ML20148H133
| ML20148H133 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/27/1978 |
| From: | Crocker L Office of Nuclear Reactor Regulation |
| To: | Case E Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7811130370 | |
| Download: ML20148H133 (45) | |
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- .. 1 EMORANDUM FOR: -EdsonG.Caie,DeputyDirector
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FROM: - ..L. P.'Crocker -
- c. Technical Assistant to the Director Divisales of Project Management '
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SUBJECTi - ACRS CDNCapts REGARDING DAVIS BESSE, UNIT 1 1
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a During the 201st ACMS uneting to Jamuary 1977, the Committee reviewed'the _ f c.
~0L application for' Davis Besse 1.f The Committee. issued a favorable < t- %
~ letter on the~ application ~en. January 14, 1977; hauever, in the letter, 1
the Committee identiffed a number of matters which it felt warranted 3
- , , further effert. A copy of the letter is at Enclosure 2. 2 ,
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~ qq . J . . i .: .:f..' W .:p- S . K; i The' staff met with thel Committee derfag the 216th segting in April 1978
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?to discuss the-status 'of reseletion of the matters mentioned in th(ACRS :;r -
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letter.; Following this status report. 'a Committee letter of August 25. J-
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" 1978 (Enclosure 3) to 'the NRC Chatrues reported on Cassittee activities - ; "
- during the preceding quarter and mentioned Davis Besse 1. The tone of
. ' the letter implies dissatisfaction with the staff. efforts to clear up the 7 Committee concerns.: #^.L 4~ <- " -
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! This subject of staff response to ACRS recommendations (presumably to
- specifically faciude Davis Besse 1) is^ scheduled to be discussed by the l Committee during its meeting.with the Commissioners on Noveder 2,1978.1 -
In preparation for that meeting, following is the current status of staff actions regarding the'ACRS recommendatices.. P
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STATUS
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i t3,y; , Gener41'- '"/ 3W5 The plant still is assigned to DpM, although'
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g~ q.# it now appears that' transfer to DDR is o 4 tuninent. A copy. of the transfer package is 2 attached as Enclosure 4.
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1 Connittee Recommendations - A totel of 9 recommendations were sade by the ) Committe in its January 14,1977 letter. ' These recommendations and the current status
- ofresolution(asperthetransferpackage) are provided as Enclosure 1.
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Asirs n3rea by
, g Lure o..Jrocher L. P. Crocker -
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Enclosures:
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- 3. ACRS letter 8/25/78 , i -
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I STATUS OF STAFF ACTIONS l l
' lw Seismic Reevaluation Committee Recommendation l
The structures and components of Davis-Besse, Unit 1, were designed for a Safe Shutdown Earthquake (SSE) acceleration of 0.15g at the foundation level. Because of changes in the regulatory approach to selection of seis-mic design bases, the Comnittee believes that an acceleration of 0.20g would be more appropriate for the SSE acceleratiors at a site'such as this in the Central Stable Region. The Applicant presented the results of preliJninary calculations concerning the safety margins of the plant for an SSE acceleration of 0.20g. The Comnittee recommends that the NRC Start review this aspect of the design in detail and assure itself that signifi- - cant margins exist in all systems required to amuglish safe shutdown of the reactor and continued shutdown heat removal, in the event of an SSE at this higher level. The Comrnittee believes that such an evaluation _ need not delay the start of operation of Davis-Besse, Unit 1. 'The Comnittee wishes to be kept informed. Status - Seismic Reanalysis License condition 2.C.(3)(r) requires the licensee to submit a seismic reanalysis and evaluation to the Commission for its review and approval of the adequacy of the facility systems needed to accomplish safe shutdown of the reactor and continued shutdown heat removal prior to startup following the first scheduled refueling outage, In performing the reanalysis, a safe shutdown earthquake acceleration of 0.20g shall be applied at the foundation level of the plant and the response spectra shall be used as specified in Regulatory Guide 1.60. Guidelines for the seismic reanalysis shall be specified by the staff. The responsibility for specifying the guidelines for the seismic reanalysis to the licensee is the Division of Systems Safety. The assignment for preparing these guidelines is specified in TACS, #4925-(May 23, 1978). Evaluation of the seismic reanalysis will be by the Division of Systems Safety. Draft guidelines were issued on September 9, 1978 and the draft guidelines were discussed with the licens'ee on September 19, 1978. The schedule for completing this task is-(1) Issuance of guidelines to licensee - November 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, (5) Staff identification of items requiring follow-up action - July 1, 1979, (6) Resolution of follow up items - October 1,1979, and (6) Final Report to ACRS - November 1, 1979. Staff reviewer will be J. Rajan (MEB). Management responsibility will be carried out by Operati'ng Reactors Branch No. 4. (
