 is different from that in the table of the 1977 Report to Congress *.
We new areas are believed to be more homogeneous in objectives, and were selected in consultation with the Office of Nuclear Regulatory Research.
For convenience of comparison, estimated expenditures are given for FY 78 as well as FY 79.
We following observations parallel and are similar to those made in the 1977 Report:
he research program amounts to about one-half of the total NRC budget.
This amount is not excessive in view of the importance and highly technical nature of the questions being addressed.
Expenditures for FY 79 are expected to be about 18 percent higher than for FY 78, but the major portion of this increase results from the transfer of operating expenses for the loss-of-Fluid-Test (LOFT) Reactor from the Depart-ment of Energy (DOE) to the NRC.
About 90 percent of the funds spent for research are for program support, that is, research contracted to outside
- Advisory Committee on Reactor Safeguards, Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program, NUREU-0392 Decemter 1977.
1-1
organizations. We remaining funds are spent for personnel compensation and benefits, travel, equipnent and general administrative support.
The programs relating t; the loss-of-coolant accident (LOCA) and emergency core cooling systems (ECCS) and to fuel behavior represeat about 60 percent of the total effort.
These programs involve the use of large and expensive test facilities such as LOFT and Power Burst Facility (PBF).
1.2 Objectives and Scope of ACRS Review he first review by the Advisory Committee on Reactor Safeguards (ACRS)* of the safety research program of the NRC was conducted in 1977 and a report submitted to the Congress in December 1977.
As in its 1977 Report. the ACRS has interpreted the words " reactor safety research" as used in Section 5 of Public Iaw 95-209 to include safety-related research in all phases of the nuclear fuel cycle, excluding only that having to do with non-radiological envirorrnental and socioeconomic concerns.
Althotr3h the ACRS had on numerous occasions reviewed specific por-tions of the NRC (or its predecessor, the Atomic Energy Commission) research programs, and had made specific recommendations to the NRC or its predecessor, it had not previously made a comprehensive review of the entire program. For this reason, the 1977 study was the first attempt by the ACRS to conduct a comprehensive review of the total safety research program.
W e projects in each area were examined in cetail to determine how they served the needs of the regulatory process, how they related to research being done by others, and the nature, timeliness, and usefulness of the results being obtained.
he 1978 ACRS review has concentrated more heavily on new or large or important projects, and other projects were reviewed only to determine what progress was being made.
We areas selected for this in-depth review, and the reasons therefor, are as follows:
LOCA/ECCS research, because of its cost, its long history, and its projected future, and because of recent proposed changes in objectives and means for achieving them.
- The Charter and Membership of the ACRS are given in Appendix C.
1-2
Advanced reactor safety research, because of its importance to the safety of types of reactors quite different from conventional light-water reactors and offering potentially quite different safety-related problems, and because of the effects of uncertainties in future developnents on the level and pace of research in this area.
Waste management research, because of the continuing, intense interest in this subject, and because of the need to define, in the near future, the information necessary for the establishment of adequate regulatory criteria on a timely basis.
Earthquake-related research, because of its importance in the siting and design of all types of nuclear facili-ties, and because of the initiation of a new and ambitious research program.
Risk assessment research, because of the growing importance of this methodology as a new tool in the regulatory process and in the research program.
1.3 Overall Assessment and Priorities 1.3.1 Recommendations of the 1977 Report In its 1977 Report, the ACRS found that the NRC safety cesearch program was, in general, responsive to regulatory needs. Within each program area, the judgment was similar, but suggestions or recommen-dations were made for specific changes in enphasis, scope, or proce-dures.
Although the ACRS made no specific recommendations for major reductions in any of the programs, the decision by the NRC not to proceed with the development of a large-scale ECC Bypass Test Facil-ity resulted chiefly from questions raised by the ACRS and its consultants during its review fot' the 1977 Report.
As a result, the conclusion of the ACRS review and report was that the programs and support levels for FY 78, and those proposed for Fi 79, were at appropriate levels, even though several recunmendations were made for moderate expansions in certain areas.
1.3.2 Recomendations of this Report No explicit attempt was made in the 1977 review and report to assign priorities within or among the major program areas.
'Ihat is, no attempt was made to determine which projects should be discontinued 1-3
if the research budget was decreased by a significant amount.
For its 1978 review, the ACRS selected certain program areas for in-depth review, as described in Section 1.2.
None of the major program areas has been found to be without merit, although they vary in content, size, and effectiveness.
Specific comments regarding elanents of each program are presented in the subsequent chapters of this re-port.
We more general findings and recommendations are summarized below for each program area listed in Table 1.
LOCA/ECCS.
A significant increase in funding will be required in FY 80 to cover the planned operation of LOFT.
However, major reductions in expenditures are expected to result from completion of various projects within the next four to eight years and from the decommissioning of LOFT in the late 1980s.
The ACRS concurs with the NRC decision that there is no need for the proposed ECC Bypass Test Facility, the large-scale multiple purpose test facility, and a full-scale integral test facility. An essential component of this research program is the inde-pendent assessment of the best estimate codes and more definitive goals are needed for this effort.
Fuel Behay_ior.
The projects in this area are impor-tant.nd :unding is at an appropriate level in view of the c of operating the PBF.
Since PBF is a unique and en
'y useful but expensive facility, the program of tests
- d. PBF should continue to be closely monitored to assure that it meets clearly defined and justified regulatory needs.
Primary System Integrity.
We Heavy Section Steel Tech-nology (HSST) project should be completed as planned. We remainder of the program takes advantage of the large amount of research done by others.
Research on piping integrity, however, should receive a higher priority than other new work being proposed.
Operational Safety.
This is a small progr'n that could pr:fitably be expanded to address the problems encountered and the lessons learned from the many operating reactors and to provide a broader and sounder data base for assess-ments of reliability and risk.
Advanced Reactor Safety.
The projects in this area are important; the current levels of funding are marginal and modest increases for the next two or three years are appropriate.
Additional emphasis should be placed on 1-4 s-
identifying and examining basic safety problans related to commercial-sized plants.
Extreme External Phenomena.
Earthquake-related
- research, which dominates this program, deserves a high priority. %e current level of funding is appropriate but will have to be increased in the years to come to support the seismic safety margins research program.
Radiological Effects.
This program is generally well focused and progressing satisfactorily.
Increased atten-tion is needed, however, to the developnent of methods for the compilation of data that can be etilized more effec-tively in evaluating possible relationships between occupa-tional radiation exposures and various health effects.
Waste Management. We problems in this area are important and will require significant amounts of research.
We current program appears to be uncoordinated and unfocused and progress to date has been inadequate, at least in part because of similar deficiencies in the nacional progran.
We immediate needs in this area are to er,tablish a plan for the research and to improve internal management con-trol.
Safeguards and Security.
%e current program addresses both physical security
',f plants and transportation, and methods for control ar.d accounting of special nuclear material.
We currcnt levels of activity and funding are adequate.
Risk M sessment.
We current research program is adequate for this stage in the develognent and use of probabilistic risk assessment.
W e ACRS recommends greater use of this methodology, but is concerned about the lpossible concen-tration of that use in the Probabilistic Analysis Staff.
Growth in this area might better be spread throughout all segments of the research and licensing programs.
Improved Reactor Safety.
We ACRS considers this research on improved safety concepts to be of high priority and recommends substantial funding ($1.5 million) in FY 79, by major reprogramming of other NRC funds if necessary, and at multimillion dollar levels in both NRC and DOE for complementary programs in FY 80.
1-5
~
1.3.3 Future Assignment of Priorities It is evident from the foregoing that the ACRS has deemed no major program area to be of low priority.
Wis is not to say, however, that each progra.n and its elements are being carried out in the most efficient manner and that all of the specific questions being ad-dressed are of equal or even of high priority.
We ACRS believes that it would be possible in its 1979 review and report to develop a hierarchy of priorities within the various program areas and grhaps even across the entire safety researen program.
In attempting t.c do this, the ACRS intends to evaluate each program element or subelement in terms of its relation to objec-tives such as:
To confirm that a proposed design may be licensed.
To better understand and -haracterize complex phenomena in order that better licensing decisions can be made.
To establish the margins that exist in approved or proposed designs, and the sensitivity of performance to specific system features.
To establish the reliability of important safety systems.
To disclose unexoacted phenomena and interactions.
To examine approaches for potential improvements to reactor safety.
To provide bases for the developnent of improved licensing crit.eria.
To maintain available a body of competent, knowledgeable experts to assist the NRC in its evaluation of complex technical proposals for licensing.
1-6
TABLE 1 ESTIMATED RESEARCH EXPENDITURES FY 78 FY 79 (Jn millions)
PROGRAM SUPPORT IDCA/ECCS
$48.6
$67.2 Fuel Behavior 24.1 23.6 Primary Systen Integrity 8.0 10.1 Operational Safety 1.5 1.9 Advanced Reactors 15.7 15.0 Extreme External Phenomena 5.1 6.8 Radiological Effects
- 7. 3 7.7 Waste Management 4.2 4.2 Safeguards and Security 6.6
- 6. 2 Risk Assessment 3.1 3.4 Improved Reactor Safety 0.0 Subtotal
$124.2
$146.1 EQUIFMENT, PERSONNEL, AD4INISTRA-TIVE SUPPORT AND NON-SAFETY-PEII.TED RESEARCH**
14.3 17.4 Total
$138.5
$163.5 The NRC plans to reprogram approximately $0.8 million in FY 79 to initiate research in this area.
- Includes non-safety-related environmental and socioeconomic re-search 1-7 e'eg
2.
LOSS-OF-COOIANT ACCIDENT / EMERGENCY CORE COOLING SYSTEMS (LOCA/ECCS) 2.1 Introduction The major effort of the NRC's safety research is concentrated in the projects involving IOCA/ECCS which account for nearly 50 percent of the safety research budget.
A significant increase in funding will be required in FY 80 to cover the planned operation of the Ioss of Fluid Test (LOFT) Reactor.
!bwever, major reductions in expendi-tures are expected to result within four to eight years and from the decommissioning of LOFT Reactor in the late 1980s.
To appreciate the.mportance of the effective operation of the emergency core cao2ing systems, one needs to recognize that the operation of a nuclear power reactor produces radioactive fission products within the fuel and, even when a reactor is shut down, the decay heat from the fission products must be removed by a coolant.
If this heat is not removed, the fuel will melt and leakage of the radioactive materials into the environment becomes more likely and constitutes a potential hazard.
The DCCS is designed to remove the decay heat in the event of a LOCA.
In reviews conducted by the licensing branch of the Atomic Energy Commission (AEC), beginning more than a decade ago, special attention was given to the development and evolution of the ECCS required for power reactors of iner aasing power density and rating.
We design basis accidents involved not only the loss of the coolant through postulated breaks in the primary piping, but also superimposed a series of other restrictions on such hypothetical accidents.
For example, the availability of offsite power was assumed to be lost, and one had to depend upon onsite emergency power sources to supply the services needed to operate the ECCS.
Concurrent with these re-strictions was the imposition of a requirement that some other single failure of equipment might render a portion of the ECCS inoperative so that the ECCS would be required to be sufficiently redundant and/or diverse to still perform its function of cooling the fuel.
W e complexities of following in detail the course of the postulated loss-of~ coolant accidents were apparent and considerable emphasis was given to improving the analytical methods and to developing further the experimental thermal and hydraulic information needed to confirm the adequacies of the bases being used in the evaluation of these safety systems.
2-1
Major changes in the ECCS were initiated as a result of reconnenda-tions by the ACRS in 1966 and in succeeding yearc for pressurized water reactors and for boiling water reactors, and significant increases in LOCA/ECCS research and development were undertaken by the reactor vendors.
%e AEC Staff imposed overall conservative re-strictions on the nethods used to evaluate the effectiveness of the ECCS.
We IDFT program was redirected.
Instead of a nuclear test facility to study primarily the transport of fission products due to loss of adequate cooling of the fuel, the test objectives were fo-cused on LOCA/ECCS. A non-nuclear test facility, known as 5fMISCALE, became a primary focus for studying in a small scale facility the system interaction effects that might occur in LOFT and in the nuclear power reactors.
%ese facilities provided, for the first time, the opportunities for studying simulated loss-of-coolant ac-cidents in an integral system. % e burden of the mathematical models and computer cos was shif ted from conservative evaluations to rea-listic or best estimate evaluations.
Even by early 1970 to 1971, the overall safety research and the de-velognent of independent analytical models for IOCA/ECCS evaluations were not full.y responsive to the needs and recommendations.
New insights into overall IOCA/ECCS phenomena, including ECC bypass, steam binding, and delayed core reflooding, resulted from ongoing research.
Interim Acceptance Criteria were set forth to impose con-servative evaluations for ECCS.
A prolonged public hearing on LOCA/ECCS brought about changes in the evaluation.nodels anJ commit-ments by the AEC for increased research to ascertain the margins of conservatism and to seek improvements in the ECCS.
Although the bacic designs for the ECCS have remained unchanged, changes in fuel designs by the reactor vendors and suppliers of core reloads, have introduced improvements in ECCS effectiveness.
With the termination of the AEC and the creation of the Nuclear Regulatory Conmission in 1975 the confirmatory safety research programs were restructured and significantly augmerted.
It should be noted that, in addition to the public hearings, peer group reviews such as that conducted by a study group of the American Physical Society focused on the safety research progran and provided recommendations to the NRC.
Public critics also F:ve challenged what they perceive to be deficiencies of the researc.. programs and have questioned the effectiveness of the ECCS.
We present LOCA/ECCS safety research program has been formulated to be responsivc to the needs of the NRC licensing staff, to the recommendations of the ACRS, and to a broad segment of public input.
