ML19308C794
| ML19308C794 | |
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| Site: | Crane  |
| Issue date: | 08/31/1979 |
| From: | Lawroski S, Moeller D Advisory Committee on Reactor Safeguards |
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zw:n : cc, ARTICLE FROM NUCLEAR SAFETY, Vol. 20-4, July-Aug. 1979 387 General Safety Considerations Edited by J. R. Buchanan Advisory Committee on Reactor Safeguards:
Its Role in Nuclear Safety By S. Lawroski' and D. W. Moellert (Editor's None: Nuclear Safety Vol. I3. No.1. contained an Abstract: For over 23 } ears the Advisory Committee on article entitled "The Role of the Advisory Committee on Reactor Safeguards (ACRS) has had a continuing responsibility Reactor Safsguards in the Reactor Licensing Ptocess." The for conducting independent reviews and evaluattons of the foUowing article, coauthored by two previous Chairmen of the health and safety aspects of nuclear powr reactors, spent fuel Advisory Committee on Reactor Safeguards (ACRS) reflects reprocessing plants. and associated activities, which include not only the continuing ACRS role in reactorlicensing but also evaluation of abnormal occurrences and proposed changes at its expanding role in safety-related matters. including generic operating facilities. the adequacy of related safety standards safety issues and the Nucleat Regulatory Commissios's safety and criteria. the adequacy of the related safety research research program.}
programs, and specific generic questions. suchas the reliability of reactor pressure vessels. The ACRS normally assues 40 to 30 reports on specifsc nuclear facilities and safety-related ques-
- Dt. Stephen Lasioski graduated from Pennsylvania tions each year. Topics discussed in this article include the State University in 1934 and teceived the Ph.D. degree in views and though,rs of the ACRS with respect to emergency chemical engtnecting in 1943 from the same university. He was core. cooling systems, anticipated transtents without scram.
a Research Assistant at the Petroleum Reiming bboratory at reactor pressure vessel failure. turbine missiles, steamline State College. Pennsylvania, from 1934 to 1943 and was breaks. seismicity, environmental monitoring. emergency plan.
employed from 1943 to 1947 by the Standard Od Develop.
ning. waste mar:agement. ssnng. and reactor safety research.
ment Company (now EXXON Research and Engmeering Co.)
as Research Chemical Engineer. He was an Advanced Profes-sional Trainee (1946-1947) in the School of Reactor Engi-Since 1957 the group having statutory responsibility neering at Chnton Laboratones (now Oak Ridge National for perfotming independent renews and providing bboratory) and was on loan to the Manhattan Project's advice to the nuclear regulatoty agency on health and Metauurgical bboratory in Chicago from 1944 to 1946.
Dr. Lawtoski began his employment at Asgonne National P
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the Advisory Cornmittee on Reactor Safeguards bboratory in 1947 as Associate Duector of the Chemistry Division and in 1948 he became Director of the Chemical Engineering Division. He later served as Associate Director of the Laboratory and is presently a Senior Technical Adviser.
tDr. Dade W. MoeDer is Chairman cf the Department of Active in professional organizations, Dr. hwroski is a FeDow Environmental Health Sciences, School of Public Health,
)
of both the American Nuclear Society and the American ll.trvard University.1n addition to his teaching and research, he Institute of Chemical Engineers (AIChE) and was a member of currently serves as a member of the Nuclear Regulatory the Board of Directors of the American Nuclear Society from Comrnission's Advisory Comritittee on Reactor Safeguards,the 1957 to 1939. Dr. bwroski was a member of the General Committee on the Biological Effects of lonizing Radiation of Adytsory Committec of the U. S. Atornic Energy Commission the National Academy of Sciences, Committee 4 of the from 1964 to 1970. He was appointed to the Advisory International Commission on Radiological Protection, and the Comnuttee on Reactor Safeguards (ACRS) in 1974 and Board of Di ectors of the National Council on Radiation reappointed in 1978. He was Vice Chairman of the ACRS in Protection and Measurements. He served as Chairman of the 1977 and Chairman in 1978.
Advisory Committee on Reactor Safeguards in 1976.
NUCLE AR SAFETY. Vol. 20. No. 4, July-August 1979 8002070 627
GENERAL SAFETY CONSIDERATIONS 388 (ACRS). This article presents a summary of the history vanety of problems nociated with light. water reactors of the ACRS,its organization, and methods of opera-(LWRs), spent suel reprocessing plants, and fast t"n, and outhnes positions the ACRS has adopted breeder reactors. These melude a range of generic items with respect to some of the key safety issues facing the that develop as a result of reviews of specific projects.
nuclear profession today.
The thir( responsibility, to advise the NRC with respect to "the adequacy of proposed reactor safety
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- 5"D5'* "'I*I *"d C " *i""i"8 HISTORY effort by the ACRS in the review of codes, criteria, The Adusory Committee on Reactor Safeguards guides, and standards pertaming to reactor safety.The had its beginning in late 1947 when a Reactor fourth responsibility, to " perform such other duties as Safeguards Committee (RSC) was formed by the U. S.
the Commission [NRC] may request,"is the basis for Atomic Energy Commission (AEC) to serve as an ACRS reviews of such problem areas as criteria and advisory body to its Division of Research.' 2 The first packages for air transportation of plutonium, safe-task of the RSC was to conduct a survey of reactors guards problems associated with use of mixed oxide already in operation and to evaluate problems peculiar fuel, and the handling and disposal of radioactive to their design and location.Upon completion of these wastes. More recently the Congress has direcH the tasks, the RSC was to address itself to questions raised ACRS to conduct an ongoing review of NRC safety by new projects in late 1950 a related committee,the research and provide to it an annual report on this Industrial Committee on Reactor Location Problems subject.s (ICRI.P), with members from the industrial and busi-ness communities, was formed to conduct an examina-Membership and Organization tion. of the hazards associated with the operation of The current membership of ACRS (see Table 1)
AEC production facilities.
includes men experienced in chemical engineering.
