NUREG-0153, NRC Staff Response to ASLB Questions Concerning Generic Safety Issues.Conclusions in Perkins SER Re Issuance of Const Permits Will Not Be Affected by Results of Generic Tasks.W/Certificate of Svc.Supporting Documents Encl
| ML19259B360 | |
| Person / Time | |
|---|---|
| Site: | Perkins  |
| Issue date: | 01/18/1979 |
| From: | Barth C NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | |
| Shared Package | |
| ML19220A237 | List: |
| References | |
| RTR-NUREG-0153, RTR-NUREG-0371, RTR-NUREG-0410, RTR-REGGD-01.139 NUDOCS 7902090021 | |
| Download: ML19259B360 (47) | |
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CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF RESPONSES TO BOARD QUESTIC'lS CONCERNING GENERIC SAFETY ISSUES" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or, as indicated by an asterisk, through deposit in the
'fuclear Regulatory Commission's internal mail system, this 17th day of January, 1979:
- Elizabeth S. Bowers, Esq., Chaiman William A. Raney, Jr., Esq.
Atomic Safety and Licensing Board Special Deputy Attorney General U.S. Nuclear Regulatory Commission P.O. Box 629 Washington, D.C.
20555 Raleigh, North Carolina 27602 Dr. Donald P. deSylva William L. Porter, Esq.
Associate Professor of Marine Associate General Counsel Science Duke Power Company Rosenstiel School of Marine 422 South Church Street and Atmospheric Science Charlotte, North Carolina 28242 University of Miami Miami, Florida 33149 William G. Pfefferkorn, Esc.
P.O. Box 43 Dr. Walter H. Jordan Winston-Salem, North Carolina 27102 881 W. Outer Drive Oak Ridge, Tennessee 37830 Mrs. Mary Davis Route 4 J. Michael McGarry, III, Esq.
Box 261 Debevoise and Liberman M ocksville, North Carolina 27023 1200 Seventeenth Street, N.W.
Washington, D.C.
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- Atomic Safety and Licensing Board Panel U.S. fluclear Regulatory Commission Washington, D.C.
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- Atomic S3fety and Licensing Appeal Board Panel U.S. luclear Regulatory Commission Washington, D.C.
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- 0ccketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Charles A. Rartn Counsel for f!RC Staff
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STN 50-490 Perkins Nuclear Station, Units,1, 2 and 3 )
NRC STAFF RESPONSES TO BOARD QUESTIONS CONCERNING GENERIC SAFETY ISSUES Board Ouestions By orcer of September 7, 1978 the Board asked the coinion Of the parties concerning the adequacy of the Perkins record in meeting Appeal Board criteria on generic safety issues.
By orcer of October 13, 1978 the Board called to the attention of all parties the recent document NUREG-0460, " Anticipated Transients Without Scram For Light Water Reactors". April 1978.
The Board then questioned whether the comitment on the part of the Aoplicant in answer to Board Question 1 will provide an adequate solution to the issue of anticipated transients without scram and requested the staff to review its testimony in reoly to the Board Ouestion.
This resconse acdresses generic safety issues except for anticipated transients without scram. That
issue is addressed in "Supolement to NRC Staff Resconses to Board Questions Concerning Health and Safety Ascects of Cherokee and Perkins".
Background
The NRC staff continuously evaluates the safety requirements 'Jsed in its reviews against new information as it becomes available.
informaticn re-lated to the safety of nuclear power plants comes from a variety of saurces.
Obvious sources of such information are experience from coerating reactors, research results, NRC staff and Advisory Committee on Reactc.- Safeguards safety reviews, and vendor, architect / engineer and utility design reviews.
Each time a new ccncern or safety issue is identified frca one or more of these sources, the need for immediate action to assure saf e operation is assessed. This assessment includes consideration of the generic imoli-cations of the issue.
In some cases, immediate action is taken to assure safety, e.g., the derating of boiling water reactors as a result of the channel box wear pro-blems in 1975.
In other cases, interim measures, such as modifications to operating procedures, may be sufficient to allow further study of the issue prior to making licensing decisions.
In most cases, however, the initial assessment indicates that immediate licensing actions or changes in licensing criteria are not necessary.
In any event, further study may be deemed appropriate to maka judgments as to whether existing NRC staff requirements should be modified to address the issue for new plants or if backfitting is apprcoriate for the long term operation of plants already under con-struction or in operation.
These issues are sometimes called " generic safety issues" because they are related to a particular class or tyce of nuclear facility rather than a specific plant. These issues have also been referred to as " unresolved safety issues". Hcwever, as discussed above, such issues are considered on a generic basis only after the staff has made an initial assessment fer individual plants and has made a determination that the safety significance of the issue does not prohibit continued operation or recuire licensing actions while the longer term generic review is underway.
These longer term generic studies were the subject of a decision by the Atcmic Safety and Licensing Appeal Board of the Nuclear Regulatory Commission. The decision was issued on November 23, 1977 (ALAB-JJJ) in connection with the Appeal Board's consideration of the Gulf States Utility Company application for the River Bend Station, Units Nos. I and 2.
In the view of the Appeal Board, (pp. 25-29):
"The resconsibilities of a licensing board in the radio-logical health and safety sphere are not confined to the consideration and disposition of those issues which may have been presented to it by a party or an " interested State ' with the required degree of specificity. To tha contrary, irrespective of what matters may or may not have been properly placed in controversy, prior to authorizing the issuance of a construction permit the board must make the finding, inter alia, that there is " reasonable assur-ance" that "the proposed facility can be constructed and operated at the proposed location without undue risk to the health and safety of the oublic".
10 CFR 50.35(a)
....Cf necessity, this determination will entail an incuiry into wnether the staff review satisfactorily has -
come to grips with any unresolved generic safety pro-blems which might have an imoact upon oceration of the nuclear facility under consideration."
"The SER is, of course, the principal document before the licensing board which reflects the content and outcome of the staff's safety review. The board should therefore be able to look to that document to ascertain the extent to which generic unresolved safety problems whicn have been previously identified in a TSAR item, a Task Action Plan, an ACRS report or elsewhere have been factored into the staff's analysis for the particular reactor -- and with what result. To this end, in our view, each SER snould contain a summary descriptinn of those generic problems under con.tinuing study which have both relevance to facilities of the type under review and potentially signi-ficant public safety implications."
"This summary description should include information of the kind now contained in most Task Action Plans. More specifi-cally, there should be an indication of the investigative program which has been or will be undertaken with regard to the problem, the program's anticipated time-scan, whether (and if so, what) interim measures have been devised for dealing with the problem pending the completion of the investigation, and what alternative courses of action might be available should the program not produce the envisaged result."
"In short, the board (and the public as well) should be in a position to ascertain from the SER itself -- without the need to resort to extrinsic documents -- the staff's per-ception of the nature and extent of the relationship between each significant unresolved ceneric safety question and the eventual operation of the reactor under scrutiny.
Once again, this assessment might well have a direct bear-ing upon the ability of the licensing board to make the safety findings required of it on the construction permit level even though the generic answer to the question remains in the offing. Among other th ngs, the furnished i
information would likely shed light on such alternatively important considerations as whether (1) the problem has already been resolved for the reactor under study; (2)
J there is a reasonable basis for concluding that a satis-factory solution will be obtained before the reactor is cut in operation; or (3) the problem would have no safety implications until after several years of reactor coera-tion and, should it not be resolved by then, alternative means will be available to insure that continued operation (if cermitted at all) would not pose an undue risk to the public."