2-2
2.2 Research objectives
%e ACRS concurs with the NRC staff that the research should ac-complish the following goals:
a) Confirm, in a reasonable way, the safety margins embedded in the current licensing evaluation models used to determine the acceptablity of the ECCS.
b)
Provide the technical input and tools needed for evaluating potential improvements in the ECCS and for better optimization of present systems.
c)
Provide the bases for extendit.g the methods used in the thermal-hydraulics and in the neutronics for treating other off-normal and transient re-actor responses.
d)
Provide the NRC staff with independent analytical models for audit purposes and for increasing their technical and practical awareness of ECCS features.
e)
Provide a basis for later reevaluation of the li-censing methods and restrictions to detarmine what changes or relaxations are justified.
2.3 Are Aiditional Majorlest Facilities Necessary?
As noted above, the LCCA/ECCS safety concerns have become a matter of wide pi ilic input, of repeated ACRS reports, and of commitments by the Comissioners.
%e ACRS concurs that a thorough, open, and substantial base of information must be provided to establish in a reasonable manner just how large the current safety margins are and what improvements to ECCS are desirable or necessary.
With few exceptions, the technical community believes that the cur-rent licensing methods do provide an overall conservative estimate of the effectiveness of the ECCS.
here remains, however, uncertainty in the magnitude of the safety margins.
Test facilities which are scaled down from actual nuclear power reactors provide data for studying ECCS performance. To what extent are full-scale tests re-quired? Obviously, if full-scale testing was to be done on a suffi-ciently extensive basis, the ' uncertainties in the safety margins could be reduced.
Such a program would be very costly, and the ACRS does not believe that this approach is necessary.
A program 2-3
that addresses all the major phenomena in the course of a LOCA and in the interactions of the ECCS should provide all the infomation that is really needed.
To achieve this, it will of course he necessary that the analytical tools be independently assessed, that proper attention be given to the problem of scaling, and that adequate safety margins be retained.
Acceptable values for the magnitude of the safety margins and the un-certainties therein remain to be defined.
We framework for achiev-ing this is in progress.
Both the NRC and the ACRS believe that the safety research progran need not be augmented to include the following costly items:
a)
The ECC Bypass Test Facility (EBTF), particularly now that the Federal Republic of Germany (FRG) Upper Plenum Test will use a nearly full size vessel.
b) A new major program such as the multi-purpose test facility at a cost of over $100 million.
c) A full-scale integral test facility.
We ACRS has concluded that the LOCA/ECCS research is necessary, and is being restrained within reasonable limits.
The features of the program are addressed in the sections that follow.
2.4 Computer Code Development Computer codes are used to predict the perfornance of nuclear power reactors and to investigate the consequences of postulated accidents.
For postulated LOCA, the NRC and the worldwide technical connunity depend upon codes for evaluating the effectiveness of the ECS.
These codes are based upon mathematical modelling of physical phenom-ena involved in LOCA/ECCS.
Confidence in the capability of these codes is derived from extensive experimental programs.
An insight into the experimental programs is given in the next section.
Since no full-scale testing of LOCA/ECCS will be carried out using a nu-clear power reactor, the procedures used in code developnent and code testing must be scrupulously reviewed.
Currently, the licensing of nuclear power reactors is based upon using conservative overall estimates in the code applications, and the LOCA/ECCS licensing codes are called Evaluation Models (EM).
Models that attempt to predict realistic results are called Best Estimate (BE) models.
For example, if an end calculation is the 2-4
prediction of the maximum cladding temperature, then the difference between the D4 calculation and the BE calculation gives an indication of the safety margin embedded in the D4 calculation.
We ACRS has reviewed the NRC research programs involving codes and has noted significant advances in the development of BE models.
BE codes are built up in a two-step process. We first step is known as code develognent, and at this stage, changes are permitted in the physical modelling, in the specific thermal and hydraulic character-istics, and in the numerical procedures.
By exercising the codes, one can ascertain whether any required features have been omitted or incorrectly incorporated.
By comparing with experimental results, one can determine how effective the code is in predicting the actual behavior.
We section on Experimental Prog 1am notes the wide range of tests needed in the code developaent.
For the BE codes being developed, there should be no artificial features that have been included so as to improve the matching of the code predictions with a specific experiment. Furthermore, any adjustments in parameters as a means of improving predictions to match specific experiments should not vio-late basic physical principles.
%e end product of the code de-veloper is a code that describes the physical phenomena to an ade-quate degree.
he second step is a check on the BE code that has been released by the code developer.
Ris step is called independent code assessment and is performed by a group that is independent of the code develop-ment g roup.
In the independent assessment, tests different from those already employed in the code development are used to determine how accurately the code can predict their results. We characteriza-tion of the code's applicability to nuclear power reactors is also a part of independent code assessment.
We NRC, at the urging of the ACrlS and other groups, has begun to spell out more clearly what needs to be done for independent code assessment.
%e ACRS is aware of the criticisms inplying that code developers continue to tune their models to fit experimental results, and thus a code may fit particular experiments but still not be suitable for ACRS members and con-final application to a nuclear power reactor.
sultants have reviewed these matters.
%e present two-step system, code development and independent code assessment, with the require-ments of prediction of test results, has been developed in an attempt to address this concern.
It should be noted that an independent code assessment has not yet been completed, and thus special attention needs to be focused on this mattor in future reviews.
Each najor test, where appropriate, should
..e preceded by blind predictions 2-5
from best estimate codes; blind predictions should also be made with the advanced code, TF-for the 3D-2D international test series.
he NRC has planned an approach that would characterize BE code un-certainties and would expose any inherent code errors or bias.
It is envisioned that a peak clad temperature probability surface could be obtained from assignment of probabilities for various combinations of plant conditions.
Wis is an important area for continued ACRS involvement.
2.5 Experimental Program te expecimental program involves basic experiments, separate effects experiments, and integral experiments.
Basic experiments are designed to provide further knowledge on two-phase flow phenonena, to provide insight on what features should be included in the ma thematical models, a v to help in establishing constitutive relationships needed for the modelling in the advanced computer codes.
%e programs that develop advanced instrumentation will provide basic details of possible flow structures and accurate measurements of flow phenomena.
Separate effects experiments permit detailed investigations of simu-lated physical features of the reactor system for various aspects of the postulated accidents, including depressurization or blowdown, refill, and reflood.
Experiments are conducted at various scales, including, for example, full-scale for simulated fuel elenents.
Tests include blowdown heat transfer, reflood heat transfer, ECC bypass, and steam-water mixing.
Past tests have provided bases for code development in conjunction with integral test data.
Ad-ditional tests and detailed measurements will provide, in part, data for later independent code assessments.
Integral experiments sinulate the primary system of a nuclear power reactor and thus provide for system interactions.
LOFT, which is approximately scaled at 1/60 the power of a pressurized light-water power reactor, is the only integral test facility in the m rld which employs nuclear fuel. Other integral test facilities employ electrically heated rods to simulate the fuel elements.
SD4ISCALE is approximately scaled at 1/30 the power of LOFT.
Foreign integral test facilities include the Japanese ROSA facility and the new Japanese cylindrical core test facility, the EURATOM LOBI test facility which is about three times the size of SEMISCALE, and the Federal Republic of Germany (FRG) PKL facility.
We only integral test facility specifically for a boiling water reactor (BWR) config-uration is the 'No-Loop-Test-Apparatus, TLTA-2.
%is facility will use a simulated full-scale single BWR fuel bundle.
2-6
IDFT !s the most costly facility to operate and represents about half the LOCA/ECCS budget.
As previously noted, the LOFT program has been redirected during the past decade, culminating in the pres-sent LOCA/ECCS objectives. Management of the program also has under-gone changes. Whereas progress in bringing LOFT into operation was slow, the non-nuclear tests now have been successfully completed, and the first nuclear test was performed on December 9,1978.
Al-though SEMISCALE and IDFT have provided considerable information on the phenomena involved in LOCA/ECCS, on special features needed in code develognents, and on scaling, the confirmation of safety margins must also depend upon other tests with other facilities. The nuclear tests do provide unique opportunities for testing large, intermediate and snall IOCAs, alternate ECCS, and other postulated accident con-tions and transients.
Predictions of the nuclear tests will provide further checks on the analytical methods and on how thoroughly the accident phenomena are understood. W e present NRC direction and the management of the test facilities indicate that the projects are being carried out competently.
The ACRS concurs with the NRC that these nuclear tests should proceed, and that the limits placed on the the program are reasonable, including the scheduled decommissioning by the late 1980s.
We entire IDCA/ECCS experimental program thus consists of many ele-ments:
basic research; small, intermediate and large-scale separate effects tests; and only small and intermediate integral tests, with IDFT being the only integral facility employing nuclear fuel.
In the review of these progrsms, the ACRS has been impressed with the many outstanding engineers and scientists who are involved in these stud-ie:.,.
Exchange programs with other countries significantly mmert the resources both in personnel and in facilities.
Included in the experimental program are the new international projects involving Japan, the FRG, and the United States.
These projects are directed at studying safety margins in pressurized water Over $100 million will be expended by Japan and the FRG in reactors.
major facilities.
A total expenditure of approximately $50 million over several years is planned by the NP.C for advanced instrumentation and for the analytical coordination of the Japanese reflood tests and the FRG upper plenum tests.
We proposed NRC budget includes support of a new facility for tests boiling water reactor LOCA/ECCS.
The facility wili irvesti-on gate counter-current flew limitations for sprays in a steam environ-in a full-sca'e 30-degree sector, with allowance, if necessary, ment to provide heated fuel elements.
2-7
There is a large industrial effort to resolve safety margins for tWling water reactor containmnts, and the NRC research projects in this area are directed toward providing independent assessmnts.
The ACRS agrees that the present mat ix of experiments should be sufficient to provide the input needea for code development and for independent assessnents.
2.6 Findings and Recommendations
- 1) The ACRS cormundr the NRC on developing the LOCA/ECCS safety re.e: arch program in a thorough and open nunner and on being responsive to the needs as determined from NRC liceasing re-quirements.
- 2) The ACRS recommnds that the bounds now being placed upon the experimental program be followed so that an orderly reduction in the projects and costs can be achieved in the 1980s.
Specif i-cally, the ACRS concurs that the following costly items are riot necessary:
The ECC Bypass Test Facility A new major multi-purpose test facility A full-scale integral test facility
- 3) The trost costly facility to operate is LOPP, the only integral test facility in the world that employs nuclear fuel.
The LOFT program is being directed and managed conpetently and effec-tively, and the ACRS recommnds that it be continued.
The lim-its to in placed upon LOPP tests should be reviewed so that the scheduled decomissioning by the late 1980s need not be de-ferred.
- 4) The ACRS concurs with the NRC Staff that the present matrix of LOGA/ECCS experimental projects should be sufficient to provide the input needed for code development and for independent code assessent.
- 5) The ACRS considers the independent assessment program for best estimate codes to be essential.
The ACRS recommends that more effort be placed on defining the requirements for such indepen-dent assessments, that the results be widely discussed in the technical comunity, and that detailed plans be made for assur-ing that sufficient virgin tests be reserved for the independent code assessments.
2-8
~
In spite of the fact that a consensus was reached by the ACRS in regard to the above reconnendations, sone concern has been expressed that continued funding of the expensive projects, such as LOPT, might encroach upon the adequacy of the support of other important safety research programs.
2-9
3.
FUEL BEHAVIOR 3.1 Objectives he heat and fission products in a reactor are generated in th + fuel.
" Defense in depth" includes the use of multiple barriers against any escape of fission products. We first barrier is the claiding which surrounds the fuel.
We overall objectives of the fuel program are to provide an independent assessment of the integrity of the clad-ding during the wide range of transients and accidents considered in reactor licensing, and of the behavior of fuel and fission prod-ucts if the fuel rod ruptures or melts.
The areas being studied are:
Properties of fuel and clad Developnent of fuel behavior codes Power Burst Facility (PBF) experiments Fuel irradiation (other than PBF)
Fuel meltdown and fission product release and transport 3.2 Properties of Fuel ano Clad An essential prerequisite for any prediction of fuel rod behavior is a knowledge of the properties of its component parts - fuel and cladding.
%e NRC had a substantial research program in this area during FY 78, especially on cladding.
We number of studies has decreased substant.lally in FY 79, and will decrease again in FY 80, since the information base seems adequate and higher priority proj-ects will ricaive attention.
We continuing projects will primarily involve integral tests, such as clusters of fuel rods, and in-reactor tests.
Several of the complete.d programs will be of substantial value in considering revisions in the emergency core cooling systems (ECCS?
evaluation models contained in the NRC regulations. Studies of the metal-water reaction and fission product decay heat have better defined the conservatisms in current regulations.
We cladding 3-1
program is pr-
.sJ a more quantitai.Iva basis for calculating clad-d&J oxidativn and deformation during a postulated loss-of-coolant accident (LOCA).
3.3 Development of Fuel Behavior Codes In the process of licensing reactors, the NRC requires analyses of fuel behavior under postulated and unlikely accident conditions.
In order to predict what might happen to the fuel, the NRC must use computer codes that model fuel behavior in sach cases.
Valid codes of this sort are highly desirable to check vendor assertions regard-ing fuel behavior during transient and accident conditions.
We codes are also used to model the results of test reactor experi-ments in order to predict the behavior of power reactor fuel in various hypothetical situations.
We NRC program on fuel behavior codes is adequate and essential for NRC needs, and provides useful guidance to the overall fuel behavior research program.
3.4 Power Burst Facility
%e Power Burst Facility (PBF) is a water-cooled, water-moderated test reactor located at the Idaho National Engineering Laboratory (INEL) in Idaho.