Reference:
SER Supplement No. 1, pgs. 2-3, 3-1, 18-1, 18-2, and E-2)
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'2 . ECCS-Committee' Recommendation
. L The performance of the Emergency Core Cooling System (ECCS) has been evaluated using a Babcock and Wilcox evaluation model applicable to the raised-loop configuration. _ The NRC Staff has reviewed these evaluations and has determined that certain assumptions 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 reviewiry; several other areas ' r
- relating to ECG performance. These matters should be resolved in' ;
a manner satisfactory to the NRC Staff. , ,
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This matter has been resolved and documented in amendments to the license. See, also, SER. Supplement-No.1, pp. 6-3 through 6-10, i t e g 9 y m e m-racem e e ,- av.v<-=- - -er- - . e r.w ow -,w - w , - m er w.-+ pr-a,-,v,wwew,-ww---w.,,ye.mer , , .e we m e.w a w -y e -e ist.v f , wer- y w a -w y .- mwr rs.*-c
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- 3. State of Ohio 4 ,
Committee Recommendation - In conjunction with tbe evaluation and assessment of the impact of ; routine waste releases from this plant, the' Consnittee recommends ~ that the NRC Staff provide leadership in encouraging the development- - of improved environmental radiation surveillance. capabilities on the-part of the State'of Ohio and appropriate local regulatory agencies. Status . During our April 1978 meeting with the ACRS, we stated that the State environmental radiological protection program was not sufficiently -i funded to warrant NRC funding support. Since April, the staff has had continuing dialogue with the State representatives, and presently plans to send staff representatives to - the State in the near future to discuss several alternatives.which may . allow some NRC funding support for the State program. The problem has been and continues to be a lack of adequate funding by the State. . . P l (;
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- 4. Hermetic Seals-
. Committee Recommendation The Comnittee notes that post-accident operation of the plant to maintain safe shutdown conditions may be dependent on instrument: tion - ,
and electrical equipnent within containment which is susceptible to l ingress of steam or water if the hermetic seals are either initially
~ 1 defective or should' become defective as' a resditiof damage or aging.
The Committee believes that appropriate test and maintenance procedures should be developed to assure continuous lonc-term seal capability. i Status This is an ACRS generic concern, number IId-2. It also is incorporated in the staff's Technical Activities Program as Task C-1. At such time as results of Task C-1 indicate that some action ,s ! necessary for Davis Besse 1, we plan to react to this Conmittee ' recommendation. .!
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- 5. Instrumentation to Follow the Course of an ' Accident ~ l
' Committee Recommendation l The Committee recommends that, prior to comercial power operation of Davis-Besse, Unit 1, additional means for evaluating the cause and likely course of various accidents,, including those 'of very low probability, should be ~in hand in order: to. provide improved bases for timely decisions-concerning possible off-site emergency measures. 'Ihe Committee wishes to be kept informed..
Status J This. matter has-to do with implementation of Regulatory Guide.l.97. At such time as a decision is made regarding implementation of this Regulatory Guide on operating plants, we will' implement it on Davis
. Besse 1.
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- 6. ATHS C_ommittee Recommendation The question of whether.the design of this plant must be nodified in order to comply with the requirements of WASH-1270, " Technical Report on Anticipated Transients Without Scram' (A'IWS) for ' dater-Cooled- Reactors," - <
remains an' outstanding issue pending the NRC Staff completion of its review of the Babcock and Wilcox generic analyses of ATHS. -The Committee recommends that the NRC Staff, the Applicant, and the Babcock and Wilcox Company continue to strive for an early resolution of this matter in a manner acceptable to the NRC Staff. The Committee wishes to be kept inferned. Status , This matter is addressed in the SER, p. 7-3, and in SER' Supplement No. 1,
- p. 18-4. The licensee has made appropriate commitments regarding ATWS to allow licensing.