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In a related catter, as a result of Congressional action on the Nuclear Regulatory Commission budget for Fiscal Year 1970, the Energy Reorganiza-tion Act of 1974 was imended (PL 95-209) on December 13, 1977 to include, among other things, a new Section 210 as follows:
" UNRESOLVED SAFETY ISSUES PLAN" "SEC. 210. The Commission shall develop a plan providing for specification and analysis of unresolved safety issues relat-ing to nuclear reactors and shall take such action as may be necessary to implement co-rective measures with respect to such issues.
Such plan shall be submitted to the Congress on or before January 1,1978 and progress reports shall be in-cluded in the annual report of the Commission thereafter."
The Joint Explanatory Statement of the House-Senate Conference Committee for the FY 1978 Appropriations Bill (Bill S.1131) provided the following additional information regarding the Committee's deliberations on this portion of the bill:
"SECTION 3 - UNRESOLVED SAFETY ISSUES" "The House amendment required development of a plan to resolve generic safety issues.
The conferees agreed to a requirement that the plan be submitted to the Congress on or before January 1, 1978. The conferees also excressed the intent that this plan should identify anc describe those safety issues, relating to nuclear power eactors, wnicn are unresolved on the date of enactment.
It sho.ld set forth:
(1) Commission actions taken directly or indirectly to develop and imolement corrective measures; (2) further actions planned concerning such measures; and (3) timetables and cost estimates of such actions. The Commission should indicate the priority it has assigned to each issue, and the basis on which priorities have been assigned."
In response to the reporting recuirements of the new Section 210, the NRC staff submitted to Congress on January 1,1978, a report on tne "NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants",(NUREG-0410). The NRC program described in NUREG-0410 was already in place when PL 95-209 was enacted and is of considerably broader scope than the " Unresolved Safety Issues Plan" required by Section 210. Al thougn the NRC program does include plans for the resolution of generic technical issues of varying degrees of safety significance, it ciso includes generic tasks for the resolution of environmental issues, for the develocment of imorovements in the reactor licensing process, and for consideration of less conservative design criteria or operating limitations in areas where over-conservatism may be unnecessarily restrictive. The major elements of the NRC program are described below.
The NRC Generic Issues Program A Technical Activities Steering Committee was established to increase high level management involvement in and to improve management oversight of generic technical activities. The Steering Committee is chaired by the Deputy Director, Office of Nuclear Reactor Regulation (NRR), and includes, as members, 'he four Division Directors in NRR. The Comgittee's functions include assigning proposed generic tasks to priority categories, assigning lead responsibility to an NRR division for defining and executing each generic task, approving Task Action Plans, and regularly reviewing the progress of ongoing tasks.
The Steering cc:nmittee's judgmental decisions regarding priorities and other matters, such as the assignment of an NRR division with lead 6
responsibility and approval of the Task Action Plan for each task, are based upon recommendations resulting from an extensive internal review process. This process begins in the NRR line organizations through their development, review, comment and concurrence on proposals for high priority tasks and Task Action Plans.
Ir addition, specific recommendations regarding these proposals are provided by the Steering Committee's Advisory Group following its detailed review. The Advisory Group is made up of five senior technical staff representing each of the NRR divisions and the Director, NRR.
Implementation of the program cegan by making the judgments referred to above regarding the relative priority of a large number of ongoing, planned or suggested generic efforts. The generic issues that were con-sidered included those from the Advisory Committee on Reactor Safeguard's listing, those listed in NRR's former Technical Safety Activities Reaort, the 27 issues discussed in " Staff Discussion of Fifteen Technical Issues Listed in Attachment to November 3, 1976 Memorandum from Director, NRR to NRR Staff" (NUREG-0138) and " Staff Discussion of Twelve Additional Tech-nical Issues Raised by Responses to Novemoer 3,1976 Memorandum from Director, NRR to NRR Staff" (NUREG-0153), E and a number of other generic issues that were identified from a variety of sources as described earlier A
in this response (ptge 2). The Steering Committee adopted four priority category definitions as descriptive of the various eeneric technical issues. These definitions are presented in Table 1.
As indicated by 1/ NUREG-0138 and NUREG-0153 ouolisned in November and Decemoer 1976 respectively, provided the staff's discussion of 27 technical issues identified by one or more memoers of the NRR staff as problems whose priority, progress or resolution was, in their opinion, unsatisfactory.
these definitions, issues were assigned to the various criority categories based on their judged safety, environmental or safeguards 'mportance or their potential for improving the efficiency Jr effectiveness of the licensing procass.
Initially, each of the four NRR divisions described and croposed to the Technical Activities Steering Committee those generic issuos it considered to warrant the nignest priority effort (Category A and Category 3 tasks).
Proposals were received for over 130 Category A tasks and over 225 Category 3 tasks in April and May 1977, respectively.
Many of there prooosals were duplicates or could De readily combined as a part of an-other proposed generic issue. These proposals were reviewed in detail by the Steering Committee's Advisory Group.
Following its review, the Advisory Group made recommendations to the Steering Committee for each task regarding the Priority Category to which it should be assigned anc the NRR division that should be assigned lead responsibility. The Steerirg Comittee reviewed the division prcposals and the recomendations of its Advisory Group, assigned each task to a Priority Category and designated an NRR division with lead res::ensibility (Lead Divisicn) for each task.
Implementation of the NRC program has been a major effort that has recuired the carticipation of virtually every working and management level in NRR.
Decisions regarding the relative priorities of the hundreds of generic issues that have beer suggestec, although based on agreed ucon criteria
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(Table 1), in the final analysis are the product of +5e collective judgments of the individuals participating in the decision making process, 11 this case the proposing line organizations, the Steering Committee's Advisory Group, and the Steering Committee itself.
Table 2 provides a listing of the issues assigned to Driority Categories A, B, C and D by the Steering Committee from among those issues originally croposed as Category A and 3 tasks. A Task Action Plan has been approved by the Technical Activities Steerir.g Committee for each of the Category A Tasks listed in Table 2.
Category A Tasks Copies cf Task Action Plans for the Category A tasks listed in Table 2 are contained in " Task Action Plans for Generic Activities, Category A",
(NUREG-0371), oublished in November 1978. This is a r.tised version of the previous NUREG-0371 published in December 1977.
It is the staff's intention to maintain this document up to date by issuing revisions to the task action plans as they become available and adding new task action plans as new Category A issues are identified.
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TADLE 1 PRIORITY CATEGCR) DEFINITIONS Category A:
Those generic technical activ ties jucged by the staff to warrant i
priority attention in ter.ns of manpower and/or funds to attain early resolution. These uatters include those the resolution of.vhicn coulc (1) provide a significant increase ir issurance of the he31:n and safety of the public, or (2) have a significant impact upon the reactcr licensing process.
Category B:
Those generic technical activities judged by the staff to be important in assuring the continued health and safety of the puDlic but for wnicn early resolution is not requirec or for which the staff perceives a lesser safety, safeguarcs or environmental significance than Category A matters.
Category C:
TSose generic technical activities judged by the staff to have little direct or immediate safety, safeguarcs or environmental significance, but which could lead to improvec staff understanding of particuiar technical issues or refinements in the licensing process.