On July 7,1953, the RSC and ICRLP were electrical engineering, materials engineering, mechani-combined, and the new organization was designated cal engineering, structural engineering, reactor opera-the Advisory Committee on Reactor. Safeguards tions, reactor physics, and environmental health. All (ACRS). With the revision in 1957 of the Atomic members have extensive backgrounds in various fields Energy Act s he ACRS was estabbshed as a statutory needed to review matters important to public health t
committee by the U.S. Congress and given the and safety.
following official charge:
Because of the large number of projects and There is hereby established an Advisory Com-subjects that must be reviewed and evaluated, the mittee on Reactor Safeguards consisting of a maxi-ACRS has established a large number of subcommittees mum of 15 members appomted by the Commission for terms of four years each. The Committee shall to follow particular problems. A separate subcommit-review safety studies and facility license applica-tee, for example,is established for the review of each tions referred to it and shall make reports thereon, nuclear power reactor project and each major generic shall advise the Commission with regard to the area. At any one time, about two to three dozen hazards of proposed or existing reactor facilities and subcommittees will be active.
the adequacy of proposed reactor safety standards, in general, the ACRS rebes on the relevant subcom-and shall perform such other duties as the Commis-mittees to determine when projects are ready for sion may request.
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Although legislation passed in 1974 by the U.S.
example, are responsible for initial reviews of specific Congress' has transferred operations of the ACRS reactor projects as well as preliminary evaluations of from the AEC to the Nuclear Regulatory Commission generic and/or topical matters and of safety related (NRC), this continues to be essentially the charge of questions pertuning to specific reactor vendors. Activi-ties of the subcommittees include site and plant tours, the ACRS today. The first responsibihty, to " review safety studies and facility license applications referred and as many subcommittee meetings are held as are to it" and to "make reports thereon," is discussed in judged to be needed to provide the necessary back-more detail in the section entitled " Project Reviews."
ground information and to resolve as many items as The second responsibility, to " advise the Commission possible before full committee action. In some cases with regard to the hazards of proposed or existing only one subcommittee meeting is needed;in others, as reactor ' facilities," requires that the ACRS review a many as 14 subcommittee meetings and two site visits NUCLEAft SAFETY, Vol 20. No. 4, Jutv-August 1979 4
CENERAL' SAFETY CONSIDERATIONS 389 Table 1 Membership of the Advisory Committee on Reactor Safeguards,'1979 CHAIRMAN: Dr. Max W. Carbon, Professor and Chairman of Dr. J. Carson Mark, Dmston trader, Los Atamos Scientific Nuclear Engineering Department, University of Wisconsin, bboratory los Alamqs. N.Mex. (reured)
Madison, Wis.
Mr. William M. Mathis, Duector, Planning. United Nuclear VICE CHAIRMAN: Dr. Milton S. Plesset, Professor of Engi-Ind ustries, Inc., Richland, Wash. (retired)
. neering Science Emeritus, Cahfornia Institute of Technology, Pa sadena, Catif.
Dr. Dade W. Moeller, Chairman, Department of Envuonmental Health Sciences, School of Public Health, Harvard Universit).
Mr. Myer Bender, Duector of Engineering Division, Oak Ridge Boston, Ma ss.
National bboratory, Oak Ridge,Tenn.
Mr. Jesse Ebersole, Head Nuclear Engineer, Division of Dr. David Okrent, Professor, School of Engineering and Engineenng Design, Tennessee Valley Authority, Knoxville, Apphed Science, University of Cahfornia, les Angeles, Caltf.
Tenn. (retired)
Mr. Jeremiah J. Ra), Chief Electrical Engmeer, Philadelphia Mr. Harold Etherington, Consulting Engineer (Mechanical and Electric Company, Philadelphia, Pa. (retired)
Reactor Engineering), Jupiter, Fla.
Dr. Paul G. Shewmon, Professor and Chairman of Metallurgical Prof. Wilham Kerr, Professor of Nuclear Engineering and Duector, Michigan Memorial-Phoenix Project, University of Engineering, Department, Ohio State Universir), Columbus, Ohio Michigan, Ann Arbor,Mich.
Dr. Stephen bwroski, Senior Engineer, Chemical Engineering Dr. Chester P. Siess, Professor Emeritus Department of Civil Division, Argonne National bboratory, Argonne,Ill.