It is designed for both steady-state and pulsed-mode operation, and can accommodate 1 to 25 fuel test rods.
We experiments are well instrumented and the reactor is generally op-erating satisfactorily.
We PBF program includes tests in the following areas: power-cooling mismatch (PCM), LOCA, flow blockage, reactivity-insertion accidents, and gap conductance and stored energy. Seven tests were cerformed in the PBF in FY 77 and four in FY 78. Many of the tests completed to date have been of an exploratory nature to define the corditions which can be simulated with fuel in PBF.
To date, tests have been performed that simulated PCM and LOCA conditions.
We PCM and II)CA simulation experiments run to date indicate that PBF can he used for tests that simulate accident conditions. Several tests have been valuable in demonstrating the conservatism of NRC licensing positions.
Future experiments in this area will better define the safety margins in hypothetical accidents which are im-portant to many liceasing decisions.
It is less clear to the ACRS why the fuel research program puts a high priority on reactivity-insertion accident (RIA) experiments.
%ere is general agreement that in current commercial power rcactors, the principal accident of concern here (rod ejection or rod drop 3-2
accident) has a very low probability of occurrence. In view of this, it would appear that this effort might better be devoted to other problems.
In general, the PBF program is producing valuable, timely results, and should be continued. W e establishment of priorities within this program is a matter that merits continuing NRC management attention.
3.5 Fuel Irradiation Studies Other Than in PBF Research on fuel behavior is underway in several countries besides the United States.
W e NRC maintains close liaison with these pro-grams, designing its own program to complement them, and joining with others to share the costs of some projects.
'Iwo merit comment. We Halden experimental reactor in Norway has, over twenty years, con-ducted an excellent program on individual fuel rod performance funded by a large number of countries.
We United States now participates in the Halden study to better define the heat conductance across the gap between fuel and cladding.
Following these tests, the fuel rods will be sent to the INEL for PBF experiments.
To provide an in-pile check of the extensive out-of-pile clad bal-looning experiments, the NRC is planning a series of tests in the NRU reactor located at Chalk River, Canada. Wese tests are designed to simulate WCA refill and reflood conditions on clusters of 32 full-length reactor fuel rods.
This will be the first such test on full-length commercial fuel rods.
Tests on fuel rods of this length are a unique capability of the NRU facility.
3.6 Release and Transport of Molten Fuel and Fission Products Although accidents that could cause a significant fraction of the reactor core to melt are extremely improbable, the potential danger to the public is substantial if och an accident should occur,. As a result, the NRC believes it prudent to obtain better information In addition to to scope the consequences of accidents of this type.
the NRC program, the NRC Staff follows the work of others, in par-ticular. the substantial FRG research program in core-melt / concrete interactions, fission product release from nulten fuel, and the fuel-coolant interaction work in many countries.
%e NRC is cur-
%e ACRS rently planning a reduced level of effort in this program.
believes this area of work is important and is concerned that the NRC effort in this area not be phased out.
We ACRS believes that careful consideration should be given to the needs of the licens-the usefulness of such studies for future risk ing staff, and to assessments and studies on improved light-water reactor safety.
3-3
3.7 Reconinendations
- 1) The high priority now assigned by the NRC staff to experiments relating to reactivity-insertion accidents should be reassessed in view of the very low probability of such accidents.
- 2) Work on phenomena important to the course of postulated core melt accidents should continue to have high priority with due consid-eration given to the ongoing program in the Federal Republic of Germany.
3-4
4.
PRIMARY SYSTEM INTEGRITY 4.1-Objectives he primary system, comprised of the reactor pressure vessel, other vessels, and piping, contains the cooling water under high pressure.
his system constitutes the second line of defense against release of radioactivity to the environment.
Failure of the reactor pressure vessel is considered by the NRC to be of such low probability that it need not be postulated as a de--
sign basis accident.
Failure of the primary system piping, however, is considered to be more probable and is in fact assumed as the de-sign basis loss-of-coolant accident.
The objective of research in this area is to assure that the proba-bility of failure in the primary system is suitably low.
We pro-gram addresses the basis for design, fabrication, inspection, and operational use of the primary coolant system loop through experi-mental and analytical investigations.
We areas of current inter-est are:
Fracture mechanics of vessels and piping Control of radiation-induced loss of ductility Steam generators and the effects of corrosion Detection and evaluation of flaws in the materials of construction Integrity of mechanical systems under transient and dynamic loadings.
4.2 Fracture Mechanics Before a pipe or vessel can suddenly rupture, a substantial crack must develop through physical deformation, fatigue loading, stress-induced corrosion cracking, or combinations of these.
We major effort of the fracture mechanics program is directed toward defin-ing the conditions under which a flaw could grow sufficiently to cause gross fracture of the pressure boundary.
4-1
ne pressure vessel codes for assuring the integrity of the primary system are premised on limiting the size of flaws, controlling the fabrication quality, and constraining the operational corditions for pressure-containing components. %e bases for flaw acceptability in the ASME Boiler and Pressure Vessel Code have been validated by the Heavy Section Steel Technology (HSST) program funded jointly by NRC and industrial participants with support from the Electric Power Research Institute (EPRI) and DOE.
Wis program has been underway for more than ten years and has led to well-verified procedures for avoiding catastrophic rupture.
he vessel program is now aimed toward better definition of the types of cracks in tough ductile steel that will result in leaks prior to gross rupture of the system.
The situation with respect to primary system piping is less well defined. By its nature, piping is more flexible than the heavy-walled vessels.
It expands and contracts during temperature cycling but must be restrained against vibration during earthquakes. The current basis for design is that piping may rupture instantaneously at any point and that the ECCS be capable of bringing the reactor to a safe condition. This requirement poses considerable difficulty for the plant designer and may be unrealistically conservative.
A new program is being initiated by NRC to define better the design re-quirements and to establish the design safety margins. At this time, the program plan is still in the formative stage.
It will be co-ordinated with international programs in Japan and the Federal Republic of Gennany, and with programs of U.S. nuclear steam supply system vendors and EPRI.
Wis program is responsive to the ACRS recommendation in the 1977 Report to Congress.
Reactor water chemistry can induce cracks and affect the rate at which cracks grow in the primary system.
%is effect is of particu-lar concern in boiling water reactors (BWR) where intergranular stress corrosion cracking has appeared repeatedly in sensitized stainless steel piping of generally small diameter pipes. We NRC is planning little or no research in this area; instead, it is depending almost entirely on the work of EPRI and of the General Electric Co., the BWR vendor. W e ACRS questions whether the NRC can maintain the expertise needed to properly carry out its regulatory function if all research is carried out by the reactor vendor.
We effect of coolant chemistry on crack growth in pressurized water reactor (NR) vessels and piping is of less pressing concern, but is an acknowledged uncertainty in the conservatism of the NRC regulatory position. The current NRC and international programs are quite modest.
4-2
4.3 Irradiation Embeittlement Fast neutron radiation from the reactor core tends to reduce the fracture toughness of the reactor pressure vessel material and in-creases the temperatures at which the vessel must be maintained during heat-up and cool-down in order to assure adequate resistance to britt1.e rupture.
The conservative projection of this trend into the future has raised concern abou* whether the older pressure vessels will be usable for their projected life.
Annealing the reactor vessel in-place by heating to temperatures in the range of 700-800 F is one possibility for recovering fracture toughness should the required operating temperature exceed practical limits.
Both the NRC and the EPRI have programs to investigate the feas;-
bility of this procedure.
Recent results from the pressure vessel surveillance capsules placed in the older licensed power reactors indicate that the embrittlement effect may saturate relatively early rather than continue to increase as assumed in the NRC licensing position.
If confirmed, s~0h satur-ation would provide greater operating margin than is now assumed, and the impact on the vendors, the utilities, and the NRC procedures might be substantial. Accordingly, the NRC should initiate a program to determine how these results could be confirmed adequately for regulatory purposes.
An essential part of an irradiation prou am is the ability to measure and compare the irradiation dose received in various reactor environ-ments.
We NRC is developing a new project which will provide an expanded, more precisely controlled set of baseline data.
%is should lead to significant improvement in the accurac1 of dosimetry measurements and in comparison of radiation effects li: light-water reactors.
4.4 Steam Generators and Corrosion Corrosion induced by water chemistry in many PWR steam generators has led to thinning of the tube walls, deformation of the tubes by denting in the regions where the tubes pass through support plates, and other relatively serious defects.
Tube-thinning has been ade-quately controlled by improved water chemistry control.
We tube denting question is still unresolved, although the staff has ap-proved vendor modification programs to alleviate this problen.
The problem is so severe in a few reactors that the owners are planning to replace the steam generators at considerable expense.
Rese corrosion problems have safety significance only under certain highly improbable loss-of-coolant accident conditions.
Interest in this problem, plus questions regarding consequences, have caused the NRC 4-3
to fund several projects in this area and expansion of this effort is being contemplated. Were is great economic incentive for vendors and plant owners to solve this problem (EPRI has a multimillion dollar program) and appreciable progress is being made.
While the NRC needs to continue some level of effort in order to provide an ade-quate independent knowledge and capability in tnis area, the pro-gram effort should be directed to complement the work already under-way under the sponsorship of EPRI.
4.5 Flaw Detection and Evaluation Critical portions of pressure vessels and piping are inspected carefully both prior to and periodically during service in order to assure that there are no significant cracks or flaws in the primary system.
Unfortunately, the techniques currently available are less reliable than desired for effective in-service inspection. We NRC has a small program aimed at improved in-service inspection.
The DOE, EPRI, and agencies in several foreign countries have substan-tial programs aimed at improving techniques for finding and evalu-ating such flaws.
Wese programs are being menitored by the NRC staff to provide knowledge of technological developnents, and should satisfy the immediate needs for regulatory purposes.
4.6 Mechanical Engineering he newly established Mechanical Engineering Research Branch in the Office of the Assistant Director for General Reactor Research is expected in the future to undertake research relating to primary system integrity. At present, however, its activities relate chiefly to the ability of mechanical components to resist earthquake-induced forces.
In the future, the program may be expanded in a number of other problem areas.
Rose given the highest priority are develop-ment of a suitable basis for combining dynamic loads, and methods for assuring the operability of pumps and valves under transient conditions of importance to safety.
Rese activities would be of considerable help in resolving the generic questions concerning systems interaction and thus deserve to be included in the research program at an earlier time.
4.7 Recormendations his program is important and deserves an overall high priority.
In general, it is being conducted in a manner suitable to the needs of NRC.
".mver, the following reconnendations are made regarding relative priorities and emphasis within the program:
4-4
- 1) The HSST program should be completed as planned.
2)
The portion of the program devoted to piping relia-bility deserves a high priority because it may, in the end, enhance the safety of nuclear power plants by simplifying the engineering requirements.
- 3) The research on the effects of coolant chemistry on crack growth in piping and pressure vessels is inade-quate.
A significantly expanded ef fort is warranted to better define the safety margins.
- 4) The NRC should initiate a project to evaluate the findings from operating experience regarding possible saturation effects of radiation embrittlement.
- 5) The project associated with stean generator tube integrity should be reviewed further to make certain that the planned expenditures are for research that complements that performed by the nuclear induztry or, if intended to duplicate industry effort, pursues research avenues that would enhance the NRC's under-standing of the problem.
4-5
5.
OPERATIONAL SAFETY 5.1 Objectives and Scope Much of the NRC safety research program is concerned with long-term research to confirm that systems will perform in accordance with predictions.
However, from time to time, new problems or needs are perceived as a result of review, or develop from operating experi-For example, questions may arise concerning the adequacy of ence.
available data to predict with a high level of confidence the be-havior of particular materials and components after long service and in hostile environnents.
Also, during normal and slightly abnormal operating conditions, particular malfunctions, failures, and operator errors may occur with greater frequency than appears reasonable.
Research programs are initiated, as the need arises, to solve such problems, and conclusions are factored into new or existing stan-dards, regulations, or other safety criteria.
The current projects in the operational safety research program are:
Evaluation of qualification testing Fire protection Noise diagnostics Man-machine interfaces 5.2 Evaluation of Qualification Testing Components that would have to function in the event of a liX'A are khere tested in the laboratory under simulated accident conditions.
appropriate, the test procedures are those specified in IEEE (Insti-of Electrical and Electronics Engineers) Std 323-1974.
The tute tests include an accelerated aging test to simulate the deterioration expected over a lifetime of operation, and a sequential test repre-senting LOCA conditions of r6diation, followed by exposure at ele-vated temperature to the applicable corrosive mediun (e.g. steam or chemical spray) for a specified time-temperature cycle.
5-1
Rere is some question concerning whether an accelerated aging test that truly represents the deterioration in actual service, and whether sequential exposure to radiation and the corrosive medium will have the same effect as simultaneous exposure.
We radiation environment used in testing also needs to be correlated more pre-cisely with LOCA conditions. A testing project on electrical cables, coatings, connector assemblies, cable splices, and lubricants has been completed at Sandia Laboratories, and additional projects have been initiated at Sandia and elsewhere.
At present, there does not appear to be any significant synergistic effect of simultaneous testing, and the much simpler current sequential testing procedure therefore appears to be satisfactory.
5.3 Fir. Protection Industry standards and NRC guidelines have been greatly improved and expanded sinco the Browns Ferry fire.
'Ibpics addressed by current fire-protection criteria include minimum separation distances between trays carrying cables of redundant systems, physical barriers to prevent spread of fire, fi re-retardant cable coatings, fire de-tection systems, fire extinguishing agents, and flame retardancy test procedures for cables (including procedures for artificial aging).
Research in this area explores the adequacy of all of these criteria by tests both for fires generated electrically in the cable trays, and for fires initiated externally.
The project incitxles analysis of the consequences of a fire on sa fety related equipnent where complete separation is not possible.
