ML19254H032
| ML19254H032 | |
| Person / Time | |
|---|---|
| Issue date: | 02/22/1980 |
| From: | Mattson R NRC - NRC THREE MILE ISLAND TASK FORCE |
| To: | Plesset M Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML19239A281 | List: |
| References | |
| FOIA-81-405 NUDOCS 8005281118 | |
| Download: ML19254H032 (3) | |
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- Et:0RANDUM FOR: Milton Plesset, Chairman Adviscry Cazittcc on Reactor Safejutras FRG :
Roger J. Hattson Ttti-2 Acticn Plan Steering 6e eup E
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IINCLUSION OF ACRS RECOMMELDATIOf;S IN THE NRC ACTION PLANS
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As part of the development of ' draft 2 of the TMI Action Plan (f;UREG-0660) '
we carefully reviewec the extent to which the reco,,nendations provided oy the ACRS in letters since March 28, 1979 were incluaed. To facilitate the reading of draf t 2 by the Cc;nittea members, we have prepared a cross.
index showing hwere the various ACRS recomendations are addressed in the Action Plan. The cross index is provided in the enclosed taole. The _
table is arranged by subject area, with the pertinent ACRS letters and (
primary iction Plan sections referenced accordingly.. You have already "1-been provided copies of draft 2 of tiUREG-0660.
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. Also, be~ staff is scheduled"to brief the' Comittee'ohlMarch 6 on~ thd,
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%. C matters raised'in your'lette'r of' January 15,1980 ~to~ C'hairman Ahearne '.g ~ <
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in which the Comittee forwarded its initial coments on~ tiUREG-0660.
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By March 5, we will provide an overall sumary and chapter sumaries of L,
draft 3 to NUREG-0660 to facilitate that briefing.
In addition, we will,
provide a detailed breakout of the action item priorities. These priorities were provided in draft 2 of NUREG-0660, consistent with the Cocnittee's advice in the January 15 letter. The detailed breakout;of priorities will be updated to draft 3 of the plan and'will be cross w.
s referenced to the Comittee's recomendations, the Comission's approved list of near term operating license (NT0L) licensing requirements, and an industry prioritization of portions of the plan which we expect to Y_
receive on January 25.
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,C on 492-7517.
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Roger J. Mattson, Director TliI-2 Action Plan Steering Group
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Enclosure:
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TABLE I Cross Reference of ACRS Recommendations with NUREG-0660 SUBJECT ACRS LETTER DATES NUREG-0660 SECTIONS Analysis of LOCA's and Anamolous 4/7/79 4/18/79, 5/16/79 II.E.2 Transients 8/14/7$
Additional Instrumentation and 4/7/79, 4/18/79, 5/16/79, II.F Increased Use of Existing 8/13/79 Instrumentation Natural Circulation Analysis, 4/18/79, 5/16/79 1.C.1, I.G., II.E.3 Experimentation and Procedures Reliability of Power Sources 4/18/79, 5/16/79 II.E.3, II.G Monitoring EFS Status 4/18/79, 5/16/79 I.D.3 Operator Training & Qualification 5/16/79 I.A.2 Evaluation of LER's 5/16/79 I.E Operating Procedures 5/16/79,8/14/79,12/13/79 I.C Emeraency Planning and NRC Role 5/16/79 III Decontamination & Recovery 5/16/79 II.H.3 Staff Capability, Review 5/16/79,12/13/79,12/17/79 IV Procedures, Rules and Regulations, Regulatory Organization and Goals Single Failure Criteria 5/16/79 Implicit in II.C Controlled Filtered Venting 5/16/79, 12/13/79 II.B.9 Limiting Conditions of Operation 8/13/79 I.B.1 Systems Interation; IREP; Safety 8/14/79, 10/12/79, 12/13/79 II.C.1 Implications of Shared Systems; 12/17/79 Common Mode Failures; Reliability Environmental Qualifications 8/14/79 II.F.3 RHR Design Basis, Dedicated SHRS 8/14/79,12/13/79 II.E.3 Direct Safety Signal 8/14/79
' Implicit in II.E.4 Analyses of Steam Generator 8/14/79 None Overfilling, and Equivalent BWR Transient
_ SUBJECT ACRS LETTER DATES N1JREG-0660 SECTIONS 0.