Engineering. University of Illinois, Urbana. Ill.
have been held. Normally, one or two full committee of the generic subcommittees currently active in meetings are adequate to handle a project. For special support of the work of the ACRS are listed in Table 2.
matters invoinns novel features, new designs, or new To supply the required administrative and technical sites, several full committee meetings may be necessary support, the ACRS is provided with a small, highly to discuss all relevant issues.
qualified, full time professional staff based in Washing-The number of members on each subcommittee ton, D.C. This staff coordinates the work of the usually ranges from three to five. The members are consultants as well as the ACRS members, reviews for supported by an appropriate team of consultants who safety significance an increasingly large volume of are specialists in the areas ofimportance to the subject reports on plants under construction or operating,and under review. In total, there are about one hundred provides condse and technically correct records of all consultants assisting the ACRS. Their areas of special.
meetings of the subcommittees and the full committee.
ization cover as wide a range as the breadth of the In addition to the preceding, the ACRS also has re. views conducted by the ACRS. The names of some provision for the appointment of up to 15 ACRS Table 2 Examples of ACRS Generic Subcommittees l
- 1. Advanced Reactors
- 15. Procedures and Admmistration
- 2. Anticipated Transients Without Scram 16.Radiolopcal Effects and Site Evaluation
- 3. Concrete and Concrete Structures
- 17. Reactor Fuel
- 4. Core Performance
- 18. Reactor Operations
- 5. Emergency Core Coolmg Systems
- 19. Reactor Safety Research
- 6. Enrichment Plants 20 Regulatory Activities
- 7. Evaluation of Licensee Event Reports
- 21. Rehability and Probabilistic Assessmsat
- 8. Estreme External Phenomena
- 22. Requests and Recommendations
- 9. Fue Protection
- 23. Safeguards and Security
- 10. Fluid Dynamics
- 24. Smgle-Failure Criterion 11, Genenc items
- 25. Spent-Fur! Storage Pool Design
- 12. Metal Components
- 26. Transportation of R.dioactive Materials
- 13. Plant Arrangements
- 27. Waste Management
- 14. Power and Electrical Systems NUCLE AR S AFETY. Vol. 20. P.O. 4. July-August 1979 I
GENERAL SAFETY CONSIDERATIONS 300
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Fig. I The Advisory Committee on Reactor Safeguards is pictured at its special meeting of Apr.16, 1979,on the Three Mile Island incident (see the article "A Prehminary Report on the Three Mile Island inciden t," this issue). Seated lett to right are Dr. David Okrent, Mr. Jeremiah Ray, Dr. Max W. Carbon.
and Dr. Milton S. Plesset; standing left to right are Mr. Harold EtherinFton, Mr. Wifam Mathis, Dr. Chester Siess, Dr. Wilham Kerr, Dr. Paul Shewrnon, and Dr. Stephen Lawtoski. Absent when picture was taken were Mr. My er Bender, Mr. Jesse Ebersole, Dr. J. Carson Marl, and Dr. Dade W. Moeller.
Fellows who periorm analyses and reviews of specific devoted to evaluations of codes, standards and enteria.
technological matters related to nuclear safety.5 regulatory guides, generic matters, reactor safety re.
search, and standard plant resiews. In addition, mem.
bers of the ACRS conducted seven site and faciht'y COMMITTEE MEETINGS visits.
The full ACRS meets monthly for 3 days in Dunng its first two decades of operation, all i
Washington, D. C., normally on Thursday, Friday, and meetings of the ACRS were closed. Beginning with the Saturday of the first fullweek of each month. Meetings 155th meetmg on Mar. 5 and 9.1973, however, with the apphcants and the NRC staff are usually portions of the meetings were opened to public scheduled for the first 2 days. Such meetmgs and the attendance and participation in accordance with the meetings of ACRS subcommittees represent the pri.
requirements of the Federal Advisory Committee Act mary channel of direct communication between the (FACA).' Meetings of the ACRS were opened still ACRS and the applicants. The third day is used chiefly more to the pubhc in early 1977. accordance with for the preparation of the ACRS's reports to the NRC.
the Government in the Sunshine Act.' Similar actions Special meetings may be scheduled to deal with have been followed with respect to meetings of the specific problems or cases. In 1978 there were 12 various subcommittees. During 1978 over 96% of the j
regular ACRS meetings,2 special ACRS meetings with nectings of the full committee and its subcommittees foreign safety groups, and 82 subcommittee meetings.
were open to the pubhc in whole or in part. Atten-About half of the subcommittee meetings were de.
dance has included members of the press, representa-voted to specific project resiews, and about half were tives of the nuclear mdustry, and members of the NUCLE AR S AFETY. Vol. 20. No. 4. July-August 1979
CENERAL SAFETY CONSIDERATIONS 391 public, All meetings are announced in the Federal safety, of the operation of the nuclear facility under Register, and members of the public are invited to consideration and to advise the NRC of this opinion.
submit wntten comments concerning the topics and/or The ACRS is not so large as to make this process facihties under consideration. Members of the public unwieldy, nor is it so small as to preclude a balanced may also orally address the ACRS or its subcommittees approach. The consensus judgment of the ACRS is if they desire Even before enactment of the FACA, summarized in a report to the NRC Chairman to which however. the ACRS was receptive to comments from individual - members may append additional written members of the public and to information made comments if they disagree with the majority opinion.
available to it from various technical sources.
