ML19261D852

Requests Plant Status Rept.Forwards Questions

ML19261D852
Person / Time
Site:	Crane, Davis Besse
Issue date:	03/29/1979
From:	Ahearne J NRC COMMISSION (OCM)
To:	Jennifer Davis, Harold Denton NRC OFFICE OF INSPECTION & ENFORCEMENT (IE), Office of Nuclear Reactor Regulation
Shared Package
ML19261D848	List: ML19261D847 ML19261D852
References
FOIA-79-98 NUDOCS 7906260523
Download: ML19261D852 (2)

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MEMO TO:

Harold Denton, NRR John Davis, I&E kj FROM:

John F. Ahearne~

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SUBJECT:

STATUS BSPORT ON DAVIS-BESSE UNIT NO. 1 In the December 20th memo to me discussing the actions taken by Davis-Besse Unit No.

1, I noted that the unit was still completing power ascension testing and had not yet operated for sustained periods above 90% pouer, even though the OL was issued in April 1977.

Recently, the response to Question 17 to Congressman Dingell's February 6, 1979 letter indicated that the Davis-Besse Unit 1 availability and capacity factor for the year 1978 was the lowest of any reactor that operated during the year, with the exception of Duane Arnold.

I am aware of the problems associated with the Duane Arnold reactor; however, I am not aware of the difficulties that are occurring at Davis-Besse.

In light of the above and the f act that the Davis-Besse design is similar to that of the Three Mile Island Unit 2 design, please p,rovide me with a detailed status report on this plant.

Attached are some specific questions I would like addressed in that response.

Attachment:

As stated cc:

Chairman Hendrie Commissioner Gilinsky Commissioner Kennedy 25f3 l[2 Co:amissioner Bradf ord Lee V. Gossickbr" Samuel J. Chilk Al Kenneke Len Bickwit 3./3c 7 I

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Attachment 1.

What are the problems that created the difficulty in completing the power ascension testing for this reactor?

Have they been adequately resolved?

Are there any test results which the staff has not deter-4 mined to be satisfactoty?

2.

What outstanding licensing issues are there?

Have any exe:cptions been granted for this plant?

If so, uhat are the justifications for those exemptions?

3.

Is part of the reason for the lcw capacity factor attributable to the plant operator personnel?

How do the plant ope rators co:.. pare v ith other plant operators in Regicn III?

In particular, I would note the recent notifications of the Commission about problems with the valving errors on the ECCS for both the low pressure and high prescure saf ety injection systems.

4.

In view of the recent Three Mile Island accident, are there any plant systems related to the safe operation of the plant which have experienced specific difficulties?

In particular, please provide me with the licensing

' staf f 's analysis of the Se.ptember 24, 1977 event and all major subsequent events that have occurred at the plant.

5.

Have any safety systems been subject to a large number of LER's?

Please provide details of the staff's revieu of these items.

6.

Ecw do the facility change requests for this station compare with other operating reactors which have been licensed in about the same timeframe?

7.

Are defects which have been identified in safety-related systems analyzed and corrected in a timely fashion by the licensce?

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MEMORA' mum FOR: Co c.issioner Ahearne (3 p m,s.FSchroeder EHayden HDenton THPV Executive Director for Operations EGCase Harold R. Denton, Director. Office of Nuclear FRGM-Reactor Regulation Your letter dated Novenher 3,1978 're;uested a chronology of the staff actions on nine specific recommandations which the ACRS made concerning Davis.Sesse Unit No.1 (DB-1) and on the ACRS reccaren-dation concerning procpt implenentation of Reactor Punp Trip (RPT) for B'J. 's. You also requested a description of the general procedure -

fellowed in response to ACRS recomendations. provides the status including a chronology of the staff actions on the nine specific reco n ndations of the ACRS ccrcerning DE-1.

To place this

~ chronology in context. the following infomation is provided as back-ground.