At such time as the ATWS issue is resolved for operating plants, we will ~ require appropriate modifications for Davis Besse 1. e
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- 7. Cy-Pass Loop 4
Committee Recommendation Davis-Besse, Unit 1, has installed a bypass loop containing two manually operated valves around the decay heat rem.nl system suction line iso-lation valves. The nor: rally closed bypass valves muld be opened in the event of a spurious closure of one of the decay heat remval system suction line isolation valves during systen operation. The Comnittee recommends that further attention be given to the means employed for iscr lation of the low pressure residual heat remval system from the pri:rary system while the latter is pressurized, and that reliable means be developed to assure such isolaticn. 'Ihis matter should be resolved in a manner sat- - isfactory to the-NRC Staff. The Comittee wishes to be kept informed. Status Design Modification Alternatives to the Present Key Lock Control in , Manual Bypass Valves DH 21 and DH 23 Licensee condition 2.C.(3)(p) requires that the licensee submit an analysis of design modification alternatives for tne 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 installa-tion of approved desfgn modifications shall be completed prior to startup following the first scheduled refueling outage. Evaluation will be made by the Division of Operating Reactors. Management responsibility will be carried out by Operating Reactors Branch No. 4. (
Reference:
SER Supplement No. 1, p. 5-5 and p. E-3) l l
' 8. Fire Protection Co mittee Recomme,ndation The Committee supports the NRC Staff program for evaluation of fire pro-tection in accordance with Appendix A to Auxiliary and Power Conversion Systems Branch Technical Position 9.5-1, " Guidelines for Fire Protection for Nuclear Power Plants." The Committee recommends that the NRC Staff give high priority to the completion of both owner and staff evaluations and to recommendations for Da'ris-Besse, Unit 1, and for other plants nearing com-pletion of construction in order to maximize the opportunity for improving fire protection while areas are still accessible and changes are more feasible.
Status Reevaluation of Fire Protection Program 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 Ruview 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 date of issuance of NPF-3. License condition 2.C.(3)(h) also requires that the licensee implement Section B of Appendix A, " Administrative Procedures, Controls, and Fire Brigade," and Section C of Appendix A, " Quality Assurance Programs," prior to startup following the first regularly scheduled fueling outage. l By letter dated August 29, 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. l i On October 13, 1977, a meeting was held with the licensee where the staff addressed the inadequacles of the licensee's Fire Hazard Analysis Report for Davis-Besse, Unit 1 submitted on Feburary 11, 1977. The licensee stated that they would resubmit an amended Fire Hazard Analysis Report in November 1977. l A meeting was held on December 6, 1977 where the licensee and the Auxiliary System Branch discussed the facility design for fire pro-tection in the cable spreading room.
L -. 9 On. January 11, 1978 the. licensee submitted-Revision No. 11to the Fire Hazard Analysis Report for Davis Besse, Unit 1. Revision No. 1;was ' ' foundLto be acceptable and is presently under detailed review by the- i staff and staff's. consultants.: A fire protection site visit was : completed on May 23-25, 1978, and staff requests for information-were-
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issued to the licensee on July 6, 1978. Om August'1, 1978, a meeting. ' was held with the licensee to clarify certain staff requests .for. information. the Davis - Major milectones Fire Protection Review scheduled are: (1) licensee for completing's to response-Besse, staff l requests- Unit 1,: , for information. received on or before September 27, 1978; (2) meeting
- with licensee'to resolve any remaining open items - October 25, 1978; !
(3) DSS and DPM SER input to ORB No. 4 - December 1, 1978; and (4) l issuance of Davis-Besse, Unit 1, Fire Protection Evaluation Report - (' January 15, 1979. Staff review and evaluation required to complete this task will be by
- Division of Systems Safety and Division of Project' Management. DSS -
reviewer is V. Leung (ASB) and DPM reviewers are F. Allenspach and-J. Conway (QAB) and J. Holman (0LB). (Note: The licensee states that in thethey Fire ~have Hazards committed Analysis Report,'to meeting TableBTP 4.1, sheet 9.5.1,.