On the average. 25 to 30 reports (letters) of this type in addition to affording an increased opportunity are issued each year.
for pubhc input, one of the basic benefits of the open meetings of the ACRS is that they serve as useful ACTIVITIES AND RECOMMENDATIONS ON forums for public interaction between the ACRS SELECTED NUCLEAR SAFETY MATTERS members and the NRC staff, reactor vendors,architec-tural engmeering groups, the utilities, consultants, and Although the ACRS believes that the health and interested members of the public. Although members safety of the public have been adequately protected to of the ACRS are appointed by the Nuclear Regulatory date, the *dommittee also believes that the increased Commission, the reviews made by the committee are number of reactors anticipated in the future make it independent of the NRC since members of the commit-necessary that there be continuing improvement in the tee are not full time NRC career employees. The knowledge available for the resolution of safety issues.
approach of ACRS to reactor safety is based primarily Summarized in the sections that follow are statements
- on the technical problems associated with public health of recent positions or actions that the ACRS has taken and safety, taking into account the public acceptance with respect to selected nuclear safety matters. The of risk in a variety of nonnuclear as well as nuclear material is based on reports written by the ACRS as well as testimony presented by Committee representa-areas.
tives to committees of the U. S. Congress.8 3 PROJECT REVIEWS Emergency Core Cooling Systems Capability The ACRS has two focal points for its reviews of The ACRS has addressed the subject of the need nuclear facilities: the construction permit stage and the for emergency core cooling systems (ECCSs) for L%Rs operating-license stage.2 Reviews are made by the in letters relating to specific projects,in letters relating ACRS at both these stages on all commercial,research-directly to ECCS capability, and in letters relating to and test reactors except, normally, those with power reactor safety research. Major changes in emergency levels below 10 MW'(t). A sinitar two step procedure is core cooling systems were initiated as a result of followed in the review of fuel-reprocessing plants.The recommendations by the ACRS in connection with the ACRS has also taken part in reviews related to initial construction permit reviews of Dresden, Unit 3 [a site selection, changes in operating organization, con
- boiling water reactor (B%R)] 8 ' and Indian Point, version of provisional to full term operating licenses, Unit 2 [a pressurized water reactor (P%R)].8 8 These and the evaluation of significant plant changes or recommendations, contained in reports written in modifications that appear to represent an unreviewed August 1966. called for much greater emphasis on safety question.
prevention of a loss of coolant accident (LOCA)
The basic documents that support these reviews are through improved quality in design, fabrication, and the Preliminary and Final Safety Analysis and Environ' installation, by the institution of adequate in service mental Reports submitted by the applicant and the inspection, and by the development and application of Safety Evaluation Reports and Environmental State-adequate leak-detection methods.
ments prepared by the NRC staff. In addition, each The new approach was approved by the ACRS in ACRS member is provided with pertinent published its report on Dresden, Unit 3, and the committee documents and staff and consultant reviews on each recommended that it and succeeding BWRs use both project or facility.
core spray and core flooding systems, either of which The goal of the ACRS reviews is to arrive at a was to be capable of coping with a LOCA, including collective opinion as to the acceptability,in regard to postulated ruptures of the largest primary-system pipe.
NUCLE AR S AFETY, Vol. 2o. No. 4. July-August 1979 l
t
CENERAt. SAFETY CONSIDERATIONS 392 The new ECCS proposed for and used in PWRs, tive of whether the plant design without such changes namely, the use of accumulators plus a pumping appears to meet the provisions of the 'Ir.terim Accep-system to flood the reactor vessel, was endorsed in tance Criteria.' " This position was reaffirmed in ?. letter ACRS reports of January 1967 on the Palisades to the AEC Chairman m September 1073 (Ref.20),
Plant and Turkey Point, Units 3 and 4 (Ref.13). It which called for additional design, development, and ia was the addition of high-pressure accumulators in the research on this problem and urged,in particular,that Indian Pomt Nuclear Power Piant, Unit 2 (Ref.14) study be directed to: (1) improved rehability of the which permitted the elimination of a water-cooled, ECCS and system components, includmg approaches refractory lined stainless-steel tank beneath the reactor intended to minimize the potential for common failure pressure vessel which was originally proposed as a modes; (2) reactor core design and operating modes
. backup to the ECCS.
that reduce the potential for high temperatures and Concurrent with its recommendations for an im-clad swelling or perforation in postulated LOCAs; proved ECCS, the ACRS recommended an acceleration (3)an ECCS whose proper functioning is relatively of research on the various aspects of LOCA-ECCS in a insensitive to reactor or ECCS design parameters and to report on reac'or safety research on Oct. 12, 1966 proper functioning of such other components as steam (Ref.15). In testimony during hearings conducted by generators or reactor containment:(4) an ECCS having the Joint Committee on Atomic Energy in 1967 redundanc// diversity, and abundance of flow such (Ref.16), the ACRS Chairman called for additional that its adequacy is subject to evaluation without experimental studies to confirm both the adequacy of undue requirement for complex evaluation techniques; analytical models to predict flow characteristics associ-and (5)other measures that further reduce the proba-ated with a loss of coolant and the effectiveness of bilities and consequences of a LOCA. These matters some of the emergency core-cooling methods used. In were further addressed m ACRS reports on the NRC its report on the Browns Ferry Nuclear Power Station Safety Research Program which were issued in 1977 in March 1967 (Ref.17),the ACRS identified the need and 1975 (Refs. 21a. 21bh for both analytical and experimental studies for all Significant changes have been made in the core
' large water-cooled reactors on the effects of fuel failure design and the ECCS desiens since.this matter first on safety, including localized failure of the clad durin!
began to receive attention.' Changes have included the a LOCA Reinforced by preliminary experiments on use of 8 bv 8 fuel bundle designs in BWRs and clad embrittlement at Argonne National Laboratory in 17 by 17 designs in PWRs, the use of accumulators to 1967, the ACRS questioned the then current ECCS supplement pumpmg systems in PWRs.a direct. diesel-design basis of "no clad-melt." and m a report on driven, high pressure coolant injection (HPCI) pump. a Surry, Units 1 and 2 in April 1968 (Ref.18), the steam-driven reactor core isolation cooling (RCIC)
ACRS noted: "In connection with postulated loss of-pump and upper-head injection in PWRs with ice-coolant accidents, the Applicant stated that, using condenser contamment designs. A major effort is conservative assumptions and allowing appropriately currentiv under way bv the NRC staff to determine the for fuel element distortion from the original core margins to failure pmvided by exisung ECCS designs.
geometry, the emergency core cooling systems will be designed to keep fuel-clad temperatures below the Anticipated Transients Without Scram point at which clad may disintegrate upon subsequent cooling."