Noble Gas Releases 10/9/79 III.D.1 1.
Reevaluate SEP 10/11/79 IV.D.3 2.
Licensee Technical Capability 12/13/79 I.B.1 3.
Control Roon Design Requirements; 12/13/79, 12/17/79 I.D Safety Classifications and Qualifications 4.
Containment Inerting; Densely 12/13/79 II.B Populated Sites; Accidents Beyond DBA 5.
Effect of a Release on A Second Unit 12/13/79 Implicit in II.B.8 6.
Station Blackout Analysis 5/16/79 II.C 7.
JointReviewofNSSS.B0Pand 8/14/79,12/17/79 I.C; I.F, II.C.1 utility ; QA of Entire Industry 8.
Seismic Implications of TMI 12/13/79 None 9.
Quantitative Safety Goals 05/16/79 None 30.
Shift Technical Advisor 08/13/79,12/13/79 I.A.1 31.
Role of the ACRS 12/10/79 IV.A 32.
Low Power Testing 12/11/79 I.G 33.
Siting 12/17/79 II.A
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SUBJECT:
DRAFT NURS3-0660, " ACTION PIANS BJR IMPLEMENTING RECCMMENDATIONS OF 'IEE PRESIDENT'S C04 MISSION AND OIEER S'IUDIES OF ME WI-2 ACCIDENr"
Dear Dr. Ahearne:
During its 237th meeting, January 10-12, 1980, the Advisory Committee on Reactor Safeguards reviewed Draf t NUREI3-0660, dated December 10, 1979.
W e draft had previously been discussed at an ACRS Sibcommittee meeting in Washington, D.C., on January 7, 1980.
During its review, the Committee had the benefit of discussions with the NRC Staff.
We draft is a compilation of recommendations made by the several organi-zations and commissions that have investigated the MI-2 accident. Se Com-mittee understands that a primary purpose of the document is to establish criteria for termination of the pause in licensing.
Other purposes are to provide a complete action plan relating to all the unresolved issues and un-implemented recommendations from the lessons learned from the mI-2 accident, and to establish priorities and requirements of funds and manpower. We draft gives preliminary target dates and estimates of the necessary resources, but does not yet reccrnmend priorities.
W e Committee believes the Plan is comprehensive, but not selective; this com-
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prehensiveness serves to dilute the items important to safety, and therefore important to termination of the licensing pause.
In the absence of priorities and identification of the items that the NRC Staff considers important, the ACRS finds it difficult to make objective comments on the Plan. % e Committee understands that the Staff is proceeding to develop priorities and identifi-cation of items of primary importance, and the Committee will expect to review the important aspects of the Plan when this has been done.
De Committee is also concerned that preoccupation with the Plan may lead to neglect of pre-MI-2 accident safety concerns, some of which are of lorg stand-ing and of greater importance than some of the listed items.
It is important to establish priorities on an overall consideration of both "old" and "new" itens.
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Honorable John F. Ahearne January 15, 1980 The Plan lists a large number of proposed charges in plant equipment, plant staffing, operating procedures, and licensing requirements.
'Ihe ACRS believes that the scheduled time for establishing a complete plan setting detailed re-quirements for all items is too short to give reasonable assurance that all changes will be in the direction of greater safety.
In illustration of this concern, the Committee points to the controversy that arose over the directive prohibiting tripping of the reactor coolant pumps following high pressure injection initiation.
The Committee _ believes that a two s_tep_ process is more app,ropriate in develop _ing, r
the Action PlaD. OriWexpedit'ed~ basis, the Staff should develop those recom-mendations for safety improvement that it oelieves can and should be adopted as requirements for a terminaticn in the pause in licensing.
Ch a longer but de-fined time schedule, the Staff should develop a plan for dealing with other issues and implications of the N I-2 accident.
Additional connents by member H. Lewis are presented below.