DR-1 was issued an operating license in April 1977. The unit is now co-pleting pnser ascension testing but has not yet operated for sus-tained periods above 90". pcwer. The licensee contained four conditions which related to the ACRS recommendations concerning seisnic reanalysis, ECCS, a bypass loop in the decay heat suction line and fire protection.

When these conditions are satisfied, the ACRS reccc 2ndations will be fulfilled.

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Between the time when the construction penait application was submitted for Unit 1 and when the construction permit application was submitted for Units 2 and 3. the method of correlating earthquakes with the values of applied acceleration was changed in the Standard Review Pl.an.

This was recognized at the time of the ACRS review of Unit 1, prior to the issuance of the operating license. The staff believes there is sufficient conservatism in the ' design analyses and construction practices to compensate for this change in method of correlation.

However, as a result of the ACRS concern the licensee is required to provide a seismic reanalysis before startup after the first refueling now estimated for March 1980.

During our April 1978 meeting with the

• ACRS we indicated that we would have guidelines for the seismic re-analysis to the licensee within a month. However, because of other natters of greater priority we have not yet provided the guidelines to the licensee.

k'e subnitted the guideline's to the ACRS for comment on November 1. 1978 and when we receive their ccoments we will submit the cuidelines to the licensee.

Since the licensee has already started all ce acte ;o compiei.e u ex in tnis arqa, we believe.the licensee am"it

..s e is ni c..re afalys !.s..and..ap ;nopriate.rodifi cations,..i.f.fne,te.sla ry.,,,

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At the time of the issuance of an operating license for Unit I we deter-nined that the ECCS analysis for Unit I was in accordance with Appendix K.

Mcwever, that deteruination.was based upon various separate computations.

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These.ere needed to correct for two different input errors in the original analysis. The license condition relating to the ECCS required the licensee to quantify the actual nargins that exist when all the changes are considered together.

It also required the reporting of coolant system flow data within 30 days following two weeks of sustained power operation at a power level of 90', or greater rated pcwer level. The first requirement, relating to quanti-fication of margin has been satisified however, the second requirement relating to data submittal has not been co~pleted because Unit I has not yet had sus-tained operation at 90". or greater for two weeks.

As part of our progran for all operating plants, we have evaluated the Unit I fire protection program in'accordance with current staff positions.

We are preparing Technical Specification changes and a Safety Evaluation Report which is scheduled for issuance in March 1979.

The recommendations of the ACRS (the concerns regarding hernetic seals, instrurentation for folicwing the course of an accident and ATWS) are generic to all nuclear plants and are awaiting the generic resolution. The nost significant of these concerns is ATWS.

The ATWS issue was first brought into focus with the issuance of WASH 1270, " Technical Report on Anticipated Transients Without Scram for '.Jater Cooled T'ower Reactors", September 1973.

Staff action on this issue with the industry, and the ACRS has been an ongoing activity. NUREG 0460, " Staff Report, Anticipated Transients Without Scram" dated April 1978 provides the current status of the issue. A supplement to NUREG 0460 is planned to be issued in December 1978.

The ACRS recommendation concerning the development of a radiation surveillance capability on the part of the State of Ohio is dependent upon the State's commitment of funds for a program for environnental monitoring. The Office of Inspection and Enforcenent has had an exchange

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.of letters with the State of Ohio in anticipation of their establishing

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a radiation monitoring program in the area of the Davis-Besse site.

The ACPS recommendation concerning sabotage is being considered in the

. ongoing review of the Davis-Besse revised nodified security plan. The licensee is required by regulation (10 CFR 73.55) to have an approved

- security plan by February 23, 1979.

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Fr.issionce Ahearne Enclosure 2 provides the status and a chronology of the staff actions cn the recowendation of the ACRS concerning RPT for BWR's. A letter with revised criteria citing specific acceptable designs and urging pro,pt installation of such syster.:s will be issued to the affected licensees in the near future.

. provides the general procedure folic ed in response to 4

ACPS cci,cerns.