Category "-
Those prop ned generic technical activities judged by the staff not to warrar* the expenditure of manpower or funds because little or no importance to the safety, environmental or safeguards aspects of nuclear reactors or to improving the licensing procesr can be attributed to the activity.
1 TABLE 2 LIST OF TECHNICAL ACTIVITIES Task No.
Title Category A Tasks A-1 Water Hanmer A-2 AsyTr.etric Blowdown Loads on DWR Primary Coolant Systems A-3 Westingncuse Steam Generator Tube Integrity A4 Comoustion Engineering Steam 7enerator Tube Integrity A-5 Babcock 3 Wilcox Steam Generator Tube Integrity A-6 Mark I Short Term 3rogram A-7 Mark I Lcng Term Program A-8 Mark II Containment Pool Dynamic Loads A-9 ATWS A-10 BWR Nozzle Cracking A-li Reactor Vessel Materials Toughness A-12 Fracture Tcughness of Steam Generator and Reactor Cociant Pump Suoports A-13 Snubber Ocerability Assurance A-14 Fla Detection A-15 Prir. iry Coolant System Decontamination and Steam Generator Chemical Cleaning A-16 Steam Effects on BWP Core Spray Distr-n A-17 Systems Interaction in Nuclear ?cwer F '.c A-18 Pipe Rupture Design Criteria A-19 Digital Computer Protection Systems A-20 Impacts of the Coal Fuel Cycle A-21 Main Steam Line Break Inside Containment - Evaluation of Environmental Conditions for Equipment Qualification A-22 PWR Main Steam Line Break - Core, Reactor Vessel and Containment Response A-23 Containment Leak Testing A-24 Qualification of Class IE Safety-Related Equipment A-25 Nonsafety Loads on Class IE Pcwer Sources A-26 Reactor Vessel Pressure Transient Protection (Overpressure Protection)
A-27 Reload Acolications 2
A-28 Increase in Spent Fuel Pool Storage Capacity A-29 Nuclear Pcwer Plant Design for the Recuction of Vulneracility to Incustrial Sabotage
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Table 2 (Cont'd)
Task No.
Title A-30 Adequacy of Safety-Related DC Power Supplies A-31 RHR Shutdown Requirements A-32 Missile Effects A-33 NEPA Reviews of Accident Risks A-34 Instruments for Monitoring Radiation and Process Variables During Accidents A-35 Adequacy of Offsite Power Systems A-36 Control of Heavy Loads Near Spent f uel A-37 Turbine Missiles A-38 Tornado Missiles A-39 Determination of Safety Relief Valve (SRV) Pool Dynamic Loads and Temperature Limits for BWR Containments A 40 Seismic Design Criteria - Short Term Program Category B Tasks B-1 Environmental Technical Specifications B-2 Forecasting Electricity Demand By State in tne United States on an Annual Basis B-3 Event Categorization B4 ECCS Reliability B-5 Ductility of Two-Way Slabs and Shells and Buckling Behavior of Steel Containment B-6 Loads, Load Combinations, Stress Limits B-7 Secondary Accident Consequence Modeling B-8 Locking Out of ECCS' Power Operated Valves B-9 Electrical Cable Penetrations of Conta'nment B-10 Behavior of BWR Mark III Containment B-ll Subcompartment Standard Problems B-12 Containment Cooling Requirements (Ncn-LCCA, B-13 Marviken Test Data Evaluations B-14 Study of Hydrogen Mixing Capability in Containment Post-LCCA B-15 CCNTEMPT Comcuter Code Maintenance B-16 Protection Against Postulated Piping Failures in Fluid Systems Outside Containment B-17 Criteria for Safety-Related Ocerator Actions B-18 Vortex Suppression Requirements for Containment Sumps B-19 Thermal-Hydraulic Stability Table 2 (Cont'd)
Task No.
Title 3-20 Standard Problem Analysis B-21 Core Physics B-22 LWR Fue; 3-23 LMF3R Fuel 3-24 Seismic Qualification of Electrical and Mecnanical Components 3-25 Piping Benchmark Problems B-25 Structural Integrity of Containment Penetrations 3-27 Implementation and Use of Subsection NF B-28 Radionuclide/Sedimer.. Transport Program B-29 Effectiveness of Ultimate Heat Sinks 3-30 Design Basis Floods and Probability B-31 Dam Failure Model B-32 Ice Effects on Safety-Related Water Supplies B-33 Dose Assessrent Methodology S-34 Occupational Radiation Exposure Reduction B-35 Confirmation of Appendix I Mcdels for "Calcu!ations of Releases of Radioactive Materials in Gasecus and Liquid Effluents From Light-Water-Cooled Pcwer Reactors" 3-36 Develop Design, Testing and Maintenance Criteria for Atmaschere Cleanup Systen Air Filtration and Adsorption Units for Engineered Safety Feature Systems and for Normal Ventilation Systems B-37 Chemical Discharges to Receiving Waters B-38 Reconnaissance Level Investigations B-39 Transmission Lines B-4C Effects of Power 31 ant Entrainr.ent on Plankton 3-41 Imcacts on Fisheries 3-42 Sccioeconcmic Environmental Imcacts B-43 Value of Aerial Photographs for Site Evaluation 3-44 Forecasts of Generating Costs of Coal and Nuclear Plants 3 45 Need for acwer - Energy Conservation 3-46 Costs of Alternatives in Environ ~1ntal Design B-47 Inservice Inspection Criteria fcr Suppcrts and Bolting of Class 1, 2, 3 and MC Com!anents 3 43 B'aR CR0 Mechanical Failure (Collet Housing) 3 49 Inservice Inspection Criteria for Containment Table 2 (Cont'd)
Task No.
Title B-50 Requirements for Post-CBE Inspection B-51 Assessment of Inelastic Analysis Technioues B-52 Fuel Assembly Seismic and LOCA Responses B-53 Load Break Switch B-54 Ice Condenser Containments B-55 Imorovec Reliacility of Target-Rock Safety-Relief Valves B-56 Diesel Reliability B-57 Station Blackout B-53 Passive "echanical "tilures B-59 Review of (N-1) Loc; Operation in SWRs and PWRs B-60 Loose Parts Monitoring Systems B-61 Allcwable ECCS Equipment Outage Periods B-62 Reexamination of Technical Bases for Establishing SLs, LSSSs, etc.
B-63 Isolation of Low Pressure Systems Connected to RCPB B-64 Decommissiong of Reactors B-65 Icdine Spiking B-66 Control Room Infiltration Measurements B-67 Effluent and Process Monitoring Instrumentation B-68 Pump Overspeed During a LOCA B-69 ECCS Le2kage Ex-containment B-70 Power Grid Frequency Degradation and Ef fect on Primary Coolant Pumos B-71 Incident Response B-72 Development of Models for Assessing Risk of Health Effects and Life Shortening from Uranium and Coal Fuel Cycles B-73 Mcnitoring for Excessive Vibration Inside the Reactor Pressure Vessel Category C Tasks C-1 Assurance of Continuous Long-Term Intagrity of Seals on Instrume..tation and Electric Equipment C-2 Study of Containment Depressurization by Inadvertent Spray Coeration to Determine Adequacy of Containment External Design Pressure C-3 Insulation Usage Within Containment C4 Statistical Methods for ECCS Analysis C-5 Decay Heat Update C-6 LCCA Heat Sources Table 2 (Cont'd)
Task No.