Sincerely, Milton S. Plesset Chairman Additional Comments by Member H. Lewis The letter of January 5,1980 from L. V. Gossick, Executive Director for Oper-ations, to the Commissioners describes the Action Plan as the complete list of all actions necessary as a result of the accident at 'IMI-2, and states that complete approval of the Plan, in its entirety, by the Commission, should be regarded as a prerequisite for the resumption of licensing. % e Staff has further told us that, though they plan to assign priority scores to the items on the list (through a scoring system of dubious relevance), it is expected that all items on the list will be accomplished, in time.
It is my view that such an unselective approach to the lessons of mI-2 is inappropriate, and that the Plan consists of an uncritical listing of anything anyone has suggested be done in the aftermath of (not necessarily as a result of) the accident at MI-2.
In particular, the Plan provides no guidance, and reflects no analysis, with respect to the safety relevance of the items, or even whether they would enhance safety.
I believe adoption of the Plan would make no demonstrated contribution to a reordering of NRC priorities toward those safety weaknesses highlighted in the various reports on mI-2.
It would be preferable to bite the bullet, and identify those twenty items that need attention, in terms of their impact on safety, as determined by any reason-able analysis.
'Ihis has not been done, nor is it contemplated.
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j December 13, 1979 Honorable John F. Ahearne Chairman U.S. Nuclear Regulatory Comission Washington, DC 20555
SUBJECT:
REPORT ON 'IMI-2 LESSONS LEARNED TASK FORCE FINAL REPORT
Dear Dr. Ahearne:
The 'IMI-2 Lessons Learned Task Force has issued its Final Report, NUREG-0585.
The ACRS provides comments herein both on the specific recommendations made by the Task Force and on related subjects.
The Committee will first address the recommendations made in NUREG-0585.
1.
Personnel Qualifications and Training.
The ACRS gives general support to the recomendations made in this category.
The ACRS believes that, although a broader technical background should be required of Shift Supervisors, it may be neither necessary nor practical to require that all Shift Supervisors have a Bachelor of Science Degree.
The Comittee recommends that the NRC define its criteria for " equivalent training and experience in engineering or the related physical sciences."
The ACRS believes that a training program tailored to the requirements of reactor operation, possibly of less than four years duration, may pro-vide a practical alternative to a formal degree program. The Committee believes that the NRC should define the scope and duration of a training program that may be considered as an acceptable alternative to a degree curriculum. The ACRS also recommends that, if the Technical Advisor system proves satisfactory, consideration should be given to offering licensees the option of retaining that system instead of upgrading the academic education of Shift Supervisors to the specified level.
The ACRS recomends that the adequacy of staffing in the NRC Operator Licensing Branch be reevaluated with respect to the number of personnel and breadth of their background.
The Committee believes that additional emphasis must be given to the determination of what constitutes an adequate degree of in-house tech-nical capability for each licensee and assurance of the continuing de-velopnent of such capabilities. The ACRS also believes that attention must be given to providing, on a continuing basis, technical backup to review safety-related design changes or to provide assistance under
Honorable John F. Ahearne December 13, 1979 accident conditions by a group having the depth of technical knowledge which exists in the organization of the nuclear steam system supplier and a well-qualified architect-engineer during the period while the plant is being designed.
2.
Staffing of Control Room.
The ACRS supports this recommendation.
3.
Working Hours.
The ACRS supp3rts this recommendation.
4.
Emergency Procedures.
W e ACRS, in general, gives strong support to this recommendation. How-ever, the Comittee believes that the emergency procedures at licensed power reactors should receive priority. h e ACRS recommends that the licensees should give priority to the developaent of improved emergency procedures with the aid of expert, interdisciplinary review groups and that the NRC Staff should review, in depth, the existing and proposed, emergency procedures for a large sample of licensed reactors on a priority basis.
The knowledge developed from the concurrent industry and NRC efforts should be used to revise, in a timely fashion, the emergency procedures of all operating plants.
5.
Verification of Correct Performance of Operating Activities.
The ACRS gives general support to this recommendation.
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6.
Evaluation of Operating Experience.
The ACRS gives general support to these recommendations.
Additional Committee comments on this subject are contained in NURD3-0572,
" Review of Licensee Event Rep 3rts (1976-1978)."
7.