- 8. Fire
Management Brig)ade Training, responsibility will be carried out.by' Operating Reactors Branch No. 4. (
Reference:
SER, p. 9-14 and.9-15, SER Supplement No. 1,
- p. 9-1)
Interim Fire Protection Technical Specifications The licensee was notified on November 21, 1977 by telephone and telecopy of our requirements for implementing interim fire protection Technical Specifications for plant systems and administrative proce-dures. A letter dated November 28, 1977 was sent to licensee providing the the sample standard Technical Specifications and requesting the licensee's response by December 7, 1977 for an application for license amendment with the plant specific interim technical specifications. The licensee submitted proposed interim Technical Specifications on December 12, 1977 and on March 22, 1978 (Amendment No. 9) interim fire technical specification was issued for the presently installed fire protection equipment at the facility, as has been done for other I operating facilities. The interim Technical Specifications deal only with administrative surveillance and corrective steps to reduce the likelihood of damaging fires pending our final review of the fire protection of Davis-Besse, Unit 1. Resolution of this matter will be by Division of Systems Safety. 055 reviewer is V. Leung (ASB), and the schedule for completing this task , is the same as specified for completing the Fire Protection Review (see Item 5, Enclosure 1). Management responsibility will be carried ; out by Operating Reactors Branch No. 4. w m e, we r w ee ,. -. , *- + a "w+ +>+=e* -re-m+ v e = + r e %- r==-*-%3w-e=,w--- :o , w-- , .- ur p re o w - e ar = == #wr,- - o s w w-i-+ m--* mrm e w e= -~ e w =wser wm n w c V e *
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. . 9. . Sabotage Committee Recommendation The Ccanittee believes that the Applicants and the NRC Staff should fur--
ther review a rity provisions for Davis-Besse, Unit 1, for measures-that could significantly reduce the possibility and consequences of sabo-tage, and that such measures should be implemented where practical.. Status The Licensee's 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 Mcdified. Amended 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. 8 e t 9 j l I
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gj - ADVISORY COMMITTEE ON REACTOR SAFEGUAP')S
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..... January 14, 1977' 4
Bonorable Marcus A. Rowden
. Chairman i U.S.' Nuclear Regulatory Commission Washington, DC 20555 ,
Subject:
REPOle CN DAVIS-BESSE NUCLEAR POWER STATICN, UNIT.1 .
Dear Mr. Rowden:
At its 201st meeting, January 6-8, 1977, the Advisory Comittee~ on Reactor ; Safeguards completed its review of the application by the Toledo' Edison Company and the Cleveland Electric. Illuminating Company for a license to , operate the Davis-Besse Nuclear Power Station, Unit 1. Members of the. , Comittee' visited the plant on May 18,.1976, and a subcomittee meeting was held in Washington, D.C. on December. 21,:1976. During:its review, the Comittee had the benefit of discussions with. representatives and
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consultants of the Applicant, .the Brlocock and Wilcox Company, the Bechtel.. Corporation, .and the NRC Staff. The Comittee -also had the benefit'of , the documents listed. The Committee reported:on.the application for: a construction' permit for this. unit on August 20, 1970. The Davis-Besse Nuclear Power Station,. Unit-1, is located on the south - ' western shore of Lake Erie "about midway between the ' cities of Toledo and ' Sandusky, Ohio. The mini == exclusion distance is 2400 ft.. The low population zone, with a radius of two miles, included about 870.~ people in the 1970 census. The nearest population' centers'are Toledo (1970 popula- ; tion 383,818) and Sandusky (1970 population 32,674), both;about 20 miles - .j from the plant. .1 1 I' The nuclear steam supply system employs a Babcock and Wilcox pressurized water reactor similar in most respects to those first used in the Oconee Nuclear Station. This system differs fra the Oconee units' and several other similar units in that the steam generator loops aie raised about 30 ft above the'1evel in the original plant arrangement. Although this: change was made to eliminate the need for internal vent valves, four such-valves are provided because of their beneficial effect in reducing steam binding following a postulated loss-of-coolant ~ accident. ' \ _,,,,,,,_._,,,.___._,,.._.__..,_...;,__.._..__._.._,_..,_._..__,__....,.._,_....,_,~.a_ t_ _ _ , , , _ . , _ . . . . . . . . .