5.4 Noise-Diagnostics his research includes studies of " noise," i.e. superimposed pertur-bations of nuclear or process instrumentation readout caused by vibration of internal parts, and studies of special instrumentation to detect and locate loose parts in a nuclear system by measuring the impact, noise.
Neutron signal perturbations have, in the past, given early warning of unacceptable internal vibration, and the current research is intended to refine and develop a diagnostic technique for interpreta-tions of signal anomalies of smaller magnitude.
There have been numerous instances of loose parts in reactor systems, and it is possible to detect such incidents and sometimes to locate the part by monitoring for impact on the system boundary.
A Regu-latory Guide has been issued concerning requirements for loose parts 5-2
monitoring systems.
However, the state-of-the-art needs to be in-proved, and commercially available systems need further develcprent.
The research program includes construction of a test loop and de-velopnent of criteria.
Equipment manufacturers are actively de-veloping loose parts monitoring systems.
5.5 Man-Machine Interfaces Human error has been an important cause of abnonnal occurrences in nuclear plants.
Such occurrences are reported systematically in the " Licensee Event Reports" (LERs) and are currently being analyzed on a probabilistic basis.
Programs will be developed to apply the principles of ergonometrics to the licensing and operational review processes.
Errors made during maintinance appear to be the most numerous, but operational errors also occur.
Since most operator action is taken in the control room, it is important that the design of the control room and control-room equipnent emphasize diag 70stic infonnation that will simplify decision-making.
Future control rooms can meet these objectives by the expected much greater use of computers and graphic displays.
Independent FMC research is necessary to support the licensing review of these ad-vanced control room designs, and to develop criteria, guides, and standards.
5.6 Recommendations
'Ihis is a small program that could profitably be expanded to address the problems encountered and the lessons learned from the many operating reactors and to provide a broader and sounder data base for assessments of reliability and risk.
In particular, the ACRS offers the following recommendations:
- 1) The man-machine interface ef fort should be given high priority, with examination of the potential for as the and consequences of human errors as well mitigating aspects of man's intervention.
- 2) The advantages and disadvantages of a greater degree of computer-controlled automation should be explored.
- 3) A more systematic review and evaluation of operational experiences and operational incidents in U.S.
plants and in similar plants in other couritries should be undertaken.
5-3
6.
ADVANCED REACTOR SAFETY 6.1 Introduction Advanced reactors include all reactors except current pressurized water and boiling water reactors; i.e.,
fast breeder reactors, both liquid-metal and gas-cooled; thermal breeders; and advanced convert-ers including spectral-shift and heavy-water reactors.
%e advanced reactor safety research (ARSR) efforts are concentrated on the liquid metal fast breeder reactor (MFBR) and thermal and fast gas-cooled reactors (GCR).
Until about a year ago, much MFBR work was related to the Clinch River Breeder Reactor (CRBR); it is now directed toward more generic problems. Most attention in the GCR work has been given to the high-temperature gas-cooled reactor (H'IGP) with a lesser amount to the gas-cooled fast breeder reactor (GCFR).
We program for FY 79 includes about $12 million for MEER research and about $3 nillion for GCR research. We DOE budget is about $39 million for advanced reacter (LMFBR, H'IGR, and GCFR) safety research and about $500 million for advanced reactor development.
In its review of the ARSR program, the ACRS had the benefit of a comprehen-sive report, " Status of Advanced Reactor Safety Research 1978,"
prepared by the NRC and dated July 20, 1978, which provides an extensive outline of NRC, DOE, and foreign advanced reactor safety work.
6.2 Current Programs
%e current LMFBR work is divided into four areas, viz analysis, materials interactions, effluent release, and structures.
We analysis work is aimed at defining analytical models and develop-ing computer codes that will (a) help the licensor to predict and understand the physical changes which may occur in a reactor system when off-normal or accidental situations develop, and (b) enable him to predict the levels of pressures, temperatures, and other physical parameters which will result from such situations.
In particular, considerable attention is being given to analyzing the course and consequences of a Core Disruptive Accident (CDA),
an extremely unlikely accident but one which might conceivably lead to core nelt and serious radiation releases unless adequate safety systems are established.
6-1
We materials interaction work involves primarily the performance of experimental tests which simulate various parts or phases of the Cm and of the events which might follow the Cm.
Many of these tests are perforned in a special research reactor, and the purposes are (a) to obtain insight into and answer questions about the physical events that occur in and after a CDA and (b) to support the analy-tical studies just mentioned.
We effluent release work involves both analytical and experimental efforts to define the kinds and amounts of radionuclides which might be released in an accident, and to study the mechanisms affecting the release and transport of this material to surrounding areas.
The structural work is aimed at evaluating the adequacy of the reactor structure under both normal and off-normal conditions.
The structural problems are unique because of the presence of sodiun and because of the high temperatures at which IWFBRs operate.
Considerable progress and impo rtant advances have been nade during the past year.
The ACRS believes, for example, that some of the analytical work has a more solid physical foundation than was the case a year ago.
Modification of an important test reactor has been completed, and experimental tests have, on the one hand, reopened an old concern about the potential for a fuel-coolant thermal interac-tion during a CDA and, on the other hand, indicated that the cooling of core debris following a CDA may be sonewhat less difficult than previously anticipated.
As indicated above, the gas-cooled reactor work is directed largely toward IITGR application and specifically toward the concept of which the Fort St. Vrain reactor is an example.
Increased attention has been given to fuel and fission product behavior, coolant depressurization, structural graphite properties, and noise analysis.
Were have been worthwhile accomplishments in the areas, for example, of fatigue correlations for high tenperature alloys, graphite oxida-tion, review or developnent of various computer codes needed to describe such processes, and developnent of a materials test loop.
6.3 Recommendations Regarding the ARSR Budget Unless the LMFBR and other advanced reactor development prograns are to be deferred for an extended period of tine, the ACRS reconnends that the Congress continue to regard advanced reactor safety t'esearch as a high national priority because of the time required to resolve important safety questions.
Many of the current safety problens 6-2
associated with light-water reactors have resulted from the fact that safety research lagged behind reactor developnent. If an advanced reactor program is to be pursued in the U.S., related safety research should be carried out concurrently with develognent.
Wis will pennit licensing to proceed in an orderly fashion when specific projects for advanced reactors are submitted.
Since the LMFBR and the H'IGR appear to be the most likely advanced reactor candidates, those concepts are emphasized below.
We ACRS believes that an increase in ARSR funding is imp 3 rtant at this time to permit continuation of present programs and to provide funding to implement recommendations for the new wrk out-lined below. It is recommended that ARSR funds be identified in the NRC budget separately from those for research on current reactor concepts.
%e ACRS believes that the NRC should be encouraged to follow foreign research prograns closely and to participate in cooperative ventures when practical. Such efforts can be expected to result in considera-ble financial savings for the United States.
6.4 Reconnendations he ACRS believes that the ARSR program is directed toward important topics, involves competent subcontractors, and is producing many worthwhile and important results.
However, the ACRS has several recommendations to make which are in the general nature of broadening the scope of the program and of endeavoring to look at needs farther down the road.
he first six recommendations relate to new programs and the last two concern changes in emphasis.
1)
It is recomended that NRC undertake a comprehensive study of the safety questions that are likely to arise for commercial LMFBRs.
The current program concentrates on a few important topics (especially the CDA) and has been directed at plants of the general FFTP or CRBR size.
%e ACRS believes that there is a hic)h-priority need to review all possible sources of serious accidents (e.g., loss of shutdown-heat renoval capability), their probabilities, and their level of seriousness in plants of commercial size.
Considerable use of probabilistic-an6 lysis techniques should be made.
Preliminary conceptual designs should be utilized in the studies as a means for focusing on an inte-grated approach to the solution of problems such as post-accident heat removal.
6-3
%e ACRS believes that, amore the results of the study, there will be a clear indication of the need to devote greater effo to the study of the structural integrity of the reactor vessel internals and the primary and s6condary coolant systems including those aspects related to seismic considerations. %is work should be important both in guiding future research prograns and in the developnent of specific regulatory criteria for MFBRs.
It is understood that DOE plans to redirect its FrmR efforts from the steam cycle to the direct cycle, that its budget for such work is undergoing a nederate rise, and that consideration is being given to demonstration plants.
Furthur, considerable study is being given to the GCFR.
The ACRS recommends that the NRC initiate scoping studies on such GCRs, similar to those described for LMFBRs in the paragraph above, as required in view of the DOE effort.
2)
It is recommended that the NRC initiate studies which place em-phasis on prevention of the CDA.
Current NRC research efforts on the CDA place primary emphasis on understanding the event and its consequences.
Industry and DOE place the greatest emphasis on prevention in their safety research efforts. The ACRS believes the NRC should be in a position to evaluate critically the results of the industry and DOE approach.
3)
It is recommended that the ARSR program study the advantages and disadvantages of alternate containment designs for the U4FBR, incorporating such features as filtered and vented containment.
The research program on aerosol release and transport can provide important input to the design of such containment systems.
We ACRS urges that a suitable program in this area be continued with this objective in nind.
If successful, it could provide data over the next few years that will be extremely useful not only in the design of containment systens but also in assessing the full range of potential radionuclide release from accidents both in W FBRs and LWRs.
4)
It is recommended that the NRC carry out a study to deternine whether new experimental facilities or programs will be needed to denonstrate the validity of natural convection cooling on com-nercial-sized LMPBRs for both pool-and loop-type reactors.
6-4
5)
It is recomended that the NRC evaluate on a continuing basis the need for new large-scal (experimental apparatus,Jarticularly as a result of any new init atives which nay result from the studies recommended in 1, 2, and 4 above.
Because of national budgetary limitations, consideration should be given to building such expensive f acilities on an inter-national cost-sharing basis, should they be needed.
- 6) It is recomnended that the NRC make a strong effort to keep abreast of licensing criteria for advanced reactors in other natio2
- 7) The ACRS recommends continued study of the CDA and the resolution of problems associated with it, such as those related to post-accident heat removal.
%e ACRS anticipates that the ultimate approach to U4FBR safety will involve both a high degree of prevention and some degree of mitigation, and it is desirable that the NRC be knowledgeable in the various phases of the accident as well as its mitigation.
This will require continued development of the SIMMER code.
However, it is doubtful that the code can ever be validated in the sense of precise calculations of such parameters as pressure, temperature, energy release, etc. Rather, the ACRS believes that
'he primary value of the code will lead to increased under-st.anding of the event. For this and other reasons, the ACRS does not believe the code development and " validation" work should carry a priority higher than other work such as that in recom-mendations 1 and 2 above. The ACRS expects that reduction of the code development goals will lead to more modest experimen-tal needs and lower costs than previously anticipated.
It is recommended that greater emphasis be placed on developing a 8) planned, methodical program to keep abreast of and profit from safety research performed in other nations.
%e NRC is to be commended for its efforts to develop interna-tional exchange agreements, and it should now place increased attention on maximizing the benefit from such programs.
6-5
- 7. EXTR WE EXTERNAL PHENO 4ENA 7.1 Objective and Scope Research in this area in intended to provide the NRC Staff with information regarding the frequency and magnitude of extreme exter-nal thenomena that may affect the integrity and safety of nuclear power plants, and with bases for evaluating the resistance of nuclear plants to these phenomena.
Extreme external phenomena are of two kinds:
natural and man-made.
Extreme natural phenomena include earthquakes, tornadoes, hur-ricanes, floods, tsunamis (seismic-induced su rc:M. and light-ning.
Extreme man-made phenomena include aircraft crashes, missiles from turbine rotor failures, and explosions from nearby fixed installations or transportation facilities.
We NRC research effort is devoted chiefly to earthquakes, with a much smaller effort devoted to other natural phenomena; none of the current effort is devoted to the effects of man-made phenomena.
We ACRS considers this distribution of effort to be appropriate in view of the much greater uncertainties associated with the frequency and magnitWe of earthcuakes and their effects on the structures and components of a nuclear power plant, and because of the poten-tial for earthquakes to seek out inadequacies or mistakes in design or construction of all pcrtions of a plant.
We program support funds for FY 79 are about $6.8 million, and represent about 5 percent of the total research budget.
About three-fourths of this amount is devoted to earthquake related re-search.
%is level of funding is appropriate at this time but will have to be increased markedly in future years if the research on earthquakes and their effects is to be carried to a stx:cessful and timely conclusion.
7.2 Earthquake-Related Research Research related to earthquakes can be divided roughly into the following four categories:
a) Research on the quantification of the safety margins in current seismic design.
7-1
b) Research on the ability of structures and components to withstand earthquake-induced motion and fo rces, c) Geological and seismological studies of regional seis-micity in the eastern United States and the Pacific No rthwest.
d)
Miscellaneous studies of soil properties, etc.
The first item, conducted under the Seismic Safety Margins Research Program (SSMRP), is both new and comprehensive and will, if success-ful, provide a framework or methodology for integrating the results from the other work now in progress, especially that under item (b).
At this time, however, the work under item (c) must be considered a separate and major effort.
%e research in these areas is discussed below.
7.2.1 Seismic Safety Margins Research Program This program is new, comprehensive, 1:ry, aM important.
Its objective is "to develop mathematical models that realistically predict the probability of radioactive releases from seismically induced events in nuclear plants."
Ine principal characteristics of this program are:
a)
Its comprehensiveness, involving all phenomena, from the frequency, magnitude, and location of an earth-quake to the response of the plant structures arid components.
b)
Its use of large complex computer codes that require extensive inputs of physical data and extensive experimental research for code assessment and ver-ification.
c)
Its emphasis on a probabilistic approach and on the establishment and quantification of uncertainties or confidence limits.