Title C-7 PWR System Piping C-8 Main Steam Line Leakage Control System C-9 RHR Heat Exchanger Tube Failure C-10 Effective Operation of Containment Sprays in a LOCA C-11 Assessment of Failure and Reliability of Dumps and Valves C-12 Pri.marj System Vibration Assessment C-13 Non-Random Failures C-14 Storm Surca Model for Coastal Sites C-15 NUREG Report for Liquid Tank Fai'ure Analysis C-16 Assessment of Agricultural Land in Relation to Pcwer Plant Siting and Cooling System Selection C-17 Interim Acceptance Criteria for Solidification Agents for Radioactive Solid Wastes Category 0 Tasks 0-1 Advisability of a Seismic Scram 0-2 Emergency Core Cooling System Capability for Future Plants 0-3 Control Rod Drop Accident (BWRs)
NUREG-0371 meets most of the informational requirements of ALAB 444 for the Category A tasks applicable and relevant to the Perkins proceeding.
The Task Action Plans in NUREG-0371 provide a description of the pro-blems; the staff's approacnes to resolution; general discussion of the bases upon which continued plant licensing or operation can proceed pend-ing completion of the task; technical organizations involved in the review and estimates of the manpower required; descriptions of the interactions with other NRC offices, the Advisory Committee on Reactor Safeguards and outsR organizations; estimates of any funding required for contractor supplied technical assistance; prospective dates for completing the tasks; and descriptions of any potential problems that could impact the plans.
Although 133 Category A, B, C and D generic tasks were identified by the staff and approved by the Steering Committee as listed in Table 2, not all are applicable to each type or vintage of plant.
For example, in the case of Perkins Nuclear Station, Units 1, 2 and 3, the following Category A tasks are not applicable because they are peculiar to boilir' w-*
reactors:
A-6 Mark I Containment Short Term Program A-7 Mark I Containment Long Term Program A-8 Mark II Containment Program A-10 SWR Nozzle Cracking A-16 Steam Effects on SWR Core Spray Distribution A-39 Deternination of Safety Relief Valve (SRV)
Pool Dynamic Loads and Temcerature Limits for SWR Containments
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The following tasks are not applicable because they are peculiar to pressurized water reactors supplied by other NSSS vendors:
A-3 Westinghouse Steam Generator Tube Integrity A-5 Babcock & Wilcox Steam Generator Tube Integrity Other generic tasks are not relevant to the licensing actions for t particular facility because they deal with iinproving the efficiene and/or effectiveness of the licensing process rather than plant safety.
These types of tasks include (1) efforts to impron guidance to accli-cants, licensees or staff reviewers, and (2) efforts to consider the relaxation of certain staff requirements that may be overly conservative. 2/
Category A tasks that fall into the first group are:
A-15 Primary Coolant System Decentamination and Steam Generator Chemical Cleaning A-19 Digital Computer Protection Systems A-27 Reload Apolication Guide A-28 Increase in Spent Fuel Storage Capacity A-34 Instruments for Monitoring Radiation and Process Variables During Accidents A-37 Turbine Missiles Category A tasks that fall into the second group are:
A-25 Non-safety Loads on Class IE Power Sources A-38 Tornado Missiles A-10 Seismic Design Criteria - Short Term Program 2/ Secause the staff's view is that these types of generic tasks are not relevant to the licensing actions for any particular facility, they have not been addressed specifically for the Perkins facility in this resconse. Hcwever, the Task Action Plans for these tasks are included in NUREG-0371.
Other Category A tasks listed in Table 2 are not relevant to the informational requirements of ALAB-444 because they deal with environmental issues rather than safety issues. These include:
A-20 Impacts of Coal Fuel Cycle A-33 NEPA Reviews of Accident Risks The remaining 21 Category A ceneric tasks are, to varying degrees, related to plant safety and are applicable to the Perkins Nuclear Station.
A discussion of each of the issues addressed by these 21 tasks as they relate to the Perkins Muclear Station is provided in a subsequent section of this response.
Catecary B, C and 0 Tasks Initially, the Category B tasks listed in Table 2 were judged by the Steering Ccanittee to be of lesser safety, safeguards or environmental significance than Category A tasks. Category C tasks listed in Table 2 were judged to have little direct or immediate safety, safeguards or environmental significance and Category D tasks listed in Table 2 were judged to have little or no imoortance to the safety, safeguards cr environmental aspects of nuclear reactors or to improving the licensing process. The staff has compiled a brief descriotion of the Category B, C and 0 tasks listed in Table 2 in " Generic Task Problem Description, Category 3, C and 0 Tasks", (NUREG-0471) published in June 1973. No Task Action Plans have as yet been approved by the Technical Activities Steering Conmittee for these tasks.
As indicated by the Priority Category definitions in Table 1, those issues assigned to Category A are the *.ost important generic tasks, includ-ing tt 1se most important in terms of safety significance. The term
" lesser... safety significance" in the Category 8 definition has not been defined. However, the report of the NRR Task Force (see NUREG-0410, Section 1.0) that originally developed the Priority Categcry definitions does offer some further insight.
In addition to developing the Priority Category definitions, the Task Force developed a set of criteria to be used to test each identified activity for assignment to the proper category. The intent was that an activity, meeting one or more of the test criteria for a given category, would be assigned to that category. The tests for Cat 3ry A are:
1.
Resolution could remedy significant deficiencies in facility design or operation.
2.
Early resolution of issue could significantly improve the existing regulatory process.
3.
Other activities that are judged to require high level management attention and oversight.
The tests for Category 3 are:
l.
Issue is important to safety, safeguards or environmental protection, but of smaller scope that does not reouire NRR wide coordination to obtain timely resolution.
2.
Resolution needed to confirm adequacy of previous staff judgments.
3.
Issue has potential of becoming a Category A issue.
From these tests it is important to note that any issue whose resolution is needed to " remedy significant deficiencies in facility design or ooeration" would be assigned to Category A.
Although some issues "important to safety"'could be assigned to Category 3, the intent was clearly not to assign issues that met the first Category A test to Category B.
Issues that are judged to meet the first Category A test are those that have the "potentially significant public safety implication (s)'
referred to in River Bend.
Since such issues are not assigned to Category B, it is not necessary to meet all of the informational require-ments of the River Send decision for Category B, or lower category generic issues.
The third test for Category B indicates that issues assigned to Category B could be elevated to Category A if new information indicates that the issue is of greater importance to safety than originally judged.
None of the issues originally assigned to Category 8, C or D had beer, elevated to Category A until recently.
Recent staff efforts have resulted in the elevation of several of these lower priority tasks as discussed later in this response.
Of those remaining Category 3, C and D tasks that are related to olant safety and are acclicable to the perkins Nuclear Station, we have not identified any that could not be resolved either by system alterations using available techniques and equipment or by operational modifications in the event that our review of the issue revealed that current require-ments required uograding during construction or operation. On this basis and the Steering Committee's judgment that the Category 3, C and D issues are of lesser safety significance than Category A issues, detailed information on these tasks is not necessary and we have not included any such information in this response.
" Unresolved Safety Issues Plan" The Commission now has updated its discussicn of " Unresolved Safety Issues" "sr inclusion in the 1978 NRC Annual Report. This matter is discussed earlier on pages 5 and 6 of this response. As noted there, the initial response of the NRC to the new Section 210 of the Energy Reorganization Act was submission of a report, NUREG-0410, which described the overall generic issues program of the staff and was considerably broader in scope than the " Unresolved Safety Issues Plan" that is required by Section 210.