In addition, we have one person whose principal assignrants is coortfinating NRR activities with the ACRS. He is in frequent contact with the ACRS staff, atterds all conthly meetings of the ACRS and aids in providing t1.ely responses to ACRS concerns.

He hepe that t!.is answers your concerns with res,tect to our respon-siver.ess to ACRS concerns.

If there are further questions please advise us.

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-*n Harold R. Dcoton. Director 9

Office of P.'uclear Reactor Regulation

Enclosures:

1.

Status & Chronology of Staff Actions on

.f Race,mendations 2.

Status & Chronology of Staff Actions on RPT's 3.

General Procedures 23l3 }[b Followed cc w/ enclosures:

Chairman Hendrie Conaissioner Gilinsky Conaissioner Kennedy Commissioner Bradford Mr. Chilk

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%..... y STAFF ACTION ON ACRS RECOMMENDATIONS OF JANUARY 14, 1977 CONCERNING DAVIS-BESSE UNIT I

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1.

Seismic Raevaluation COT.ittee Recomenda tion The structures and components of Davis-Sesse, Unit 1, were designed for a Safe Shutdown Earthquake (SSE) acceleration of 0.15g at the foundation level.

Because of changes in the regylatory approach to selection of seismic design bases, the Com.ittee believes that an acceleration of 0.20g would be core appropriate for the SSE acceleration at a site such as this in the Central Stable Region.

The applicant presented the results of prelimir.ary calculations concerning the safety mrgins of the plant for an SSE acceleration of 0.20g.

The Ccanittee recomends that the NRC Staff review this aspect of the design in detail and assure itself that significant margins exist in all systems required to accomplish safe shutdann of the reactor and contir,ued shutdown heat removal, in the event of an SSE ct this higher level.

The Cc m ittee believes that such an evaluation need not delay the start of operation of Davis-Besse, Unit 1 The Committee wishes to be kept infonned.

_S ta tus By letter dated April 22, 1977 we issued an operating license for the Davis-Besse Nuclear Power Staticn, Unit No.1.

The license contained condition 2.C(3)(r) which requires the licensee to submit prior to startup following the first scheduled refueling outage a seismic reanalysis and evaluation to the Comission for its ieview

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and approval of the adequacy of the facility systems needed to acccmplish safe shutdown of the reactor and continued shutdown heat rccoval.

k'e require that in perfonning the reanalysis, a safe shutdcwn earthquake acceleration of 0,20g be applied at the foundation level of the plant and that the response spectra specified in Regulatory Guide 1.60 be used.

Draft guidelines for the seismic reanalysis were discussed with the

' licensee on September 19, 1978. The draft guidelines were also sent to the ACRS on November 17, 1978, for their review and comments. The

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schedule for completing this task is (1)' Issuance of guidelines to licensee - December 1978, (2) Submittal by licensee to staff guidelines

- March 1, 1979, (3) Staff's review of licensee's submittal complete -

May 1, 1979, (4) Staff site visit for seismic audit - June 1,1979, 2313 177 f

_2 (5) Staff identification of items requirin9 follow-up action -

July 1, 1979, (6) Resolution of follow-up items - October 1,1979, and (7) Final Report to ACRS - November 1,1979.

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2.

ECCS Comnittee Reconi.enda tion The perfomance of the Emergency Core Cooling System (ECCS) has been

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evaluated using a Babcock & Wilcox evaluation nodel applicable to the raised-loop configuration. The NRC Staff has reviewed these evaluations and has determined that certain assurption regarding return to nucleate boiling do not comply strictly with the provisions of Appendix K to 10 CFR part 50.

The NRC Staff is also revie.ving se. oral other areas relating to ECCS perfomance. These matters should be resolved in a manner satisfactory to the NRC Staff.

Sta tus These matters were included as conditions to the license which '..as issued April 22, 1977 and as requirements in the precperational and start up tests prior to full pcaer cperation.

