ML100260591

Proposed Pages 247,251-253 & 259 to Tech Specs Re Mgt Titles

ML100260591
Person / Time
Site:	Indian Point, FitzPatrick
Issue date:	09/28/1979
From:	Power Authority of the State of New York
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ML100260592	List: ML100260591 ML19305A735 ML19305A742
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NUDOCS 8002130117
Download: ML100260591 (124)

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ATTACHMENT I PROPOSED TECHNICAL SPECIFICATIONS CHANGES RELATED TO MANAGEMENT TITLE CHANGES (APPENDIX A)

POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 SEPTEMBER 28, 1979

_Po o 2/3c/

6.0 60 __ADMINISTRAT IVE CONTROLS Administrative Controls are the means by which plant opera tions are subject to management control. Measures specified in this section provide for the assignment of responsibilities, plant organization, staffing qualifications and related re quirements, review and audit mechanisms, procedural controls and reporting requirements. Each of these measures are neces sary to ensure safe and efficient facility operation.

6.1 RESPONSIBILITY The Resident Manager is responsible for safe operation of the plant. During periods when the Resident Manager is un available, the Superintendent of Power will assume his respon sibilities. in the event both are unavailable, the Resident Manager may delegate this responsibility to other qualified supervisory personnel. The Resident Manager reports directly to the Executive V. P. and Director of Power Operations for administrative matters and functionally to the Manager Nuclear Operations for operational related matters, as shown in Fig. 6.1-1.

6.2 PLANT 'STAFF ORGANIZATION The plant staff organization is shown graphically in Fig.

6.2-1 and functions as follows:

1. A licensed senior reactor operator shall be on site at all times when there is fuel in the reactor.

2. In addition to item 1 above, a licensed reactor operator shall be in the control room at all times when there is fuel in the reactor.

3. In addition to items 1 & 2 above, a licensed reactor operator shall be readily available on site whenever the reactor is in other than cold condition.

4. Two licensed reactor operators shall be in the con trol room during start-ups and scheduled shutdowns.

5. A licensed senior reactor operator shall be respon sible for all movement of new and irradiated fuel within the site boundary. A licensed reactor operator will be required to manipulate or directly supervise the manipulation of the controls of all fuel moving equipment, except the reactor building crane. All fuel movements by the reactor building crane, except new fuel movements from receipt through dry storage, shall be under the direct supervision of a licensed react~or operator. All fuel movements within the core shall be directly monitored by a member of the reactor analyst group.

Amendment No.24 247

It is recognized that expertise of the SRC members col lectively may not encompass all of the areas listed in ANSI 18.7-1972. Therefore, special consultants to provide expert advice may be utilized when the nature of a particular problem dictates. The Chairman shall be appointed by the President and Chief Operating Officer. The Vice Chairman shall be appointed by the Chairman.

(B) Alternates Alternates shall be appointed in writing by the Chairman.

No more than two alternates shall participate in SRC activities at any one time.

(C) Meeting Frequency Meetings of the SRC will be called as the occasions for review arise. Meetings will be held at least once per six months.

(D) Quorum Chairman or Vice Chairman and two members, including designated alternates, shall consitute a quorum.

(E) Responsibilities

1. Review proposed changes and/or modifications to procedures, equipment or systems which may involve an unreviewed safety question as defined in 10 CFR 50.59 as identified to SRC by the Resident Manager or the headquarters' Technical Staff.

2. Review proposed tests and experiments which involve an unreviewed safety question as defined in 10 CFR 50.59 as identified to SRC by the Resident Manager or the headquarters'Technical Staff.

3. Review the safety evaluations for changes and/or modifications to procedures, equipment or systems completed under the provisions of 10 CFR 50.59 to verify that such actions did not constitute an unreviewed safety question.

4. Review the safety evaluations for test and/or experiments conducted under the provisions of 10 CFR 50.59 to verify that such actions did not constitute an unreviewed safety question.

Amendment No. 251

5. Review proposed changes in the operating License and Technical Specifications.

6. Make or cause to be made periodic audits of plant operation to verify conformance with the facility operating license and other regulatory requirements.

7. Review reports and minutes of PORC.

8. Review violations of applicable statutes, codes, regulations, orders., Technical Specifications, license requirements or of internal procedures or instructions having nuclear safety significance.

9. Review NRC inspection reports, reporatable occurrence submittals, and related correspondence.

10. Review aspects of plant design and operation which may result in an unacceptable environmental effect.

11. Review the Facility Fire Protection Program and implementing procedures at least once per two years.

(F) Audits The SRC shall provide an independent review and audit function of safety-related aspects of plant activities which shall encompass:

1. The conformance of facility operation to all pro visions contained within the Technical Specifications and applicable license requirements.

2. The performance of the entire facility staff relative to nuclear safety.

3. The results of all actions taken to correct anomalies occurring in the facility, equipment, structures, systems or method of operations.

4. The adequacy of the Quality Assurance Program to meet the criteria specified in 10 CFR 50, Appendix B.

5. The Emergency Plan and implementing procedures.

6. The Security Plan and implementing procedures.

7. Any other area of facility operation considered appropriate by the SRC or the President.

(G) Authority and The Safety Review Committee shall be advisory to the President Chief Operating Officer.

Amendment No. 252

(H) Records Records will be maintained in accordance with ANSI 18.7-1972 and in accordance with the SRC Charter.

(I) Charter Conduct of the committee will be in accordance with a charter, approved by the President and Chief Operating Officer setting forth the mechanism for implementation of the committee's responsibilities and authority.

REPORTABLE OCCURRENCE ACTION (A) In the event of a Reportable Occurrence, the NRC shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.

(B) Each Reportable Occurrence requiring 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />' notification to the NRC shall be reviewed timely by the PORC and a report submitted by the Resident Manager to the Manager-Nuclear Operations and the SRC.

SAFETY LIMIT VIOLATION (A) If a safety limit is exceeded, the reactor shall be shut down and reactor operation shall only be resumed in accordance with the provisions of 10 CFR 50.36 (c) (1) (i).

.(B) An immediate report of each safety limit violation shall be made to the NRC by the Resident Manager. The Manager-Nuclear Operations and Chairman of the SRC will be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(C) The PORC shall prepare a complete investigative report of each safety limit violation and include appropriate analysis and evaluations of: (l) applicable circumstances preceding the occurrences, (2) effects of the occurrence upon facility components, system or structures and (3) corrective action required to prevent recurrence. The Resident Manager shall forward this report to the Manager-Nuclear Operations, Chairman of the SRC and the NRC.

PROCEDU.ES (A) Written procedures and administrative policies shall be established, implemented and maintained that meet or exceed the require ments and recommendations of Section 5 "Facility Administrative Policies and Procedures" of ANSI 18.7-1972 and Appendix A of Regulatory Guide 1.33, November 1972. In addition, procedures shall be established implemented and maintained for the Fire Protection Program.

(B) Those procedures affecting nuclear safety shall be reviewed by PORC and approved by the Resident Manager prior to implementation.

(C) Temporary changes to nuclear safety related procedures may be made provided:

1. The intent of the original procedure is not altered.

Amendment No. 253

PRESIDENT &

CHIEF OPERATING OFFICER m

-- I 0

EXECUTIVE V.P.

& CIIIEF ENGINEER

,& DIRECTOR OF EXECUTIVE POWER VP, s OPERATIONS DIRECTOR OF SAFETY/SECURITY Ul MANAGERI

10) NUCLEAR OPERATIONS, RESIDENT

• RESPONSIBLE FOR POWER AUTHORITY MANAGER DIRECTION OF FIRE PROTECTION PROGRAM FIGURE 6.1-1

- -ADMINISTRATIVE POWER AUTHORITY OF THE STATE OF NEW YORK FUNCTIONAL JAMES A. FITZPATRICK NUCLEAR POWER PLANT MANAGEMENT ORGANIZATION CHART

ATTACHMENT II PROPOSED ENVIRONMENTAL TECHNICAL SPECIFICATION CHANGES MANAGEMENT TITLE CHANGES (APPENDIX B)

POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 SEPTEMBER 28 , 1979

> PRESIDENT & CHIEF OPERATING OFFICER o SAFETY REVIEW EXECUTIVEVICEPRESIDEN EXECUTIVE VICE PRESIDENT CHIE ENGNEER& DIRECTOR OF POWER OPERATIONS DIRECTOR OF 3z ENVIRONMENTAL PROGRAMS STAFF FIGURE .2-2 POWER AUTHORITY OF THE STATE OF NEW YORK MANAGEMENT ORGANIZATION ENV IRONMENTAL

ATTACHMENT III SAFETY EVALUATION RELATED TO MANAGEMENT TITLE CHANGES POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 SEPTEMBER 28, 1979

Section I- Descition of the Modification The position of General Manager and Chief Engineer has been retitled President and Chief Operating Officer. The position of Assistant General Manager-Engineering has been retitled Executive Vice President and Chief Engineer.

The position of Director of Power Operations has been changed to Executive Vice President and Director of Power Operations.

The Resident Managers will now report administratively to the Executive Vice President and Director of Power Operations.

The functional relationship between the Resident Manager and the Manager-Nuclear Operations remains unchanged.

Section II - Purpose of Modification The purpose of this modification is to comply with the require ment of 10 CFR 50.59(c) to report all changes to the operating license.

Section III - Impact of the Change These changes in management titles will not alter the conclusion reached in the FSAR and SER Accident Analysis.

Section IV - Implementation of the Modification The modification as proposed will not impact the ALARA or Fire Protection Program at JAF.

Section V - Conclusions The incorporation of this modification: a) will not increase the probability nor the consequences of an accident or mal function of equipment important to safety as previously evalu ated in the Safety Analysis Report; b) will not increase the possibility for an accident or malfuncticn of a different type than any evaluated previously in the Safety Analysis Report; and c) will not reduce the margin of safety as defined in the basis for any Technical Specification.

Section VI - References (a) JAF FSAR (b) JAF SER

6 _0&k4

--- I -, 7 -, F1' a M 0 1980 RE~~ TERA ASchwenter Docket ft (50-295/304 L01shan DZiemanp 50-24 /9 6)

[leeves Tlppolito NRC PDR .-

Linsees RReld Local POR plus short 0ELD DOR Reading Docket Hos. ,10-29513 list - OI&E(3)- NRR Reading 1 andqe!!

7 ACRS(16) HRDenton CParrlsh EGCase I4EMORANDUN FOR:, Darrell . Eisenhut, Acting. DRWj nan RVollmer ODvisi on of Operating Reactors' BGrimes w1. W6anunll FROM; Gia yN. Zech,- Technical Assistant LShao Oivisfn of Operating  : - '. Reactors-. ... -.. J'l1ler DGEisenhut SUB3JECT:, SU A.?ARY OF MEETING HELD ON DECEMt ER 20, 1979, _ T.desco COMONWEALTH EDISON CO, (COECO), CONSOLIDATED EDISON,,

OCNPANY OF NEd YORK (CONED) AND POWER AUTHORITY OF THE STATE OF HER YORK' (PASNY) REGARDING THELZION STATION UWITS 1 AND 2 AND THE INDIAN. POINT' UN3L-TS 2 IVD 3 FACILITIES On December 20, 1979, we met tith representatives of CECo, Con"Ed and PASHY to further discuss the reviews of the Zion Station Units 1 and 2 and Indian Point Units 2 and 3 sites that were nfitiAlly addressed in our meeting Of'December 5, 1979. In that earlier meetlng- had adivisedthe licensees of a staff review, that had comenced of these sites to determine iwhat additional measures might be taken in terms of both intern aisuresahd desigi changes to further -reduce the probability of a severe reactor accident and to reduce the consequences of such an event should one occur. "'At that meeting the licensees agreed to return in about two weeks time to"discuss their proposed revieW to achieve the same objectives as the- staff'S' ei W. Attendance list is attached. -

Following the presentation the staff and licensees agreed that separate meetings should be held within about 1 rnth to initially discuss VariOus technical aspects of the reviews. Also, 'Tt was agreed that a final meeting would be held in early to mid February to di-scuss th tco0clusions 'eached by the licensees regarding possible system design'chenges at their faciflities The staff indicated that we would also be working withhd 1censees to oMpee certain actions to further reduce the probability ojf "a-seVere.reactor accident." These actions xeve discussed in our December 5, 1'979-meetngind ould include improved interim operational items and completion of current licensing actions on a priority basis.

0riginal Signed By.

Gary G. e2-Odbi, Technical Assistant Division of Operating Reactors'

Enclosure:

List of Attendees

~LLS2Z~1/ ,kl OFFICE ... .. . ...

SURNAM................................. ..

NRCFOR M,

318 (9-76) NRCM 0240- 1U.S. GOVERNMENT PRI NTING OFFICE: 1979-289-369'

UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555 JANJUARY,2318

'Docket Nos. 50-295/304 and 50-286/247 MEMORANDUM FOR: Darrell G. Eisenhut, Acting Director Division of Operating Reactors FROM: Gary G. Zech, Technical Assistant Division of Operating Reactors

SUMMARY

OF MEETING HELD-ON DFCEMBER 20, .1979, WITH COMMONWEALTH EDISON CO. (CECo)., CONSOLIDATED EDISON COMPANY.OF NEW YORK (CONED) ANDPOWER AUTHORITY OF THE'STATE OF NEW YORK (PASNY).REGARDING.THE ZION STATION UNITS 1AND 2 AND THE INDIAN POINT,UNITS 2 AND 3 FACILITIES On December 20, 1979, we met with representatives of CECo, Con Ed further discuss the reviews of the Zion Station Units land 2 and and-PASNY-to Units 2 and 3 sites that were initially addressed in our meeting Indian. Point of December 5, 1979. In that earlier meeting we had advised the licensees of a-staff that had commenced of these sites to determine.what additional measures review be taken in.-terms of both interim measures and design changes to'further might the probability of a severe reactor accident and to reduce the consequences reduce of such an event should one occur. :At that meeting the licensees return in about two weeks time to discuss their proposed review agreed to to same objectives as the staff's review. Attendance list is attached.achieve the Following the presentation the staff and licensees agreed that separate should be held within about 1 month to initially discuss various meetings aspects of the reviews. Also, it was agreed that a final technical meeting would be held in early to mid February to discuss the conclusions reached regarding possible system design changes at their facilities.by the licensees The staff indicated that we would also'be working with the licensees to complete certain actions to further reduce the probability of a severe reactor accident.

discussed in our December 5, 1979 meeting and would include These-actions were improved interim operational items and completion of current ],ensing actions on a priority basis.

Gary (2--ech, Technical-Assistant Division of Operating Reactors

Enclosure:

List of Attendees

. NCLOSURE LIST OF ATTENDEES MEETING WITH ZION. & INDIAN POINT DECEMBER 20, 1979 NRC Power Authority of the State of New. York G. Zech P. W. Lyon L. Olshan P. J. Early E. Reeves J. R. Schmiler R. Denise J. F. Davis P. Collins H. Sayed A. Schwencer S. -Zulla E. Adensam N. P. Mathur J. E. Kohler, Region III R. L..Goyette F. Kantor W. A. Josiger L. Soffer M. Shaughnessy T. Speis P. Bayne S. Acharya A . Martin R. DiSalvo J. Lamberski R. Sherry M. Silberberg Commonwealth Edison Co.

C. Kelber R. Young C. Reed J. Zudans W. Wogsland J. Meyer W. F. Naughton R. Licciardo G. T. Klopp R. J. Goddard W. F. Pasedag Westinghouse T. Rebelowski D. L. Basdekas D. Walker R. Slember Consolidated Edison D. F. Paddleford T. M. Rudein J. Makepeace W. E. Kortieu D. M. Speyer C. Belston M. J. Scott G. Toto W. Cahill B. Bennett Sandia M. L. Lee R. P. Remshaw B. Varnado J. P. Davis A. S. Benjamin Bechtel VE&C N. B'. Willowhby R. F. Duerr R. W. Barton Sheldone, Harmon & Weiss Philadelphia Electric E. R. Weiss V..S. Boyer I

0 i Indian Point/Zion ear Site Studies Presentation to NRC December 20, 1979 Commonwealth Edison Company Consolidated Edison Company Power Authority. of the State of New York

0 Indian Poi.nt/Zion Near Site Studies PRESENTATION TO NRC December 20, 1979 On December 5', 1979, the NRC identified a potential concern for the effects of avery-low probability core melt event and possible subsequent radiation release to populations near nuclear s-ites. Zion and Indian Point were selected as the two sites for immediate s.tudy because of their proximity to areas of higher population density.