Eonorable Marcus A. Fc den Janua ry 14, 1977 The proposed power level for the unit is 2772 Kdt, as compared to 2633 Kdt proposed at the construction permit stage. This higher pcver level is the same as that proposed for the Pwho Seco ard Three Mile Island, Unit 2 reactors, both of which have been reviewed by the NRC Staff and the Comittee and found acceptable. The structures and components of Davis-Besse, Unit 1, were designed for a Safe Shutdown Earthquake (SSE) acceleration of 0.15g at the fourdation level. Because or changes in the regulatory approach to selection of seis-mic design bases, the Comittee believes that an acceleration of 0.20g e would be note appropriate for the SSE acceleration at a site such as this in the Central Stable Region. The Applicant presented the results of preliminary calculations concerning the safety margins of the plant for an SSE acceleration of 0.20g. The Comittee recommends that the NRC Staff review this aspect of the design in detail and assure itself that signifie cant nargins exist in all systems required to acconplish safe shutdown of the reactor and continued shutdown heat ret. oval, in the event of an SSE . at this higher level. The Ccmittee believes that such an evaluation need not delay the start of operation of Davis-Besse, Unit 1. The Comittee wishes to be kept informd. The performance 'of the Emergency Core Cooling, System (ECCS) has been evaluated using a Babcock and Wilcox evaluation nodel applicable to the raised-loop configuration. The NRC Staff has reviewed these evaluations and has determined that certain assumptions regarding return to nucleate boiling do not comply strictly with tne provisions of Appendix K to 10 CFR Part 50. The NRC Staff is also reviewing several other areas relating to ECCS performance. These matters should be resolved in a manner satisfactory to the NRC Staff. In conjunction with the evaluation and assessment of the impact of routine waste releases from this plant, the' Cenmittee recomends that the NRC Staff provide leadership in encouraging the development of improved environmental radiation surveillance capabilities on the part of the State of Ohio and appropriate local regulatory agencies. The Comittee notes that post-accident operation of the plant to maintain safe shutdown conditions may be dependent on instrumentation and electrical equipment within containment which is susceptible to ingress of stea:n or water if the hermetic seals are either initially
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' Honorable Marcus A. Powden . January - 14, 1977 defective or should become defective as a result of darage or aging.
The Comittee believes that appropriate test and raintenance procedires , should be developed to assure continuous long-term seal capability. The Comittee recorrnends that, prior to comercial power operation of Davis-Besse, Unit 1, additional means for evaluating the cause and likely course of various accidents, including L.ose of very low probability, should be in hand in order to provide improved bases for timely decisions . concerning possible off-site emergency measures. We Comittee wishes to be kept informed. The question of whether the design of this plant must be nodified in order to comply with the requirements of L'SH-1270, " Technical Report on Anticipated Transients Without Scram (AWS) for Water-Cooled Reactors," remains an outstanding issue pending the NRC Staff conpletion of its , review of the Babcock and Wilcox generic analyses of AThs. The Comittee recommends that the NRC Staff, the Applicant, and the Babcock and Wilcox Company continue to strive for an early resolution of this matter in a manner acceptable to the NRC Staff. The Conmittee wishes to be kept
, informed.
Davis-Besse, Unit 1, has installed a bypass loop containing two ranua11y operated valves around the decay heat removal system suction line iso-lation valves. The normally closed bypass valves would be opened in-the event of a spurious closure of one of the decay heat removal system suction line isolation valves during system operation. The Con:rittee recommends that further attention be given to the means exployed for iso-lation of the icw pressure residual heat renoval system from the primary system while the latter is pressurized, and that reliable means be developed to assure such isolation. n is matter should be resolved in a manner sat-isfactory to the NRC Staff. The Conmittee wishes to be kept informed. The Comittee supports the NRC Staff program for evaluation of fire pro-
- tection in accordance with Appendix A to Auxiliary and Pcwer Conversion Systers Branch Technical Position 9.5-1, " Guidelines for Fire Protection for Nuclear Power Plants." The Connittee recon
- aends thati the NRC Staff give high priority to the completion of both owner and staff evaluations and to recomendations for Davis-Besse, Unit 1, and for other plants nearing com-pletion of construction in order to maximize the opportunity for i. proving -
fire protection while areas are still accessible and changes are note feasible. ! I 1
Bonorable Marcus A. Rowden January 14, 1977 l 1 l l
'Ihe Comnittee believes that the Applicants and the NRC Staff should fur- !