The staff program includes other generic tasks of importance to accomplishing the NRC's mission such as tasks for the resolution of environmental issues, for the development of improvements in the reactor licensing process, for consideration of less consersative design criteria or operating limitations in areas where overly conservative recuirements may be unnecessarily restrictive or costly, for the maintenance and cevelopment of the NRC staff's capabilities to cerform indeoendent audit calculations, and for the actual performance of independent addit calcu-lations. The section being included in the 1978 NRC Annual Report is limited to describing the progress on those issues in the overall NRR program that have been determined to be " Unresolved Safety Issues".
It is the Cormission's belief that the intent of Section 210 was to assure that plans were developed and implemented on issues that posed substan-tive questions about the adequacy of current safety requirements or current plant designs. For this reason, the following definition Of an
" Unresolved Safety Issue" was developed for use in identifying those generic issues in the broader NRR staff program, that should be reported to Congress pursuant to Section 210.
"An Unresolved Safety Issue is a matter affecting a number of nuclear power plants that poses important questions con-cerning the adequacy of existing safety requirements for which a final resolution has not yet been developed and that involves conditions not likely to be acceptable over the lifetime of the plants affected."
All of the generic issues reported to the Congress last year in NUREG-0410 were considered as candidates for " Unresolved Safety Issues". A systematic review of these issues was undertaken. As an aid in conducting this review, the topics addressed by these issues were evaluated from the standpoint of their relative contribution to public risk. This risk-based characterization was utilized in conjunction with a substantial body of additional information (e.g., heavy weight was given to issues that resulted from events that have been recorted to the Congress as Abnormal Occurrences pursuant to Section 208 of the Energy Reorganizatien Act of 1974, as amended) to determine which issues met the definition of an
" Unresolved Safety Issue".
During the course of this review effort, the Technical Activities Steering Committee acted to elevate Tasks 3-18, 3-57 and C-3 to Category A tasks.
Tasks 3-18 and C-3 were combined and designated as Task A-43 " Containment Emergency Sumo Reliability", and Task 3-57 was designated as Task A-44
" Station Blackout".
In addition, in view the continuing problems witn cracks in the piping of boiling water reactors, the Technical Activities Steering Committee also established a new Task A-42 "Pioe Cracks at Boiling Water Reactors".
As a result of this review effort the seventeen generic issues in the following listing were determined to be " Unresolved Safety Issues".
These generic issues are addressed by twenty-two generic tasks in the NRR Pro-gram for the Resolution of Generic Issues. The task numbers of the applicable generic tasks are provided in parentheses following the title of each issue. Three of the twenty-two generic tasks addressing these seventeen issues have been completed. Generic Task A-6 was comoleted and documented in a report, NUREG-0408, " Mark I Containment Short Term Pro-gram Safety Evaluation Report", in December 1977; Generic Task A-26 was completed and documented in NUREG-0224, " Reactor Vessel Pressure Tran-sient Protection for Pressurized Water Reactors", in September 1978; and Generic Task A-31 was comoleted and documented in Regulatory Guide 1.139,
" Guidance for Residual Heat Removal", in May 1978.
"U"RESCLVED SAFETY ISSUES" (APPLICABLE TASK NOS.)
1.
Water Hammer - (A-1) 2.
Asymmetric Slowdown Loads on the Reactor Coolant System -
(A-2) 3.
Pressurized Water Reactor Ste:m Generator Tube Integri ty - ( A-3, A-4, A-5)
~
4.
BWR Mark I and Mark II Pressure Suppression Containments -
(A-6,A-7,A-8,A-39) 5.
Anticipated Transients Without Scram - ( A-9) 6.
BWR Nozzle Cracking - (A-10) 7.
Reactor '/essel Mate-ials Toughness - (A-ll) 8.
Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports - (A-12) 9.
System Interactions in Nuclear Power Plants - (A-17) 10.
E,vironmental Qualification of Safety-Related Electrical Equipment - (A-24) 11.
Reactor 'iessel Pressure Transient Protection - (A-26) 12.
Residual Heat Removal Requirements - (5-31) 13.
Control of Heavy Loads Near Spent Fuel - (A-36) 14.
Seismic Design Criteria - (A-42) 15.
Pipe Cracks in Boiling Water Reactors - (A-42) 16.
Containment Emergency Sumo Reliability - (A43) 17.
Station Blackout - (A-44)
The " Unresolved Safety Issues" listed above are being reported to Congress as those issues that pose important questions concerning the adequacy of existing safety requirements.
"onetheless, for the purpose of this par-ticular resocnse, we have elected to discuss each of the Category A issues that are relevant and applicable to the Perkins facility.
In this connection it should be noted that recently identified Task A-42 " Pipe Cracks in Boiling Water Reactors" is not apolicable to the Perkins units which use cressurized water reactors.
Thus, t.bere are ncw a total of 23 Category A generic tasks that are apolicable and relevant to tne Perkins Nuclear Station. These tasks are listed below and are discussed in the following section:
CATEGORY A TASKS APPLICABLE AND RELEVANT TO THE PERKINS NUCLEAR STATION 1.
A-1 Water Hammer 2.
A-2 Asymmetric Blowdown Loads on PWR Primary Coolant Systems 3.
A4 Combustion Engineering Steam Generator Tube Integrity 4.
A-9 AT'.iS 5.
A-ll Reactor Vessel Materials Toughness 6.
A-12 Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports 7.
A-13 Snubber Operability Assurance 8.
A-14 Flaw Detection 9.
A-17 Systems Interaction in Nuclear Power Plants 10.
A-18 Pipe Rupture Design Criteria 11.
A-21 Main Steam Line Break Inside Containment -
Evaluation of Environmental Conditions for Equipment Qualification 12.
A-22 PWR Main Steam Line Break - Core, Reactor Vessel and Containment Response 13.
A-23 Containment Leak Testing 14.
A-24 Qualification of Class IE Safety-Related Equipment 15 A-26 Reactor Vessel Pressure Transient Protection (Overpressure Protection) 16.
A-29 Nuclear Power Plant Design for the Reduction of Vulnerability to Industrial Sabotage 17.
A-30 Adequacy of Safety-Related DC Power Supplies 18.
A-31 RHR Shutdown Requirements 19.
A-32 Missile Effects 20.
A-35 Adequacy of Offsite Power Systems 21.
A-36 Control of Heavy Loads Near Scent Fuel 22.
A-43 Containment Emergency Sump Reliability 23.
A-44 Station Blackout Category A Tasks Acolicable and Relevant to the Perkins Nuclear Station Facilities We have reviewed those 23 Category A tasks that are applicable and relevant to our safety findings regarding issuance of construction per-mits for the Perkins Nuclear Station facilities and the related staff discussion in the Perkins SER. Discussions of the impact of each of these tasks on our findings are provided below.
In some instances, additional discussion related to specific tasks is provided in separate responses.
In such instances, we have provided an appropriate reference.
Based on our review of these items, we have concluded that the conclusions regard-ing the issuance of construction permits presented in the Perkins Nuclear Station SER are unaffected by these generic tasks.