Commonwealth Edison, Consolidated Edison, and.The Power Authority of the State of New York have joined forces to review this potential concern. This presentation will address the'evaluations conducted to-date and plans for future work in two areas:

1) Mitigation of the effects of core melt

.2) Reduction of the probability of a severe accident.

1n addition, we wil.1 define a plan of action and a schedule which will lead to a thorough evaluation of the risk- of core melt and methods to reduce this risk. Finally, we will address considerations involved in plant design and operation which continue to, assure publ: health and safet'- while detailed studies' are in Progress and- the ev. luated results are being implemented.

t s, in our opinion, extremely important to recognize hat t "s S e', eview is consisent ith the ACRS, NRC, and S

sees ece on standino hiloso :y of providi-ng added safety margin

0

-l2-- . .

for the Zion and Tndian Po int sites. Such philo.s ophy, evolved dur ing the licensing process and resulted in features such as: containment liner weld channel -pressurization systems; containment penetration pressurization systems; and the isolation valve seal water systems.

We have in place and underway a program which we feel is fully responsive to the points you raised in our last meeting.

(Screen 1 & 2, Slide 1).. This program.contains studi-es and evaluations of the means to mitigate the effects: of an unlikely severe accident, by reducing radiation release and increasing the available evacuation time. It also includes evaluations of potential means to further reduce the already low probability of a core melt. A preliminary task sequence and schedule has been develop..ed which is resoonsive to the 60 day timetable established by the NRC in our last meeting.

Utilizing the resources of the three utilities and Westinghouse, we have already performed scoping studies to better define our current, substantial capabilities in the areas of core melt prevention and mitigation This work includes a review of existing plant design, a review of our status relative to interim concerns, and preliminary probabilistic risk assessments.

Mi t idaint n the Effects of Core Melt The first area, mitlgation of the effects of the very cr e Ie, is centered around the deveopment. of means to ncre~s Ce a ln n- eph defer es so as to reduce or delay 3e MSs ,f radizacivity from SUh a core mel-t I

3 The task ..of developing alternative. ways to mitigate the effects of a severe accident and to properly evaluate these means would normally be a lengthy and very-complex engineering effort.

However, we appreciate the need to be responsive to the timetable set forth in our last meeting. We are very fortunate in that a great deal of work in this area has already been performed. We have found that we can draw on work performed by Westinghouse, Off-Shore Power Systems, Clinch River; FFTF, Sandia, and others-and tailor much of it to our pllant specific considerations.

The-overal.l program for developing mitigating features (Slide 2, Screen 1), involves four (4) basic steps:

1) defining the engineering problem

2) developing alternatives

3) evaluation

4) decision Definino the Task The first steo is to define the task in realistic engineering terms, i.e., time to core melt, mechanism for radiation release, amount of radiation released, etc. To do this, we need to develop design conditions based on selected severe transient seluences from a plant specific oreliminary probabilistic risk assessmen: utilizing WASH- 4O0 methodology (a mini-WASH-1400review

. h is.task, the accident sequence and component failure

. ase evejooed in WASH-1400 xnd in follow-on, studies on a 4-tlto es. imate failure probabili ties

-4 for those accident sequences which are the dominant risk contributors for core melt accidents. Estimates have been derived for the plants under consideration by comparing the Zion and Indian Point systems and components to the systems and components in the reference plants already evaluated in detail. This comparison has been-performed for all of the dominant accident sequences.

Engineeringjudgment has been used to modify the reference study so that it reasonably models the actual systems design of Zion and 7ndian .Point.

As part of this task, estimates of fission product release have been made for each of the dominant accident sequences and associated containment failure modes. For the mini-assessment, differences between fission product removal processes for the reference 'plants and Zion or Indian Point have been examined. Where differences, are not significant, releases for a sequence have been assigned to the same release category employed for the reference plants.. Where differences in fission product removal capabilities are significant for a particular sequence and associated failure mode, engineering judgment has been used to assign the sequence to the appropriate release category. For a PWR, seven (7) release categories were utilized in WASH-1400. They ranged in severity from

he large atmosheric release associated with an early containment fai.u-e to a small atmospheric release associated with melt through

- t e i nm ent base mat.

-:ns .ini-assessment t.k -has been performed to provide R r J i i ct~ n of which accider sequence contributes most to core

-5 melt accident risk.. *The prelimi.nary severe acci.dent, risk spectrum for eachplant will be utilized to assess the relative potential for risk reduction offered by various system modifications or by new containment features or systems.

A detailed quantitative WASH-1400 type evaluation of the Zion and rndian Point plants is planned as part of the longer term follow-on studies. This detailed study will define risk estimates for the dominant accident sequences and indicate whether sequences with substantial contributions to risk have been omitted in the mini study. rt is expected that risk values derived in the preliminary assessment 'will be within a factor of 2 of those derived from the quantitative study. In view of the detailed, studies conducted by NRC for two o'ther Westinghouse PWR.'.s (a 3-loop PWR with a high pressure containment and a 4-loop PWR with an ice containment), the likelihood of missing an accident sequence.which is a significant contributor to risk is regarded as small. Thus the mini-study is regarded as an adequate basis for ear.ly decision processes.

Develooment of Alternatives The sequences which are major contributors to risk of a core melt accident will be utilized to identify one or more scenarios as design bases for the new features. Included in this orocess ,vwill be a definition of the accident transient associated.

,V h t-the tes n boases. The transient evaluation will lead directly t .1 on of one or more sets of design parameters to be 1 te *eeoomen

.sed t and evalu tion of new tii.iatinoaeat.ures.

-6 Our development and evaluation process will include those features commonly associated with the mitigation of the effects of a severe accident leading to core melt. In addition, we will have the benefit of plant specific engineering input and of our preliminary probabilistic risk analysis to ensure that other features of possible merit are considered. At this time, it is appropriate to consider the engineering factors involved in this effort (Slide 2, Screen 2). Our review will include preliminary or scoping design layouts for features or, at least, a close enough examination to determine that a feature should or should not be given further consideration. A number of factors will be considered in this work. These include determinations that the features:

- address the design goal (i.e., the reduction of risk to the population)

-can be added to the plant employ a sound technological base

-. do not degrade other safety aspects of the plant

- can be implemented on a reasonable timetable, and are most effective compared to alternatives being considered.

Some of the features that will be considered are:

- Core ladles Controlled, filtered containment venting

- Hydrogen control measures, and

-iuzmented containment cool ng measures.

-7 Core Ladle The core ladle has historically been associated with core melt postulations and-has a substantial background of development. This device, is constructed by lining the bottom and side walls of the cavity beneath the reactor vessel with a layer of refractory material such as magnesium oxide. The purpose of the ladle is to retain the debris resulting from core melt for a period of time (a few days) while appropriate actions are taken to reduce the potential consequences of such a severe accident. In concept, the refractory is a sacrificial material; that is, the refractory material i.s not cooled.

Incorporation of a-core ladle was required by NRC for the Floating Nuclear Plant for the purpose of delaying releases to liquid pathways so that liquid-pathways interdiction could be initiated. It is not at all clear that a ladle provides similar benefits with respect to reduction of radioactivity release and dose consequences via air pathways following a core melt accident-. This aspect of incorporating a core ladle into the Indian Point or Zion design will be evaluated as part of the task.

One r.elated concept that will also be evaluated is coo'k-ing either the existing base mat material or a new e'r ctory mater.ial following core melt. The purpose of co:,lIn, vou1d be to inc.'ease the time which the core coul d

.be retained in the region below the vessel without melt through. Permanent retention while difficult to ensure, might be possible as part of this evaluation, both active and passive cooling: schemes will be investigated.

Containment Venting Controlled, filtered venting of the containment is a promising feature that has received attention recently. Such venting serves the purpose of relieving containment pressure through radioactivity. removal systems at a controlled rate. The venting thereby further reduces the already small potential for overpressure failure of the containment and subsequent uncontrolled release of radioactive material We intend to look very closely at such systems with a variety of treatment systems including sand filters and scrubbers. System capaci.ties will be' conservatively established and the effectiveness of the systems i.n addressing the dominant containment failure modes will be given careful consideration. Considerat-ion.

may be given to elevated stack releases to augment these systems should the results of evaluations show that worthwhile improvements can be obtained.

The venting systems themselves will be evaluated

'1i:h particular attention to ensuring that no degradation f er pl nt safety asp4cts occurs. Interactions with

ystems wi .be vry. carefuly considered. The

-9 venting systems will be considered with respect to those mechanisms leading to containment overpressure which are determined to be dominant risk contributors for these plants.

Hydrogen Control One possible contributor in this regard is hydrogen evolution and combustion. A number of activities and alternatives for consideration have been planned to complete our evaluation of hydrogen control. (Slide 3, Screen 2). These are all considerations well beyond the hydrogen control required by current design basis accidents.

The first action will be to establish the need, if any, for specific hydrogen control features. Loads from postulated severe accident consequences will be applied to the containment to determine its response.

Computer codes will be used to evaluate the limiting conditions the containment can withstand. These loads will be compared with those predicted in other parts of the program and an overall picture of capability will be developed. Should the load picture and hydrogen failure node warrant, added features will be considered. These inc ude:

100

- 10

1. Containment Venting 2 Multiple Ignitions Sources

3. Recombination

4. Containment Inerting

5. Removal of Containment 02 on Demand (Aqueous Chemical Spray)

Augmented Containment Cooling The possibility of augmenting existing containment cooling capability with systems of a diverse nature and/or power source will also be given some consideration. At first glance, this alternative appears least likely to be successful given space limitations, known technology, and other considerations. We, will, however, give-serious consideration to any concepts which appear to have merit in this regard.

Ev.aluation of Relative Effectiveness of Feasible Alternatives The next step in our program, given a reasonable selection of competing: alternatives, is to evaluate the effectiveness of these alternatives. To accomplish this, we plan to evaluate the relative reduction in dose to the-population for each alternative and to comoare these values to the values for the plants without any r .itga i n features.

Iiis -. ork will be done using the CRAC code. As part of z"e . I u ion ev.iua

-aH- the CR V, code was developed. for estimating

,i ,e n-d r-isks (the= product .)f ~probabili.ty and consequences)

~- 11 for releases of radioactivity to air pathways following core melt accidents. Tne form of the code utilized for WASH-1400 performs these estimates for generic sites. As part of this task, specific demographic and meteorological data for the Indian Point and Zion sites will be incorporated into the code.

Population distribution information for Zion based on 1971 statistics, including projections out to 1985, have been collected for use with evacuation plan and meteorological data in establishing baseline risk values. Meteorological data for Zion through 1975 has also been collected. Reviews are underway to establish the need for and feasibility of updating this information.

The most recent docketed meteorological information for the indian. Point site is found in the Site Appendix I Evaluation Report submitted to NRC on March 14, 1977. This report utilizes "average annual meteorological data" which was developed from actual meteorological data collected during 1974 and 1975.

The most recent demographic information is still that information 'ncluded in the Indian Point 3 FSAR. It should be noted that even the most recently submitted Con Edison and PASNY revised Emergency Plans, which were submitted to NRC in November, 1979, utilized the demographic data in the Indian Point 3 FSAR. This democraphic data is based on the 1970 population census and was projected to the year 20i.0.

Th e modified version of CRAC will then be employed for coe emt ac,- d en t consequence esimates for these soecific sites.

0

- 12 The code will first be utilized for baseline risk estimates for the plants as now designed. Risk calculations will also be performed for the plants, with the proposed plant and system modifications, to estimate the magnitude of risk reduction available from the proposed modi fi cati ons.

Parameters reflecting current evacuation plans for each of the two sites will also be incorporated into the CRAC code and calculations will be performed to estimate dose reduction from existing evacuation procedures to add to the perspective of the study.

It might be in order to touch briefly on the current status of the individua.l evacuation plans.

Commonwealth Edison has and will continue to work with the State of illinois.to develop, evacuation plans which will receive NRC concurrence. An outside consulting firm has also been commissioned to independently develop a detailed study of times required to evacuate, meeting NRC requirements, for all of our nuclear plants.

The fir st plant to be studied will be Zion. Such information as is available-.in the 60 day period will be factored into our study.

The most recent Indian Point 2 Emergency Plans were submitted to NRC in November, 1979, and were based on the new guidelines and. criteria issued by NRC. These emergency plans are revised periodically to aahere to new criteria and regulations.

At the present time, Con Edison representatives are working with state and county officials to assist them in fulfilling their responsibilities to formulate local emergency plans in accordance with NRC guidelines. The county emergency plans will include planning for evacuation, should such a step be determined advisable by local or state governmental agencies.

The Power Authority of the State of New York has a similar program underway. In December, 1979, a joint effort:was established to complete the evacuation analysis. This joint effort consisted of the Consoldiated Edison Company, The Power Authority of the State of New York, the New York State Nuclear Civil Protection Planning Section and the consulting firm of Parsons, Brinckerhoff, Quade &

Douglas, Inc.. The objective of this joint effort was to concentrate resources to ensure the completion of the evacuation analysis to the requirements of the NRC in their November 29, 1979, letter by January 31., 1980.

Dec isi ons Once the relative effectiveness of feasible alternatives has been determined, decisions can be made to implement those features which would significantly reduce risk for the two sites.

Given the constraints of time, we feel this program is as thorough and resoonsive as possible. (Slide 1, Screen I).

Reduct-ion *f -obability of Core Melt The next major por: i.on f the program is to evaluate 1- tio ds 'hn ch may reduce, the pro ibility of severe accidents I

including core melt. Probabilistic risk assessments will be accomplished in two phases. In Phase A, WASH-1400 event sequences will be investigated in detail for Indian Point and Zion.. The goal will be to identify the more likely sequences and those sequences leading to a rapid core melt. Phase 8 will expand the Phase A effort and result in comprehensive, plant specific, event trees and.

fault trees, which will be quantitatively evaluated.

The Phase A effort will be accomplished within the near term by two teams; one team for Zion and one team for Indian Point.

Each team will consist of experts in probabilistic risk assessment and experts on plant systems and operations.

These teams will construct event trees specific to each plant. They will utilize relevant, previous work. This will include WASH-1400, work at the Electric Power Research Institute (EPRI), event trees for Oiablo Canyon seismic analysis, and other recent work.

These models will utilize the best available data to det ermine for each plant the "important" sequences. Reliability data sources w-ill include WASH-1400, EPRI,-and recent NRC assessments. Plant specific experience will be used where aporoori ate.

This effort will allow us to identify maj.or contributors to risk and to investigate the feasibility of plant modi.fications that lfqit yie~d major reductions in risk. We will also employ this wora.s a check on the results of the mini-WASH-MOO review conducte as . pirt of our evalua:ion of alternatives which would

,a' e a core melt

Phase B of this work will be an effort to make the Phase A models more complete and to allow further evaluation of any design changes identified in Phase A. Phase A work will be modified with additional fault trees and, where necessary, more complete versions of-Phase A event/fault trees. The data will be! structured for evaluation by appropriate computer codes. The data utilized during this phase will be of the same sources as Phase A, but more time will be available to further consider plant experiences. During this phase, the baseline models (plant "as is" models) will be modified for any design changes which.result from the design efforts paralleling Phase A. This revised model will. be used to evaluate the improvement (i.e., reduction) in risk due to. the design changes.

An additional benefit to be gained from Phase"B will be utility versions of the Integrated Reliability Evaluation Program (rREP) models for each specific plant. Since the IREP for these plants will most likely be in progress during the same time' frame, the utility effort and the NRC effort can be used as a basis for interchange which would assure comprehensive models of the specific p ants. (Slide 4, Screen 2).

Plan of Action and Task Sequence A plan of acti-on and first cut at a task sequence are shown on th is viewgrap h. We are serious about meeting the program 7 1.,, recuested. We are already hard at work on these tasks.

n didton to resources at The .wer Authority of the State of New York, _on soi aia:ed Edison, and C-;nmonwealth Edison, we have retained

Westinghouse and Argonne National Laboratory. We are considering other consultants in specialty areas. In the capacity of architect engineers, United Engineers & Constructors and Bechtel are both assisting Consolidated Edison and The Power Authority of the State of New York, and Sargent & Lundy is assisting Commonwealth Edison.