We ACRS believes that this research program is of great importance and should receive a high priority.
In importance and scope, it is similar to the existing program of LOCA/ECCS research.
It is similar to that program also in its methodology.
It differs chiefly in the fact that the need to quantify uncertainties on a probabilistic basis has been recognized from the beginning and that the program is 7-2
being planned with the hope of avoiding the need to obtain a solution involving every conceivable variable.
Instead, an attempt will be made to develop an approximate best estimate solution that considers those phenomena that contribute most significantly to risk or to the tacertainties of prediction.
%ese objectives are consnendable and have some chance of being achieved on a time scale and at a cost significantly less than that of the LOCA/ECCS research program.
%e ACRS believes that the SSMRP and its adjunct programs must be monitored closely to assure progress toward their objectives and to reassess and redefine, as necessary, those objectives. %e NRC Staff has recognized this need and has established a Senior Research Review Group for this project, assisted by four experts, in addition to the usual Research Review Group. % e ACRS will continue to review the SSMRP through appropriate subcomittees and with the help of its consultants and will report further in subsequent years.
7.2.2 Regional Seismicity We research on the seismicity of the Eastern U.S. was discussed in I* was deemed important Chapter 5 of the 1977 Report to Congress.
Wese and was considered to be progressing at an acceptable rate.
conclusions have not changed.
Studies have been initiated, in cooperation with the Army Corps of Engineers, to evaluate the seismicity of the Pacific Northwest region. Wis is in accordance with needs perceived by the ACRS.
We relation of this research to the SSMRP has not yet been defined.
% is should be done within the next year or two.
7.2.3 Other Earthquake-Related Research '
This category includes several small to moderate sized projects, all related to earthquake effects, most preceding the SSMRP, and all eventually to be integrated into that program. Projects are underway areas relating to:
seismic qualification tests for in in FY 79 analytical and experimental studies of nonlinear scaling equipnent, and modeling, dynamic response and resistance of containments, soil properties and response, and earthquake hazards evaluation.
All of these programs have been developed to meet well-defined user Some are short-term and will be completed well befora they needs.
can be related to the SSMRP.
%e ACRS will assess progress on the integration of these programs in a future report.
7-3 1
7.3 Non-Earthouake-Related Research This research relates to tornadoes, hurricanes, floods and tsunamis, with the greatest emphasis on tornadoes.
All of this research was discussed in the 1977 Report and the comments and recommendations in that report remain unchanged.
We work on tornadoes is expected to be completed in FY 80, at which time the ACRS will review the results and their utilization in the NRC's licensing activities.
Research on the effects of aircraft crashes on the equipnent housed inside structures and research on various effects of turbine missiles is proposed for FY 80.
Ae latter is complementary to the extensive work being done by industrial organizations such as the Electric Power Research Institute.
Wese proposed projects are considered worthy of support at an intermediate priority level.
7.4 Findings and Recommendations 1)
Because research on extreme external phenomena addresses ques-tions relating to the siting of all types of reactors and fuel-cycle facilities, it should be assigned a high priority in the NRC safety research program, and should be funded at appropri-ately increasing levels over the next few years.
- 2) The ACRS believes that the major emphasis given to earthquake-related research is appropriate and undoubtedly will constitute the most substantial demand for additional funding in future years.
- 3) The recently initiated Seismic Safety Margins Research Program (SSMRP) is highly desirable and should become the keystone of the earthquake research program.
- 4) The other earthquake-related programs, including the scientifi-cally oriented studies of regional seismicity, continue to be important but should be reviewed periodically to determine their relation to the SSMRP.
7-4
8.
RADIOLOGICAL EFFECTS 8.1 Introduction Wis chapter concerns research associated with the control of radi-iation exposures resulting from the operation of commercial nuclear power plants and certain supporting activities.
Primary attention is directed to studies necessary for improved understanding and con-trol of exposures to workers inside tha plants, and for the evalua-tion and control of radionuclide releGas from the plants under both routine and accident conditions. Ancillary subjects include research to improve understanding of the health effects of low-level radiation exposures, refinements in quantifying the Imrameters governing the movement of radionuclides through the environment, improvements in the methodology for energency planning and decontamination and re-entry procedures following an accident, and radiation exposures associated with the transportation of radioactive materials.
8.2 Occupational Exposures 8.2.1 Improved Da_ta Base Operating experience has shown that occupational doses are increasing in commercial nuclear power plants. Currently there is inadequate information to enable the development of specific guidance for exposure reduction at nuclear power plants. % e ACRS reconmends that the NRC support investigations at operating plants to identify and order the tasks according to their contribution to occupational exposure so that guidance for the control of associated exposures can be developed and issued expeditiously.
8.2.2 Exposure Sourcea and Control Re ACRS recommends increased research to determine the basic factors that govern radionuclide buildup in reactor cooling systems, includ-ing the possible influence of operating practices such as rapid tem-perature variations, load following, end-of-fuel-cycle operation, and variations in coolant chemistry.
The ACRS believes that research to develop improved means for remov-ing radioactive materials from the primary coolan circuit would be helpful.
Before such systems can be developed, however, there is 8-1
a need for the procurement of much better data on the chemical and physical properties of the deposited material, itself, and on its mechanisns of formation and deposition.
8.2.3 Research on Associated Health Effects In its 1977 report, the ACRS reconnended that the NRC not devote a significant effort to research on the health effects of low-level ionizing radiation.
In offering this recommendation, the ACRS was referring specifically to large-scale laboratory studies of health effects in animal populations exposed to known amounts of low-level external radiation or of internally deposited radionuc-lides. The ACRS reaffirms this recommendation.
On several occasions during the past year, however, the NRC Staff has been asked to comment on epidemiological studies made by investi-gators who believe that there have been significant increases in morbidity and nortality attributable to radiation exposures among people associated with NRC licensed facilities.
In several cases, however, the investigators appear to have omitted certain key factors from their calculations, either because they were not available through the given data base or because their significance had been overlooked.
Because evaluations of this type will undoubtedly continue to be nade in the future, and because dose data reported to the NRC will continue to serve as one basis for such studies, the ACRS makes the following reconnendations:
a) The NRC should arrange to have its data banks on human occupational radiation exposure examined by qualified epideniologists, specifically with the view towa rd developing data that can be more effectively utilized in determining relationships between radiation exposures and various health effects.
A portion of this exanination night include attempts to use the existing data in epideniological assessments.
Such exercises could also be an effective tool for revealing deficiencies in the current systen.
b)
If the reconnended review results in a need for an expansion in the current occupational dose recording system, the NRC should provide resources, as appropri-ate, to support the necessary changes.
8.3 Population Exposures from Routine Releases 8.3.1 Radionuclide Removal Systems Although application of the as-low-as-reasonably achievable (ALARA) criterion has resulted in what appears to be adequate control of 8-2
routine radionuclide releases from commercial nuclear power plants, problems still exist to which attention should be directed.
a)
It is recommended that research be conducted to better define the variables in the data yielded by various methods for testing the efficiency of adsorber and filter systems within nuclear power plants.
Statistical methods need to be applied to define the errors in the sampling and analytical procedures used in such tests, to establish their confidence intervals, and to specify criteria for the acceptability of the test results.
b)
The ACRS recommends a modest research effort to promote the evaluation and application of the latest in radionuclide removal, adsorption and filtering systems in those circumstances where they can be shown to be cost-effective.
8.3.2 Dose Assessment Methodology for Routine Releases The ACRS continues to believe that there is a need for improvements in the methodology for calculating radiation doses to population groups residing in the vicinity of nuclear facilities. This includes the need to verify the accuracy of dose calculational procedures; the improve criteria for assuring the quality, representative-need toand cost-effectiveness of environmental samples; and the need
- ness, for better information on the particle size distribution and chemical form of airborne releases as a function of time and distance from the point of discharge.
We ACRS notes that efforts for the improvenent of meteorological models appear to be adequate; greater ef fo rt, however, is needed in the improvement of models for assessing the transport of radionuclides through the aquatic, marine and terres-trial pathways.
Were is also a need for improved understanding of those aspects of the actions and habits of people that may modify their exposure to, or intake of, radioactive materials.
Additional comments on this subject are included in Section 9.3.
Because of increased interest in population exposures from the front end of the fuel cycle, there is a need for better data on the bio-logical half-lives of the various naturally occurring radionuclides associated with uraniun mining and milling.
In addition, there is a need for the development and application of techniques that can be routinely used to check the accuracy of the dose estimates made on the basis of various mathematical models.
This might involve, for example, more widespread use of whole body counting o people who r
live in areas where data are available on radionuclide cor centrations in their surrounding environnent.
8-3
8.3.3 Extension of AINIA Criterion The ACRS continues to believe that there is a need for research to provide basic guidelines for applying ALARA principles to a wider range of environnental problems associated with nuclear operations.
These problems include those associated with uranium mining and milling, fuel fabrication, and the development and employment of alternate fuel cycles.
During the past year, the NRC has initiated research in several of these areas; such work is to be encouraged.
8.4 Emergency Planning Recently an NRC/ EPA Task Force completed a draft of a revised docu-nent which contains proposed generic guides for the use of state and local governments in emergency planning.
We ACRS plans to review this report and offer comments.
Specific needs in this subject area include:
a) Research on steps that might be implemented in the recovery and re-entry phase following an accident.
This would include procedures for making medical decisions concerning the affected population, methods for decontaminating and reclaiming land, buildings, and equiprent, and the establishment of dose limits or guides for population groups desiring to return to areas that have been evacuated.
b) Research on the developnent of instrunentation, and nethods for quick interpretation and estimation of the timing, nature, and quantity of radionuclide releases in the event of a serious accident.
8.5 Transportation of Radioactive Materials Research currently underway is addressing the susceptibility of shipping containers to sabotage and includes the development of a series of scenarios for the full range of accidents thought to be possible in the transportation of radioactive naterials. The ACRS is impressed with these efforts and believes that this subject area is being addressed adequately.
8.6 Long-Range Planning Most of the research efforts of the NRC on radiological effects are directed to immediate problems.
We ACRS recommends that a modest portion (perhaps 10 percent) of the total research funding in this 8-4
subject area be allocated to efforts designed to address problems on a longer range basis, before they become major issues. The selection of problems for attention rught be guided by a panel which has input from a variety of organizations, including potential user groups.
8.7 Findings and Recommendations In general, the ACRS believes that the research efforts of the NRC Staff on matters pertaining to the assessment and control of occupa-tional radiation exposures and environmental radionuclide releases associated with the operation of commercial nuclear power plants are well focused and are progressing satisfactorily.
However, there are areas to which the ACRS believes increased atten-tion should be directed.
In this regard, the ACRS recommends that the NRC:
- 1) Gather better data on the identification and importance of the various operational tasks that contribute to occupational exposures so that better guidance for control measures can be developed and applied.
Spe-cific attention should be directed to a determination of the basic factors that govern radionuclide buildup in reactor cooling systems, including the influence of operating practices such as rapid temperature variations, load following, end-of-cycle operation, and variations in coolant chemistry.
- 2) Arrange to have the NRC data banks on occupational radiation exposures examined by qualified epidemi-ologists.
3)
Initiate research studies to define better the vari-ables in the data yielded by various methods used for testing the efficiency of adsorber and filter systems within nuclear power plants.
- 4) Continue to develop improved methods for, and the data base supporting, the calculation of radiction doses to population groups residing in the vicir.ity of nuclear facilities.
5)
Expand its research to develop instrumentation and methods for the rapid interpretation and estimation of radionuclide releases in the event of a serious accident, and initiate research studies on methods for decontaminating and reclaiming land, buildings, and equipnent, and for establishing dose limits for popula-tion groups desiring to return to areas that have been evacuated following a nuclear accident.
8-5
- 6) Allocate a modest portion (perhaps 10 percent) of its total research funding on radiological effects to ef-forts designed to address problems on a lorger range basis, before they become major issues.
8-6
/
9.
WASTE MANAGEMENT 9.1 Introduction he primary objective of a radioactive waste management procram should be to control and minimize, to the extent reasonably achiev-able, both the maximum individual doses and the collective population doses resulting from waste management operations, including ultimate disposal. The ACRS review of the NRC waste management research program has shown that its efforts toward achieving the above objec-tives are uncoordinated and unfocused.
Wis is due, at least in part, to similar deficiencies in the national program.
Within the NRC program, there appears to be a lack of a systematic process for identifying research needs in this field and for assigning priori-ties for their accomplishment.
There also appears to be a lack of adequate interaction and communication among the several NRC groups involved.
Wis is particularly true in the comnunication of waste management research needs to the Of fice of Nuclear Regulatory Research. %ere is also a need for better interaction and connuni-cation between the NRC and other governmental agencies having responsibilities in this area.
Covered in this chapter are the research efforts needed to:
a) Determine risks b) Establish acceptable individual and collective popula-tion dose limits.
c) Develop improved data for estimating population doses, d) Establish : ~Julatory criteria.
e) Develop criteria for decontamination and decommissioning operations.
9.2 Risk Assessment 9.2.1 Risk in Perspective Just as for the operation of connercial nuclear power p.tants, there is a need to determine an acceptable level of risk for that part of 9-1
the fuel cycle that involves the nanalenent of radioactive wastes.
Ihis involves sone narticularly inportant considerations, the first of which is the requirenent to c'evelop an acceptable nethodology for naking suitable risk assessments.
Next, there is a need for conpara-tive evaluations of the several options available within the nuclear fuel cycle as well as the health haza rds associated with various fossil fuel options.
And finally, there is a need for a broad perspective on risk in all aspects of society.
Be risks associated with a properly designed and operated radio-active waste disposal facility will be prinarily of two types:
(a) those associated with the doses incurred in handling, shipping, and emplacement of the wastes; and (b) those associated with any re-leases fron the site over the long tern.