A-1 Water Hamer As indicated in NUREG-0371, November 1978, since 1971 there have been about 100 incidents involving water hammer in pressurized water reactors and boiling water reactors. The water hamers have involved steam gen-erator feedrings and piping, the decay heat removal system, emergency core cooling system, containment spray lines, service water lines, feedwater lines and steam lines. However, the systems most frequently affected by water haniner effects are the feedwater systems of pressurized water reactors. These types of water hammer events are addressed in the staff's SER for Perkins at page 10-5 and at page 10-2 of Apoendix A to the SER vere the potential for future system design changes are recognized. No new requirements for system design modifications have to date been defined by the staff for the Perkins facilities. Any necessary modifications wnich might be identified by further staff work on this particular generic con-cern will be required during our review of an application for a license to operate the Perkins plant.
As indicated in Section 3.0 of the Task Action Plan for Task A-1, adequate protection is currently orovided against potential water hamer in other plant systems. We consider that the Perkins applicant has ful-filled the preliminary design requirements necessary at the construction permit stage of review. Accordingly, our previous conclusions in the Perkins SER regarding the issuance of construction permits are unaffected by this ongoing generic task.
A-2 Asymetric Bicwdown Loads on Reactor Primarv Coolant System The apolicant has complied with all current staff requirements regarding this generic issue (SER, Apoendix A, page 3-11).
As described in the Task Action Plan for Task A-2, the staff expects to complete its review of analytical methods for assessing asymmetric blowdown loads for Combustion Engineering Dasigns by March 1979. A final, olant-specific analysis for the Perkins facility will be comoleted during the operat-ing license review for this facility.
Accordingly, cur previous conclusions in the Perkins SER regarding the issuance of construction permits are unaffected by this ongoing generic task.
A-4 Combustion Encineerina Steam Generator Tube Intecrity The applicant has complied with current staff requirements for the construction permit stage of review. Specific measures that will be taken by the applicant to ensure tnat the tubes will not be subjected to conditions that will cause deleteriou,s wastage or cracking are described in tne SER at page 5-5, and at pages 5-10 and 5-11 of Apoendix A to the SER.
Based on our review of these measures, we concluded that construc-tion per nits for the Perkins facilities can be issued with reasonable assurance that there will be no undue risk to the health and safety of the public. The efforts under Task A-4 regarding steam generator tube integrity may result in improved criteria that provide further assurance in this regard.
However, such improvements are likely to be procedural, rather than system modifications and their application to the Perkins facilities is a matter that can reasonably be left to the operating license stage of review. Accordingly, cur previous conclusions in the Perkins SER regard-ing the issuance of construction permits are unaffected by this ongoing generic task.
A-9 ATWS As noted on pages I and 2 of'this response, this issue is discussed separately in the Supplement to NRC Staff Responses to Board Questions Con-E cerning Health and Safety Ascects of Cherokee and Perkins.
A-ll Reactor Vessel Materials Touchness Section 3 of the Task Action Plan for Task A-ll indicates that current criteria related to fracture toughness together with the materials currently employed for reactor vessel fabrication are adequate to ensure suitable safety margins for the reactor vessels throughout their design lives. As indicated in the CESSAR SET < (Appendix A to the Perkins SER) at pages 5-5 and 5-6, the reactor vessels for the Perkins facilities will be designed, fabricated, tested, and coerated in conformance with current criteria and materials.
Our conclusion in the CESSAR SER on page 5-5 indicates that confor ance with such criteria will ensure adequate safety margins during operating, testing, maintenance and postulated accident conditions.
This conclusion regarding postulated accident conditions was a judgment made by the staff based principally on empirical results from NRC experiments at the Oak Ridge National Laboratory.
Further testing is currently being considered.
As part of Task A-ll, the staff is evaluating reactor vessel material toughness under postulated accident conditions. The results of this evaluation may indicate that some operating plants have to take measures late in life to assure adequate fracture toughness for postulated E
accidents.
A preliminary review indicates that adequate safety margins can be maintained even for these older plants for postulated accidents for up to approximately 20 years of neutron irradiation.
Although this part of Task A-ll could conceivably provide information relevant to current vintage reactor vessels, we celieve that thcse vessels will have adequate safety marg'.ns related to fracture tougnness for postulated accidents throughout their design life. Accordingly, we anticipate that the task results will confirm our previous conclusions presented in the Perkins SER.
A-12 Fracture Touchness of Steam Generator and Reactor Coolant Pumo Succorts The applicant has comnitted to comply with current staff requirements.
Section 3 of the Task Action Plan for Task A-12 orovides the basis for continued licensing pending completion of this generic task.
On the basis of this discussion, we have concluded that our orevious conclusions in the Perkins SER regarding issuance of construction permits are unaffected by this ongoing generic affort.
A-13 Snubber Ooerability Assurance The applicant will be required to describe the program for snubber operability assurance at the operating license stage of review. We anticipate that prior to the operating license review, our generic efforts under Task A-13 will provide comprehensive requirements for snubber opera-bility assurance for use in the staff's review at that time.
Further, we E
expect that any future requirements that may be required for the Perkins facilities can be satisfactority implemented at the operating license stage. Accordingly our previous conclusions in the Perkins SER regard-ing the issuance of construction permits are uncffected by this ongoing tas k.
A-la Flaw Cetection As described in the Task Action Plan for Task A-14, o s generb efforts, in conjunction with those being undertaken by industry organizations, may result in improved inspection techniques for the detection of defects in tne reacter coolant pressure boundary.
However, as described in Section 3 of the Task Action Plan, such improvements, although desirable, are not necessary to maintain adequate margins of safety. Accordingly, our ore-vious conclusions in the Perkins SER regarding issuance of construction permits are unaffected by this ongoing generic task.
A-17 Systems Interactions In Nuclear Power Plants The licensing requiremer ts and crocedures used in our review address many different types of systems interactions. Task A-17 has been developed to confirm that our current review process encompasses all pctentiaily ad-verse systems interactions. As indicated in Section 3 of the Task Action Plan for Task A-17, we anticipate " at this task will c.onfirm that current licensing requirements and procedures are adequate, al.hougn some modifi-cations for improvement in the review procedures and licensing recuire-ments may be made. On this basis, we believe that our orevious conclusions in the Perkins SEP regarding issuance of construction permits ar; tqaffected by this ongoing generic task.
A-18 Pice Ruoture Desian Criteria The acclicant has complied with all current staff sa"ety requirements regarding pipe rupture design as discussed in Section 3.6 of the SER t
and in Section 3.6 of the CESSAR SER (included as Aapendix A to the Perkins SER). As indicated in Section 3 of the Task Action Plan for Task A-18, the task may result in adjustments to the current criteria to achieve a better balance between design of piping systems fer normal operation and design to assure adequate protection against postulated pipe rupture.
Such adjustments are desirable, but not necessary, to assure that plants such as the Perkins units that meet current requirements, have adequate protection against pipe breaks. Accordingly, our previous cd.iclusions in the Perkins SER regarding issuance of construction oermits are unaffected by this ongoing generic task.
A-21
'dain Steam Line Break Inside Containrent As indicated in the Task Action Plan for Task A-21, the generic task will establish the acceptability of steam generator blowdown and containment analysis models to be used to calculate the worst case main steam line break for equipment qualification. The results of the generic task are expected to be available for use in the operating license raview for Perkins.