Our plan of action for mitigation of the effects of core melt includes a detailed review of core ladles, filtered containment vents, hydrogen control, and augmented containment cooling. The analytical process for conducting this review has been described in detail earlier in this presentation.

Our plan of action for reducing the probability of a significant accident includes a determination of major, contributors to severe transient probability. This task, described earlier as Phase A of our program, is underway. Once major contributors are defined, feasible modifications will be evaluated for their effectiveness in reducing severe transient probability.

The key point, for both the mitigation and probability reduction task sequences, is that we are scheduling completion for the end of the 60 day period requested by the NRC.

Resoonse to Interim Concerns The last major area we wish to address is the plant status relative to interim concerns. (Slide 3, Screen 1).

As Mr. Eisenhut pointed out in our meeting on December 5, 1979, both we and the NRC recognized the existence of above average population densi ies near the Zil)n and [.ndian Point sites during, the

early design and licensing stages of the jobs. Over 11 years ago, during the construction permit stages, we, the NRC and the ACRS had extensive discussions regarding this matter.

Out of these discussions and our own internal assessments, a common philosophy emerged relative to both sites. That philosophy called for extraordinary design measures to reduce the risk to the public and extraordinary measures during all phases of plant operation to that same end. Hardware features are built into these plants to meet this philosophy that are not found on contemporary plants or even newer plants. Provisions were made for the incorporation of, even more features should extensive, on-going research programs show a need for such features. Plant operating and training measures were instituted which went beyond then current practice and the .continuing evolution of our practices in this area have kept pace with. or gone beyond current regulatory standards and industr-y norms. It is worth noting that many of the practices now considered routine by the industry and NRC trace their origins to work done at the Zion and Indian Point sites. These include such areas as quality assurance and single failure criteria.

We would like to present some specifics regarding these plant features which are still,. after all these years, as a total package, unique to these plants. (Slide 5, Screen 2).

A number of extra features are found on all four units.

Thiese 3e

• 18

- 18

1. Containment weld channel and weld channel pressurization system: All containment liner welds are enclosed by continuous linear channels welded to the liner to form a redundant seal at the joints of liner plates. Those channels which cover joints not buried in concrete are pressurized with air to a pressure exceeding calculated containment peak pressure. This eliminates leakage at liner pl-ate joints.

2. Penetration pressurization system: Inaddition to the normal pressurization of electrical penetrations (with dry nitrogen), mechanical penetrations are pressurized with air to a pressure above calculated containment peak pressure. This eliminates leakage through penetration assemblies.

3. Isol-ation valve seal water system: Those double isolation valves, normally closed-on a containment isolation signal, in water and small air systems have the area between valves filled (if needed) and maintained in a filled condition at a pressure exceeding calculated containment design pressure by this system. This eliminates any leakage of containment atmosohere via- an open (or ruptured) ine through the redundant isolation valves.

- 19

4. Extra containment fan cooler capacity: Each containment has 5 fan cooler units, 3 of which are required for post accident containment cooling. The added capacity provides assurance of system availability.

5. Post LOCA hydrogen control: Each unit has both recombiner and post-LOCA containment purge capability. The recombiner capability was added to provide added conservatism.

6. Third auxiliary feedwater pump: Each unit has 3 auxiliary feedwater pumps per unit. Two of these are 100% capacity motor driven pumps and the third is a 200% capacity steam turbine driven pump. All three pumps are intertied through lines and valves designed for. an active or passive failure. This extra capacity over a 2-100% capacity pump configuration provides added assurance of system availability.

7. Added containment radioactivity removal has been provided. On Zion a third, 100% capacity, diesel driven, containment spray pump is installed for each unit. This added conservatism over a conventional, 2 pump per unit, configuration gives added assurance of system availability. On ndian Point, each fan cool.er unit is equipped with HEPA an harcoal filters for post-accident particulate an.J iodine removal.

- 20

8. Confirmatory "S" signals: Confirmatory Emergency Safeguards Features (ESF) actuation signals are sent to power operated valves which are not required to change position. This ensures that, if a valve had inadvertently been placed in an incorrect position, it would move to the correct position upon ESF actuationt. This has been applied to critical safety system valves.

In addition, each unit has extra margin in service water and component cooling water capacity and availability. They have augmented auxiliary building air filtration systems and closed valve leak off systems to reduce offsite exposure due to leakage. They have redundant electrical heat tracing on vital borated systems.

Conclusion In summary, a great deal of margin already exists for these units and has existed from the day they started operating.

As noted earlier, the original design and licensing would philosophy of these two sites called for extra measures which assure the safety of the nearby populace. In addition to hardware and training, features built into these plants and plant operating ne. sures have been instituted which go beyond minimum requirements.

As valuable as this margin is, we have not ignored the c~nst't e'o.TJ~'in of the industry at large-nor have we ignored

-e ent his rt3. (5lide 6, Scree,- 2).

- 21 -

Both the Zion and. Indian Point Plants have, within the last two years, made substantial improvement.s in their safety margins. A number of these improvements are common to all three utilities. These are presented today. Other major improvements and factors contributing to-safety margins exist. Each utility will be pleased to discuss these individually with you either after this presentation or at any other time convenient to you.

The first of the common actions is the implementation of the NUREG-0578 recommendations including, specifically, the shift technical advisor.

The second action includes the existence and use of plant specific simulators for training. These simulators are at or very near each site.

The third action consists of the use of either 5 or 6 shift rotation to reduce operator fatigue and promote training.

The fourth major action is the early incorporation of the new NRC operator qualification requirements into the training programs. This has been accomplished at Zion and is being implemented on the-Indian Point units.

Other factors and actions also point to added margin.

These include actions taken some time ago to enhance control room nan-machine interfaces, constantly improving operating records, and high levels of management commitment and involvement in plant ooer at ions.

n concusion, (Slides cff) we feel we have presented you w.i:h !n .2a ressive and responsive program to address the areas you

presented in our December 5th meeting. We are already actively into this work. We are drawing heavily on expertise from all quarters of the industry. The full resources of each utility, Westinghouse Electric Corporation including, Off Shore Power Systems are active in this effort. In addition, we expect to employ the expertise of national laboratories, outside consultants, and our architect-engineers in this work. We invite close coordination with your staff.

We have also presented some of the factors which contribute to the very substantial and very real extra safety margins enjoyed by these plants even without the further actions planned.. We are convinced that our current and continued operations

.embody more than adequate extra margin at these sites.

We will be happy to expand on any of these remarks and to answer any questions you may have.

CECO, PASNY, CON ED HAVE INITIATED RESPONSIVE PROGRAM TO EXAMINE AND IMPLEMENT FEATURES WHICH ARE EFFECTIVE IN MITIGATING THE CONSEQUENCES OF SEVERE ACCIDENTS STUDIES AND EVALUATION

- MITIGATION

- REDUCTION OF PROBABILITY

- PLAN/SCHEDULE SCOPING EVALUATIONS INDICATE SUBSTANTIAL CURRENT CAPABILITIES TO PRECLUDE AND MITIGATE. SEVERE ACCIDENTS PLANT STATUS RELATED TO INTERIM CONCERNS THIS ACTIVITY IS CONSISTENT WITH LONG STANDING UTILITY PHILOSOPHY TO PROVIDE AN EXTRA MEASURE OF SAFETY MARGIN FOR THESE SITES Slide I Screen 2

PROGRAM IS COMPREHENSIVE AND RESPONSIVE TO SEVERE ACCIDENT ISSUES TOTAL PROGRAM Slide I Screen I

MITIGATING STUDIES PROVIDE A BASIS FOR DESIGN, EVALUATION, AND EFFECTIVENESS ASSESSMENT OF FEATURES TO REDUCE THE EFFECTS OF SEVERE ACCIDENTS SELECT CORE MELT SCENARIO(S)

FROM WASH-1400 MINI-REVIEW Slide 2 Screen

ENGINEERING FACTORS INVOLVED IN DEVELOPMENT OF MITIGATING FEATURES:

FEATURE ADDRESSES DESIGN GOAL

- FEATURE CAN BE-ADDED TO PLANT (Physical constraints)

- FEATURE EMPLOYS SOUND TECHNOLOGICAL BASE

- FEATURE DOES NOT DEGRADE OTHER, SAFETY ASPECTS OF PLANT

- FEATURE CAN BE IMPLEMENTED ON REASONABLE-TIMETABLE FEATURE IS MOST EFFECTIVE COMPARED TO ALTERNATIVES BEING CONSIDERED Slide 2 Screen 2

ENGINEERING FACTORS INVOLVED IN DEVELOPMENT OF MITIGATING FEATURES:

FEATURE ADDRESSES DESIGN GOAL

- FEATURE CAN BE ADDED TO PLANT (Physical constraints)

- FEATURE EMPLOYS SOUND TECHNOLOGICAL BASE

- FEATURE DOES NOT DEGRADE OTHER SAFETY ASPECTS OF PLANT

- FEATURE CAN BE IMPLEMENTED ON REASONABLE TIMETABLE

- FEATURE IS MOST EFFECTIVE COMPARED TO ALTERNATIVES BEING CONSIDERED Slide 2 Screen 2

HYDROGEN CONTROL CONSIDERATIONS

- CONTAINMENT STRUCTURAL CAPABILITY

- CONTAINMENT VENTING

- MULTIPLE IGNITION SOURCES

- RECOMBINERS

- CONTAINMENT INERTING OXYGEN REMOVAL Slide 3 Screen 2

Plan of Action and Task Sequence Developed to Meet 60 Day NRC Requirement Mi tigation Probability Reduction Action Points Action Points

1. *NRC Meeting/Initiate 1. *NRC Meeting/Initiate Study (12/5/79) Study (12/5/79)

2. *WASH-!400 mini-review 2. *Investigate and Design Sequence Consulta.nts Selection 3 *Collect Metro and Demographic Data

4. *NRC Meeting 3. *NRC Meeting

5. Containment 4. Select Consultant Transient Devel opment

6. Develop Feasible 5. Perform P'hase A MItigating Features Program

7. Evaluate Feasible 6. Evaluate Feasible Mitigating Features Modifications

8. **NRC Meeting 7. **NRC Meeting

• Item complete

-- To be completed by 2/4/30 Slide. 4.

Screen 2

INDIAN POINT AND ZION PLANTS INCLUDE ADDITIONAL SAFETY MARGINS

- ADDITIONAL FEATURES IN ORIGINAL PLANT DESIGN

- RECENT ACTIONS Slide 3 Screen 1

ZION AND INDIAN POINT PLANTS INCLUDE ADDITIONAL FEATURES TO PROVIDE SAFETY MARGIN EXAMPLES:

CONTAINMENT INTEGRITY

- WELD CHANNEL & PRESSURIZATION

- ISOLATION VALVE SEAL WATER

- EXTRA FAN COOLER CAPACITY

- HYDROGEN RECOMBINERS HEAT REMOVAL FEATURES

- AN EXTRA CAPACITY FEEDWATER PUMP

- 3RD DIVERSE PUMP RADIOACTIVITY REMOVAL

- ADDITIONAL DIVERSE SPRAY PUMP OR CHARCOAL FILTERS CONFIRMATORY SAFETY SYSTEM VALVE ACTUATION SIGNALS Slide 5 Screen 2

RECENT ACTIONS AT INDIAN POINT AND ZION PLANTS PROVIDE ADDITIONAL SAFETY MARGIN SHIFT TECHNICAL ADVISOR TRAINING / QUALIFICATIONS

- PLANT SPECIFIC SIMULATORS

- 5 OR 6 SHIFT ROTATION

- NEW NRC REQUALIFICATION STANDARDS MET OR BEING IMPLEMENTED OTHER CONSTANTLY IMPROVING OPERATING RECORD

- ENHANCED MAN-MACHINE INTERFACE CONTROL ROOM

- HIGH LEVEL OF MANAGEMENT COMMITMENT AND INVOLVEMENT Slide 5 Screen 2

REGUOLATORYZKETFILE COPY

.... JANUARY 1 0 1980 Docket Nos. 50-295/ ..

and 50-247 286jI.

MEMORANDUM FOR: Darrell G. Elsenhut, Acting Director Division of Operating Reactors FROM: Gary G. Zech, Technical Assistant Division of-Operating Reactors

SUMMARY

OFMEETING HELD ON DECEMBER 5, 1979, WITH COMMONWEALTH EDISON COMPANY (COg'ro), CONSOLIDATED EDISON. COMPANY OF NEW YORK (CONED)- AND POWER AUTHORITY OF THE STATE OF NEW YORK.(PASNY) REGARDING THE ZION.;STATION UNITS I AND 2 AND THE INDIAN POINT UNITS 2'AND3 FACILITIES On December 5, 1979, we met with ireipre seniit--atlv-e5 of CECo, ConEd and PASKY to advise them of a staff review that was being considered on the Zion Station Units 1 and 2 and Indian Point Unis 2 and 3 sites. Reference was made to our letters of November-29, 1979,, Which provided prelimlnary information regarding this review. Agenda and attendance list areattached.

As-a matter of introduction, the staff-stated that it recognizes that, .inthe event of a severe reactor accident, evacuation could present special problems should the accident occur at a factility located near a high population area and that a greater population risk'existed in.areas of high population density even with improved capability tO takeprotectlve actions. In anticipation of such an occurrence, special facility 'modifications may be appropriate-that would be inaddition to the general upgrading of the emergency preparedness capabilities of licensee and state and local plans. 'This upgrading of emergency preparedness iscurrently inprogress for a11 plants.'

The questions that exist, therefore,%are what can be done-to reduce the likelihood of a severe reactor accident and-what car te 'done to reducethe consequences of such an accident at facilities locatedin areas of high population densities.

Reference was made to Commiuss~ioners statements that shutdown of these facilities was not a precluded outcome..

The staff indicated that Zion Station Units I and 2 and Indian Point Units 2 and 3 have been selected as two sites for an hinitial review to determine what additional measures and/ordesign changes might :be appropriate inView of the location of these sites inthe areas bf the highest populatfon-densities of all operating plants. Depending onlwhat isleairned-from the current -review, additional sites may be considered for simtlar reviols'in the future, 1e" o OFFIC.. .................. ..................... ....... ...... ................ . ... .........

S SURNAME ..... ............. ......... . . . ......

• D A TE . . . ..

-3 6 9

.NRC FORM 318 (9-76) NRCM 0240 *'u.S.,GOVERNMENI FPRINTING OFFICE: 1979-289 "

L NRC FORM 318 (9-76) NRCM 0240 - ~U.S.GOVERNMENT-'PRINTING OFFICE: 1979-289-369

9 keeting Sumary a 2 ~- -JAUARY 1 0 1980 In response to a question from CECo, the staff.indicated that additional time to evacuation (delay of releases) on the order of several hours would bekbjectiWi of any additional measures that might be considered for idplementation. -However, it was also roted that the.benefits to bd gained from any design changes considered, such as a core ladle or filtered, 'Vented containment required quantification following a review.of the various accident scenarios hypothesized.'

The staff emphasized that the Zion Station and Indian Point reactors are not considered to be any less safe than other-plants but Iin view of their locations, other measures should be considered tocomperisate for the higher total population risk.

The staff's review, as outlined in the enclosed agenda. will consist of three specific areas, A.ke., 1) Improved Interim Operations, 2) Current Performance and Licensing Actions and 3) Accident Hitigation. The first two a" intended to provide additioral assurance, in the short term, that a severe accident will not occur.: Included among these areas willbd resolution and implementation on a priority basis wherever possible" of the st~ff's positions on generic Issues that are outstanding on these factlifties. Th third lrea of review will include design changes that have the potential for reducing the consequences of -an accident by either delaying a release follbWing 'anaccident to provide additional time for evacuation or by reducing the 'amount of activity released to the environment.

A followup meeting was tentatively scheduled for December 20, 1979, at which time the.icensees were requested to discuss their proposed review of any long term measures that will be studied and aAy interiw measures that could be taken to further reduce the probability of a severe acildent. A parallel reiew by both licensees and the staff will then be 'performed. Following this review decisions will be made regarding- possible futur actions.