Because of the rapid decay rate of nany of the radionuclides within such wastes, the second category of hazard will decrease substantially within the first few hundred years of storage.
9.2.2 Risk Estinates The ACRS reconnends that the NRC intensify its research ef forts to identify the dominant contributors to risk in radioactive waste nanagement operations and to quantify their associated uncertain-ties.
For deep geologic and other ultinate disposal nethods, these assessments should include estinates of the rates with which radio-active naterials will escape the repository and enter the human environnent, coupled with estinates of the nagnitude of the resul-tant consequences.
9.3 Dose Calculation Models he dose received by the public is rarely neasured directly because it is so low.
Rather, a calculation is naie based on release rates, transit nodes, and uptake from the various pathways.
In setting up conputer programs for naking population dose estinates, at least four sutriodels are used. These are:
a) The physical transport of the radionuclides through the atmosphere, hydrosphere and ]ithosphere.
b) The uptake, retention and transport of the radionuc-lides by biota (ecosystens) c) The diet, age disttibution, life-style, food distribu-tion patterns, and other relevant characteristics of the receptor population.
9-2
d) The uptake, distribution and retention of the radio-nuclides within the human body.
item has been rather thoroughly explored and Although the last relevant data tabulated, comparable guidance is not available on the multitude of parameters involved in the first three items.
For the ACRS recommends that the NRC investigate the this reason, research needs for making population done estimates, and direct the establishment of appropriate numerical values for attention to those parameters where data are lacking.
Special attention should the long-term transport of the transuranics and long-be given te lived fissi >n products in geologic structures and to the uptake and retention of such radionuclides by plants and animals.
Current research by the U.S. Department of Energy on the migration of radionuclides from past underground nuclear tests in Nevada holds some promise of providing data that will meet a portion of this need.
9.4 Waste Siting and Handling Criteria the NRC should give priority attention to
'lhe ACRS believes that research on waste management approaches and safe disposal cri-teria. Among the factors to be considered are:
a) Site criteria, from the standpoint of hydrology and geology, meteorology, and seismicity.
Criteria for limiting any potential effects that the b) wastes might have on the characte.tistics of the storage site, especially on a long-tern basis.
Criteria for the types of containers that must be c) provided for various chemical and physical foms of the wastes.
d)
Criteria for supplemental radionuclide trapping or retention systems as a function of potential changes in the chemical and biological behavior of the wastes with time, and the nature of the products resulting from the decay of the initial wastes.
Criteria for the necessary reliability of heat e) removal and shielding provisions based upon estimated thermal and radiation release rates from the wastes.
9-3
f) Criteria for determining acceptable levels of migra-tion of radioactive materials from or within a site and acceptable nonitoring procedures to detect and evaluate such migration.
9.5 Decontamination and Decomissioning Operations For some time the ACRS has stressed the need for the develognent of design criteria for nuclear facilities that will facilitate decon-tamination and decommissioning operations, but not significantly affect the operational safety of a plant.
The NRC has research studies underway on this subject.
9.6 Other Problems _
Related problems that need to be addressed include:
Evaluation of Alternate Fuel Cycles. We 7.CRS believes that a portion of the NRC research effort on waste management should be directed to wastes generated in possible alternate fuel cycles.
Containment of Airborne Releases.
01anges in the 1977 Clean Air Act (Public Law 95-95) may make it mandatory that techniques be developed for collecting, immobilizing and. storing certain volatile and gaseous radioactive wastes.
In addition to the related risk assessment research currently underway, the NRC should examine other potential research needs that may become necessary because of this legislative action.
9.7 International Cooperation he ACRS urges that the NRC Staff take maximum advantage of inter-national developments in this field through cooperative research programs with ot' er countries.
9.8 Findings and Recommendations The problems in this subject area are important and will require significant amounts of research.
We current NRC program appears to be uncoordinated and unfocused and progrees to date has been inade-quate, at least in part because of similar deficiencies in the national program.
We immediate needs in this area are to establish a plan for the research and to improve internal NRC management control.
9-4
About $4 million is being budgeted for waste management research within NRC for FY 79.
Although projections are that this will be increased substantially in FY 80, plans are to decrease the budget in subsequent years. On the basis of its review, the ACRS believes that substantia) funds may need to be directed to this problem area for some time to come.
In terms of future NRC research activities on waste management, the ACRS believes that emphasis should be directed to the following areas:
1)
Identification of the dominant contributors to risk
'n radioactive waste management operations and quanti-fication of the uncertainties in the risk estimates.
- 2) Continued development of criteria fo r the design and operation of redioactive waste disposal and storage facili* ties.
This applies to low-level as well as high-level wastes.
- 3) Development of quantitative data for the more signifi-cant parameters involved la the calculation of the physical and biological transport of radionuclides within the environnent.
- 4) Continued development and implementation of licensing criteria to facilitate the decontamination and decon-missioning of nuclear facilities.
5)
Evaluation of research needs on waste disposal problems associated with alternate fuel cycles.
9-5
10.
SAFEGUARDS AND SECURITY 10.1 Introduction NRC work on safeguards and security is generally concerned with the possibility that some individual or group might attempt to gain possession of significant amounts of special nuclear material (SNM) either by, stealth or by force from a licensed facility (as from a reprocessing or fuel fabrication plant or from a trensport operation) or might attempt to sabotage an operating reactor cr other licensed facility in such a way as to release a large quantity of ra'.ioactive material.
Related concerns such as a group gainicq cantrol of some facility and attempting to use the threat of daw to obtain ransom must also be considered.
Problems of many dif ferent types are raised by these concerns, including suitable instrumentation for detecting and measuring various materials, alarm systems and barriers to protect against intrusion, security force procedures, and methods of assessing reliability of personnel. Aspects of these questions have been under study by the Department of Energy (DOE) and its predecessor agencies for many years as a result of their having their own operat-DOE still has ing nuclear facilities and large inventories of SNM; an impressive program of studies in progress.
Closely related problems have been of concern to other government agencies.
The relevant parts of the results of such studies are available to the NRC.
Because it is engaged in the regulation, licensing, and inspection of facilities operated by others, and because it must be able to assess the efficacy of provisions and procedures proposed by others, there are many questions confronting the NRC that are different from those of direct _ interest to DOE or other goverrment agencies.
For these questions, there is a need for the NRC to develop new insights, additional information, and new and improved techniques.
10.2 Funding NRC funding for safeguards and security research for FY 79 is fore-cast at $6.2 million, which is about 5 percent smaller than it was for FY 78.
In addition to research, about $4 million will be devoted by other NRC program offices to technical assistance contracts in the 10-1
same field.
These latter generally consist of projects of a clearly defined nature to obtain info:mation for which the need is specific to the office concerned (as, for example, might be the case with respect to the quite particular needs arising in connection with inspection).
During the past year the STAR Group (STAR for Safe-guards Technical Assitance and Research) has been formed, with a chairman from NMSS and representatives from the other program of fices, which reviews all safeguards projects proposed by the various offices to coordinate such proposals so as to aveld dupli-cation.
This arrangement appears to be working effectively, al-thotgh there may also be a need to improve communication between the offices in order to maintain a fully rounded and complete. program.
10.3 Research in Progress Major problems being addressed include the development of improved methods for: material inventory control and accounting; the evalua-tion of integrated safeguards systems for fixed sites; the evaluation of syrtems and tactics for the protection of S!N in transit; and the identification, selection, and assessment of design features that would reduce vulnerability to sabotage.
Methods in the nature of
" war games" have been developed for simulating combat situations and have been applied in a few selected cases.
Some sttriies have been made to determine the minimum set of reactor components and vital areas that must be protected from damage by saboteurs in order to preclude unacceptable consequences. In connection with all these studies consideration is being given to the effects on plant safety and normal operation as well as to relative costs of the options investigated.
With respect to most of these topics, a main need of the various NRC offices will be that of having instrumentation, procedures and methods in hand to assess the effectiveness of the provisions which may be proposed, to be able to call attention to favorable options where improvements might seem to be warranted, and to ensure that installed systens and procedures are operating as intended.
Obvious examples would be to have neans of ensuring that alarm systers and personnel access provisions are properly maintained, or that proper responses on the part of security forces will in fact be made. Such questions are being given attention in the present program.
'Ihe specific matter of protection against diversion of SNM, which has generated much public concern, has also been given a great deal of technical attention, both to naintaining good materials control and accountability and to monitoring for attempts to remove small amounts of material.
For a number of years, the DOE has sponsored 10-2
large programs in these areas, resulting in greatly improved equip-ment and techniques.
These developnents are available to the NRC and are being taken into account but additional work is needed.
%e NRC program is addressing means to ensure that the calibration and handling of instruments and procedures used by licensees for the measurement of SNM will provide results which are reliably related to absolute standards.
10.4 Additional Research Needed Particular questions related to possible alternative fuel cycles are not being direc ly addressed in the present program.
An example of such a problem is that of the measurement standards and procedures which may be needed for material control and inspection.
Although work undertaken now may never be used if a particular fuel cycle option is not adopted, some of the requirements in this field involve quite finicky and necessarily slow-moving work which will take several years to complete after a program has been initiated. Wough it may be the proper policy at the moment to defer such sttriies until decisions are reached as to which new fuel cycles are to be con-sidered for actual use, there will be a st rong need to add such studies to the present progran as soon as fuel cycle decisions are made.
In last year's report, the ACRS recommended that additional effort be directed to several items, including:
(1) alternative fuel cycles; (2) alternative (below grade) locations for spent fuel pools; (3) dedicated heat removal systems; (4) increased separation of redundant safety-related systems; and (5) studies related to means for locating diverted SNM.
Of these, item (1) has been discussed above; (2), which is not currently under investigation, still should be examined when feasible; (3) is included in NRC's new program of research on improved safety systems; (4) is receiving some attention in connection with new plant designs; and (5) has been and continues to be a function of the DOE.
10.5 Evaluation and Comment he particular problems being addressed, and those planned for EY 79, are all relevant to NRC needs.
Though it has not been possible to accommodate all the items deserving attention within the present program, the selection of topics for current funding would seen to be quite reasonable. The contractors chosen appear to be fully capable.
Good use is being made of information availab]e from other sources.
With one potentially worrisome exception, discussed below, the research program on safeguards and security appears to be well fonnu-lated and well managed.
10-3
In the opinion of the ACRS, the current funding level for this program is adequate unless either the perceived threat level of forcible attempts at theft or sabotage should increase sharply, or the nation's policy should call for the early adoption of new fuel cycles or new reactor types.
In either of these events, it might become appropriate to increase the funds available.
In connection with the protection of reactors against sabotage there is a natural tendency to include all " safety-related equipaent" among the items to be defended.
Some attention has been given in the present progran to the question of just what night constitute the absolutely essential set of components requiring protection. It is by no means obvious that all of the equipnent defined as safety-related is essential to ensure public safety in the event of sabotage.
If, after thorough study, it should be confirmed that some particular set of components is necessary and sufficient for that purpose, such a finding could be of great value -- particularly with respect to existing plants -- in concentrating and focusing attention if any additional protective measures should be decided to be necessary.
An appreciable fraction of the work in progress is devoted to the formulation of computer nodels to sinulate one or another physical situation -- an activity often, though obscurely, described as "develognent of nethodology."
For some purposes, such as that of obtaining a convenient material accounting system, the computer routine itself may constitute the desired end-product, and may be of direct practical use.
For some other purposes, such as that of simulating a combat situation, the con.puter model, if it has value at dll, has value only as a tool.
Some of these war-ganc-type codes have features, such as logical or arithnetic structure, in nany respects similar to features found in codes for probabilistic risk assessnent of reactor accidents.
Bere are, however, at least two cignificant differences. One is that the data base for reactor accident assessment, no matter how inadequate it may be on sone points at any particular time, is in principle capable of being filled in and inproved.
% is is less obvious with respect to conbat models where some paraneters are at best narginally quantifiable.
Perl.aps more significant is the point that, for the reactor case, the systen whose behavior is being studied is well specified (so nany pumps, valves, etc.) and even a single calculation may provide results of interest.
In the present casa, the results of interest are not so much the indicated outcome af any particular combat as the comparisor. of the effects of alternative systen characteristics on the outcome of a whole spectrum of possible, sombat situations.
It is possible to use code nodels to obtain useful infornation of 10-4
this sort; but it requires the thoughtful comparison of the results of series of calculations in which parameters have been varied to elucidate relative effects.
In this way, for example, the relative value of increasing alarm capability versus improving barrier fea-tures could be assessed, and ultimately the elements of an optimized and balanced system could be ascertained.
It is not evident to the ACRS that, in supporting the develorrient of some of the codes of this type, the NRC staff has given sufficient attention to the particular questions that the code might enabb them to answer, nor to the difficulties which may be entailed in using the codes to obtain the desired assessments which, in turn, would consti-tute the only significant end-product of the develorrient.
On the same general matter, there are the following two additional comments.
The first is that the NRC will indeed require some means of gauging the relative usefulness of particular security system items (such as some new and more sensitive alarm device proposed by some supplier) to serve as a basis for deciding whether or not the proposed item improves the system in an overall sense to the extent that the use of it or some equivalent should be required of licen-Should computer codes prove to be the most effective tool for sees.
such a purpose they would probably have to be applied, and their results assessed, by or on behalf of the staff. As a second point --
and again assuming that the present or similar computer madels can serve to provide useful information -- although the schema of the codes will no doubt be in the public domain, the details of input and results of applications to specific situations will almost certainly deserve protection, since meaningful studies of this sort will disclose weaknesses as well as strengths, and will also indicate available margins.