As discussed in the SER at paces 3-18 and 6-3, the apolicant has committed to use a temperature higner than the c?lculated ceak temcerature resultina from the worst case main steam line break for Perkins tor the qualification of safety-related electrical ecuipment located inside con-tainment. We have concluded that this commitment is acceptable at the const,uction permit stage of review. Accordingly our con:lusions in the SER regarding issuance of construction permits are unaffected by this ongoing generic task. -
A-22 pWR Main Steam Line Break - Core, Reactor Vessel and Containment Buildino Resoonse This generic task is expected to confirm our prior determination that the postulated main steam line break accident has been conservatively evaluated using current staff licensing requirement;. The generic study, when com-pleted, will include an evaluation of the reliability of non-safety grade equipment and an evaluation of the adequacy of certain safety systems and operator actions necessary to mitigate the consequences of the main steam line break.
Further discussion of this matter is orovided in the Task Action Plan for Task A-22.
No increases in requirements are foreseen that would require design changas or other modifications to the Perkins facilities as currently proposed by the apolicant.
However, if desig, modifications are warranted, we will require that such changes are made.
Accordingly, our previous conclusions in the Perkins SER regarding the issuance of construction permits are unaffected by this generic task.
A-23 Ccntainment Leak Testing As reported in the SER at page 6-2, we have reviewed the applicant's E
containment leak testing program description and have concluded that it is acceptable for the construction permit stage of review.
Details of the acclicant's containment leak testing program will be reviewed for conformance to Appendix J to 10 CFR Part 50 at the acerating license stage of review. As indicated in Section 3 of the Task Action Plan for Task A-23, tne generic effort is to deve. lop proposed changes to Accendix J to clarify its application and to resolve any conflicting or imoractical requi rements.
Even if such modifications are not available at the operat-ing license stage of review, the staff's review will cortinue to orovide the necessary interpretation of the existing regulation to assure that containment integrity will be effec;ively monitored. Accordingly, our previous conclusions in the Perkins SER regarding issuance of construc-tion permits are unaffected by this ongoing generic task.
A-24 Oualification of Class IE Safety-Related Ecuioment The acplicant has comolied with all staff requirements at the construction permit stage of review, as reported in the SER at pages 3-18 and 3-19.
In Appendix A to tha Perkins SER, at page 7-13, Combustion Engineering has stated that all Class IE equipment in Combustion Engineering's scoce will be qualified in accordance with IEEE Std. 323-1974 without exception. We concluded that these commitments were acceptable at the construction per-mit stage of review, but that we would conduct a review of a technical report on khis matter, required to be submitted by the apolicant approxi-mately six..eaths after issuance of construction permits for the Perkins facility.
4 Task A-24 was developed to provide a mechanism for conducting the generic review of equipment qualificatinn programs of the major NSSS vendors and balance-of-plant equi:: ment suppliers. The review of the Combustion Engineering qualification program discussed above is encom::asseo by this task and is a convenient and efficient means of completing the review by the operating license stage for tne Perkins facilities. Accordingly, our previous conclusions in the Perkins SER regarcing issuance of construction permits are unaffected by this ongoing generi-task.
Constrt.ction permits were issued on December 30, 1977 for Cherokee Nuclear Station, Units 1, 2 and 3.
On June 29, 1978, in response to its commit-ment stated in the SER, Duke Power Company submitted a report on its Class IE Equipment Qualification Program.
By letter of October 26, 1978 we advised Duke Power Company of the results of our interim evaluation of this recort and reconnended that the contents of the recort ce revised to conform to the information requirements that we enclosed. The June 29, 1978 report, our interim evaluation, future revisions to the report and our final evaluation will also be applicable to the Perkins facility which is a duplicate of the Cherokee facility.
A-26 Reactor '/essel Pressure Transient Protection,'0veroressure)
The applicant's commitment to provide an overpressure protection system, and the staff aporoval of that commitment, is documented in '.opendix A to the SER at pages 5-2 and 5-3.
Generic Task A-26 to define the criteria for overpressure protection system design and operation is comolete and the results have been documented in NUREG-0224, " Reactor '/essel Pressure Transient Protection for Pressurized Water Reactors", issued in Sectember 1978. The applicant's previous commitments are consistent with the criteria resulting from the task. As stated in the SER, we will review the details of the design and procedural provisions during tne operating license stage of review. Accordingly, our previous conclusions in the Perkins SER regarding the issaance of construction permits are unaffected by this generic task.
A-29 Nuclear Pcwer Plant Cesien for the Reduction of Vulnerability to Industrial Sabotace The applicant has met all current requirements for a construction permit application, as discussed in the SER at page 13-4 At the operating license stage of review we will require that the applicant demonstrate comoliance with 10 CFR Part 73.55.
As indicated in Section 3 of the Task Action Plan for Task A-29, the implementation of 10 CFR Part 73.55 provides high assurance of protection of the health and safety of the public and accordingly, is an adequate basis on which to license the Ferkins facilities. Although Task A-29 may identify design concepts that could provide alternative or more effective means of achieving protec-tion against industrial sabotage, the imolementation of such design improvements are not necessary to provide adequate protection of the Perkins facilities. Accordingly, our previous conclusions in the Perkins SER regarding the issuance of construction permits are unaffected by this ongoing generic task.
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A-30 Adecuacy of Safety-Related DC Power Socolies The staff's effort on this generic task is expected to confirm that the simultaneous and independent failure of redundant direct current (dc) power supplies is so unlikely as to be incredible, and that their failure from a comon event is judged to have a low enough probability w t a
adequate protection presently exists. This is the current staff view as discussed in Section 3 of the Task Action Plan for Task A-30.
Therefore, although the generic study will provide a quantitative assessment of reliabilities of de power supplies, particularly with respect to coman mode failures, the staff has concluded that continued licensing and operation of nuclear power plants with the dc power supply system designs now in use and proposed does not present an undue risk to the health and safety of the public. Accordingly, our previous conclusions in the Perkins SER regarding issuance of construction permits are unaffected by this ongoing generic task.
A-31 RHR Shutdown Requirements The staff evaluation of the Combustion Engineering Standard Safety Analysis Report (CESSAR) concluded with issuance of a Safety Evaluation Report (which is included in the Perkins Safety Evaluation Report as Appendix A) and a Preliminary Design Approval No. PDA-2 dated A
December 31, 1975.
Since issuance of PDA-2, staff requirements for the residual heat removal system have been re-evaluated as described in the Task Action Plan for Task A-31.
This generic task has been completed. and revis-1 recuirements have been deieloped and approvec by the staff.
These requirements were embodied in Regulatory Guide 1.139, issued for comment in May 1978. As indicated in the implementation section of the guide, implementation of these require-ments will be reviewed on a case-by-case basis for all plants for which construction permit or Preliminary Design Approval applications were docketed before January 1, 1978.
Both the Perkins construction permit application and the CESSAR Preliminary Design Approval apolication are among those w.1ich must conform to the revised requirements.
The additional requirements to be imposed on the Perkins 'acilities are such that design changes, if necessary, are technically feasible and can be made during the period of reactor plant construction. We conclude that this issue can be acceptably resolved during the post-construction permit period.
Final approv.11 of the design would be necessary prior to a decision on issuance of operating licenses for the Perkins facilities.
Accordingly, our previous conclusicns in the Perkins SER regarding issuance of construction permits are unaffected by the results of tnis generic task.
A A-32 Missile Effects As indicated in Section 3 of the Task Action Plan for Task A-32, the results of this generic task are expected to confirm the edecuacy and conservctism of the current NRC licensing critaria for protection frca
a the effects of impact from missiles caused by accidents, tornados, or failure of the main turbine.