Original Signed Bp, .:

Gary S. Zech, Technical Assistant

• Division4 of Operating.Reactors

Enclosures:

DISTRIBUTION

1. Agenda Dockets (50-295/304 & -247/286)

2. List .of Attendees NRC PDR , J iller.

.. -Llshan

'Local PDR LShao EReeves DOR Reading ASchwencer Licensee plus NRR Reading 'DZiemann - cc short lI st HRDenton TIppol ito

-EGCase RReid DGEisenhut OELD RLTedesco O&E(3)

GGZech ACRS(16)

RVollmer" JRBuchanan BGrimes TERA WGammill CParrish

• . .iIIO"

-"OFFCE OF IC. E . I ... ..... ...... .... ........

SURNAME 0 -, lb

- "e ............... .

'DATEK / .. ./ D . ... . .. . . . . ... . . . . . . .

NRC FORM 318 (9-76) NRCM-0240 *U.S. GOVERNMENT PRINTING -OFFICE: 1979-289-369

UNITED STATES J NUCLEAR REGULATORY COMMISSION WASHINGTON., 0. C. 20555 JA.UARY 0 *TaO Docket Nos. 50-295/304 and 50-247/286 MEMORANDUM FOR: Darrell G. Eisenhut, Acting Director Division of Operating Reactors FROM: Gary G. Zech, Technical Assistant Division of Operating Reactors

SUMMARY

OF MEETING HELD ON DECEMBER 5, 1979, WITH.COMMONWEALTH EDISON COMPANY (CECo ), CONSOLIDATED EDISON COMPANY OF NEW YORK (CONED) AND POWER AUTHORITY OF THE STATE OF NEW YORK (PASNY) REGARDING THE ZION STATION UNITS 1 AND 2 AND THE INDIAN POINT UNITS 2 AND 3 FACILITIES On December 5, 1979, we met with representatives of CECo, ConEd and PASNY to advise them of a staff review that was being considered on the Zion Station Units 1 and 2 and Indian Point Units 2 and 3 sites. Reference was made to our letters of November 29, 1979, which provided preliminary information regarding this review. Agenda and attendance list are attached.

As a matter of introduction, the staff stated that it recognizes that, in the event of a severe reactor accident, evacuation could present special problems should the accident occur at a facility located near a high population area and that a greater population risk existed in areas of high population density even with improved capability to take protective actions. In anticipation of such an occurrence, special facility modifications may be appropriate that would be in addition to the general upgrading of the emergency preparedness capabilities of licensee and state and local plans. This upgrading of emergency preparedness is currently in progress for all plants.

The questions that exist, .therefore, are what can be done to reduce the likelihood of a severe reactor accident and what can be done to reduce:the consequences of such an accident at facilities located in areas of high population densities.

Reference was made to Commissioners'statements that shutdown of these facilities was not a precluded outcome.

The staff indicated that Zion Station Units 1 and 2 and Indian Point Units 2 and 3 have been selected as two sites for an initial review to determine what additional measures and/or design changes might be appropriate in view of the location of these sites in the areas of the highest population densities of all operating plants. Depending on what is learned from the current review, additional sites may be considered for similar reviews in the future.

Meeting Summary 2 , . A In response to a question from CECo, the staff indicated that additional time to evacuation (delay of releases) on the order of several hours would be an objection of any additional measures that might be considered for implementation. However, it was also noted that the benefits to be gained from any design changes considered, such as a core ladle or filtered, vented containment required quantification following a review of the various accident scenarios hypothesized.

The staff emphasized that the Zion Station and Indian.Point reactors are not considered to be any less safe than other plants but in view of their locations, other measures should be considered to compensate for the higher total population risk.

The staff's review, as outlined in the enclosed agenda, will consist of three specific areas, i.e., 1) Improved Interim Operations, 2) Current Performance and Licensing Actions and 3) Accident Mitigation. The first two are intended to provide additional assurance, in the short term, that a severe accident will not occur. Included among these areas will be resolution and implementation on a priority basis wherever possible, of the staff's positions on generic issues that are outstanding on these facilities. The third area of review-will include design changes that have the potential for reducing the consequences of an accident by either delaying a release following an accident to provide additional time for evacuation or by reducing the amount of activity released to the environment.

A followup meeting was tentatively scheduled for December 20, 1979, at which time the licensees were requested to discuss their proposed review of any long term measures that will be studied and any interim measures that could be taken to further reduce the probability of a severe accident. A parallel review by both licensees and the staff will then be performed. Following this review decisions will be made regarding possible future actions..

Gary G. Zch, Technical Assistant Division of Operating Reactors

Enclosures:

1. Agenda

2. List of Attendees

AGENDA CLOSURE 1 o PURPOSE To discuss possible measures to reduce the orobability of a severe reactor accident and to reduce the consequences of such an accident.

o INTRODUCTION o BACKGROUND o SPECIFIC AREAS OF REVIEW Improved Interim Operations

- TMI-2 Lessons Learned Actions

- Bulletins and Orders

- Operator Training/Qualifications Emergency Procedures

- Human Factors Engineering Considerations

- Integrated. Reliability Evaluation Program (IREP)

Current Performance and Licensing Action

- License Amendment Applications

- Generic Issues

- Unresolved Safety Issues

- Resident Inspector's Role

- Operating Experience Accident Mitigation EmergencyPreparedness Considerations Degraded Core and Core Melt Accident Scenarios Capability of Present Systems Site/Population Data Requirements Dose/Risk Evaluations With and Without System Design Changes

0 ENCLOSURE 2_

LIST OF ATTENDEES MEETING WITH ZION & INDIAN POINT DECEMBER 5, 1979 NRC Power Authority State of New York Eisenhut Bayne Zech J. Early Grimes M. Pratt Olshinski F. Davis Schwencer P. Mathur P. Speis L. Goyette K. Long M. Wilverding Collins Gammi 11 Westinghouse Olshan Soffer R. J. Lutz, Jr.

E. Kohler, Region III A. Reeves Limerick Ecology Action A. Murphy C. Glynn L. FUfour Kelber Silbergerg Illinois Attorney General Muller VanNiel D. Hansell Mann Burns New York State Energy Office Goddard W. McPeek T. DeBoer E. Martin Craig Sheldon, Harmon & Weiss DiSalvo Rebelowski E. R. Weiss Commonwealth Edison W. F. Naughton D. L. Peoples G. T. Klopp C. Reed Consolidated Edison J. Cahill, Jr.

D. O'Toole Makepeace A. Monti L.. Lee Brandenbury P. Davis P. Remshaw

MITIGATION OF SEVERE ACC-DNT EXAMINE AND IMPLEMENT FEATURES THAT MAY BE EFFECTIVE IN MITIGATING THE CONSEQUENCES OF MAJOR ACCIDENTS, INCLUDING CORE MELT THE OBJECTIVE. ISTO DELAY AND/OR PREVENT CONTAINMENT FAILURE FROM THE CONSEQUENCES OF SEVERE ACCIDENTS; THIS WILL PROVIDE ADDITIONAL TIME FOR IMPLEMENTING 'EFFECTIVE EMERGENCY PLANS AND ACTIONS AROUND AFFECTED PLANTS; AND IN GENERAL, REDUCE FURTHER THE RESIDUAL RISKS FROM DEGRADED CORE AND CORE MELT ACCIDENTS,

• ACCIDENTS INVOLVING CORE DEGRADATION, INCLUDING CORE MELT

0 MITIGATION OF SEVERE ACCIDENTS ACCIDENT CONSIDERATIONS POTENTIAL MITIGATIVE FEATURES/SYSTEMS STUDIES/EVALUATIONS NEEDED.

IMPLEMENTATION CONSIDERATIONS

0 ACC IDENT CONS IDEAT IONS CONSEQUENCES OF DEGRADED CORE AND CORE MELT ACCIDENTS TO BE CONSIDERED:

• PARTIAL/EXTENSIVE FUEL MELTING

• HYDROGEN

• CONSEQUENCES OF HYDROGEN BURNING/DETONATION

• HYDROGEN CONTROL; TO PREVENT AND/OR MITIGATE FOR UNDESIRED CONSEQUENCES/EFFECTS

• STEAM EXPLOSIONS e POTENTIAL FOR STEAM EXPLOSIONS* DURING THE EVOLUTION OF A MELTDOWN ACCIDENT

• POTEN TIAL MAGNITUDE

• CAPABILITY OF REACTOR VESSEL/REACTOR CONTAINMENT

• "SLOW" CONTAINMENT OVERPRESSURIZATION, I.E., FEW TO MANY HOURS

ACCIDENT CONSIDERATIIONS -2

. ACCIDENT SCENARIOS INTERACTION OF CONTAINMENT ENGINEERING FEATURES WITH ACCIDENT EVOLUTION, I.E., FAILURE OF SYSTEMS SUCH AS CONTAINMENT COOLERS AND SUBSEQUENT EVOLUTION OF SCENARIO OR POTENTIAL CONSEQUENCES FAILURES (E.G., MECHANICAL) THAT CAN RESULT, INA LOSS OF EFFECTIVE CONTAINMENT ISOLATION

• DISPERSAL OF LARGE AMOUNTS OF RADIOACTIVITY IN CONTAINMENT RELEASE OF RADIOACTIVITY FROM CONTAINMENT (PARTIAL/GROSS FAILURE), DURATION OF RELEASE

• ATMOSPHERIC

• GROUND

POTENTIAL MITIGATIVE FEATURES/SYSTEMS IMPROVED/DIVERSE CONTAINMENT HEAT REMOVAL

• CONTROLLED VENT FILTERED CONTAINMENT

• SACRIFICIAL MATERIAL BASEMAT

• A COMBINATION OF THE ABOVE TWO

• HYDROGEN CONTROL, E.G., BURNING, INERTING, PURGING

• HIGH STACK RELEASE FOR NOBLE GASES PROGRAMMED RELEASE

STUDIES/EVALUATIONS NEEDED CAPABILITY OF INDIAN POINT 2/3 - ZION 1/2 SYSTEMS (I.E., PRESSURE VESSEL/PRIMARY BOUNDARY, CONTAINMENT)

TO ACCOMMODATE/DELAY CONSEQUENCES (OR SOME OF THEM) OF DEGRADED CORE, CORE MELT ACCIDENTS ANALYZE/EVALUATE P-T HISTORY INTHE CONTAINMENT FOR A NUMBER OF KEY ACCIDENT SCENARIOS THIS "INTEGRAL" ANALYSIS SHOULD CONSIDER A BASEMAT MADE WITH EITHER CONCRETE (ALREADY INPLACE) OR SACRIFICIAL MATERIAL THIS EVALUATION ISNEEDED FOR THE DESIGN OF A VENTED-FILTERED SYSTEM (ASSUMING ONE ISNEEDED), AND THE SIGNALS TO ACTUATE IT.

RADIOLOGICAL SOURCE TERM ESCAPING CONTAINMENT WITH/WITHOUT FEATURES/SYSTEMS SUCH AS VENTED-FILTERED CONTAINMENT

• CONSEQUENCE EVALUATIONS (USING THE CRAC MODEL ADOPTED FOR SPECIFIC PLANT SITES) WITH/WITHOUT FEATURES/SYSTEMS SUCH AS VENTED-FILTERED CONTAINMENT

0 IMPLEMENTATION CONSIDERATIONS

• EFFECTIVENESS* OF FEATURES/SYSTEMS INREDUCING THE RESIDUAL RISKS OF DEGRADED CORE, CORE MELT ACCIDENTS

• CONSIDERATIONS IN IMPLEMENTING EFFECTIVE EIRGENCY PLANS AND ACTIONS AROUND SUBJECT PLANTS

• SCHEDULE

• STATUS OF TECHNOLOGY-(FILTERS)

RELIABILITY "NEGATIVE" SYSTEMS INTERACTION ASPECTS COST VS RISK REDUCTION -

.EGO.*TOR. DOC FI UNITED STATES

• NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555 JANUARY 7 19.

Docket Nos.: 50"3

• LICENSEES:- Consolidated Edison Company of New York Power Authority of the State,of:New York (PASNY)

"FACILITIES: Indian Point, UnitNoS. 1, 2 and 3

SUMMARY

OF A MEETING ON DECEMBER 18, 1979-TO-DISCUSS THE INDIAN POINT SITE. EMERGENCY PREPAREDNESS A meeting to.discuss:.emergency.preparedness was held on December 18, 1979 near-the Indian Point site..',Attendees included the NRC, the. licensees, state and local officials, and members of the public. See Attachment.l.

In-the morning,. the acceptance criteria were reviewed for thestate and:

-local.officials,and the public. (These criteria had been previously dis cussed.with the licensees during-the September,25 and 26, 1979- meeting.

-See meeting summary dated October 18, 1979.) A New York State official stated. that it might not be possible tosatisfy our requirement to have the capability of notifying everyone withina 10-mile radius of the. plant within .15, minutes of receiving notiffication from the facility.operator'.

" TheNRC felt that- this requirement could be met and,.-accordingly, asked the

-State to, re-examine- this issue.

Inthe afternoon, our review of the licensees' revised plans was discussed.

Attachments 2 and 3.were distributed and formed the bases for.the discussion.

The licensees agreed to resubmit, their plans within two months,.revised to reflect our concerns..

At 7:15,p.m.,-public comments were invited and continued until the meeting

.was-adjourned at-12:45 a.m. on December 19, 1979., Most' of the comments.

asked for'the shutdown of the Indian Point facil.ities. .Written comments received were examined and are attached asAttachment 4.

1 *.

6 ..

1.. i~, I .,:, : -i;, i * , ...

4C..

2 JANUARY

'- 7 198Q' Two specific.public comments of particular value were. received. One relating to provisions for the handicapped'and one relating to radio comnunications

for the Ossining Police Department. These will be followed up: on during.

.subsequent staff review of the.emergency plans.

Olshan, Project Manager

• Operating Reactors Branch #1 Division of Operating Reactors:

Office of Nuclear Reactor Regulation Attachments:

As stated

Meeting Summary for Indian Point:.1', 2 & 3 Docket Files White. Plains Public Library NRC PDR .100 Martine Avenue Local PbR White Plains, New York 10601' ORbI Reading NRR Reading Joseph D. Block, ,Esquire H., Denton ExecutivelVice President E. Case Admi nistrative:

D. Eisenhut ConsolidatedlEdison Company R. Tedesco of New York, Inc.

G. Zech 4 Irving Place B. Grimes New York, New York 10003 W. Gammill L. Shao Joyce P. Davis, Esquire.

J. Miller Law Department R. Vollmer Consolidated Edison Company T. J. Carter of New York, Inc.

A. Schwencer. 4 Irving Place D. Ziemann New York, New York 10003 P. Check G. Lainas. Richard Remshaw D. Crutchfield Nuclear Licensing Engineer S . Grimes Consolidated Edison Company

..Ippolito of New York,. Inc.

R. Reid 4 Irving Place:;

V.: Noonan New York, NewYork. 10003 :

G.-Knighton D. Brinkman Anthony Z. Roisman .

Project Manager Natural Resources Defense Council OELD 917 15th Street, N.W.

OI&E (3) Washington, D. C. 2.0005.

C. Parrish/P. Kreutzer ACRS (16) Dr. Lawrence R. Quarles NRC Participants Apartment 51 NSIC Kendal at Longwood, TERA Kennett Square,- Pennsylvania 19348 Mr. J.:P. Bayne, Resident Manager' Theodore A. -Rebelowski IndianPoint 3 Nuclear Power Plant U. So Nuclear Regulatory. Commission P. 0. ox.:215 .,P. 0. Box 38 Buchanan, New York _10511

Buchanan, New York 1051 .

Mr. J. W. Blake, Ph.D., Director, John D. OToole Envi ronmental 'Programs Assistant :Vice President Power Authority of the Consolidated Edison Company..

of.New York, Inc.

State!of New York  : .4 I rvi ng Pl ace:',

10: Columbus ICircl e New York., New -York .10019 New: York, New York. 10003,,

h

-4

--Dr. LawrenceD.; Quarles Apartment, 51

. Kendal.- at Longwood

• i.Kennett Square, Pennsylvania 19348 Mr. George M. WilVerding Licensing Supervisor Power Authority: of the

• State of New:York 10 Col umbus Ci rcl e New York, New York 10019 Mr. P. W. Lyon, Manager.- Nuclear Operations

-Power Authority of -the State-of New York 10 Columbus Circle New.York, New York 10019 Mr. Vito J. Cassan Assistant General: Counsel Power Authority of -,the State" of, New York 10: Columbus Circle%

New York,. New York 10019 Ms. Ellyn Weiss Sheldon, Harmon and Weiss 1725 I Street, N.W.