License condition 2.C.(3)(1) requires additional supporting analyses for the large break spectrum. The Toledo Edison Company met with the NRC staff on June 28, 1977 at which time the staff specified the larce break analyses required to meet the requirements of license conditon 2.C(3)(1). These requirements were confimed in our letter of August 25, 1977 to the Toledo Edison Company.

By letter, dated October 21, 1977, the Toledo Edison Company provided the analyses for the reactor coolant piping large break spectrum as specified in our letter of August 25, 1977. Based upon review of the Toledo Edison Company's submittal, we conclude that the stipulations of license condition 2.C(3)(1) for the submittal of large break spectrum analyses were satisfied.

Amendment No. 7, to the license was issued November 28, 1977, removing the requirement for the large break analysis.

Item F.4 of Attachment 2 to the license issued April 22, 1977, required a verification of a minimum sump water flow of 40 gallons per minute.

These tests were completed by August 26, 1977.

These actions were reviewed with the ACRS April 7,1978.

License condition 2.3(3)(m) required installation of flow measuring

. devices.

By letter dated May 19, 1978, the Office of Inspection'and Enforcement infomed us that they had verified that the finw measuring devices had been installed and.' tested as required. By letter dated May 26, 1978, Amendment No. 10 was issued which removed the license condition 2.C.3(n).

License condition 2.C.(3)(1) requires the reporting of cooiant system flow data within 30 days following two weeks of sustained power operation at a power level of 90% or greater rated pcwer level. Upon completion of this item we will consider this issue complete.

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State of Ohio C_opJ ltee Recom endation

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i In cor. junction with the evaluation and assessment of the frcpact of routine waste releases from this plant, the Comnittee recommends that the NRC Staff provide leadership in encouraging the developmant of improved environmental radiation surveillance capabilities on the part of the State of Ohio and appropriate local regulatory agancies.

Sta tus We discussed this item during our meeting with the ACRS in April 1978.

As we stated in that meeting this item was discussed in Section 18, item 3 of Supplcment 1 of Safety Evaluation Report. As we stated in Supplement 1, the NRC at that time could not provide technical aid to the State of Ohio without adequate funding, and a commitment on the part of the State of Ohio.

Since April 1978 we have been informed that the State of Ohio has started an environmental monitoring program in the area of the Davis-Sesse facility. Their pres nt program censists of collecting and ar,alyzing samples of milk, surface water and ground water. With the addition of one person to their staff the State of Ohio plans to eventually add the collection and analysis of vegetation, crops and meat to their program.

By letter dated October 20, 1978, the State of Ohio recuested criteria and an example contract operating an environmental radiation Quality Assurance program as an inaepenaent check on the licensee's program. By letter dated November 7, 1978, we provided the State of Ohio the suggested scope for such a program in the form of a copy of a typical contract which the Office of Inspection and Enforcement has entered into with participating states.

If the state shows further interest, we will plan to meet with the State of Ohio's Department of Health to review their capability and program and possible to negotiate a contract with them to provide some environmental data collection for us.

We have not directly informed the ACRS of these later developments because we have not concluded the action on this item.

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Hennetic Seals Cor ni.t_ tee Reconnendation The Comr.ittee notes tha t post-accident opera tion of the plant to Ir.aintain safe shutdown conditions may be dependent on instrumentation and electrical equipment within containment which is susceptible to ingress of steam or water if the hermetic seals are either initially defcctive or should become defective as a result of damage or aging.

The Cc,.nittee believes that appropriate test and maintenance procedures should be developed to assure continuous long-term seal capability.

Status The staff reviewed this item with the ACRS on April 7,1978.

At that time this item was identified as ACRS generic concern number Ild-2 which efines the problem as:

Certain classes of instrumentation incer-porate hermetic seals. 'dhen safety related components within containment must function during post LOCA accident ccnditions, their cparability is sensitive to the ingress of steam or water.

If the hernetic seals should become defective as a result of personnel errors in the maintenance of such equipment, such errors could lead to the loss of effec-tive hermetic seals and the resultant loss of equipment operability.