Commercial water-cooled pover reactors have long in a report in January 1972 (Ref.19), the ACRS been designed to accommodate safely any " anticipated accepted the Interim Acceptance Criteria for ECCS as transients" that might arise during operation. Antici-being useful and helpful to the AEC Regulatory Staff pated transients are those abnormal occurrences ex-in the licensing process but indicated that such criteria pected to occur one or more times during the life of.
should be considered for interim use only and that plant. Examples are loss of feedwater, loss ofload with more work was required on code development, safety turbine trip, loss of off site power, and inadvertent research oriented to the ECCS, and the development control-rod withdrawa!. Before 1969, analyses used in and design of an improved ECCS. The ACRS further licensing assumed that the reactor would automatically recommended in that report "that design changes to scram whenever required by occurrence of a transient.
improve ECCS capability be sought and, to the extent However, failure to scram during certain transients practical, employed in plants for which construction could result in serious consequences to the plant and permit applications are received in the future,irrespec-possible hazards to the public.
NUCLE AR SAFETY, Vol. 2o. No. 4, July-August 1979
i CENE2AL SAFETY CONSIDERATIONS 393 In numerous reports on license applications since contractors, by industry, and by others. The ACRS 1969, the ACRS has recommended additional study of Subcommittee on Reactor Pressure Vessels has con.
- design features triat could be incorporated in reactor ducted an extensive review of the present state of the systems to make tolerable the consequences of failure art to evaluate the adequacy of current requirements to scram following an anticipated transient. Similarly, relating to light wa'ter power reactor vessels and,as far ACRS has recommended study of means for increasing as practical with existing data,io estimate the proba-the reliability of scram syst. ems. During this time,20 bility of disruptive failure of a reactor vessel. Although subcommittee meetings have been held to discuss these faced with the problem that such a determination subjects with the NRC staff and the vendors of would be difficult because the number of reactor water-cooled power reactors.
vessels or vessel years was too small to permit valid Considerable investigative work in the Treviously statistical inferences (and the fact that the much mentioned areas and in scram system reliability has greater body of statistics covering nonnuclear vessels been accomplished by bth the NRC staff and the includes many vessels of poor or unknown quainty), the reactor vendors. Reports analyzing the consequences of ACRS concluded that there is reasonable assurance anticipated transients without scram (ATWS) have that the probability of disruptive failure in normal been submitted by all the vendors. Some vendors have service of a reactor vessel designed, constructed, and also supplied quantitative estimates of the reliability of Operated,in accordance with current standards is less their scram systems (based primarily on analysis) and than 10-' per vessel year (less than one disruptive of the effect of certain design changes that could be failure per million vessel years). In Jaauary 1974 the incorporated to ameliorate the consequences of an ACRS published its findings in Report on theIntegrity ATWS.
of Reactor l'essels for Light Water Power Reactors.'"
The ACRS is now (April 1979) reviewing the NRC In the opinion of the ACES, such estimates of staff's proposed position on ATWS as stated in the failure probabilities of essential reactor components are report Anticipated 7ransients WithoutScramforLight highly useful in the evaluative process related to the /
Water Reactors (NUREG-0460) (Ref. 22). The' ACRS overall safety of such systems. The ACRS has encour.
Subcommittee on Anticipated Transients Without aged similar studies on many aspects of nuclear reactor Scram has been meeting with representatives of the operations. With respect to pressure vessels, the report NRC staff and the nuclear industry and gathering cited in the preceding paragraphs recommends con-information for the full committee.The ACRS expects tinued support of research and deve'opment, particu-to prepare its recommendations on the ATWS resolu-larly on the properties of heavily i adiated pressure-tion during 1979.
vessel steel and on new techniques for in-service inspection.
Failure Probabilities of Reactor Pressure Vessels The remote possibility of a reactor pressure vessel failure has long been an area of interest to the ACRS.
There have been several instances of failure of in a report on this subject issued in November 1965 turbines in which large rotor fragments have been (Ref. 23), the ACRS suggested that the industry and ejected as high-energy missiles.The probability of such the AEC give further attention to methods and details missiles striking a portion of a nuclear plant and of stress analysis, to the development and implementa-causing damage that might endanger the health and tion of improved methods of inspection during fabrica-safety of the public has been considered by the ACRS tion and vessel service life, and to the improvement of for some time. Since reactor subsystems are normally means for evaluating the factors that may affect the nil housed in substantial concrete structures,in order for a ductility transition temperature and the propagation of turbine failure to cause an intolerable release of flaws during vessel life. The ACRS recommended in radioactive material, a missile would have to penetrate this report that means be developed to ameliorate the a structure and strike a particular target or combina-consequences of a major pressure vesselrupture, tion of targets in a manner that would negate safe Since that letter was written, improvements have shutdown capability or breach the containment coinci-been made in reactor vessel technology, and the dent with a radionuclide release.