ther review security provisions for Davis-Besse, Unit 1, for measures l that could significantly reduce the possibility and consequences of sabo-tage, and that such measures should be implemented where practical. Other generic problems are discussed in the Courrittee's report, " Status of Generic Items Relating to Light Water Reactors: Report No. 4," dated April 16, 1976 (Attached). Those problems relevant to the Davis-Besse, Unit 1, should be dealt wi1 by the NBC Staff and the Applicant as solu-tions are found. The relevant items are: II-1, 2, 3, 4, 6, 7, 9, 11; " II.A-1, 4, 5, 7, 8; II.C-1, 2, 3, 4, 5, 6. The Advisory Connittee on Reactor Safeguards believes that, if due re-gard is given to the items mentioned above, a:d subject to satisfactory coupletion of construction and pre-operational testing, there is reason-able assurance that the Davis-Besse Nuclear Power Station, Unit 1, can be operated at power levels up to 2772 MNt without undue risk to the " health and safety of the public. Sincerely ours, 9 M. Bender Chairman
Attachment:
Status of Generic Items Relating to Light Water Reactors: Report No. 4 dated April 16, 1976
References:
- 1. Davis-Besse Nuclear Power Station, Unit 1, Final Safety Analysis Report (March 1973) with Revisions 1 through 24.
- 2. Safety Evaluation Report (NUREG-0136) in the mtter of the Davis-Besse Nuclear Power Station, Unit 1 (December 1976) .-
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.g 7 ?/i 7, NUCLEAR REGULATORY COf."/. ES!ON
- . s;[A 3.g/.:, j ADV :;ORY COMMITTEE CJ 7.EscioR 5 T E GU'.RDS
% f . msw:cTos. o. c. miss August 25, 1978 1
l I Honorable Joseph M. Hendrie Chairman U. S. Nuclear Regulatory Commission
. Washington, DC 20555 ]
SUBJECT:
REPORT Ot1 ACRS ACT1VI"'1ES: MAY-AUGUST, 1978
Dear Dr. Hendrie:
This is a brief report of ACRS activities during May, June, July, and August 1978. Selected topics in this report will be discussed during the next joint NRC-ACRS meeting. Procosed Use of CRAC Code in Site Comcarisons The Committee has considered the NRC Staff's proposed use of the con-sequence model (the "CRAC" Code), which was developed for the Reactor Safety Study (WASH-1400), for evaluating environmental impacts of al-ternatives to sites with relatively high population density. Studies to date have shown that the CRAC Code can provide additional understanding of the public health impacts of accidents exceeding the limits of 10 CFR 100. However, there are many factors influencing the application of the code that have an imtertant bearing on the cceputa-tional results but which the code does not address adequately. These include regional meteorology (particularly for coastal and river valley sites), plume geometry, and effluent particle size distribution. ~ In addition, the code does not address the oehavior of radionuclides within containment prior to release. Also, the probability of various release magnitudes remains a factor having considerable uncertainty. For this reason, the ACRS recommends caution in the use of the code as a determining basis for judgment in alternative site evaluations. Nevertheless, the CRAC Code is one of the more useful rrethods currently available for evaluating environmental impacts of alternative sites,
;nd efforts should be continued to develop improved input data for the .l Code.
,5;>norable Jcceph M. Badrie . ./. gust 25, 1973 Dynamic Icadina Combinations _ .