By letter to the ACRS dated May 4,1978, we reported on the status of all generic items relating to light water reactors including the subject item. ACRS generic concern, number IId-2 is incorporated in the staff's Technical Activities Program as Task C-1.

At the present time there has not been any action on this issue.

b'e do not plan any action in the immediate future.

A category "C" Task is considered by the staff to be one of low priority.

In general a category "C" activity is one judged by the staff to have little direct or immediate safety, safe-guards or environm2ntal significance, but which could lead to improved staff understanding of particular technical issues or refinemants in the licensing process.

In the ACRS Status. Report No. 6 dated November 15, 1977, concerning ACRS Generic Items'the ACRS adopted the staff definitions of priority l categories which assigned IId-2 to Category C.

Unless results of Task C-1 indicate that some action is necessary for Davis-Besse 1, we plan no further action on this Ccmmittee recomacndation.

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Instrumentation to Follow the Course of an Accident Committee Reconcsndation The Committee reccanends that, prior to core.ercial power operation of Davis-Besse, Unit 1, additional means for evaluating the cause and likely course of various accidents, including those of very 10w probability, should be in hand in order to provide improved bases for timely decisions concerning possible off-site emergency m:2sures.

The Committee wishes to be kept informed.

Status This matter pertains to the implementation of Ecgulatory Guide 1.97.

Regulatory Guide 1.97 was first issued for comment in DecemSar 1975 and was revised and reissued for comment in f ugust 19/7.

The staff's Task Action Plan A-34 initially called for implementation of the guide on 4 lead plants. Af ter caining experience on the pilot imple-mentation program, the plan was to be icplemented on all other operating plants consistent with the experience gained in the pilot program.

However, issuance of the guide for ccament generated opposition as reflected in the letter dated June 13, 1978, frcm the Atomic Industrial Forum (AIF). On November 9,1978, the staff met with an ad hoc committee of the AIF. As a result of that meeting the staff has agreed to revise its action plan and work directly with the ad hoc coccaittee of the AIF. The staff is preparing a response to the AIF letter dated June 13, 1973, and will be revising its task action plan and schedule.

Regulatory Guide on operating plants, we will imp Besse 1 2313 182 G

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ATWS

_C_o:.U ittee Re. cac.cnda tion _

The question of whether the design of this plant must be modified in

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order to comply with the requirements of WASH-1270, " Technical Report on Anticipated Transients Without Scram (ATWS) for Water-issue pending the NRC Staff Cooled Reactors," remains an outstanding & Wilcox general analyses completion of its review of the BabcockThe Coiaaittee recommends that t and the EaScock & Wilcox cc~pany continue to strive for an early of ATWS.

resolution of this matter in a manner acceptable to the NRC Staff.

The Committee wishes to be kept infonned.

Status This matter is addressed in the SER, p. 7-3, and in SER Supplemant No. 1, p. 18 4 At such time as the ATWS issue is resolved for operating plants, we will require appropriate modifications for Davis-Sesse 1.

Staff recccc.endations for resolution of the ATWS issue are currently under review by the ACRS and the RRRC. NRR expects to submit its recer.n.endations to the Commission within the next few months.

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Ey-Pass Loop Ccanitte_e Recommenda tion Cavis-Besse, Unit 1, has installed a bypass loop containing two manually operated valves around the decay heat removal system suction line iso-lation valves.

The nonnally closed bypass valves would be opened in the event of a spurious closure of one of the decay heat removal system suction line isolaticn valves during system operation.

The Concittee recor rends that further attention be given to the means employed for iso-lation of the low pressure residual heat renoval system from the primary system while the latter is pressurized, and that reliable means be developed to assure such isolation.

This matter should be resolved in a manner satisfactory to the NRC Staff.

The Conmittee wishes to be kept informed.