10.6 Findings and Recommendations
- 1) Plans for the research on safeguards and security should provide for a progran at about the present level of effort for at least the next several years, with some allowance for the possibility that it may be necessary to increase the level should the national policy call for early adoption of new fuel cycles or new reactor types.
- 2) Studies should be made to determine whether the use of alternative fuel cycles would change significantly the nature or importance of the types of safeguards measures now being studied.
'Itese studies should include also an estimate of how soon new questions might arise and how long would be required to solve them.
10-5
- 3) The matter of determining the minimum number of essential com-ponents which, if fully protected, could enable a plant to withstand sabotage of other components deserves the maximum emphasis which can usefully be put on it.
4)
It is recommended that, in connection with the developnent of computer codes directed at security problems, the NRC Staff give careful prior attention to the type of question for which the code might provide answers, the use to which such answers would be put, and the amount of effort likely to be needed to obtain them.
10-6
- 11. RISK ASSESSMEMP 11.1 Objectives Risk assessment research, the responsibility of the Probabilistic Analysis Staff (PAS) of RES, has as its mission the development of methods for, and the promotion of the application of, quantitative risk assessment to assist the NRC staff in carrying out its various responsibilities.
Its activities span a spectrum from research aimed at the developing and testing of new methods, to application of these methods to problems whose solutions are needed to reach de-cisions in a number of licensing, inspection, and program-planning areas.
11.2 Scope
%e PAS has as a tool the fault-tree / event-tree methodology of the Reactor Safety Study (WASH-1400)* which can provide significant insights into the behavior of reactor systems from a probabilistic risk viewpoint; however, this methodology and the results obtained from it are only beginning to be used in the regulatory process. The PAS thus finds itself initiating new activities for which it sees a need, providing guidance and assistance to those divisions of NRC that are attempting to apply the methods already developed, and specific applicaulons of immediate import to some NRC working on staff responsibility.
Although a significant fraction of PAS activity is research, much of what it does is a direct application of earlier research to immediate problems.
Wis situation is de-sirable, but requires continuing oversight to ensure that a proper balance is maintained between research and application.
We pres-ent balance seems appropriate.
A significant fraction of the research and develornent for which PAS has responsibility is done by its own professional staff.
- U.S.
Nuclear Regulatory Commission, Reactor Safety Study: An As-sessment of Accident Risks in U.S. Commercial Nuclear Power Plants, WASH-1400 (NUREG-75/014), October 1975.
11-1
Typical of research and developnent efforts are:
Developnent of a risk assessment method aimed at quanti-fying fire risks and consequences.
Developnent of computer codes dealing with fault tree manipulation, the effects of testing and maintenance on system and component reliability, and a systematic treatment of common cause failures.
Description and analysis of human errors observed in connection with operating reactors.
Efforts to define an appropriate program of research to examine the question of acceptable risk.
Typical of work that is primarily application of risk assessment to existing or anticipated problems are:
The application of the WASH-1400 risk assessment methods to four different IMR plant designs.
We develoinent of criteria for outage times and surveil-lance intervals for systems and components.
We developnent of a model to predict flood occurrence probabilities, associated system failures, and resultant consequences.
Efforts to model the behavior of parameters important to safety in the behavior of a radioactive waste depository located in deep geologic media.
Developnent of a model for calculating risks to reactor plants due to transportation of non-radioactive hazardous materials nearby.
A study ot emergency responses to reactor accident se-quences.
11.3 Relation to the Needs of the NRC he work being done by the PAS and that being planned appear to be relevant to the needs and responsibilities of NRC.
The PAS is taking the initiative in defining and developing new areas of 11-2
investigation.
Computer code developnent programs, although even-tually responding to NRC needs, is primarily in this category.
The PAS work to collect, correlate, and evaluate performance data is also being done primarily as a result of PAS initiative.
Research on flood risk analysis, fire risk assessment, and the analyses of Class 3-8 accidents for use in environmental reviews is in direct response to requests from various other groups within NRC.
The recent increase in professional staff represents an increase in level of activity commensurate with increasing applications of risk assessment in the licensing and regulatory activities of the NRC. These applications are likely to increase.
It is important that the PAS continue to recognize that risk assessment is not an end in itself and that, although the PAS will continue to be responsible for initiating and assisting in the developnent of new projects, methods must be taken over and used by other divisions as soon as feasible.
11.4 Progress and Results of special note are the activities of the PAS in improving the methods first developed in WASH-1400 for predicting consequences of the release of radioactive materials in reactor accidents.
Various aspects of this part of the Reactor Safety Study have received serious criticisms, and a major effort is being made by the PAS to improve the method.
The basic vehicle now being developed for consequence prediction is called the CRAC Code.
It is designed to sample statistically a large population of atmospheric situations and to model a large number of atmospheric phenomena and site character-istics.
Results are expected to predict consequences in some repre-sentative situations.
Although progress is being made in improving the model, there are indications that it still has deficiencies that require further effort. This is an activity which should be pursued with diligence.
The PAS is nearing completion of a study that extends the effect of liquid-borne activity on reactor accident consequences beyond that carried out in WASH-1400.
Another study extends the WASH-3400 study to light-water reactors of different designs.
This new study includes a reanalysis of the dominant accident sequences using improved models. Special attention is given to analysis of systems designed to mitigate accident conse-quences and to accident analyses which provide a more advanced treatment to release magnitudes.
In addition, attention should be called to the beginning of a program to assess the risks associated with deep sea bed disposal of wastes.
11-3
Wis program is of special importance, both because of its possible long term implications and because it will require the international cooperation that is necessary for a permanent solution of the waste problem.
We ACRS believes that existing and planned programs of the PAS are responsive to the recommendations of last year's report.
11.5 Findings And Recommendations Risk assessment is an expanding area and needs for both developnent of new techniques and applications of existing methods are likely to g row.
The ACRS has not found any serious gaps in the existing program. However, a number of items deserve emphasis.
- 1) As the PAS and others have observed, and as the Risk Assessment Review Group (RARG) Report
- emphasizes, accurate risk assessment requires a data bank of performance histories of components and systems.
The PAS is working within the NRC and with others to collect and evaluate data.
It should continue to emphasize this activity and also should provide guide-lines to ensure that apprcoriate infonnation is re-ported to those responsible for collecting reactor system performance infonnation.
- 2) A point of continuing concern in connection with accident consequence prediction is the appropriate description of biological effects of radiation.
We BEIR Committee is scheduled to release a report within a few months. The ACRS recommends that the consequence calculations be re-examined in light of the recomren-dations of that report when it is released.
- 3) Many of the PAS research projects result in sophisti-cated computer codes applied to specific systems with assumption about such items as failure modes and uncertainties on data.
%e ACRS believes that there is a need for quality assurance in the methodology and application of probabilistic analyses.
The ACRS recommends that a systematic method of evaluation be developed which includes the necessary documentation of assumptions needed to enable peer review.
- H. W. Lewis, et al., Risk Assessment Review Group Report to the U.S.
Nuclear Regulatory Commission, NUREG/CR-0400, September 1978.
11-4
- 4) Many comments, including those in the RARG Report, have stressed the importance of further development of methods to eva.luate more quantitatively the contribution of human error to risk.
It is equally important that the contribution of operator adaptability be evaluated, because it may be a significant contributor in decreas-ing risk.
An accurate evaluation may well provide insights into improvemen's in operator selection and training which could be implemented to further enhance safety of reactors.
- 5) After exchanges of correspondence with the EPA, the NRC agreed to undertake a study to determine acceptable levels of risk.
This subject is of significance not only to ;he NRC but to virtually every organization making decisions that could affect the health and safety of the public.
The ACRS believes that such studies are very important and there is a need for consideration of acceptable risk by each such organization. However, the ACRS believes that there is need for a comprehensive re-search program with the goal of defining potential criteria for societal risk acceptance, conducted with broad support from the many federal departments and agencies involved in such decisions, and conducted under the auspice of an organization not tied directly to the problems of any specific activity or regulatory decision.
6)
Finally, the ACRS recommends t, hat careful consid ration be given to the recommendations of the RARG Report.
11-5
12.
IMPROVED REACTOR SAFETY 12.1 Objectives
%e basic purpose of this research is to investigate concepts that have the potential for improving the safety of light-water reactors.
%e FY 78 Budget Authorization Act for the NRC modified Section 205 of the Energy Reorganization Act to require that the NRC prepare a long-range plan for the development of new or improved safety systems for nuclear power plants.
In its 1977 Report to the Congress, the ACRS recommended that the NRC become more involved in research that has the potential of leading to the developnent of improved safety system co x:epts, and said that "It is both desirable and appropriate for the NRC to conduct research on new safety concepts, but their developnent and implementation should be carried out by the nuclear industry or the Department of Energy."
12.2 Scope The Office of Nuclear Regulatory Research submitted to the Congress on April 12, 1978 its " Plan for Research to Improve the Safety of Light-Water Nuclear Power Plants" (NUREG-0438).
This plan selected five research projects as having significant potential for improving the safety of light-water nuclear power plants, and recommended them for the initial phase of the program. We projects are:
A.
Alternate containment concepts, especially vented containments.
B.
Alternate decay heat renoval concepts, especially bunkered systems.
C.
Alternate emergency core cooling concepts.
D.
Improved in-plant accident response.
E.
Advanced seismic designs.
We report recommended also that work be initiated on two additional projects, as follows:
Improved methodology for evaluating research topics.
F.
G.
Scoping studies on eleven other topics.
12-1
The scoping studies were to be performed to determine whether re-search on these other topics would be warranted in the NRC's future program on research to improve safety.
Rese other topics include, for example, improved plant controls, reactor vessel rupture control, core retention measures, equipnent for reducing radioactivity re-leases, and improved plant layout and component protection.
NUREG-0438 estimated that most of the research projects proposed would require one to two years for completion. W e NRC noted that no funds had been budgeted for these research projects as of the time of submission of the plan and estimated that implementation of the proposed plan would require aoout $15 million over a three-year period from the time work was started. We NRC noted that additional funding is likely to be required in future years because the scoping studies as well as other efforts may identify more projects that should be undertaken.
12.3 Evaluation he ACRS reviewed the draft NRC plan for research on systems to im-prove safety in February and March,1978 and approved the plan in a letter dated March 13, 1978 to NRC Chairman Joseph M. Hendrie, which quoted below, in part:
"The proposed program has been developed in response to the requirement by Congress in the FY 78 Budget Au-thorization Act for the NRC.
Although the pertinent section of the Act bears the subheading, " Improved Safety System Research," the wording of new subsection (f) re-fers to "... projects for the developnent of new or im-proved safety systems..." %e NRC Staff has recognized, and pointed out in its report, that the requirement for "developnent," if interpreted literally, could com-promise the position of the NRC as an impartial judge of safety systems incorporated into nuclear plants.
The NRC Staff has proposed, therefo re, that its program be limited chiefly to the evaluation of new concepts for im-proving reactor safety.
We Conmittee agrees with this approach.
In its recent report to the Congress (NUREG-0392), the Committee stated:
'...Tne ACRS believes that the development, testing, and proof of efficacy of new or improved safety systems should not be the responsibility of the NRC, but should be conducted by the nuclear industry or DOE.
12-2
However, the ACRS believes that it is a proper and even necessary function of the NRC to perform or sponsor research on con-cepts that, if developed and implemented by the appropriate bodies, could lead to im-provements in safety.'
"'Ihe NRC Staff has recommended five research projects as having the greatest prospect of leading to improved safe-ty...."
"The Committee concurs in these choices and believes that these studies should be undertaken even though t!.eir risk reduction potentials are not yet clearly known.
These studies and the follow-on programs will serve to place in perspective the extent and suitability of possible safety improvements.
"The NRC Staff has stated in its report that nost of these research projects will require only one to two years for completion, the possible exceptions being Projects A and Although these five projects in themselves would not E.appear to represent the sort of "long-term plan" requested by the Congress, the NRC Staff has proposed that two addi-tional programs be undertaken, as follows:
Improvement of the methodology for evaluating F.
research topics and alternate plant designs.
G.
Scoping studies of the eleven additional re-search topics that have been suggested.
"These programs can be expected to provide a basis for a longer term effort.
"The Committee believes that Project F on the developnent of better methods for evaluating concepts proposed to improve safety is essential to the success of this new effort.
Although there will always be a large subjec-tive or judgmental element in the selection of research projects on improved safety, these selections should be made on as quantitative and factual a basis as practical.
It seems evident also that it will be extremely difficult to provide a suitable nethodology without at some point addressing the question of how safe is safe enough."
12-3
12.4 Relation to Work Being Done by Others The ACRS has been advised that the DOE is formulating and initiating a research and development program on inproved safety for light-water nuclear power reactors.
At least in its initial phase, this progran will be devoted to measures intended to reduce the probability of occurrence of serious accidents, rather than to neasures, such as modi fications in containment design, which are intended to reduce the offsite consequences of potentially serious accidents.
A study on underground siting performed on behalf of the State of California concluded that the vented, filtered containment (chosen by the NRC for study in Project A of NUREG-0438) offered more prom-ise.
%e California Energy Commission has endorsed this conclusion.
Studies on underground siting, as well as considerable research on
.tenomena related to the behavior of a molten WR core within con-tainment, have been underway for some time in the Federal Republic of Germany and are continuing.
Germany has already implemented re-quirements for a dedicated, protected-access shutdown heat removal system on its newer light-water power reactors.
12.5 Progress and Results We program plan in NUREG-0438 was sent to the Congress on April 12, 1978.
No funds were authorized or reprogranned for this program in FY 78, and there was no progress.
%e NRC FY 79 authorization bill earmarks $1.5 million for this program; however, the FY 79 appropriation bill is not specific in this regard.