Suite 506 Washington, D. C. 20006

1o#7 5~~

Flo4q NOTE TO. Darrell Ejsenhutt _ 1,t:..---,. r E

SUBJECT:

LETTERS ON LIMERICK,. IP, AND ZION .

ANL has, provi ded -the_ enclosed ,groupi~ng fr om: the .letters, you received as a result of the IP/ZION population statements. They estimate $60K for signature-ready: copi es for all letters. This seems high to me (-j $200 per letter) since the subject matter is narrow.

We could:

1. Turn on ANL having them develop standard paragraphs for your approval to :respond to all letters.- .Complete letters could then be developed.

I'd suggest you have Gary*follow the process to keep it on track.

2. Have ANL develop standard paragraphs and do the assembly'and clerical effort In youroffice. . .

3. Request Ann Savolainen's help in developing standard paragraphs and prototype letters. Because of the narrow focus of the letters, this might be the fastest and least costly way to go.

Richard H. Vollmer Distribution Copies:

• • ocket Files:

TERA SEP r/f RVollmer O FFIC E ..................... ..............................................................

SURNAME ......... .......... *S1IV-o 1De

)W6 .r/j m DATE 0 ..................

NRC FORM 318 (9-7 ri

S oil35 5.fo v 4 5o.j.36 "NOTE TO:. Darrell1 El senhut _Z, . ,.  :

SUBJECT:

LETTERS ON LIMERICK, IP, AND ZION ANL has provided the enclosed groupig from the -lettersyou received as a result of the IP/ZION population statements. They estimate $60K for signature-ready. copies for all letters. This seems high to me (-' $200 per letter) since the subject matter is narrow.

We could:

1. Turn on ANL having them develop standard paragraphs for your approval to respond to all letters.,,Complete letters could then be developed.

I'd suggest you have Gary follow the process to keep it on track.

2. -Have ANL develop standard paragraphs and do the assembly-and clerical effort inyour..office.,.

3. Request Ann Savolainen's help in developing standard paragraphs and prototype letters. Because of the narrow focus of the letters, this might be the fastest and least costly way to go.

Richard H. Vollmer Distribution Copies:

mocket Files:

TERA SEP r/f RVol Imer

' .O FF,C FIC SURNAMEI E,......................

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I.........

DATE ... .................... ....

1! " Z,

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NOPCFlORMI 210 7O2 NRO& f'AA .,A,

% , I - .1 C:InV7, 1070- -

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9 Nfill, IAOW~h- -

Docket No,. 50-286 LICENSEE:. Power Authority 6f the.State of New York (PASNY)

FACILITY: Indian Point, Unit- No;.3

SUMMARY

OF MEETING ON OCTOBER 25, 1979 TO DISCUSS STEAM GENERATOR INSPECTION A meeting was held on October 25,' 1979 between representatives of the Power Authority of the State of New York,(the licensee), Westinghouse, and the .

NRC.1 A list of attendees is attached. The meeting was held to discuss the results of the-recent steam generatdr inspection conducted at Indian Point, Unit 3.

The inspection indicated cracking of the support plates, hourglassing and, as categorized by Westinghouse, "extensive denting." Preliminary conclusions with supporting sketches, photos and tabulations are given in Attachment 2.

Because of the evidence of hourglassing of the upper support plates, Westing house recommended plugging of all the tubes in Row 1 in all foursteam generators. The licensee is followingthis recommendation, and is also plugging all tubes that did not.pass a.610 mil probe. Results-of the gauging are tabulated below.

Steam No. of Tubes No. of Tubes Passing Probe of Generator Inspected 720 mil 650 mil 6TO mil. 540 mil

.31 488 243 5 2 0 32 495 254 15 2 0 33 682 267 19 9 1 34 497-' 219 *18 - 3 1 The licensee experienced -periods.of high chloride intrusion during August and September 1979 which may have contributed to the denting. .The licensee stated that a program of boron addition to the feedwater will be instituted in an effort to slow down the rate of denting. /

The licensee proposed that inspection .of all four steam generators be'done at the next refueling. outage, scheduled for April .1981.

-)rv~kJJ OURA ...................

NRRMF 31 ( .)P 02..............

...... .. ...... ............. PP C :

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70......0 NULFORM 318 (9-76) NR(2 0240

• U.Si. 00"VERNMUNT PRINTING OFI=ICE : 1978 - 2'S$." 760

' " " SOME"ON

- '2 .-

The. NRC staff did not fihd this acceptable and requested that the licensee.

consider an earlier inspection of all steam generators.- The licensee was also requested to submit an application for a license amendment-concerning inspection of steam generators that follows the form.at of the Indian 'Point 2 license condition.

" At the conclusion of the meeting the resolution of a date for the next inspections of all-steam generators was left as an open action item as was the schedule for submittal of-a request for license amendment dealing with steam generator inspections in light of the additional degradation reported at the meeting.

L. Olshan, Project Manager Operating Reactors Branch #1 Division of Operating Reactors Attachments:

1. List of Attendees 2.. Preliminary Conclusions

°" IVLmJ R:shan: #1 DOR:ORd ORB:ms'"'lAgc~en~g I 12/;7-/79 " 121X)/79 . .. ....... .. .............

. ..............4 ...... . '. ....... 2.... ......... t ........... ...........

U.S GVE~iNON PINTNGOPC: 1"7 - 2615.,- 769 N11C PORM 318 (9-76) NRCK 0240 769

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 Docket No. 50-286 LICENSEE: Power Authority of the State of New York (PASNY)

FACILITY: Indian Point, Unit No. 3

SUMMARY

OF MEETING ON OCTOBER 25, 1979 TO DISCUSS STEAM GENERATOR INSPECTION A meeting was held on October 25, 1979 between representatives of the Power Authority of the State of New York (the licensee), Westinghouse, and the NRC. A list of attendees is attached. The meeting was held to discuss the results of the recent steam generator inspection conducted at Indian Point, Unit 3.

The inspection indicated cracking of the support plates, hourglassing and, as categorized by Westinghouse, "extensive denting." Preliminary conclusions with supporting sketches, photos and tabulations are given in Attachment 2.

Because of the evidence of hourglassing of the upper support plates, Westing house recommended plugging of all the tubes in Row 1 in all four steam generators. The licensee is following this recommendation, and is also plugging all tubes that did not passa 610 mil probe. Results-of the gauging are tabulated below.

Steam No. of Tubes No. of Tubes Passing Probe of Generator Inspected 720 mil 650 mil 610 mil 540 mi 1 31 488 243 5 2 0 32 495 254 15 2 0 682 267 19 9 1 33 1 34 497 219 18 3 The licensee experienced periods of high chloride intrusion during August and September 1979 which may have contributed to the denting. The licensee stated that a program of boron addition to the feedwater will be instituted in an effort to slow down the rate of denting.

The licensee proposed that inspection of all four steam generators be done at the next refueling outage, scheduled for April 1981.

The NRC staff did not find.this acceptable and requested that the licensee consider an earlier inspection. of all' steam generators. The' licensee-was

-also requested to submit an application for a license amendment concerning inspection, of steam generators that follows the format of the Indian ,Point 2 license condition.

At the conclusion of the meeting the resolution of a .date for the next inspections of all steam generators was left as an open action.item as was the schedule .for submittal of a request for license amendment dealing with steam generator inspections inlight of the additional degradation reported at the meeting.

L. Olshan, Project Manager Ope.ratilng Reactors Branch, #1 Division of Operating Reactors.

Attachments:

1.l List of Attendees

2. Preliminary Conclusions

0 December 28,1979 Meeting Suzmnary for -Power Authority of the State of New York (PASNY).

Docket Files, NRC PDR Local PDR L iite Plains Public Library 100 Martine Avenue

'Whi t Plains New York 10601 ORB] Reading NRR Reading Mr. Vito J. Cassan H. Denton !Asisistant General Counsel E. Case D. Eisenhut Poer Authority of the R. Tedesco. State of New York G. Zech 16 Columbus Circle New York, New York 10019 B. Grimes W. Gammill L. Shao Anthony Z. Roisman J. Miller Natural Resources Defense Council R. Vollmer 917 - 15th Street, N.W.

T. J. Carter Washington, D. C. 20005 A.- Schwencer D. Ziemann Dr. Lawrence D. Quarles P. Check Apartment 51 G. Lainas Kendal at Longwood D. Crutchfield Kennett Square, Pennsylvania 19348 B. Grimes

-. T.-Ippolito.

Mr. George M. Wilverding R. Reid Licensing Supervisor V. Noonan P'ower Authority of the G. Knighton State of New York D. Brinkman. 10 Columbus Circle Project Manager New York, New York 10019 OELD OI&E (3) Mr. P. W. Lyon C. ParrishiP. Kreutzer Manager - Nuclear Operations ACRS (16) Power Authority of the NRC Participants State of New York NSIC 10 Columbus Circle TERA New -York, New YQrk 1.001-9.

Licensee Short Service List

Mr. .J. P. Bayne, Resident Manager Theodore A. Rebelowsk-i - . ......

Indian Point 3 Nuclear Power Plant U. S. Nuclear Regulatory Commission

'P.0.-Box--215 -- . P. O. Box 38 Buchanan, New York 10511 Buchanan, ,New York 10511 Mr. J. W. Blake, Ph.D., Director Ms. Ellyn Weiss Envi ronmental Programs Sheldon, Harmon and Weiss Power Authority of the 1725 1 Street, N.W.

State of New York Suite 506 10 Columbus Circle -Washington, D. C. 20006 Newiork.,_NewY ork_, 10019.

. . . .. "4

Attachment 1 List of Attendees PASNY NRC J. F. Davis L"." Frank R. L. Goyette R.- D. Li'aw J. J. Kelly R. A. ,McBrearty, S. Zulla E. L. Murphy L. 'N.O1shan.

Westinghouse J. Strosn,ider C. Benton C. W. Hirst 0.D?.

.Malinowski

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NOV 2 9 1979 Docket No. 50-28 MEMORANDUM FOR: A. Schwencer, Chief,-Operating.Reactors Branch No. 1, DOR FROM: R. Bangart, Acting Chief, Effluent Treatment Systems Branch, DSE

SUBJECT:

INTERIM COMMITMENT FOR DEMONSTRATING COMPLIANCE WITH 40 CFR PART 190 FOR INDIAN POINT In a letter dated November 1, 1979, Paul'Early of the Power Authority of the State of New York provided an interim commitment for demonstrating compliance with 40 CFR Part,190 at Indian Point Nuclear Plant, Unit No. 3. This commitment is acceptable, however, the licensee should be informed that the report required by Technical Specifications 2.4.1.b and/or 2.4.2ic in Appendix B of DPR-64, should include a

,dose assessment with respect to conformance with 40 CFR Part 190, or whenever a variance to 40 CFR Part 190 is projected.

Original Signed By Richard L. Banart Richard.L, Bangart, Acting Chief Effluent Treatment Systems Branch.

Division of Site Safety and Environmental Analysis cc: W. Kreger J. Collins L. Olshen To Murphy F, Congel G. Knighton L..Rarrett P. Wagner

-W. Burke F. Cardile DISTRIBUTION:

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9 9 JAN 1 1 1980 5 ,.

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NOV 2,61979 MEMORANDUM FOR: A. Schwencer, Chief, Operating Reactors Branch #1, DOR FROM: G. Knighton, Chief, Environmental Evaluation Branch, DOR

SUBJECT:

FUEL HANDLING ACCIDENT INSIDE CONTAINMENT - Indian Point 3 (TAC 08761)

PLANT NAME.: Indian Point Station Unit 3 DOCKET NO.: 50-286 RESPONSIBLE BRANCH: ORB 1 !IEP~

PROJECT MANAGER: L. Olshan G"L' R1 FI,I U STATUS: EEB - Complete By letter dated January 17, 1977, the staff requested the Power Authority of New York (the licensee) to submit an evaluation of a postulated Fuel Handling Accident Inside Containment (FHAIC) at Indian Point Unit 2 and Unit 3 (Indian Point 2/3). The licensee submitted an evaluation of an FHAIC by letter dated March 21, 1977. The staff requested, in letter dated May 5, 1977, that the licensee provide a basis for his model for mixing and for isolating the contain ment before a complete release of activity occurs in Indian Point2/3. We also requested an analysis including the worst single failure during the accident. The licensee submitted, in a letter dated June 15, 1977, the assump tions used for containment mixing, and locations and descriptions of monitors which will automatically isolate the containment. The staffreviewed the licensee's June 15, 1977 submittal, and'by letter dated January 12, 1978 proposed possible means to provide adequate assurance that the consequences of the FHAIC are within the guideline of 10 CFR Part 100 for Indian Point 3.

These proposals being: (1) increase the minimum time after shutdown before refueling, (2) redundant radiation monitors on the operating floor which will automatically isolate the containment, (3) a safety grade duct and charcoal filter on the purge exhaust from the containment, (4) smoke tests or other experiments or analysis which will demonstrate that the radioactivity released from the damaged fuel assembly would be mixed in the containment, or (5) conservative analysis which demonstrates that the containment would be isolated in a timely manner by the existing monitors assuming a single failure. The licensee agreed, in a letter dated February 14, 1978, to increase the minimum time between shutdown and fuel handling from 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> in-the Indian Point 3 Technical Specifications. This increase in.shutdown time before refueling was accepted into the Indian Point 3 Technical Specifications by the Safety Evaluation dated March 22, 1978.

We have reviewed the February 14, 1978, submittal and have found that for us to conclude that the potential consequences of this postulated accident are appropriately within the guidelines of 10 CFR Part100 (less than 100 Rem thyroid), we have had to assume (1) operation of, and periodic and appropriate CONTACT: S. Baker, EEB/DOR 49-28066 ,

A. Schwencer NOV 2 6 1979 testing of the Containment Purge System charcoal filters or (2) the minimum delay between shutdown and initiation of refueling is 365 hours0.00422 days <br />0.101 hours <br />6.035053e-4 weeks <br />1.388825e-4 months <br />. Neither assumption is based on limits in the Indian Point 3 Technical Specifications.

For our evaluation to be valid suitable technical specifications must be adopted concerning the Containment Purge System charcoal filters or the minimum delay time between shutdown and refueling is 365 hours0.00422 days <br />0.101 hours <br />6.035053e-4 weeks <br />1.388825e-4 months <br />. Acceptable technical specifications on the Containment Purge System (CPS) are in Enclosure 2. We believe the most practical assumption to implement at Indian Point 3 would be the adoption of technical specifications on the Containment Purge System.

Therefore, we have based the enclosed evaluation on the implementation of these ventilation filter system technical specifications. Based on the enclosed technical specifications on the charcoal adsorbers, degradation of the adsorbers during operation of the CPS and a margin of safety to assure the charcoal radioiodine removal efficiencies are at least the efficiencies assumed in our evaluation of the FHAIC, we have only assigned a 70% charcoal radioiodine removal efficiency for the CPS.

In our review, as per the memorandum dated April 11, 1977, from J. Donohew to B. Grimes, we did not require that the CPS be safety grade and did not consider the Single Failure Criteria, IEEE Standards, seismic design and equipment quality group classification. The CPS is not safety grade. We conclude that this is acceptable because the potential consequences of the postulated FHAIC are within the exposure guidelines of 10 CFR Part 100 with no credit given for operation of the CPS. In addition, the surveillance requirements we require for the CPS filters discussed above are less than the requirements on safety grade ventilation filter systems because to have the potential consequences of this accident appropriately within the exposure guidelines of 10 CFR Part 100, more stringent surveillance requirements on the non-safety grade RBPES filters are not needed.

The enclosed evaluation also considers the failure of all fuel pins in two spent fuel assemblies indicated by a recent study, following the dropping of an assembly about 14 feet into the core and directly hitting another assembly.

DSE/DSS has been asked to review the probability and consequences of dropping a spent fuel, assembly in the core and damaging more fuel pins than the equivalent of one assembly. If, for both assemblies, we use the assumptions given in Regulatory Guide 1.25 and taking no credit for the non-ESF charcoal filters, the potential consequences of this accident are greater than' the guidelines of 10 CFR Part 100. But taking into account more realistic values for power peaking factor, clad gap activity and pool decontamination for both assemblies, we conclude that potential consequences of this postulated accident should not be greater than the exposure guidelines of 10 CFR Part 100. Because these potential

• . consequences are less than the guidelines of 10 CFR Part 100, we have concluded that no additional restrictions on fuel handling operations and plant operating procedures are needed while our review is underway. The atmospheric dispersion factor used by the staff is taken from the memorandum from L. Hulman to G. Knighton'ated September 4,.1979.