Status Desicn Modification Alternatives to the Fresent Key Lock Control in Manual _ Bypass Valves DH 21 and DH 23 Ey letter dated April 22, 1977 ue issued an operating license for the Davis-Eesse flucicar Power Station - Unit 1 with licensee condition 2.C.(3)(p) which requires that the licensee submit an analysis of design modification alternatives for the present key lock control in the manual bypass valves DH 21 and DH 23 around the decay heat removal suction line valves to decrease the likelihood of the bypass path being opened inadvertently when isolation of the decay heat removal loop is required.

The submitted analysis and installation of approved design nedifications shall be completed prior to startup following the first scheduled refueling outage.

By letter dated October 30, 1978, the licensee has scheduled submittal of the analysis and design modifi-cations for staff review by February 1979.

The first scheduled refuel-ing outage is scheduled approximately January 1980.

The A RS receives copies of licensing action and it is through these actions that the ACRS is kept informed.

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Fire Protection Co...ittee Recom;:-ndation The Caccnittee supports the NRC Staff program for evaluation of fire pro-tection in accordance with Appendix A to Auxiliary and Power Conversion Sys tens Branch Technical Position 9.5-1, " Guidelines for Fire Protection 4

for Nuclear Power Plants." The Committee recommends that the NRC Staff give high pricrity to the completion of both owner and staff evaluations and to recomrendations for Davis-Besse, Unit 1, and for other plants nearing cc:rpletion or constructicn in order to, maximize the opportunity for irpreving fire protection while areas are still accessible and changes are m:re feasible.

Status Reevaluation of Fire Protection Procram License condition 2.C.(3)(h) of NPF-3 requires the licensee to increase the level of fire protection in the facility to the levels recommended in Appendix A to the Standard Review Plan 9.5.1, Revision 2, " Fire Protection System" or with alternatives acceptable to the' staff.

The level of facility fire protection as stipulated in item 2.C.(3)(h) shall be completed within three (3) years from the dai of issuance of NPF-3.

License condition 2.C.(3)(h) also requires that the licensee impicment Section B of Appsndix A, " Administrative Procedures, Controls, and Fire Brigade," and Section C of Appendix A, " Quality Assurance Programs,"

prior to startup folicwing the first regularly scheduled fueling outage.

By letter dated August 20, 1977, the licensee was provided a copy of NRC document, " Nuclear Plant Fire Protection Functional Responsibili-ties, Administrative Controls and Quality Assurance," to be used as supplement guidance for the licensee's implementation of Sections B and C, Appendix A.

On October 13, 1977, a meeting was held with the licensce at which the staff addressed the inadequacies of the licensee's Fire Hazard Analysis Report for Davit-Besse, Unit 1 submitted on February 11, 1977.

The licensee stated that they would resubmit an amended Fire Hazard Analysis Report in November 1977.

A meeting was held on December 6, 1977 Auxiliary System Branch discussed the facility design for fire protection in the cable spreading room.

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. C' % ~ary 11, 1978 the licensee submitted Ravisicn '!o. I to the Fire

' 'rard l uijsis Fe;crt for Davis Besse, Unit 1.

Revision No. 1 was fo o.d to Le acceptable and is presently under detailed review by the Staff and staff's ccnsultants.

A fire protection site visit was com-pietc.d on May 23-25, 1978, and staff requests for information were iss,.d to the licensee on July 6, 1918. On August 1,1978, a meeting

.cas i.21d with the licensee to clarify certain staff requests for i n f a r:.3 + i c n.

?y letters dat2d Septe.:ler 7 and 20,1978 the licensee responded to the s taf f's requ w t 'or additicnal infora.ation dated July 6,1978.

A i.eeting 5.as held v.ith the licensee on October 26, 1978 to resolve the

'le h2ve scheduled the issuance of the outstanding issues.