Man-Machine Interface.
The ACRS gives general support to these recommendations.
In addition to the nine items listed in NUREG-0585, Appendix A, Section 7.1, the Committee recommends that We licensee should include in his evaluation the data recording requirements and recall capabilities of the minimum set of plant parameters that defines the safety status of a nuclear power plant.
Honorable John F. Ahearne December 13, 1979 8.
Reliability Assessments of Final Designs.
W e ACRS strongly supports the application of reliability assessments to final designs. The Committee supports the Integrated Reliability Evaluation Program (IREP) which is being initiated by the office of Nuclear Regulatory Research.
However, the Committee does not agree that the proposed IREP will fully satisfy the need. W e ACRS recom-mends that the NRC develop a program in which licensees acting indi-vidually or jointly develop reliability assessments of their plants, in addition to the NRC IREP, which should be performed concurrently.
If the reliability assessments were performed in the manner proposed above, it would accelerate obtaining potentially significant safety information and expedite the development of the basis for changes, should they be necessary.
It would also provide the operating organi-zations with better technical insight into the safety of their plants and would provide the benefits to be derived by separate studies of system reliability.
9.
Review of Safety Classifications and Qualifications.
W e ACRS supports this recommendation. A particular problem warranting early attention is the qualification of operator information systems.
More generally, the Committee believes that more than a year will be needed to accomplish the overall task, partly because of its breadth and depth, and partly because of the very considerable number of know-ledgeable personnel which would be needed.
The Committee agrees that completion of the overall task should not be made a condition for the licensing of new plants.
10.
Design Features for Core-Damage and Core-Melt Accidents.
W e ACRS supports this recommendation. However, the Committee believes that the recommendation should be augmented to require concurrent de-sign studies by each licensee of possible hydrogen control and filtered venting systems which have the potential for mitigation of accidents involving large scale core damage or core melting, including an esti-mate of the cost, the pssible schedule, and the potential for reduction in risk.
The ACRS agrees with the recommendation made by the Lessons Learned Task Force in NUREG-0578 that the Mark I and Mark II BWR containments should be inerted while further studies are made of other possible con-tainment modifications in accordance with the general recommendations in this category. W e ACRS also recommends that special attention be given to making a timely decision on possible interim measures for ice-condenser containments.
Honorable John F. Ahearne December 13, 1979 We Committee also recommends that special. attention be given to oper-ating reactors located at densely populated sites.
11.
Safety Goal for Reactor Regulation.
The ACRS supports this recommendation.
12.
Staff Review Objectives.
The ACRS supports this recommendation. However, the ACRS believes that there is a need for review of NRC safety rules, regulations, guides and philosophy on a regular basis in order to ascertain various matters including the following:
a.
Does an appropriate balance exist in the expenditure of NRC financial and manpower resources artong the various research areas, on the resolution of safety isses, on the legal requirements of licensing, and on inspection and enforcement?
b.
Is there an appropriate division of effort and responsibility between industry and the NRC?
c.
Has an undesirable inflexibility in the approach to safety developed due to previous decisions, or for other reasons?
d.
Are there any important gaps in the existing safety review process? Is there a mcchanism for searching out such gaps?
13.
NRR Emergency Response Team.
The ACRS gives general support to these recommendations. We Committee believes that the timing of implementation should be more flexible. The Committee believes that better definition of the NRC role and responsi-bilities in an emergency will have an influence on the determination of the makeup, training and abilities of an NRC emergency response team.
The ACRS wishes to make some comments and recommendations on several matters not directly addressed in NUREG-0578 or NUREX3-0585.
1.
W e ACRS believes that the lessons learned from the mI accident should be viewed in a broader perspective. The Committee agrees that the mI accident shows a need for considerable improvement
Honorable John F. Ahearne December 13, 1979 in reactor operations and in knowledge of the behavior of a plant during a wide range of transients.
However, the Committee believes that there are other potentially important contributors to the probability of a reactor accident, and they should also receive priority attention.
Reliability assessments and systems interactions studies, as discussed under recommendations 8 and 9 above, should serve this function in part.