As of December 7,1978, the ACRS has been advised that the NRC hopes to reprogram $800,000 for research to improve reactor safety in FY 79 and plans to initiate work on vented containments, alternate decay heat removal concepts, and value-impact methodology.
Work scopes have not yet been prepared for the projects on alternate emergency core cooling concepts, advanced seismic design, or the scoping studies.
As of December 7, 1978, the NRC Staff does not antici'pate beginning work on these projects in FY 79 and possibly not on improved in-plant accident response.
12.6 Findings and Reconinendations
% e ACRS approved the Program Plan presented by the NRC in NUREG-0438 and believes that this research should be given high priority. %e ACRS believes that the delay in initiating and implementing the proposed NRC program on research to improve safety is unfortunate 12-4
since the prryf ram could have benefited from the preliminary efforts poqsible with modest support in FY 78.
The ACRS recommends that the program receive substantial funding ($1. 5 million) in FY 79, by reprogramming of other NRC funds if necessary.
The ACRS recommends that in subsequent years, this program be funded at the level needed to permit effective pursuit of all the research projects and the scoping sttriies in NUREG-0438. %e ACRS recomnends further that emphasis be given to the work on alternate containment concepts, on bunkered dedicated shutdown heat removal systems, on improved in-plant response to accidents or potential accidents, on improved meth-odology for evaluating research topics; and to scoping studies on the topics reLeing to prevention or mitigation of the offsite con-sequences resulting from postulated core melt accidents via liquid pathways, and to possible design measures for protection against sabotage.
We ACRS believes that there are complenentary roles for both NRC and DOE in research to Drove light-water reactor safety and that aggressive programs at the nulti-million dollar funding level should be pursued by each agency with appropriate coordination.
12-5
APPENDIX A METHODOLOGY OF ACRS STUDY The procedures followed by the ACRS in this study and in the prepa-ration of this report differed from those for the 1977 Report in two the subcommittee organization, and the nature and extent of ways:
participation by the full committee.
Following completion of the 1977 Report, the ACRS Generic Subcommit-tees were reorganized so that each of the research program areas could be assigned to a subcommittee, as indicated in Table A-1.
In addition, the chairmen of these generic subcommittees constituted the Reactor Safety Research Subcommittee, which was responsible for the scope and direction of the study.
Although the 1977 Report was reviewed, discussed, and approved by the full ACRS, the in-depth reviews were made by the individual Task Groups only.
For the 1978 Report, the full Committee reviewed and discussed in-depth the research programs in the areas mentioned in Section 1.2 of Chapter 1.
The schedule of meetings of the subcommittees and of the full ACRS is given in Table A-2.
A-1
TABLE A-1 SUBCOMMITTEE ASSIGNMENTS Steering Comittee - Reactor Safety Research Subcomittee D. Okrent, Chairman C. P. Siess, Editor M. Bender M. W. Carbon H. Etherington H. S. Isbin W. Kerr J. C. Mark D. W. Abeller P. G. Shewmn T. G. McCreless, Staff J. H. Austin, Associate Editor CHAPTER SUBCOMMITTEE 2.
loss of Coolant Accident /
ECCS Energency Core Coolire Systens H. S. Isbin, Co-Chaiman M. S. Plesset, Co-Chairnan M. N. Carbon J. Ebersole H. Etherington D. Okrent A. L. Bates, Staff 3.
Fbel Behavior Reactor Fuel P. G. Shewmn, Chaiman H. Etherington H. S. Ishin S. LMeroski J. C. Ma rk N.
M. Mathis D. Okrent P. A. Poehnert, Staff A-2
4.
Prinary System Integrity Metal Components P. G. Shewmon, Chairnan M. Bender H. Etherington H. S. Ishin D. Okrent E. Igne, Staf f Reactor Operation 5.
Operational Safety H. Etherington, Gairman J. Ebersole W. M. %this D. W. Moeller D. Okrent R. L. Wright, Staff Advanced Reactors 6.
Advanced Reactor Safety M. W. Carbon, Chairman M. Bender W. Kerr J. C. Mrk P. G. Shewmon C. P. Siess R. P. Savio, Staff Extreme External Phenomena 7.
Extreme External Phenomena C. P. Siess, Chairman J. C. Mark D. W. Moeller D. Okrent R. P. Savio, Staff Radiological Effects and Site 8.
Radiological Effects Evaluation, D. W. Maeller, Gairman J. Ebersole H. S. Isbin S. Lawroski D. Okrent R. Muller, Staff A-3
9.
Waste Management Waste Management D. W. Moeller, Chairman M. W. Carbon W. Kerr S. Lawroski J. C. Mark W. M. Mathis M. S. Plesset R. Muller, Staff 10.
Safeguards and Security Safeguards and Security J. C. Mark, Qiaiman M. Bender H. Etherington S. Lawroski W. M. Mathis P. G. Shewnon C. P. Siess R. K. Major, Staff
- 11. Risk Assessment Reliability and Probabilistic Assessment W. Kerr, Chairman M. Bender J. Ebersole H. S. Isbin J. C. Mark D. Okrent R. L. Wright, Staff 12.
Improved Reactor Safety Improved Safety Systems D. Okrent, Chairman H. Etherington S. Lawroski D. N. Moeller M. S. Plesset C. P. Siess S. Duraiswamy, Staff A-4
TABLE A_-2_
MEETINGS IN 1978 REGARDING THIS REPORT Subcomittee May 31 Reactor Safety Research July 14 Extreme External Phenonena July 17-18 ECCS July 24-26 Waste Management August 11 Advanced Reactors August 14-15 in Idaho Falls, ID ECCS August 17-18 in Idaho Falls, ID Reactor Fuel August 25 Reliability and Probabilistic Assessment August 28-30 ECCS September 6 Reactor Safety Research September 8 Full Connittee September 11-12 Metal Components September 12-13 in Albuquerque, NM Advanced Reactors September 26 Safeguards and Security September 27-28 Radiological Effects and Site Evaluation October 4 Reactor Fuel October 4 Improved Safety System October 4 Reliability and Probabilistic Assessment
- Meetings were held in Washington, DC unless otherwise indicated.
A-5
October 6-7 Full Committee October 11-12 Waste Management November 1 Reactor Operations November 3-4 Full Committee December.6 Advanced React.r3 December 7-9 Full Committee e
A-f5
APPENDIX B GLOSSARY ACRS Advisory Comnittee on Reactor Safe-guards AEC Atomic Energy Commission AIARA As low As Reasonably Achievahle ARSR Advanced Reactor Safety Research BE Best Estinate BEIR Biological Effects of Ionizing Radiations BWR Boiling Water Reactor CDA Core Disruptive Accident CRBR Clinch River Breeder Reactor CRAC Calculation of Reactor Accident Consequences DOE Department of Energy FBTF ECC Bypass Test Facility ECC Dnergency Core Coolant ECCS Dnergency Core Cooling Systems D4 Fualuation Model EPA Environmental Protection Agency EPRI Electric Power Research Institute ETtG Federal Republic of Germany FY Fiscal Year B-1
GCR Gas-Cooled Reactor GCFR Gas-Cooled Fast Reactor lffGR High-Temperature Gas-Cooleo Teactor HSST Heavy Section. el Technology DMEL Idaho National Engineering Laboratory LMFBR Liquid-Metal East-Breeder Reactor LOCA Loss-of-Coolant Accident LOFT Loss of Fluid Test LWR Light Water Reactor NMSS Office of Nuclear Material Safety and Safeguards NRC Muclear Regulatory Commission PAS Probabilistic Analysis Staff PBF Power Burst Facility ICM Power-Cooling Mismatch RW1 Pressurized Water Reactor RARG Risk Assessment Review Group RES Office of Nuclear Regulatory Research RIA Reactivity Insertion Accident SNREP Safety Research Experiment Facilities SNM Special Nuclear Material SSMRP Seism}c Safety Margins Research Program TAC Technical Assistance Contract TRAC Transient Reactor Analysis Code B-2
APPENDIX C Tile ADVISORY CONNTTTEE CN RFACTm SAFEGUARDS The Advisory Committee on Reactor Safeguards was established as 6 statutory Committee in 1957 by revision of the Atomic Energy Act.
Tne Comnittee was charged with responsibility for review of safety studies and facility license applications submitted to it, and to make reports thereon advising the Connission with regard to the hazards of proposed or existing reactor facilities and the adequacy of proposed reactor safety standards, and to perform such other duties as the Commission night request.
Section 182b of the Atomic Energy Act requires ACRS review of the construction permit and operating license applications for power and testing reactors and spent fuel reprocessing facilities licensed under Section 103, 104b er 104c of the Atomic Energy Act; any application for a research, develoinental or medical facility licensed under Section 104a or e of the Act which is specifically referred to it by the Commission; and any request for an amendment to a construction permit or operating license under Sections 103 or 104a, b, or c which is specifically referred to it by the Commission.
'Ihe Energy Peorganization Act of 1974 transferred the AEC regulatory functions to the newly formed Nuclear Regalatory Commission and the ACRS operations were also transferred to NRC to assist in its regu-latory functions.
In 1977, Public Law 95-209 added to it; other duties a requirement for the ACRS to undertake a study, making use of all available sources, of reacDar safety research and prepare and submit annually to the United States Congress a report containing the results of this sttxly.
The first of these reports was submitted to the Congress in December of 1977.
C-1
CHAIRMAN:
Dr. Stephen Lawroski, Senior Engineer, Chenical Engi-neering Division, Argonne National Laboratory, Argonne, Illinois VICE-CHAIRMAN:
Dr. Max W. Carbon, Professor and Chairman of Nuclear Engineering Department, University of Wisconsin, Madison, Wisconsin Mr. Myer Bender, Director of Engineering Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee Mr. Jesse Ebersole, Head Nuclear Engineer, Division of Engineering Design, 'Iennessee Valley Authorit- Knoxville, Tennessee (retired)
Mr.
Harold Etherington, Consulting Engineer (Mechanical Reactor Engineering), Jupiter, Florida Dr. Herbert S.
Isbin, Professor, Chemical Engineering and Materials Science, University of Minnesota, Minneapolis, Minnesota Prof. William Kerr, Professor of Nuclear Engineering and Director, Michigan Memorial-Phoenix Project, University of Michigan, Ann Arbor, Michigan Dr.
J. Carsori Mark, Division Leader, los Alanos Scientific Labora-tory, los Ala:aos, New Mexico (retired)
Mr. Milliam M. Mathis, Director, Planning, United Nuclear Industries Inc., Richland, Washington (retired)
Dr. Dade W. Moeller, Chairman, Department of Environnental Health 9_iences, School of Public Health, Harvard University, Boston, Massachusetts Dr. David Okrent, Professor, School of Engineering and Applied Science, University of California, Ios Angeles, California Dr. Milton S. Plesset, Professor of Engineering Science - Emeritus, California Institute of Technology, Pasadena, California Mr. Jeremiah J. Ray, Chief Electrical Engineer, Philadelphia Electric Company, Philadelphia, Pennsylvania (retired)
Dr. Paul G. Shewnon, Professor and Chairman of Metallurgical Engi-neering Department, Ohio State University, Columbus, Ohio Dr. Chester P. Siess, Professor Emeritus, Department of Civil En-gineering, University of Illinois, Urbana, Illinois C-2
1 Al POR T N UY8 6 H IA p*" N DDC' NRC roau 335 U S. NUCLE AR RE GUL AToRY COMMISSloN 7p BIBLIOGRAPHIC DATA SHEET IJUREG-0496 4 TITLE AN D SUBTITLE (Add Volume No. nf appecer.are) 2 fleave b.aMI Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program For 1978. A 3 Re CieiENT S ACCESSIO:4NO Report to the Congress of the United States of America by the Q g y p -uiceu vuFesct_cr Ge % ards.
- 5. D ATE REPORT CCYPLE TE D MONTH YE AM December 1978 9 PE RF or MsNG ORGANIZATION N AVE AND M AILING ADDRESS Ilactude 2,p Cooel DATE REPORT ISSUED I*E^"
Advisory Committee on Reactor Safeguards M
Thr 1717 H Street, fM.,
Room 1016-H 6 '" '"' "'"*#
Washington, D.C.
20555 8 (L eav e blan e )
- 12. SPONSORING ORGANIZ ATION N AYE AND M AILING ADDRE SS (Inc/vae I,p Code) 10 PROJECT!TASKinORK UNIT NO l
- 11. CONT R ACT NO
- 13. TYPE OF REPORT PE RIOD CoV E AE D (19clus;ve dates)
Report to Congress CY 1978 14 (Laave blank)
- 15. SUPPL EMENTARY NOTES 16 ABSTR ACT (200 *oras or less}
Public Law 95-209 includes a requirement that the Advisory Committee on Reactor Safeguards submit an annual report to Congress on the safety research program of the 1;uclear Regulatory Commission. This report presents the results of the ACRS 1978 review and evaluation of the riRC safety research program. The report contains a number of findings and recommendations.
- 19. SE CURITY CLASS (Th,s report) 21 NO. OF P AGES
- 18. AV AILABILITY ST ATEMENT unclassified 20 SECURITY CL, ASS (Trns pap) 22J RICE public availability unclassified s
NRC FORM 335 (7-77)
UNITED ST ATES NUCLE AR REGULATORY COMMISSION I
l WASHINGTON. D. C. 20555 POST AGE AND F E ES P AID OFFICI AL SUSINESS u.S. MuCLE A R RE GuLATORY PEN ALTY FOR PRIV ATE USE, $300 COmu1SS40N L
J 12C555C11232 L R lC L XC AN US NRC ACP CIST SERV BRANCH - SHELF C1t WA3HINGTON CC 20555
. _ _