understanding of the behavior of pressure vessels has In most of the newer plants,the orientation of the been greatly enhanced by extensive research and turbine is such that it would be extremely improbable development by the Department of Energy and its for a turbine-generated missile to cause the kind of NUCLE AR SAFETY, Vol. 2o, No. 4. July-August 1979
394 CENERAL SAFETY CONSIDERATIONS damage previously described. In many older plants, problem is now being reviewed by the SRC staff for all however, the orientauon of the tt.rbine axis is such existing and proposed plants. Various solutions to this that this type of damage cannot be ruled out.
problem are available;for many plants there may be no For armage to occur, the following' events must prc,blem with the present design.
take place: the turbine must fail near operating speeds, The existence of this situation emphasizes the fact or due to overspeed, and produce a missile:the missile that new problems arise and that safety reviews in must be ejected from the turbine housing; the missile depth, both as to concept and as to details of design, must be directed toward the target in question,usually execution, and operation, are essential to the mainte-the containment, the reactor building, or the control nance of appropriate levels of safety in nuclear power l
room; the missile must penetrate. the tarEet; and the reactors. Although operating experience will undoubt-missile must strike components or systems whose edly continue to reveal new items of concern, such failure would lead to unacceptable consequences.
experience cannot be relied on as the sole means of Calculations of the probabihty of harm resulting identifying all possible hazards.
from turbine missiles, considering this chain of events, have led to widely varying results, depending on the stress Corrosion Cracking in BWR Piping degree of conservatism applied to those areas where Cracks 'have been detected and leaks have been conclusive data are lacking. In a letter to the AEC in April 1973 (Ref. 25), the ACRS recommended that observed in statnless-steel piping in several BWRs. In letters to the NRC Chairman dated Feb. 8 and 14, more attention be given to certain aspects of this problem, including, but not necessarily limited to, the 1975 (Refs. 26,27), the ACRS made recommendations reliabihty of systems to prevent overspeed of turbines with respect to actions then being undertaken and
)
and means of calculating the penetration of heavily those necessary for a longer range solution to this reinforced concrete walls by irregularly shaped turbine Problern. The ACRS pointed out that stress. corrosion missiles. Although further study is required, the infor.
cracking in sensitized stainless steel was not a new mation already gained is valuable in determining Problem, that the action taken by the NRC in ordering whether there is need for additional protection of Prompt inspections of potentially affected piping in all vulnerable structures, for improved overspeed protec-Operating BWRs was both necessary and appropriate, tion, or for modifications in the turbine orientation to and that interim measures for repairing the cracked j
P Pes were available.The ACRS concluded in assessing j
i reduce the probability of unacceptable damage by turbine missiles.
the health and safety aspects of this situation, that, l
although such cracks clearly increase the risk o,f pipe failure, the consequent increase in risk to the puohe Break in High. Energy Pipes Outside Containment did not appear to be substantial and was no cause for i
A break in a main steamline or in any other pipe or immediate alarm. This is particularly true because system containing fluid under high pressure has alw2ys inspection and leak detection procedures are available been considered a credible accident for which protec-and can be used to reduce the probabihty of a pipe l
tion must be provided. For a break inside the contain-rupture while a reactor is in operation. However, ment, such protection is generally in the form of inasmuch as the interim measures then proposed could redundant and automatically operated isolation valves not be expected to eliminate the problem, the ACRS that confime the released fluid to the containment. In recommended that programs of research and turn, the containment and the systems therein are development be pursued to develop a longer range designed to resist the resulting pressures and tempera-solution. Specific examples cited by the ACRS where tures. Originally a break outside the containment was additional information would be desirable included evaluated only in terms of the possible release of items such as: (1) development of quantitative tech-radioactive material from the ruptured pipe or system.
niques for detecting and measuring residual stresses; Several years ago, however, it was recognized that jet (2) determination of the combined role of cyclic and forces and gas pressures associated with such a break static stresses on stress corrosion cracking;(3) response could in some circumstances severely damage or disable of piping to stress-corrosion cracking as a function of adjacent components or systems housed in the same degree of sensitization;(4) assessment of the reliability building. Depending on the layout, damage could be and sensitivity of such nondestructive techniques as done to the control room or to components or systems ultrasonic testing when apphed to piping systems; required for safe shutdown of the reactor. This (5)better methods for measuring and controlhng both NUCLE AR SAFETY, Vol. 2o. No. 4. July-August 1979
395 GENERAL SAFFTY CONSIDERATIONS oxygen and other impurities in the coolam immedi-Environmental Monitoring and Emergency Procedures ately adjacent to regions of potential failure;(6)the The ACRS has also been active in recent years in role of mternal surface finish on the initiation phase of the review and evaluation of problems associated with stress corrosion crackmg; and (7) selection of alloys
- "* "**** * "U 'i" E "" * **'E*"#Y E'*"
E' not susceptible to stress-corrosion cracking which One of the recommendations of the ACRS has been could be used to replace parlicular components tr.
hM M
h h@ &
operating reactors or which might be used for the vironmental monitoring programs be readily applicable P ping systems in future reactors.
to the estimation of annual radiation doses to the i
public resulting from effluent releases. Special consid-Seismic Studies eration has been recommended in cases where severt!