Cavis-Easse, Unit No.1, fire Protection Evaluation Report and rela t.cd lii_ _'.se c...eni 2nt by March

, 1979

• f' = ria Fire.; c' 'c tion. Technical _ Leci fica tions The ;.u:nsc 2 v.as notified on Nov

' e r 21, 19 7 7 by telcphone and tele.

ct;y of cur requirer:nts for imple.nnting interim fire protection Technical Specifications for plant systems and administrative procedures.

A letter dated Ncve:+ :r 28, 1977 was sent to licensees providing the 71e st'rd ird Technical Specifications and requesting the licensee's c'

ros ?c:se by

<. : 'er 7,1977 for an applica tion for license amand. ment

..!th the clant specific interim Technical Specificctions.

The licensee submitted proposed interim Technical Specifications on Cecember 12, 1977 and on March 22, 1978 (Amend..ent No. 9) interim fire Technical Specification was issued for the presently installed fire protection equipment at the facility, as has been done for other operat-ing facilities.

The interim Technical Specifications deal with administrative, surveillance and corrective steps to reduce the likeli-hood of damaging fires pending our final review of the fire protection of Davis-Eesse, Unit 1.

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. 9.

Sabotage Comaittee Recge.mandation The Committee believes that the Applicants and the fiRC Staff should fur-ther review security provisions for Davis-Besse, Unit 1, for measures

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that could significantly reduce the possibility and consequences of

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sabotage, and that such measures should be implemented where practical.

_S ta tus The licensee's amended security plan in response to 10 CFR 73.55 was submitted on May 25, 1977.

The staff review for Phase 1 was completed on September 8,1977. A modified amendmant security plan was submitted in December 1977 and the review was completed in April 1978. A revised modified amended security plan was submitted in June 1978 and still is under review.

In general, the licensee is in full compliance with staff requirements.

Full compliance with the requirements of 10 CFR 73.55 is required by February 23, 1979.

By letter dated riovember 15, 1978, the licensee indicated that in their judgement, the Toledo Ed'ison Company will be in compliance with 10 CFR 73.55 by February 23, 1979.

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The ACRS commented to the NRC staff in March 1975 that they believed a recirculation pump trip (RPT) should be icple:nented promptly on all EWR's unless such a trip is not required to mitigate consequences of an ATUS. The staff set out to implement RPT generically separate from other ATWS nodifications

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and obtained licensee connitnents to insta ' an RPT. However, the licensees indicated their intention to delay final design and installation of the RPT, which they had earlier committed to install, due to concern that potential new requiraments resulting from the ongoing ATWS generic issue might result in required changes in the RPT syste:a's design. A letter with revised criteria citing specific acceptable designs and urging prompt installation of such systems will be issued to licensees in the near future.

The chronology of these events is:

"2rch 12, 1976 Letter to D. Moeller (ACRS) to L. Gossick reccmmanding prompt implementation of recirculation pump trip (RPT).

April 12.1976 Letter B. Rusche to D. Moeller stating that backfit determinations for RPT will L.) done en an individual case basis.

Following discussions between the staff and industry on broad ATWS requirements there was a staff decision to proceed generically on ATWS RPT separate from other ATWS modifications.

September 1,'1976 Lett..rs Staff to Licensees asking that they commit to an RPT and provide detailed design.

N 2313 188

._, i..;

.s

...,...._.......__..3 No.w-ber 15, 1976 resporces began.

De enber 28, 1976 Letter Rusche to Mociler informing ACRS of RPT implementation schedule.

Februa ry 22, 1977 Reviewers found responses inadequate and requested

[

additional informatio).

It became apparent to the staff at this time t at more detailed criteria should be developed.

Ihrch 10,1977 Standardized review criteria for RPT developed to April 3,1978 within NRR.

Because compatible RPT criteria ware desired for rewly licensed plants and operating plants, concurrence was required of review groups in DSS and DDR as well as those assigned to ATWS task management. A number of meetings were held and proposed criteria were circul.ated several times before agreement could be obtained between various concurret.ses.

New Criteria sent to licensees.