A. Schwencer - NOV 2 6 1979 is the safety evaluation of the postulated FHAIC at Indian Point 3.

Enclosure 2 are acceptable technical specifications on the Containment Purge System. It should be noted that these conclusions are based on evaluation criteria which has been applied uniformly to all operating Plants, and, therefore, do-not reflect any additional safety margin-which may result from the Task Force review of "high risk" reactor sites.

by Original signed

___GeorgeW Cyight _

G_W_Kfi-i~h-tb~, Chi-f EiWi ronment-l-EV-l-at-i-nBrch Division of Operating Reactors J3 Ehf-oI'Sures:

As stated cc:

-- jD_Ei-nhut

-MTl-jler

__W.-Ganli

________R.W_.Hus ton L. -Ho-lma'n Secti-on- EEB ER Ep Ac4 DOR iwLIa new WPasedag Lrrett GKnigh on J il I0/3 /79 o/30/79 10/6 /79 TplV/79 /79

UNITED STATES NUCLEAR REGULATORY COMMISSION

" .WASHINGTON,

" D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGARDING THE FUEL HANDLING ACCIDENT INSIDE CONTAINMENT POWER AUTHORITY OF THE STATE OF NEW YORK. INC.

INDIAN POINT STATION, UNIT 3 DOCKET NO. 50-286 Introduction By letter dated January 17, 1977, the staff requested the Power Authority of the State of New York, Inc. (the licensee) to evaluate the previously unevalu ated potential consequences of a postulated Fuel Handling Accident Inside Containment (FHAIC) at Indian Point Unit 3 (Indian Pt. 3). The licensee submitted, in a letter dated March 21, 1977, an evaluation of the FHAIC. The staff reviewed this submittal and requested in a letter dated May 5, 1977, that the licensee provide a basis for his model for mixing and for isolating the containment before a complete release of activity occurs. The staff also requested an analysis including the worst single failure during this accident.

The licensee stated, by letter dated June 15, 1977, that the potential conse quences for the worst single failure in the accident are 277.8 Rem Thyroid and 1.24 Rem Whole Body. The licensee described the assumptions used for containment mixing and locations of monitors which will automatically isolate the containment.

The staff reviewed the licervsee's June 15, 1977, submittal and concluded additional actions were needed to provide adequate assurance that the potential consequences of this accident were less than the guidelines of 10 CFR Part 100.

By letter dated January 12, 1978, the staff proposed possible means to provide adequate assurance that the consequences of the FHAIC are within the guidelines of 10 CFR Part 100 for Indian Point 3. These proposals being: (1) increase the minimum time after shutdown before refueling, (2) redundant radiation monitors on the operating floor which will automatically isolate the containment, (3) a safety grade duct and charcoal filter on the purge exhaust from the containment, (4) smoke tests or other experiments or analysis which will demonstrate that the radioactivity released from the damaged fuel assembly would be mixed in the containment, or (5) conservative analysis which demonstrates that the containment would be isolated in a timely manner by the existing monitors assuming a single failure.

The licensee, in a letter dated February 14, 1978, agreed to increase the minimum time between shutdown and fuel handling from 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> in the Indian Point 3 Technical Specifications. This Technical Specification change was evaluated and approved in the safety evaluation-dated March 22, 1978, for Indian Point 3. This change reduces the magnitude of radioactivity in the spent fuel assemblies, available for release during this accident and provides additional assurance that the potential consequences are within the Part 100 guidelines.

Evaluation We have completed our-review of the licensee's.March 21, 1977, June 15, 1977, and February 14, 1978 submittals, which address the potential consequences of an accident involving spent fuel handling inside containment. We have performed

-an independent analysis of the FHAIC. Our assumptions and the resulting potential consequences at the Exclusion Area Boundary are given in Table 1.

The conclusion of this evaluation are contingent upon the licensee adopting suitable Technical Specifications regarding the Containment Purge System (CPS) operability and periodic testing of its charcoal adsorbers iodine removal efficiency. Acceptable Containment Building Purge Filtration System Technical Specifications are included in Enclosure 2.

We conclude that testing the Purge System charcoal filters every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation for a 90% methyl'iodine removal efficiency at 95% relative humidity will provide adequate assurances that the charcoal adsorber iodine removal efficiency is at least that which we have given in Table 1 and have assumed in our evaluation of the FHAIC. Based on the above discussed technical specifications on the charcoal adsorbers, degradation of the adsorbers during operation of the CPS and a margin of safety to assure the charcoal radioiodine removal efficiencies are at least the efficiencies assumed in our evaluation of the FHAIC, we have assigned a 70% charcoal radioiodine removal efficiency for the CPS.

We conclude that the implementation of these Technical Specifications into the Indian Point 3 Technical Specifications will provide adequate assurance that the potential consequences of a postulated FHAIC are appropriately within the guidelines of 10 CFR Part 100. Appropriately within the guidelines of 10 CFR Part 100 has been defined as less than 100 Rem to the thyroid. This is based on the probability of this event relative to other events which are evaluated against 10 CFR Part 100 exposure guidelines. Whole body doses were also examined, but they are not controlling due to decay of the short-lived radio isotopes prior to fuel handling.

In our review, we did not require that the CPS be safety grade and did not consider the Single Failure Criteria, IEEE Standards, seismic design and equipment quality group classification. The CPS is not safety grade. We conclude that this is acceptable because the potential consequences of the postulated FHAIC are within the exposure guidelines of 10 CFR Part 100 with no

.credit given for operation of the CPS. In addition, the surveillance require ments we require for the CPS filters discussed above are less than the require ments on safety grade ventilation filter systems because to have the potential consequences of this accident appropriately within the exposure guidelines of 10 CFR Part 100, more stringent surveillance requirements on th'e non-safety grade CPS filters are not needed.

A recent study' has indicated that dropping a spent fuel assembly into the core during refueling operations may potentially cause damage to more fuel 1J. N. Singh, "Fuel Assembly Handling Accident Analysis," EG&G idaho Technical Report RE-A-72-227, October i978.

pins than has been assumed for evaluating the FHAIC. This stud. has intc cted that up to all of the fuel pins in two spent fuel assemblies, the one dropped and the one hit, may be damaged because of the embrittlement of fuel cladding material from radiation in the core. The radiation embrittlement would occur within the fuel's first few months of operations.

The probability of the postulated fuel handling-accident inside containment is small. Not only have there been several hundred reactor-years of plant operating experience with only a few accidents involving spent fuel being dropped into the core,-but none of these accidents has resulted in measurable releases of activity. The potential damage to spent fuel estimated by the study was based on the assumption that a spent fuel assembly falls about 14 feet directly onto one other assembly in the core; an impact which results in the greatest energy available for crushing the fuel-pins in both as'semblies. This type of impact is unlikely because the falling assembly would be subjected to drag forces in the water which should cause the assembly to skew out of a vertical, fall path.

Based on the above, we have concluded that the likelihood of a spent fuel assembly falling into the core and damaging all the fuel pins in two assemblies is sufficiently small that refueling operations inside containment are not a safety concern which requires immediate remedial action. However, because there is a chance that more than one spent fuel assembly may be damaged during refueling, we are reviewing the study and the probability and consequences of dropping a spent fuel assembly in the core and damaging more fuel pins than the equivalent of one assembly. The:objective of this review ,. to determine if any additional restrictions on fuel handling operations or plant operating procedures are needed. Any conclusions of this review which are applicable to this plant will be implemented.

We have calculated the potential radiological consequences of a fuel assembly drop onto the core assuming all the fuel pins in two spent fuel assemblies are ruptured. If, for both assemblies, we use the assumptions given in Regulatory Guide 1.25, and taking no credit for the non-ESF charcoal filters, the potential consequences of this accident are greater than the guidelines of 10 CFR Part 100.

However, the source term defined in Regulatory Guide 1.25 is conservative because (1) these two assemblies are unlikely to both have the high power peaking factor and clad'gap activity used in Regulatory Guide 1.25, and (2) the pool 'decontamination factor for inorganic iodine may well be greater than that used in Regulatory Guide 1.25. Taking into account more realistic values for power peaking factor, clad gap activity and pool decontamination, we conclude that potential consequences of this postulated accident should not be greater than the exposure guidelines of 10 CFR Part 100. Because these potential consequences are less than the guidelines of 10 CFR Part 100, we have concluded that no additienal restrictions on fuel handling operations and plant operating procedures are needed while our review is underway.

The results of this analysis warrarted an investigation of a similar accident in the spent fuel pool. For this, a drop of 2-1/2 feet was postulated and the analysis performed in the same manner as previously described. Results indicate that in this scenario damage to the missile or target is minimal. No fuel.

ins in either fuel assembly were calculated to be ruptured.

Environmental Considerations The environmental impacts of an accident involving the handling of spent fuel inside containment have been addressed in Section 6.1 of the Final Environmental Statement (FES) dated February 1975 for the operation of Indian Point 3.

Conclusion As discussed above, the staff has evaluated the licensee's analysis of the postulated FHAIC. After performing an independent analysis of the radiological consequences of an FHAIC to any individual located at the nearest exclusion boundary, the staff concludes that the doses for one assembly failure, are appropriately within the guideline values of 10 CFR Part 100 and for failure of two assemblies are within the guideline values of 10 CFR Part 100 and are, therefore, acceptable. For our conclusion to be valid, the licensee must incorporate acceptable surveillance technical specifications pertaining to the operability and the periodic testing of the Containment Purge System charcoal's iodine removal efficiency.

5 Table 1 ASSUMPTIONS FOR AND POTENTIAL CONSEQUENCES OF THE POSTULATED FUEL HANDLING ACCIDENTS AT THE EXCLUSION AREA BOUNDARY FOR INDIAN POINT STATION UNIT 3 Assumptions:

Guidance in Regulatory Guide 1.25 Power Level 3025 1MWt Fuel Exposure Time 3 years Power Peaking Factor 1.65 Equivalent Number of Assemblies Damaged 1 Number of Assemblies in Core 193 Charcoal Filter Efficiency Elemental and Organic 70 percent Decay time before moving fuel 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> 0-2 hours, X/Q Value, Exclusion Area Boundary -3 3 (ground level release) 1.1 x 10 sec/m Doses, Rem Thyroid Whole Body Exclusion Area Boundary (EAB)

Consequences from Accidents inside Containment.

• 4.

-6 Enclosure 2 REFUELING OPERATIONS CONTAINMENT BUILDING PURGE FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION The Containment Building Purge Filtration System shall be exhausting through HEPA filters and charcoal adsorbers.

APPLICABILITY: WHEN IRRADIATED FUEL IS BEING HANDLED IN THE CONTAINMENT AND IRRADIATED FUEL HAS DECAYED LESS THAN (365) HOURS ACTION: With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel within the containment until the system is restored to operating status.

SURVEILLANCE REQUIREMENTS The above required Containment Building Purge Filtration System shall be demonstrated OPERABLE:

a. At least once per 18 months and (1) after each compl te or partial replacement of a HEPA filter or charcoal adsorber bank, or (2) after any structural maintenance on the HEPA filter or charcoal adsorber housing which could effect system operation:

1. Verifying that the charcoal adsorbers remove >99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place while operating the ventilation system at a flow rate of cfm +

• 0%.

2. Verifying that the HEPA filter banks remove >99% of the DOP when they are tested in-place while operating the ventilation system at a flow rate of cfm + 10%.

b. At least once per 18 months and after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation, subject a representative sample of carbon from the charcoal adsorbers to a laboratory analysis and verify within 3l days a removal efficiency of >90% for radioactive methyl iodine at an operating air flow velocity + 20% per test 5.b in Table 2 of Regulatory Guide 1.52, July 1976.

Basis: The limitatiors on the Containment Building Purge Filtration System ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and a charcoal adsorber prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the

1, 1- . , .

-7 assumptions of the accident analyses. The representtie carbon sample will be two inches in diameter with a length equal to the thickness of the bed.

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555 November 23, 1979 Docket No. 50-286 LICENSEE: Power Authority of the State of New York FACILITY: Indian Point, Unit 3

SUMMARY

OF NOVEMBER 13, 1979 PHONE CONVERSATION REGARDING LESSONS LEARNED IMPLEMENTATION During a phone conversation on November 13, 1979 the NRC Lessons Learned Implementation Team discussed with the licensee, its October 22, 1979 response to our September 13, 1979 letter.

The team informed the licensee of those lessons learned items for which the licensee's proposed schedule for implementation is unacceptable. These items, along with the proposed and required completion dates, are listed in Enclosure 1.

The team informed-the licensee of those items for which the proposed action does not appear to comply with the lessons learned requirement. These items and their associated deficiencies are listed in Enclosure 2.

The team also informed the licensee of those items for which further clarification of the licensee's commitment is.necessary to demonstrate compliance with the lessons learned requirements. These items and the associated team questions are listed in Enclosure 3.

Items 2.1.3.6 (Instrumentation for Detection of Inadequate Core Cooling) (Procedures only), 2.1.7.a (AFW Initiation), 2.1.7.6 (AFW Flow), and 2.1.9 (Accident and Transient Analysis) were not discussed since these items are being implemented by the Bulletins and Orders Task Force.

By letter dated October 30, 1979 we provided additional clarification of the lessons learned requirements to all licensees. We also requested that within 15 days licensee's justify proposed actions not in complete agreement with the staff's requirements and improve the implementation schedule where it differed from the staff's requirements. During this phone conversation we informed the licensep that those ftems listed in Enclosure 1 and 2 should be addressed in their response.. In addition, the licensee agreed to provide the information requested-in Enclosure 3 in its response to our October 30, 1979 letter or as soon thereafter as possible.

L.N. Olshan, Project Manager Operating Reactors Branch #1 Division of Operating Reactors Enclosures (3):'

As stated cc w/enclosures: See next page

ENCLOSURE 1 INDIAN POINT, UNIT 3 ITEMS THAT 00 NOT MEET LESSONS LEARNED IMPLEMENTATION SCHEDULE

1. SECTION 2.1.4 -Containment Isolation Provisions The Indian Point 3 submittal dated October 22, 1979 stated that the implementation schedule of modifications will be dictated by "component lead time, equipment and unit availability." The Indian Point 3 modification schedules should be revised to assure compliance with the NRC schedules.

2. SECTION 2.1.6.b - Shielding Review The necessary shield modifications should be implemented by January 1, 1981 (Category B).

3. Containment Pressure, Containment Water Level, Containment Hydrogen Indicator IE Indian Point 3 submittal dated October 22, 1979 stated that the implementation schedule of modifications will be dictated by "component lead time, equipment and unit availability." The Indian Point 3 modification schedules should be revised toassure compliance with the NRC schedules.

4. SECTION 2.1l.3..a -Direct Valve Position Indication Your response indicated that you will provide a schedule for implementation on January 1, 1980. You should note that the implementation of this item is required to be completed by January 1, 1980 and is-required to be in conformance with the requirements delineated in the October 30, 1979 letter.

ENCLOSURE 2 INDIAN POINT, UNIT 3 PROPOSED-ACTIONS DO NOT APPEAR TO COMPLY WITH LESSONS LEARNED REQUIREMENTS

1. SECTION 2.1.3.b - Instrumentation For Detection of Inadequate Core Cooling The Saturation meter should provide a continuous display of saturation margin. This display is required to be in the control room. In addition to this display, a backup should be provided in the form of procedures and steam tables on curves.

2. SECTION 2.2.1(b) Shift Technical Advisor IP #3 is not responsive to this item in that their intention is to provide additional training to the shift supervisor to satisfy the accident assessment function. Page A-50 of NUREG 0578 states: ... In any event, when assigned as shift technical advisor, these personnel are to have no duties or responsibilities for manipulation of controls or commercial operations."

The use of plant engineers to perform these functions on an interview basis is acceptable as long as the STA is on shift.

The staff requires that these personnel be on duty by January 1, 1980 and that fully trained personnel be assigned by January 1, 1981.

ENCLOSURE,3 INDIAN POINT, UNIT 3 FURTHER CLARIFICATION IS NEEDED TO DEMONSTRATE COMPLIANCE WITH LESSONS LEARNED REQUIREMENTS

1. SECTION 2.1.1 - Emergency'Power Supply Requirements Your response to this section'did not address the specifics of the sub-parts of the position as described in NUREG-0578.