The Cc=ittee has attempted on several recent occasions, including its 218th meeting, June 1-2, 1978, to encourage the Office of Nuclear Reactor Regulation to reconsider the rationale for establishing design basis loadings and loading combinations in p=rformirg safety analyses. The ACRS recommends that such a reevaluation be undertaken as soon as possible. , NRC Staff Response to ACRS Recorr;tendations During its April 1978 meeting the Comittee heard a report regarding ~ action taken in connection with ACRS recommendations in its January 14, r 1977 report to you on operation of Davis-Besse Unit 1. Of the nine {
/
specific recommendations made by the Committee, action had teen recently initiated on four and only preliminary work had been started on the other five. . On March 12, 1976, the Committee recommended prompt implementation of ' Reactor Ptnp Trip (RPT) for iERS. Iatters were sent to E'G licensees in May of 1978 by the NRC Staff specifying criteria for RPr design. In order to obtain earlier responses to Committee recommendations, a follow-up system has been instituted which involves su:rrarizing Ccciait-tee requests within 6 months and a formal response by the Staff there-after. Reply to Congressman Udall The Comittee has forwarded a reply to Congressman Udall's inquiry about the need for a statutory board to review reactor safety tratters, patterned after the National Transportation Safety Board. In its reply, the Com.nittee stated that it did not believe that such a board was necessary, based on the assigned responsibilities and current activi-ties of the NRC and ACRS and on the history of operation of commercial nuclear power plants. Nuclear Plant Reliability Data System (NPRDS) The Committee has considered a proposal to make a reliability data collection system mandatory for all licensees. In a memorandum to the Executive Director for Operations dated July 12, 1978, the Ccemit-tee concluded that the existing system was providing data at a reason-able rate, but suggested also that the system could be improved by analyses of the internal consistency, usefulness, and reasonableness of the data already on hand. The Committee did not recommend that hTPDS reporting be made mandatory. .
. v----
e .. m-
' Ib:."M e Cr =eph M. MaMrie
. Jagast 25, 1973 Mic entation of RG 1.97, Revision 1, " Instr nentation to Fo13cw the Course of an Accident" Tne Committee has for some time been urging improvements in instrumen-tation to follow the course of accidents. In August 1977, Revision 1 to PG 1.97 was published but no firm schedule for its implementation hasasyethenestablished. We NRC Staff is working with selected licensees to develop approaches to implement the Guide. Se Committee ,
il.1 continue to monitor the implementation of this Guide. l Anticipated Transients Without Scram (AWS) . The Committee's ANS Subcom:nittee is proceeding with its review of the recently published NtTPSG-0460 " Anticipated Transients Without Scram." The Committee hopes to reach a conclusion in this area by the end of 1978. Generic Items . Ce Committee is planning the customary semiannual reevaluation of its list of unresolved generic items, and is planning to initiate a de-tailed review of the implementation of those items that are considered to have been resolved genericall,y. IDCA-ECCS Research Procrams The RES Staff has proposed that current IDCA-ECCS research programs be carried to completion, but that no new large LOCA-ECCS test facilities (e.g., the EBTF, Multi-Purpose Test Facility, etc.) be constructed. They have also recommended that the cooperative program with Japan, France and the Federal P,epublic of Germany be expanded. D e ACES con-curs in these suggestions. Future Schedule . 222nd ACRS MEETING OC'IOBER 5-7, 1978 l - ( No Projects Scheduled l l m
. vm
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- f:noni2e Joseph M. Es.*.5rie --
E;ds: 25, 1973
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223rd ACRS MEETING I b"JVEMBER 2-4, 1978' Zirener (OL) , 0;RPR Sarety Evaluation Report 9/1/78 ACPS Subcomittee Meeting ' KPS report 11/9/78 1 BDPPSAR/BSAR-205 (PDA) QER Safety Evaluation 10/2/78 ' ACRS Subcontnittee Meeting ACRS Report 11/9/78 GETR (Special Review) QiPR Safety Evaluation Report -* ACPS Subcomittee Meeting , ACRS Report 11/9/78 Sincerely yours, h ley A' Stephen Lawroski Chairman e i is
*'Ib be scheduled -
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- - Enclosure 4 j# 5 UNITEo ST ATEs NUCLEAR REGULATORY COMMISSION q' yw. j,4,.