However, there is a need also to consider, in some more systematic way, methods to uncover significant desiga errors, to detect system or com-ponent degradation, and to test systems under conditions more closely simulating the range of situations which might result from transients and accidents.
2.
The Task Force has not addressed the need to reexamine the adequacy of the current design basis for emergency cooling recirculating systems, as recommended by the ACRS in its report of August 14, 1979 on " Studies to Improve Reactor Safety."
There are several other specific recommendations made by the ACRS in its interim reports Nos. 2 and 3 on Three Mile Island both dated May 16, 1979 and in its report of August 14, 1979 on studies to improve reactor safety. W e Committee believes that the NRC Staff should address each such recor:mendation in formulating its overall action plan.
3.
W e ACRS recommends that a reevaluation should be made of the potential influence of a serious accident involving significant atmospheric release of radioactive materials from one unit of a multiple unit site on the ability to maintain the other units in a safe shutdown condition.
4.
W e ACRS recommends that the industry and the NRC Staff undertake studies to ascertain what contingency design measures, beyond those covered in the Task Force recommendations, may ensure improved capabilities for recovering from or mitigating the effects of accidents beyond the oesign basis.
For example, in some cases, it may be possible to provide alternative measures in the event of loss of the safety grade ultimate heat sink for an extended period of time.
5 We ACRS recommends that the NRC Staff give attention to the seismic im-plications of MI, for example, the seismic qualifications of auxiliary feedwater supplies, the acceptability of failure of nonseismic Class 1 equipnent, and the suitability of emergency procedures for earthquakes.
6.
The ACRS recommends that greater corisideration be given to the provision of dedicated shutdown heat removal sytems, and to the potential nerits of having a shutdown heat removal system capable of operating at normal system pressure.
Honorable John F. Ahearne December 13, 1979 The ACRS expects to address other considerations of reactor safety and the regulatory process in a separate report.
Sincerely, Max W. Carbon Chairman h
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'9 es August 14, 1979 Honorable Joseph M. Hendrie Chairman U. S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
STUDIES TO IMPROVE REACTOR SAFETY
Dear Dr. Hendrie:
The Advisory Committee on Reactor Safeguards has'several recommendations to make concerning studies to improve reactor safety, l.
Accident Analysis The ACRS recommends that an analysis be undertaken of Festulated accidents involving a steam line rupture followed by a small break in the primary system, arising from an open relief valve, a steam generator tube break or some other opening. 'Ihe analysis should not only review the capability of engineered safety features to cope with such an event but also examine the symptoms available to the operator and determine the adequacy of existing operating procedures.
2.
Studies to Reduce the Probability of an Accident The ACRS recommends a systematic reevaluation of the common-mode failure potential of compressed air systems used for control or service in both safety and non-safety applications. Amorg the matters to be considered in such a review should be the effect of moisture and corrosion products, and a total loss of air supply. Also of concern is any interconnection of compressed air supplies to both safety and non-safety devices and to other fluid systems. Consideration should be given to the adequacy of separation rules for air systems.
3.
Studies to Reduce the Probability of an Accident
'.he ACRS recommends that studies be made of the interrelationship between the operation and interconnection of the auxiliary feedwater system, the main feedwater system, the atmospheric dump and the control system, in-board of the isolation valves on the main feedwater and main steam lines in order to ascertain whether there are significant undesirable interactions under various postulated accident scenarios.
Honorable Joseph M. Hendrie August 14, 1979 The ACRS also recommends examination of feedwater flow during postulated ruptures of the main steam line in order to assure that adequate, but not excessive, heat removal capability is retained.
4.
Studies to Reduce the Probability of an Accident The ACRS recommends that studies be made of possible significant effects adverse to safety arising from shared systems or locations in existing 3
multiple-reactor stations, of the probability of such adverse effects, and of possible mitigating features.
Examples of potentially adverse effects L;,g to be considered include the possible loss of access to systems needed for one unit due to unexpected releases of radioactivity from a neighbor, and l'
the ptential overloading of emergency diesel capability arising from con-current IDCA signals (real or spurious) for a system using a shared diesel.
5.
Operatina Procedures he ACRS recommends that a study be made on how operating procedures should best be written.