Clarification is required.

2. SECTION 2.1.6.a - Systems Integrity The January 1, 1980 program summary report should include measured leak. rates of the existing systems as well as the results of the design review, a summary of the procedures and a description of proposed modifications. Systems containing gases must be included.

3. SECTION 2.1.8.a - Post-Accident Sampling In addition to the engineering review, by January 1, 1980 the Authority should have both procedures and a description of modifications needed to obtain and analyze the necessary samples. Results should include H concentration in containment atmosphere as well as .issolved gases (H2 , 02) and B concentration in primary coolant.

4. SECTION 2.1.8.b - High Range Radiation Monitors The Authority should, by January 1, 1980, have methods for estimating both noble gas and radioactive releases even if the existing instrumentation goes off-scale. Both high range effluent monitors and in-containment monitors should be functional by'January 1, 1981.

5. SECTION 2.2.2B - Onsite Operational Support Center Clarification is required in that you are requested to address the specifics of the position as described in NUREG-0578.

6. SECTION 2.2.2C - Onsite Operational Support Center Clarification is required in that you are requested to address the specifics of the position as described in NUREG-0578.

IOVEMBER' 2 1 1979 RELL rOay DOCKET LCopy MEMORANDUM'FOR: Harold R. Denton, 'Director Office of Nuclear Reactor Regulation FROM: ... DarrelG. Eisenhut,-Acting Director Division of Operating Reactors

SUBJECT:

PROPOSED UhSK FORCE TO REVIEW INDIAN POINT 2/3

• AND ZION 1/2 BACKGROUND:

Following the Three Mi l e Island accident, the staff recognized 'an urgent need to conduct an in-depth review of the emergency preparedness capabilities of all nuclear power plants. In July 1979, criteria were identified for upgrading the capabilities of' such facilities and the goals for implementation. When this review effor-t along wi.th implementation of the short term lessons learned actions are completed, it is believed that the ability .of plant operators to determine the severity'and follow, the course of an accident as well as the responsiveness by various levels .of government officials to an accident will be greatly enhanced.

Two operating nuclear power plant sites, however, because of their location in areas of high population" density, are considered to warranta more in-depth and separate review to determine if the need exists for additional measures to further 'reduce the probability of a severe accident and to reduce the consequences of such an event. These two sites are. at the Indian Point 2/3 and Zion 1/2 facil-ities located near New York City and Chicago, respectively. A risk.to the public, which is high relative to other operating nuclear power plant sites, is believed to exist because of the difficulties of evacuating large numbers of people in a short time period during an emergency and because of the additional potential health effects resulting from accidents of very low likelihood for which emergency act'ons would-not be completely successful at any site.

PROPOSED PLAN In response to your request, we have outlined 'a review of the Indian Point 2/3 and Zion 1/2 facilities to study various alternatives that would be effective in reducing the probabil-ity and consequences of a severe accident. We propose that such areview be conducted by a small task group and would include a consideration of both preventive means as well as those for mitigating the consequences of a severe accident.. Accord'ingly, we have developed a preliminary Action Plan outlining such a review effort. The plan is.based upon the utilization of a multidiscipline review group involving other NRC Offices (principally RES and IE). The results of the group's efforts, while initially directed at Indian Point 2/3 and Zion 1/, would have general applicability to other

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0 NOVEMBER 2 1 1979 Harold R. Denton -2" The basic structure of the plan for these specific plants, is'three fold:

(1) improved operational actions, (2) curreft licensing actions,"and (3) an evaluation of various mitigating design features.. The 'actions -that are involved in items .- (1) and '(2) are mainly directed toward accident prevention, whereas (3) deals principally with mitigation -of the potential consequences of an unlikely but severe'accident'.

.() Improved'Interim'Operational Actidns: These actions involve,. but are' not necessarily limited-to,, such considerations'as increased IE-inspections,

• additional resident inspectors, augmented control room staffing and

• i~mproved-operat6r(; qualifications and training. Selected aspects from the Long Term Lessons Learned Report (NUREG-0585) may also be included as measures to further reduce-the likelihood of a severe'

-accident-, e.g. corporate management-accountability. In:addition, a review of the emergency procedures would be included to improve operational capabilities. This'effort will also include an examination of the derating of the plant during an interimperiod while the actions in items (2) and (3)-below are reviewed.

• (2) Current Licensing Actions: The actions involved in this aspect of the plan will include implementation on a priority basis of those TMI-2 short term lessons learned-actions stated in NUREG-0578 and Bulletin and-Order review matters. In-addition prompt attention will be given to processing of outstandilng .amendments for these plants.

Subsequent actions would include implementation of any recommendations resulting .from the resolution of unresolved safety items .on a priority basis.

(3) Severe Accident Mitigation Features:- In addition to the increased preventive actions described in (1) and (2)*above, there are.several

-. measures that,may be effective in mitigating the ,consequences of a major reactor accident involving core melt. Additional means-such as derating and measures to augment the existing ECCS via an independent system are considerations that fall between prevention and mitigation ani may be effective agents against a major-nuclear accident. Such means also would include a bunkered system with a dedicated heat removal system.,

Present considerations for severe accident mitigation include such features as a core catcher. (e.g., magnesium oxide), hydrogen control measures, and filtered containment venting. These',features, either singly or in combination could delay or prevent containment failure..

and.provide varying degrees of .protection against an early release of large quantities of fission products from the containment if a major accident involving a core melt should occur. Delaying the SURNAN flA* .U. . ....... ...

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-Harold R. Denton . - 3"- NOVEMBER 2 1 1979 release of-large quantities of fission products from-the containme.nt provides additional time to'implei-tent effective emergency plans and actions .in regions aroung the subject plants as well as other plants

.where evacuations may become necessary in the event of a major nuclear accident. While all of the mitigation measures described have neither peen fully evaluated nor developed sufficiently to quantify either their feasibility or effectiveness, the state-ofthe-art technology appears adequate so that they may be given preliminary consideration for the indicated applications. To better assess their effectiveness, analyses would ,be necessary tabound such related-core melt phenomena as steam and hydrogen explosions and the release of noncondensibles from the interaction between core melt products and the concrete

- basemat'. These aspects as well as containment failure modes would be-included in the overall studies of selected scenarios that would.

apply to several,core melt type accidents.

Over the last week, we have had several NRC staff discussions and have developed the foregoing discussion that is in effect a proposed program. We have also called the three affected licensees to inform them that-this peeliminary review is under way. We wil.l also consider ACRS involvement as the program develops further.

We anticipate ueeting with the three licensees involved to discuss this program in general during the week of November 26 or December 3, 1979. We-will keep you.

informed as our program becomes more refined S.. fI Signed By.

Darrell G. Eisenhut, Acting Director

• Division of. Operating .Reactors cc: E. Case D. Vassallo DISTRIBUTION R. Mattson Central Files S. Hanauer- NRC PDR.

U. Ross TERA D; Muller JRBuchanan /

S. Levine NRR Reading T. Murley DOR Reading N. Moseley GGZech V. Stello R. Tedesco T. Telford..

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aOCTOBER 1 0 1979 MEMORANDUM FOR: Herbert 4N. Berkow, Acting Chief Manager-and Administrative Support Branch, NRR FROM: Darrell Eisenhut, Acting Director Director of Operating Reactors, NRR

SUBJECT:

DATA REQUIRED FOR INDIAN POINT 3 OPERATING LICENSE FEE CALCULATION Your memorandum of September 18, 1979 asked DOR to calculate the costs associated with the review and-issuance of the Indian Point 3 Operating License from the issuance of the CP to the 100% power authorization on, August 18, 1979.

From a comptlter printout of the "NRC Manpower System" a minimum of 13,986.8 hburs of staff review is justified (see .attached sheets).

Using, 1800 houi~s per year and $70,000 per man-year, as you suggested to L.,Olshan, the present IP3 project.manager, this equates to $543,931 as a minimum.

Additional-ly, the December 1976Monthly Report to the Division of Site-Safety and Environmental Analysi,s, prepared by Oak Ridge National Laboratory (ORNL) shows that the Indian Point 3 environmental review by. ORNL cost $525,166. -An excerpt from that report is attached for your information-. Thus the total cost was in excess of $1,069,097.

Since the total expenditure Of .NRC man-hour and contractual funds, associated with the above 100 percent power licensing activities exceedthe 1,024,500 fee LFMB cited in its July 5, 1978 memorandum to OELD on this subject, no further NRR staff effort is considered necessary to resolve this matter. Your September18 memorandum and attachments are returned as Attachment 3 to this memorandum.

Darrell G, E senhut, ACting Director Division of Operating-Reactors, NRR Attachments:

1. NRC Staff Man Hours

2. Contractor Dollar Expenditure YELLOW

3. Mem'o, Ierkow to EiFsenhut - __ - i. ,I '. . - - _

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NEMORANDUM FOR: Herbert N. Berkow, Acting Chief lanager and Administrative Support Branch, NRR FROM: Darrell Eisenhut, Acting Director Director of OperatingReactors, NRR

SUBJECT:

DATA REQUIRED FOR INDIAN POINT3 OPERATING LICENSE FEE CALCULATION Your memorandum of September 18, 1979 asked DOR to calculate. the costs associated with the review and.issuance of.the Indian Point 3 Operating License 'from .the issuance of the.CP to the OU%. power authorization on August 1b, 1979.

From a computer printout of the "NRC Manpower System" a minimum of 13,986.8-hours of staff review is justiiied (see attached sheets).

Using 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br /> per year and $70,000 per man-year, as you suggested to L. Olshan, the present IP3 project manager, this equates to $543,931 as a mi ni mum.

Additionally, the December 1976 Monthly Report to the Division of Site Safety and Environmental Analysis, prepared by Oak Ridge National Laboratory ()ORNL) shows that the Indian Point 3.envi~ronmental review by ORNL cost $525,166. An excerpt from that report is attached for your information. Thus the total cost was in excess of $1,069,097.

Since the total expenditure of NRC man-hour and contractual funds associated with the above 1OU percent power licensing activities exceed the 1,024,500 fee LF14B cited in its July 5, 1978 memorandum to OELD on this subject, no further NRR staff effort is considered.

necessary to resolve this matter:. Your September 18 memorandum and attachments are returnea as Attachment 3 to this memorandum.

//, ,,/* ., .

.' Darrell G. Eisenhut, Ating Director ,"

Division..of OperatinqgReactors, NRR '

//

Attachments:

1. NRC Staff ian Hours

2. Contractor Dollar Expenditure -

3. Memo, Berkow to Eisenhut, dated 9-18-790 w/attachf;ients

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.- ETHRU: LeonReiter, leader Geology and Seismology Section Geosciences Branch, DSS FROM: Phyllis Sobel, Geophysicist, Geology and Seismology Section Geosciences-Branch, USS

SUBJECT:

STAFF RESPONSE TO',,PETITIONS. FOR-COMMISSION REVIEW INDIAN..'POINT - SEISMIC MONITORING Enclosed is the Staff Response to petitions for Commission review of the Indian Point Unit 3 issue on the expanded seismicmonitoring network. I was assisted in this review by Richard McMullen, Geologist, Leon Reiter, Section Leader and David Budge, Geologist and Consultant to NRC, SD. I have reviewed the previous documentation in this proceeding and current seismicity data collected near the Ramapo Fault. I have concluded that-due to increased monitoring of the Ramapo Fault in recent years by Lamont-Doherty Geological Observatory (Lamont), there is now no need-to require the licensees to expand their network.

-However, there continues to be sufficient uncertainty whether the Ramapo is a localizer of seismicity and therefore monitoring in the area must be continued at least at the current level. Therefore, I recommend that thestations Con Ed plans to donate to Lamont be ep oyed in an expanded/Arrangement to increase coverage of the Ramapo Fault and other faults in the High--ds--.I -am currently preparing.an NRR User Need Statement for RES that will ask RES to fund this expanded seismic network. Driginal S L' d b1 SSob~l Phyllis Sobel, Geophysicist Geology and Seismology Section Geosciences Branch,-DSS

Enclosure:

As stated cc: w/enclosure J. Knight J. Harbour, RES

.R. Jackson B. Smith, ELD L. Reiter J. Moore, ELD P..5obel R. McMullen Mew 0q LOlshan, ORBIi".. R. Rothman

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INDIAN POINT - STAFF RESPONSE TO PETITIONS FOR COMMISSION REVIEW Summary This Staff review addresses the two petitions for Commission review from the New York State Energy Office and the Citizens' Committee for Protection of the Environment. Both of these petitions request the Commission to condition the Licensees' operating license with the requirement that the Licensees expand their seismic monitoring network.

The first discussion of this issue was held in March 1977 at the ASLAB Indian Point 3 hearings on an expanded microseismic monitoring network along the Ramapo Fault. The ASLAB majority decided that the enlarged network was unnecessary. In 1978 the ACRS reviewed an article by Aggarwal and Sykes which contained new information on the Ramapo Fault System. On August 3, 1979, Mr. Farrar, Chairman of the-Indian Point ASLAB..issued his dissenting opinion on the expanded microseismic monitoring network.

On September 6, 1979, the Indian Point majority issued their supplemental opinion, supporting their original decision. Then on September 21, 1979, the New York State Energy Office filed a petition for Commission review of the majority and dissenting opinions, especially the issue of the expanded seismic network. On September 26, 1979, the Citizens' Committee for Protection of the Environment filed a similar petition.

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The NRC Staff then evaluated documents in the Indian Point 3 proceeding, evaluated the seismic data collected near the Ramapo Fault since the March 1977 ASLAB hearing, and reassessed the need for an expanded network. We conclude that due to increased monitoring of the Ramapo Fault by Lamont-Doherty Geological Observatory, partly funded by the NRC, there is now no need to require the Licensees to expand their network. However, it is our position that microearthquake monitoring in the. area must be continued at least at the current level. Therefore, we recommend that Con Ed donate their stations to Lamont so that the stations can be incorporated into Lamont's monitoring of the Ramapo Fault.

Indian Point Unit 3 ASLAB Hearing Indian Point Units 1, 2 and 3 became operational in 1962, 1974, and 1976, respectively. The Staff first considered the significance of the Ramapo Fault in the review of Indian Point 3. The Ramapo Fault is the major geologic structure in the region, extending for a distance of about 50 miles from Peapack, New Jersey to Stony Point, New York, where it enters into the Hudson Highlands in a highly diffuse pattern. The Ramapo Fault proper and its northern extension into the Hudson Highlands have been referred to collectively by Ratcliffe (1971) as the Ramapo Fault System. Until 1975 the Staff had accepted the conclusions of a 1969 USGS report for the AEC that stated "there are no known active faults or other young geologic structures in the area that could be expected to localize earthquakes in the immediate vicinity of the site."

-3 The AEC Staff review of the Ramapo Fault was initiated in response to a request by the New York State Geological Survey (the State). The AEC subsequently'met with representatives of the Licensee, Consolidated. Edison (Con Ed), to express the safety concerns raised by the State and to indicate the need for more precise geologic and seismic data in the Indian Point site region. Con Ed responded by initiating additional geologic studies of the Ramapo Fault system and by installing a dense network of seismograph stations to obtain more accurate-locations of earthquakes in the. site region sufficient to permit unambiguous conclusions to be drawn about the relationship between earthquake occurrence and geologic structure. The 12 seismic stations are located around the Ramapo Fault on the east and west sides of the Hudson River clustered around the Indian Point site. Several stations of this network were in opera tion by June 1975 and the network was considered to be fully operational by September 1975. Information on the seismic recordings of the network through May 1979 have been transmitted to the NRC. The operators of the Con.Ed network are Woodward-Clyde Consultants.

In Appendix C to Supplement No. 1 to the Safety Evaluation of Indian Point 3 (1975), the Staff summarized their geologic and seismic evaluation of the Indian Point site. Based on the applicant's investigations, a review of the professional literature, and consultations with geologists and geophysicists, the Staff concluded that the Ramapo Fault was not capable, and the Safe Shutdown Earthquake (SSE) design value of .15g was determined to. be adequately conservative.