Based on this and the other information provided in Section 3 of the Task Action Plan, we have concluded that our previous conclusions in the Perkins SER regarding the issuance of construction permits are unaffected by this ongoing generic task.
A-35 Adecuacy af Offsite Power Systems The applicant has complied with ail current staff requirements regarding offsite power systems as discussed in the SER at page 8-2.
As indicated in Saction 3 of the Task Action Plan for Task A-35, if the ta'k identifies areas where current criteria should be modified to increase safety margins, such modifications are not exoected to be extensive.
Further, for current construction permit applications such as Perkins, any forthcoming require-ments are expected to be available for consideration for these applications well in advance of a decision on issuance of operating licenses. Accord-ingly, our previcus conclusions in the Perkins SER regarding issuance of construction cennits are unaffected by this ongoing generic task.
A-36 Control of Heavy loads Near Scent Fuel The applicant has complied with our current criteria for fuel handling systems, as discussed in the SER at pages 9-2 and 9-4 As indicated in Section 3 of the Task Action Plan for Task A-36, any revisions to current requirements that might evolve from this generic study will be available
,, ell in advance of the planned operation of the Perkins facility.
Further, the types of changes in current requirements, if any, that could v
-w result are expected to be procedural and, therefore, their implementation can reasonably be left to the operating license stage of review.' Accord-ingly, our previous conclusions in the *-. ins SER regarding the issuance of construction permits are unaffected by this ongoing generic task.
A-43 Containment Emeroency Sumo Reliability The NRC staff initially planned to study the issue of containment emergency sump blockage from insulation as part of Generic Task C-3, " Insulation Usage Within Containment."
In addition, initial plans were to study the vortex formation issue as part of Generic Task B-18, " Vortex Suppression Requirements for Containments." However, containment emergency sumo operability is fundamental to the successful operation of both the emergency core cooling system (needed to ecol the core) and the contain-ment spray system (needed to assure containment integrity) following a loss-of-coolant accident. For this reason, these portions of Tasks C-3 and 8-18 have been combined and elevated to Category A as Generic Task A-43 under the more general title of " Containment Emergency Sump Reliability." Because this action has only recently been taken, a Task Action Plan and schedule for this task have not yet been developed.
The Task Action Plan is scheduled to be developed by.4 arch 1,1979.
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W Cne postulated means of losing the ability to draw water from the emergency sump could be blockage by cebris. A principal source of such debris could be the thermal insulation on the reactor coolant system pioing.
In the event of a piping break, the subseauent violent release of the hign pressure water in the reactor coolant system could rip off the insulation in the area of the break. This debris coul'd then be swept into the cump, potentially causing damage.
Currently, regulatory positions regarding sump design are presented in Regulatory Guide 1.82, " Sumps for Emergency Core Cooling and Containment Spray Systems," which addresses debris (insulation). The Regulatory C-uide recommends, in addition to providing redundant separated sumas, that two protective screens be provided. A low approach velocity in the vicinity of the sump is required to allow insulation to settle out before reachir.g the sump screening; and it is required that the sump remain functional assuming that one-half of the screen surface area is blocked.
The Perkins units are designea with two functionally and physically in-dependent containment sumps wnich are in full conformance with the criteria of Regulatory Guide 1.82
' amps for Emergency Core Cooling and Containment Spray Systems,"
(see ?s-kins PSAR Table 1.7-1, Sheet 12 of 19).
Conformance with this Regulatory Guide provides a high degree of assurance that the sumps are not susceptible to L :ockage Ly debris.
.y,
.s A second postulated means of losing the ability to draw water frcm the emergency sump could te abnormal conditions in the sump or at the pump inlet such as air entrainment, vortices, or excessive pressure drops.
These.:enditions could result in pump cavitation, reduced flow and possible dimage to the pumps.
Currently, regulatory positions regarding sump testing are contained in Regulatory Guide 1.79, " Pre-Operational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors," which addresses the testing of the recirculation function. Both in-plant and scale model tests have been performed by applicants to demonstrate that circulation through the sump can be re'liably accomplished.
As stated in the Perkins SER, page 6-19 of Appendix A, tests will be performed in conformance with the guidance Regulatory Guide 1.79.
Sump testing in conformance with Regulatory Guide 1.79 provides a high degree of assurance that emergency sump operation
11-not be impaired by the formation of vortices.
Task A-43 is principally concerned with sumo reliability for clants witn containment emergency sumps desigr.ed and/or tested prior to the NRC staff's current guidance and acceptance criteria.
It is anticipated J
that the task will confirm that current requirements and guidance are adequate, althotgh some minor modifications could conceivably result.
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5 Because the Perkins facility was reviewed using these current recuire-ments, additional recuirements for Derkins are not anticipated.
Accordingly, our previous conclusions in the Perkins SER regarding the issuance of construction permits are not expected to be affected by the results of this generic task.
A,-44 Station Blackout This safety issue was previously included in the NRC Program for the Resolution of Generic Issues as Generic Task B-57, but has recently been elevated to Category A as Generic Task A-44 Because this action has only recently been taken, a Task Action Plan and schedule for this task have not yet been developed.
Electrical power for safety systems at nuclear power plants is supplied by two redundant and independent divisions. The systems used to remove decay heat to cool the reactor core following a reactor shutdown are in-cluded among the safety systems that must meet these requirements.
Each electrical division for safety systems includes an offsite alternating current (a.c.) power connection, a standby emergency diesel generator 3.c. cower supoly, and direct current (d.c.) sources.
Task A-44 involves a study of whether or not nuclear cower piants should be designed to accommodate a complete loss of all a.c. power, i.e., a loss of all a.c. for an extended period of time in pressurized water reactors accompanied by loss of the auxiliary feedwater pumps (usually one of two
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A redundant pumps is a steam turbine driven pumo tnat is not dependent on a.c. power for actuation or operation) could result in an inacility to cool the reactor core with potentially serious consequences.
If the auxiliary feedwater ? umps are dependent on a.c. power to function, then a loss of all a.c. power for an extended period could of itself result in an inability to cool the reactor core.
Current NRC safety requirements require as a minimum that diverse power drives be provided for the redundant auxiliary feedwater pumps. As noted above, this is normally accomplished by utilizing an a.c. powered electric motor driven pump and a redundant steam turbine criven pump.
Of primary concern is the design adequacy of plants licensed prior to adoption of the current requirements, although an initial survey of operating plants performed uy the NRC staff indicates that all operat-ing pressurized water reactors have either steam turbine driven or diesel engine driven auxiliary feedwater pumps (neither of which are dependent on a.c. power.
Further study regarding this issue will include determining if requirements beyond diverse power drives for the auxiliary feedwater pumos are needed.
Such requirements mignt include specific time requirements for which the plant must be capable of accommodating a station blackout.
4 As noted in Section 10.5 (page 10-3) of the Perkins SER, the Perkins facility is specifically designed with redundant emergency feedwater
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ea e
systems which include steam driven pumps as one of the diverse subsystems to provide emergency feedwater in the event of a loss of onsite and offsite power.
The results of Task A-44 are expected to confirm the present staff view I
that a loss of all a.c. power does not have a signi ficant safety impact on plants such as Perkins. Accordingly, our conclusions in the Perkins SER regarding the issuance of construction permits are not excected to be affected by the results of this generic task.
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