May 15,1978 to May 20,1978 1

July 1978 to Letters from licensees indicating their September 1978 intention to delay final design and instal-

~

lation of the RPT, which they had earlier committed to install, due to concern that potential new requirements resulting from the ongoing ATWS generic issue might result in required changes in the RPT system's design.

2313 189'_

In The Near Future Letters to Bh'R licensees offering approval of specific designs for RPT and emphasizing importance of installation will be issued after issuance of the supplemant to fiUREG-0460.

2313 190 O

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, PROCEDURE FOLLOWED IN RESPONSE TO ACRS CONCERNS The staff receives advice, comments and reconmendations from the ACRS in three forms.

a.

The Cor -ittee's letters to the Chainnan folicwing Committee review of a project normally contain a number of c0mments and recommenda-tions for further staff action.

b.

The Committee en occasion writes a special letter to the staff containing recommendations or expressing Conmittee concern regard-ing a particular issue, generally of a generic nature.

Quastions arise during the monthly Committee meetings in ccnnection c.

with discussion of the various agenda topics.

!n addition, the Comnittee writes an annual letter to the staff itemizing the various generic issues of concern to the Committee.

Subjects initially discussed in the special letters (item b, above) often are subsequently incorporated in the Committee's generic items list.

The procedure for staff response to the issues raised by the Cennittee varies according to the source of the Connittee concern.

The comments and recommendations in the Concittee letter on a project are subsequently addressed by the staff in a supplement to the Safety Evalua-tion Report.

The supplement describes each Committee concern and indicates either the action the staff has taken or intends to take in response to the issue raised.

The Project Manager is responsible for follow-up on each of these plant specific issues to assure ultimate resolution, with reports back to the Committte as appropriate.

Special topics raised by the Committee are responded to by the staff by means of either an oral or written report or, quite often, a written report folicwed by an oral presentation.

Resolution of the generic issues is handled through the staff's program for reso'ation of generic issues. A report is forwarded to the Coqaittee semi-annually which provides the updated status of resolution of each of the Committee's generic concerns.

Procedures for handling responses to miscella,neous questions have improved considerably during the last few years.

The Committee new 2313 191 manyme.

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provides a listing on an approximate semi-annual basis which itemizes the questions that have been raised during the preceding six months by by Co.uittee r. embers.

The staff responds to these letters with letter reports indicating the resolution or the status of resolution of each matter.

For the past two years, the staff has prepared a brief sunnary of each ACP,5 meeting which attempts to identify questions or concents raised dur-ing the meeting which appear to require staff action. Appropriate staff me.-icrs are then requested to pursue these natters by written or oral reports to the Cc=.ittee.

Starting two months ago, the Com7.ittee staff began furnishing the NPP. staff copies of material indicating the questicns and issues it felt nad bcen raised by Committee mambers.

These items are handled in the same fashion as those t; pics identified by the staff.

2313 192

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,csicner :.r.earne cu'A ct^;ure il/Z/78 4 49 n c,

n cF oc..:NT

.w.c.. r w. c a n o en n. uE n.v 11/3/78 TC:

Po.tP M E F O R SIG';A T',

IE Executive Dir. for 0jnrations O F '-

FIN At., R EP( Y O C H A t R ?.t A N FILE LOCAT:ON O EX ECUTIVE CI A ECTOR crscR,_ D.enton

' EE033rilON O t.&.T 1 L R C 'ACMO O RCPORT O OTHER SP ECI AL IP.ST A UCTIOP.S 08 a C.'. AR K S MSe O "'

%N Req a chronolocy of staff action on recomenda- // fgh.d #< M""t-M

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tions in ACRS report 1/14/77 on Davis-Besse f

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M' Unit,1 and 3/12/76 on Reactor Pump Trip

.m ror Ba,,,s, & a ceneral description of pro-n cedures followed in response to ACRS encm W.i ons_ _ _

CLACIFIED D ATA U C.. U b.N T/CC.~ Y f.O CIMS trtC A TICN N U '.' 3 E R C F P A G ES

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