In 1976 the Staff was given new information concerning the locations of two relatively small earthquakes which occurred near the Ramapo Fault -- the 1951 Rockland County, New York earthquake (Modified Mercalli intensity V) and the magnitude 2.5 March 11, 1976 Pompton Lakes, New Jersey earthquake (Modified Mercalli intensity IV-V). The Staff considered the lack of evidence of geologically young movement and the absence of any obvious clustering of historic earthquake activity along the Ramapo Fault System to support the conclusion that the fault was not capable.. Nevertheless, because of the uncertain location of the two earthquakesinear the fault and the fact that acknowledged authorities had suggested that some earthquakes could be associated with the fault system, the Staff was not able to conclude that the Ramapo Fault might not play a role in localizing earthquake activity. The Staff considered a confirmatory program directed toward a more definitive determina tion of the age of most recent movement and a determination of the potential for earthquake activity on the fault system to be necessary. Accordingly, in Supplement No. 3 (1976), the Staff conditioned the, amendment to the operating license authorizing power operation of Indian Point 3 to require the Licensees to conduct a program of geological and seismological investigations. One part of the program was to extend the existing Con Ed earthquake monitoring network southward to include the Pompton Lakes, New Jersey epicenter area and northward to include the Fahnstock region. This expanded network was to be operated at least two full years following complete installation of all stations to see if the seismicity pattern of the region was random in nature. The Licensees. on August 27, 1976 filed a Motion to Modify License Condition to delay the implementation of the expanded seismic network until after the Atomic Safety

-5 and Licensing Appeal. Board (ASLAB) hearing on the seismic questions. On November 10; 1976, the Appeal Board granted a.stay in the implementation of the expanded seismic network program until an evidentiary hearing was held on the need for the network. The Commission allowed the stay of the expanded seismic network to continue because, the-Staff considered there to be no great threat to public health or safety if it was not expanded. The ASLAB in ALAB-436, dated October 12, 1977, concluded that the Ramapo Fault was not a capable fault under Appendix A, 10 CFR Part 100. The ASLAB decision also addressed the issue "Is the operating license condition for Indian Point 3, requiring an expanded microseismic monitoring network along the Ramapo fault warranted?"

This issue was addressed at an evidentiary hearing conducted during six days between March 8 and March 16, 1977. Written testimony was supplied by the Licensees, the State of New York and the NRC Staff. CCPE submitted no testimony but participated in cross-examination.

In the hearing, the Licensees stated that:

1. The requirements of Appendix A had. been satisfied. The value of the expanded network is only in terms of research and development.

2. The relationship between small earthquakes and the potential for larger earthquakes is questionable.

3. Due to the low rate of seismic activity, there is little chance of obtaining sufficient data in a period of two years to determine a relationship between small and larger earthquakes.

The licensees projected the cost of installing and operating the expanded network to be $1,071,000.

The Staff stated three reasons for requiring the expanded network:

1. A wider network was required to focus on the Ramapo and other nearby fault systems rather than a geologic feature in the immediate locale of the plant site.

2. Instrumental data are more accurate-than historical data for determining the locations of earthquakes. The expanded network would show whether or not the Ramapo Fault plays a role in localizing earthquakes.

3. The expanded network would provide a greater potential for gathering data on the nature of earthquake activity in the area (for example focal mechanism solutions, the distribution of microearthquakes and the relation ship between earthquakes and geologic structures).

From these kinds of data the Staff could possibly determine the stress environ ment and assess whether there is the potential for larger earthquakes or.

whether conditions were favorable for movement along-a particular fault.

Although microearthquake data in many cases may not be useful for prediction of larger earthquakes, microearthquakes reflect current tectonic activity and may be used for evaluating the geological causes of earthquakes. Should it be determined as a result of these studies that larger earthquakes couldoccur on the Ramapo Fault, the Staff would have to consider the near field effects of an earthquake localized on a fault near the site, which may result in greater impact on the design response spectra than the design earthquake assumed to

occur randomly within the prbvince. Near field effects, such as. possible high acceleration values, were not considered in the case of the Indian Point site.

The State, which had recommended the expanded network, stated that instrumental data was needed to determine if the earthquakes were related to local faults.

If such is the case, then historical events might also be associated with faults in the area. The State's thesis then was that an expanded network could provide more focal mechanism solutions which would "link a particular fault with a given earthquake."

Because of the following reasons the ASLAB majority halted the expansion of the network (ASLAB decision ALAB-436). The ASLAB majority determined that "the enlarged monitoring network would not contribute to the assurance of health and safety of the public and is therefore unnecessary." The ASLAB majority stated that the Staff's license condition was not a proper one under their interpretation of Appendix A. Instead, it was their view that it was a research project with dubious ability to predict a larger event than the SSE.

The ASLAB view was that the Staff had ignored that part of Appendix A which states that "structural association of a fault with geologic structural features which are geologically old (at least pre-Quaternary) such as many of those found in the Eastern region of the United States shall, in the absence of conflicting evidence, demonstrate that the fault is not a capable fault within this definition." Also that in the 18 months the 13 station Con Ed network had been in operation, 18 earthquakes were recorded, 4 of which were near the Ramapo Fault; thus the instrumental data showed few microearthquakes in a large region.

Mr. Farrar, Chairman of the ASLAB, dissented from the majority opinion. A few details of his dissenting opinion were presented in the decision ALAB-436. On the issue of the expanded network, Mr. Farrar agreed with the Staff, the State, and CCPE that the Ramapo Fault might "play a possible role in localizing earthquakes" and that the expanded microseismic monitoring network is therefore warranted. Mr Farrar planned to submit in writing a full response at a later date.

The Staff did not appeal the majority decision.. The Staff believed the informa tion would be helpful, but there was no technical basis for stating that the public health and safety would be jeopardized by not having the network. The Staff's data base supporting its reasons for wanting the expanded network remained the same, and it was felt that little would be gained by reopening the case. Instead, the Staff planned to explore whether funding should be sought for a research project involving a similar network.

ACRS Review of Aggarwal - Sykes Article On May 11, 1978, the Staff notified the ASLAB of an, article which appeared in the April 28, 1978 issue of Science magazine by Drs. Aggarwal and Sykes concerning the Ramapo Fault System. On June 16, 1978, the Indian Point 3 Subcommittee and the Seismic Activity Subcommittee of the Advisory Committee on Reactor Safeguards (ACRS) met to discuss the significance of the Aggarwal-Sykes article. Presentations were made by the staff (Dr. John Kelleher), the Licensee, and Drs. Sykes and Aggarwal on material in. the Aggarwal-Sykes article that might have a bearing on the'ASLAB decision. In the Aggarwal-Sykes paper, about

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-9 10 new focal mechanism solutions near the, Ramapo Fault were shown. On the bases of the new and old data, the article's authors also calculated the probability of occurrence of an intensity greater than the SSE at the Indian Point site to be about 5 to 11 percent. The Staff and ACRS after considering all the evidence concluded that the information contained in the Aggarwal-Sykes article did not warrant a change in the SSE value of .15g for Indian Point, Units 2 and 3.

Specifically, the Staff and the ACRS consultants stated that:

T. The data'base is not sufficient to be considered a valid study.

2. The data used to calculate the probability values is questionable because of the small number-of events. For example, some of the historical earthquakes used in the calculations are not associated with the Ramapo Fault.

3. The historical data does not support the recurrence relationship.

Presentations by Con Ed and the Power Authority of the State of New York stated that:

1. None of the larger earthquakes (greater than or equal to intensity V) predicted by the Aggarwal and Sykes relationships have occurred near the Ramapo Fault.

2. Some of the earthquake data-used in the calculations are not associated with the Ramapo Fault.

3. More larger events are needed to establish the parameter b (the relationship between frequency of earthquakes per year and magnitude).

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At its July 6-8, 1978 meetings, the full ACRS considered this matter and reported in a July 13, 1978 report to Chairman Hendrie that. the review of the Aggarwal Sykes article "found insufficient basis for suggesting a change in the current seismic criteria."

Mr. Farrar's Dissenting Opinions On August 3, 1979, Mr. Farrar, Chairman of the Indian Point ASLAB, issued his detailed dissenting opinions about portions of the majority, opinions related to seismology. Mr. Farrar agreed with the Staff that with regard to the role of the Ramapo Fault in possibly relating to the Indian Point site, "something in the general vicinity appears to be localizing earthquake activity."

Mr. Farrar stated that "safety would'be enhanced by pursuing further investiga tions in the vicinity," that is, by reinstating the expanded microseismic monitoring network which the Staff had wanted the applicants to install.

ASLAB Majority Supplemental Opinion On September 6, 1979,. Drs. Buck and Quarles, members of the Indian Point ASLAB, issued their supplemental opinion. They state that Mr. Farrar did not support his view that "something in the general vicinity appears to be localizing earthquake activity." Drs. Buck and Quarles determined that the Licensees and their ratepayers should not be saddled with the expense of a seismic research project when there is no indication that "the enlarged network is necessary to provide reasonable assurance that operation of the Indian Point reactors will not endanger the public health and safety."

- 11 04 Petitions for Commission Review' The intervenors in the Indian Point 3 proceeding continued to argue for an expanded network. On September 21, 1979, the New York State Energy Office (the State) filed a petition for Commission review of ASLAB decisions ALAB-436 (the majorityopinion) and ALAB-561 (Farrar's dissenting opinion) in the matter of the Indian Point 3 proceeding. The petition for review is directed primarily to the Board majority with regard to the issue of the expanded seismic network. The petition mentions the 1975 and 1976 earthquakes in the vicinity of Indian Point which were discussed in the Aggarwal and Sykes article (mentioned above). The State argues that "the monitoring system should be expanded and maintained until there is some clearer-indication of the meaning of the seismic, activity in the vicinity of the Ramapo Fault." Additionally they argue that it is a small cost to ratepayers (estimated at lN0 to 36¢ per household for an initial-outlay) to pay for this scientific tool for analyzing stress and seismic hazard. In the State's view.the capability of the Ramapo Fault muststill be considered an open question until the significance of recent seismic activity in the vicinity of the Ramapo Fault can be determined.

This view was their position at the end of the. Indian Point 3 hearings.

The State also mentions several other issues in which it differed from the ASLAB decision. These were:

1. Farrar's method of isolating the 1897 Giles County, Virginia earthquake from the Indian Point site is considered inconsistent with his method in the Seabrook matter.

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2. Reasonably large tectonic provinces,.such as those suggested by the State, should be used.

The NRC Staff agrees with the majority decision on the first issue. On the second issue, the Staff disagrees with the tectonic provinces the Board used, and instead continues to endorse the tectonic provinces described by the Staff in the ASLAB hearing.

On September,26, 1979 the Citizens' Committee for Protection of the Environment (CCPE) filed a petition for Commission reviewof ASLAB decisions ALAB-436 and ALAB-561, relating to the Indian Point seismic monitoring network. CCPE

.argued that in deleting the monitoring requirement, the Appeal Board found against "a substantial body of expertopinion" (that is, the Staff, the State and CCPE). In the CCPE view "the seismic monitoring network brings additional knowledge about a potential threat to the safety of the Indian Point plant."

Recent Staff Review The matters relating to the necessity for an expanded seismic network have not been formally reviewed by the Staff since the March 1977 ASLAB hearing.

During the time span since then the key Staff seismologists who participated in the Indian Point 3 review have left the Geosciences Branch and/or the NRC.

The current Staff seismologists have reviewed the previous documents in the Indian Point 3 proceeding, the majority and dissenting opinions of the Appeal Board, the petitions for Commission review, microearthquake activity recorded near the Ramapo Fault since the March 1977 hearings, and any changes in the

configuration of seismic stations in the area since March 1977. The Staff has also reviewed the Aggarwal-Sykes article, ACRS's view on their work, and contacted seismologists now monitoring seismicity near the Ramapo Fault.

Significant new information has developed from the operation since 1970 of a network of seismic stations in New York and New Jersey by the Lamont-Doherty Geological Observatory of Columbia University (Lamont). In 1978, Lamont's seismic monitoring program was funded by the New York State Energy Research and Development Authority, the U.S. Geological Survey, the New York State Geological Survey, and the NRC Office of Nuclear Regulatory Research (RES).

During 1976, 1977 and 1978, Lamont increased coverage of the southern end of the Ramapo Fault by establishing five stations in New Jersey (.Station names GMTN, GPD, LVNJ, PRIN, RAMA). In 1978, Lamont increased the network coverage of the northern end of the Ramapo Fault by establishing two stations (station names - CLIN and WPNY). Thus while the Con Ed network was not expanded to the extent suggested by the State and the Staff in the Indian Point 3 hearings, coverage of the Ramapo Fault System has been increased in recent years through Lamont's efforts by the addition of seven stations.

Seismologists at Lamont have reanalyzed seismograms of earthquakes that occurred in the last 20 years in southern New York and northern New Jersey. Epicentral locations and fault plane solutions have been determined where possible.

Lamont seismologists, Aggarwal and Sykes, in their 1978 Science article, state that these fault plane solutions indicate that high-angle reverse faulting on N to NE trending planes is the predominant fault plane solution near the

Ramapo Fault. The detection threshold of the Lamont network currently extends down to magnitude 1-1/2 to 2 near the Ramapo Fault. The ability to locate epicenters is within a few km. Seismologists at Lamont and Con Ed have exchanged P readings and seismograms for earthquakes near the Ramapo, so that both networks use the other network's data to determine epicentral locations. 'The operators of the Lamont and Con Ed networks are in almost daily contact to work on. the problem of distinguishing earthquakes from the more numerous quarry blasts (personal communication from operators of Lamont and Con Ed networks). Woodward-Clyde Consultants, operator of the Con Ed network, records all the station data on magnetic tape, with three of the station records being on paper. The paper records are kept indefinitely, but the tape records are destroyed after 6 months if there are no interesting earthquakes. Since the conclusion of the Indian Point 3 hearings in 1977, the Commission Staff, the State, CCPE, and the ASLAB have been receiving quarterly reports from Con Ed and PASNY on microseismic events recorded by the Con Ed seismic network.

The pattern of microearthquake epicenters recorded since July 1976 and published in the Lamont bulletins generally conforms to the broad (60 to 70 km wide) northeast trending band of seismicity defined by the historical earthquake epicenters in the Hudson Highlands. This band follows the regional northeast southwest structural grain of the Appalachian Mountains in the Highlands.

There is an aseismic region northwest of the Highlands in northwestern New Jersey and southern New York.

-15 The Aggarwal-Sykes article mentioned in the State's petition for Commission review was discussed at the 1978 ACRS meeting. As mentioned previously, the Staff and the ACRS consultants considered the information presented in the Aggarwal-Sykes article and found that it.did not change any prior conclusions pertaining to the seismic criteria applied to Indian Point Units 2 and 3.

The Staff does not support the CCPE and State petitions asking for. Commission review. We do not believe .that requiring the utility to operate an expanded network is assuring public health and safety since increased monitoring.of the.

.Ramapo Fault in the last few years has not shown that the Ramapo Fault is capable. However, it is the Staff view that the historical earthquakes near the Ramapo Fault. prior to,1970 have poorly determined locations. The present combined.Con Ed and Lamont seismic networks are providing valuable information on the nature of seismicity near the Ramapo Fault system. This is the type of information envisioned by the Staff when it.imposed the license condition..

However, the Staff finds that it is inappropriate to basea-serious conclusion on a series of earthquakes in a region of low activity being monitored in a period of only a few years. The Staff believes that the NRC (RES) should continue to fund the operation of the Lamont.stations near the Ramapo Fault.

The Staff continues to believe that there is no geological evidence to indicate that the Ramapo Fault is capable within the definition of Appendix A to Part 100.

However, there continues to be sufficient uncertainty whether earthquake activity may be associated with the Ramapo Fault and neighboring faults. Continued microearthquake monitoring.would help resolve these uncertainties. We recommend that the monitoring network which Con Ed plans to donate to

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- 16 Lamont be deployed in an expanded arrangement in the Highlands so as to decrease the detection threshold and increase the accuracy of determining epicentral locations near faults in the Highlands. The NRR staff is currently preparing an NRR User Need Statement for RES which will ask that the following tasks be performed:

1. Incorporate the Con Ed stations in the Lamont network in the region around the Ramapo Fault. Lamont and the NRC Staff should work together to determine optimal deployment of the Con Ed stations to monitor seismicity in the Highlands.

2. Determine the velocity structure of the crust in the Highlands Region.

Reevaluate instrumental data using the velocity structure to determine more accurate epicentral locations.

3. In the quarterly Lamont bulletin include maps for the seismicity occurring in the reporting period and for cumulative instrumentally determined seismicity in the Highlands.