nuclear facilities are in close proximity. The ACRS has One of the major needs in terms of the evaluation recommended that consideration be given to periodic of reactor sites is improved methods for determining evaluation of the combined liquid and airbome radio-the intensity of the ground movement' which might nuclide releases as they may affect the health and accompany an earthquake affecting a given site. This safety of the public.a' The ACRS has also urged that need is most obuous in those areas of the United P ans for *ernergency procedures tn the event of an l
States west of the Rocky Mountains where large be well coordinated with state and local accident earthquakes have occurred at relatively frequent inter.
agencies whose actions would be essential in dealing vals. In this region, earthquakes can usually be related with the population and that the adequacy of arrange-to known geological features which have been or can ments for implementing such plans be confirmed be explored extensively, and sufficient information can before initial operation of a plant.80 Again, where be obtained to permit a data based determination of other nuclear facilities are in the same general area, the the magnitude of the earthquake and the related ACRS has urged that effective cooperative emergency intensity of ground motion against which a given. plant response plans be developed.
must be protected.
The ACRS also recommended that the Regulatory Seismic conditions in the eastern United States, Staff provide numerical dose limits as a guide to the however, are poorly understood in comparison to those reactor industry in applying the "as low as practicable" in the West Coast region, and some of the major criterion.* Implementation of this recommendation earthquakes of the past in the east have not been was reflected in the development of Appendix 1, definitely related to geological formations or subsur.
Title 10, Part 50 of the Code ofFederalRegulations.8' face conditions. For these reasons, the determination of seismic design levels in the eastern United States has In applying this criterion to the control of routine releases from power reactors, the ACRS has for some involved extrapolation based on limited evidence with tine encouraged the use of more complete treatment a concomitant trend toward increased conservatism to systems for both liquid and gaseous wastes. Where compensate for these uncertainties. As a result, the P ants without up to-date gaseous waste treatment l
ACRS believe that steps should be taken by the systems have come into operation, the ACRS has appropriate federal agencies to develop information endorsed incorporation of augmented systems gener-required to understand more clearly the seismic poten.
ally no later than the first refueling outage.
tial of the eastern United States. Recommendations for the initiation of such a program were made b) the Waste Management ACRS in a letter to the AEC Chairman in May 1973 (Ref.28). Similar research has been urged by other While stressing improved retention of waste radio.
such groups as the National Academy of Engineering.
nuclides within the plant, the ACRS has also recog-The present seismic design approach required by nized that such procedures could lead to increased the NRC includes many conservative factors, and the operating problems. In a letter in September 1973 NRC staff believes that the overall procedures are (Ref.32), the ACRS urged that careful eva!uation be conservative for essentially all plant sites. There may, made of the possible effects of the gradual bui' dup of however,be parts of this design sequence which are not tritium in plants designed to process and reuse liquids conservative for specific appbcations. Research pro.
removed from various reactor systems. Factors in need grams have recently been implemented to provide a better understanding of seismic events and character.
.Now called the "as low as reasonably achievable" f
(ALARA) enterion.
istics.
M/CLE AR S AFETY, Vol. 2o. No. 4. July-August 1979
396
, CENERAt. SAFETY CONSIDERATIONS of assessment included potential increases in radiation has been expressed in efforts by the ACRS to develop exposures of operating personnel,possible difficulty in an index for companng the relative acceptability of plant maintenance, and the possible influence of sites and in debberations on the interrelationship increased tritium concentrations on the consequences between siting and engineered safety features. Title 10, 8
of unanticipated releases.
Part 100 of the Code of Federal Regulations ' states:
In two reports on the management of high-level Where unfavorable physical characteristics of the radioactive wastes, issued ir ~1976 (Refs. 33,34), the site exist, the proposed site may nevertheless be ACRS endorsed the concept of retaining the provision found acceptable tf the design of the facility for retrievability in the first one or two disposal sites, includes appropriate and adequate compensating since this would allow time for' development and engineering safeguards.
validation of methods for ultimate disposal.The ACRS Ahhough recognizing this concept and,indeed, endors-urged that the NRC establish appropriate interim ng its application. the ACRS has been concerned with criteria for licensing such facilities at an early date and the need to better specify the degree to which that these enteria include definition of the forms of engmeered safety features can be used to compensate waste acceptable for storage and the conditions of for site deficiencies. This concern was expressed in a
~
waste durability and integnty required through some report issveti in December 1975 (Ref. 35)in which the stated minimum period, at the end of which retrievabil-ACRS suggested that:
ity must still be possible. The ACRS also recommended Studies be conducted on the degree to which that, in developing thesc criteria, the NRC give engineered safety features or alterations in plant consideration to the volumes of wastes that it might be design should be used to compensate for specific necessary to retrieve and the actions to be anticipated site deficiencies. In particular, it would be useful to following such retrieval, including contingency plans determine whether there are characteristics for for storage of the wastes through some interim period which compensating engmeering changes should not while more permanent arrangements were being de.
be apphed.
fined.
The ACRS also succested that:
The ACRS also pointed out that there is a need to pop,j,. tion dose be estimated for a broad range of recognize the differences in the magnitude of the site characteristics and for a broad spectrum of immediate risks associated with the various compon-accidents, including Class 9, using probabilistic data entr or portions of the nuclear fuel cycle. For example, and methods of the types applied in the Reactor most nuclear safety experts would agree that the Safety Study (WASH 1400) including sensitivity potential for impact c & bealth and safety of the studies and, allowances for uncertainties. An impor-public from accMents is greate for nuclear power tant objective of this effort should be to determine plants than it is for a facility properly designed for %
the relative importance of specific site character-long term storage of solidified high.leven wastes istics in terms of their impact on population doses and risks under accident conditions.