For example, should procedures be characterized in tenns of events or in terms of symptoms, or both? Should the operator actions be keyed to changes in symptoms? Is the priority of operator action fixed so as to optimize public safety? How does one determine whether a given pro-cedure is understood by the operator, and that the operator will carry it out properly?
The ACRS also recommends that a systematic examination be made of steps that the operator should be advised not to take, when and why.
One example of such a step could arise in connection with the operator's capability to iso-late certain LOCAs in BW'Rs and some PW"As.
It is passible that such isolation following a relatively large loss of original primary system inventory could lead to reactor rapressurization and an inability of the available high pres-sure capacity to keep the core adequately covered for a period long enotx3h to cause significant core damage before the situation was recognized and remedied.
6.
Environmental Qualification of Systems in Containment he ACRS recommends a review and reevaluation of the current basis for judging
- 7 envirormental qualification requirements for equipnent in containment and in other buildings where a hostile envirorment might result.
W e same review should be made of the locations of vital sensors and other measurement devices.
The pros and cons of modified envirormental qualification and equipuent loca-tion should be examined with due consideration given to the difficulties of modifying existing equipnent.
Honorable Joseph M. Hendrie August 14, 1979 7.
Design, Construction, and Ooeration Review The ACRS reccrnmends.that consideration be given to the need for joint review by the nuclear steam system supplier, the architect-engineer, and the operat-ing utility, prior to operation of a reactor, to consider, among other things, g
the adequacy of interfaces and other features developed under the aegis of multiple suppliers, the acceptability of technical specifications and other safety-related operational limits, and the adequacy of operational and acci-dent procedures.
8.
Decay Heat Removal Systems in ECCS The ACR$ recommends a reevaluation of the design basis of the low pressure recirculation heat removal system of the ECCS, including system capability y
and lorg-term reliability when circulating highly radioactive fluid contain-ing particulates which might jeopardize certain components.
9.
Direct Rather than Derived Safety Signals The ACRS recommends that a review be made of the passible improvement in actua-tion reliability that could be achieved by employing a safety signal directly b
related to the matter of concern; e.g., radiation level for containment iso-lation. Derived signals have merit but they may be tied to specific scenarios for a transient or accident and hence subject to failure if some unanticipated course of events were to transpire.
10.
Systems Interact.ons Involving Air, Instrument, or Hydraulic Lines The ACRS recommends that each licensee be requested to review and evaluate his
...e as-built plant for possible significant systems interactions wherein rupture
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..s in a medium or high pressure line could cause loss of vitally important air,
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'P(, 'a instrument, and hydraulic lines and electrical circuits and equipment.
Fo r boiling water reactors, attention should be given particularly to the iines S
related to actuation of the scram system.
- 11. Accident and Transient Analyses
$ C The ACRS recommends that further analyses be made of the course, consequence,
\\ p' and probability of transients which would lead to gross overfillc.g of the secondary side of the steam generator in PdRs and the equivalent event in ks0
.y EWRs in order to ascertain whetner any additional measures are appropriate
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to provide additional protection of the public health and safety.
- 12. Studies to Imorove Safety The ACRS has previously recommended on several occasions that the NRC Staff hy utilize the methodology of probabilistic analysis to examine the reliability
Honorable Joseph M. Hendrie August 14, 1979 of the design of systems imprtant to safety, both for existing reactors and as it might apply to re. actors to be constructed.
Ebr example, on July 11, 1978 the ACRS, in a letter from R. F. Fraley to L. V. Gossick, recommended that the NRC Staff provide an evaluation of the reliability of the auxiliary feedwater systems of current PdRs and PARS.n terms of various transients and incidents.
Recently, following the tree M.ile Island Accident, the NRC Staff performed a short-term, intensive review of the auxiliary feedwater system for Combustion Engineering and Westinghouse operating PdRs and found many items of interest, including some which suggested a need for early regulatory action to remedy deficiencies. We ACRS recommends that this same procedure be applied, as expeditiously as practical, to each of the other, systems of importance to safety in order to ascertain whether there are other features on operating reactors warranting early or near-term improvements.
Sincerely, Max W. Carbon Chairman e
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