The potential health problems associated with a disposal facihty were regarded primarily a' being of a
^
chronic loolevel nature that would decrease substan-Resolution of Generic Matters tially during the first few hundred years of decay. The ACRS was unable to develop or postulate any event in As most people familiar with the work of the such a facility that would be comparable to a Class 9 ACRS realize, the committee does not confme its accident. Although this does not mean that there are attention to mdividual licensing apphcations. A major no associated hazards, the ACRS expressed the opinion effort is devoted to generic matters involving issues which affect general classes or types of nuclear that it does mean that the period of time during which responses to problems must be implemented could facilities. For example, a substantial and continuing generally be relatively long (months to years). The effort is maintained by the ACRS and its individual option of retrievabihty should greatly facilitate the members in the review and development of codes, application of corrective measures.
criteria, standards, and guides. ihe ACRS seeks also to keep abreast of significant, current safety incidents and W
Reactor Siting Policy operations. Updates of the progress toward resolution For some years the ACRS has been concerned with of generic matters are provided in periodic letters to policies for the siting of nuclear facilities.This concern the Chairman of the NRC.8' Topics recently reviewed, NUCLE AR SAFETY. Vol. 20. No. 4. July-August 1979
, e. a. m m u n e w s c c:u w... m n.:.a.. s. w w..: h.. x.,
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i (ENERAL SAFETY CONSt!E2ATIONS 397 or for which review is now under way, include of controversial and complex technical issues that are monitoring for excessive vibration or for loose parts considered in the review and evaluations conducted by
-inside pressure vessels, instrumentation to follow the the committee.
course of an accident, and steam generator tube Other areas ofinterest to the ACRS include quality leakage.
assurance and the effective application of operating exp rience. Although the health and safety of nuclear Review of NRC Safety Research operations can be enhanced by proper desigr' and e nstructi n f facihties, the committee believes that In accordance with the Nuclear Regulatory Com-the degree of safety actually aclueved is largely mission Appropriation Authorization for 1978 (Public dependent upon the people who operate the plant. The Law 95 209) (Ref. 5), the ACRS was requested to ACRS is pleased to note increased attention by such prepare annual reports on the NRC Safety Research Program. The first such report,a' which was issued in gr ups as the American Nuclear Society,the American National Standards Institute, and the U.S. Nuclear December 1977, covered eight basic technical areas:
Regulatory Commission to standards, guides, and Systems Engineering and Analysis Development, Fuel criteria for effective quality assurance programs. With Behavior, Metallurgy and Materials, Site Safety, Ad-vanced Reactors, Fuel Cycle and Environmental Re-respect t,o reviews of reportable occurrences in operat-ing nuc' lear facilities,the ACRS believes that additiona!
search Safeguards, and Risk Assessment. The ACRS attention should be given to evaluating this informa.
concluded that the overall effort appropriately involves research necessary to provide the NRC with suitable tion and applying it to the licensing process. In response to a request from the NRC Chairman, the bases for carrying out its regulatory responsibilities, c mmittee has under way a study of all Ucensee Event but recommended that,in addition to continuing such Reports submitted by licensed nuclear power plants work, the NRC should become more involved in during the period, 1976-1978.
research leading to improved safety system concepts. A The ACRS has also recommended that additional second report on the NRC Safety Research Program attention be given in the regulatory process to a variety was issued in December 1978 (Ref. 21b).
o a manus,includmg (1)the physical protection Recently the ACRS reviewed the NRC report, Plan f nuclear facilities; (2) reliability of decay heat re-for Resecrch to Improve the Safety of Light kater
, m val systems and auxiliary power suppbes; Nuclear Pour Plants. NUREG-0438 (Ref. 37). The
( ) *Y"'** N""*
"* " I plan was developed in response to a requirement by UM an phncal anangunents of adjacent synnns as Congress in the 1978 Budget Authorization Act tha, weH as nacu ns unucmected synant the NRC develop a long range plan for new E'
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- Y 'Y"#**
"E improved safety systems for nuclear power plants In a tainment; (5) reduction of radiation doses to plant report dated Mar. 13,1978 (Ref. 38) the ACRS con.
personnel as well as the general public; and curred with the program proposed in NUREG-0438 (6):mprovements in the basic understanding of physi-and indicated that "the development of better methods cal phenomena including the ability to contatn and for evaluating concepts proposed to improve safety is c ntr i the consequences of accidents, such as a major essential to the success of this new effort. Although core melt, which are larger than the design basis there will always be a large subjective or judgmental accident.
element in the selection of research projects on improved safety, tiese selections should be made on as quantitative and facu l a basis as practical. It seems gg g
evident also that it wn. ve extremely difficult to provide a suitable methodology without at some point We gratefully acknowledge the assistance of our addressing the question of how safe is safe enough."
fellow committee members, including former ACRS member Spencer H. Bush, and ACRS staff members in the preparation of this article.
As previousiv noted, the ACRS has had a long and REFERENCES active interest in a wide range of safety related matters.
The items discussed in this article do not represent a
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NUCLE AR S AFETY. Vol. 20. No. 4. July-Augwst 1979
~
i CCEZAL SAFETY CONSIDERATIONS 398
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