WASHINGTON, D. C. 2055s j l.k..C ks.,kbh,]JJ
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Docket No. 50-346 MEMORANDUM FOR: Victor Stello, Jr., Director Division of Operating Reactors, NRR FROM: Roger S. Boyd, Director Division of Project Management, NRR j
SUBJECT:
TRANSFER OF DAVIS-BESSE, UNIT 1 FROM DIVISION OF PROJECT MANAGEMENT,-LIGHT WATER REACTORS BRANCH NO. 1, TO DIVISION OF OPERATING REACTORS, BRANCH NO. 4 i Effective on the date of this memorandum, the project management responsi- - bility for Davis-Besse, Unit 1, is transferred from Light Water Reactors Branch No. 1, Division of Project Management, to Operating Reactors Branch No. 4, Division of Operating Reactors. Also, the environmental responsi-bility for Davis-Besse, Unit 1, is transferred from the Division of Site Safety and Environmental Analysis to the Division of Operating Reactors as delineated in Enclosure No. 4 to this memorandum.- The licensees, the Toledo Edison Company and the Cleveland Electric Illuminating Company, received a facility operating license (NPF-3) on April 21, 1977 which authorized full power operation at 2772 MWt. However, the operation of the facility was restricted to a sequence of operational modes until preoperational tests, startup tests, and other items were completed to the satisfaction of I&E and NRR. A chronology for the issuance of the amendments to license NPF-3 and the date of authorization for proceeding to sequential operational modes is presented below. Also, a brief description is provided of the amendments presently issued for license NPF-3. CHRONOLOGY
~
- 1. License NPF-3 (Mode 6-Fuel Loading) Apri1 22, 1977
- 2. Mode 5 - Cold Shutdown May 10, 1977
- 3. Amendment No. 1 May 27, 1977
- 4. Amendment No. 2 June 14, 1977 t
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1 a i 1 I j 1 I I 1 Amendment No. 3 June 24, 1977
- 5. -
June 30, 1977
- 6. Mode 4'- Hot Shutdown.
- 7. Amendment No. 4 July 8, 1977
- 8. ' Mode'3 - Ho't !Standby . July 8, 1977.
s
- 9. Amendment No. 5 July 21, 1977 !
- 10. Mode 2 --Startup. August.9, 1977 -
- 11. Reactor Critical August 12, 1977 j
* .j August 26', 1977
-12. Amendment No. 6 ,
- 13. Mode 1 - Power Operation August 30,-1977 -
r November 29, 1977 14 Amendment No. 7 , i February 28, 1978-
- 15. Amendment No. 8 March 23, 1978
- 16. Amendment No. 9 ,
Amendment No. 10 May 26; 1978
- 17. '
June 16, 1978 t
- 18. Amendment No.111 Amendment No. 12 August 18, 1978 19.
September 29, 1978
'20. Amendment No. 13 Amendment No. 1 for NPF-3 revised the Technical Specifications.to allow !
decay heat removal train pump switching operations for the purpose of testing in operational Modes 3, 4 ana 5. . Amendment No. 2 revised license NPF-3 by removing license condition ;
'2.C.(3)(o) which temporarily restricted facility operation.to Mode 3 until '
the necessary modifications were implemented for ensuring that the decay
~
heat removal relief valve would activate prior to automatic closure of the decay heat removal system isolation valves. (1) JAmendment No. 3 revised the Technical Specifications as follows: correcting administrative errors, (2) allowing surveillance testing of atmospheric vent valves ICS MA and ICS MB in Mode 4, (3) allowing surveil- i
;1ance testing for verification that the annulus space can be depressurized to one quarter inch. negative pressure within four seconds from the time that the emergency ventilation system fans attain 8000 cubic feet per w -. . - _ _ _ . , - . . - . - , _ , ~ _ _ . ~ . . . _ _ ,. _ . . _ . - . _ - . ~ . . _ - . . . _ _ . . . _ . , ,
3 -- P minute flow, ~(4) verifying that automatic isolation and interlock action of
'the' decay heat removal system from the reactor coolant. system wi l l occur :
with a simulated reactor coolant system pressure.of: greater than or-equal to 413 pounds per square inch gauge, and (5). allowing the interlock on either decay heat isciation valve to be taken out of service in order to perform the Channel Calibration or Channel Functional Test Requirements ' required in Modes 4 and 5. Amendm'ent No. 3 also revised license'NPF-3 by-revising license condition 2.C.(3)(j) to. allow the performance of specific or.preoperational tests-on decay heat removal valves DH li.and DH:12 and ; requiring that_an operator be stationed.in.the control room and' assigned to monitor' flow rates in the: decay-heat removal trains for-those periods of time when surveillance testing is being performed or.when a standby decay heat removal train is being brought on line. Amendment No. 4 revised the Technical. Specifications to reflect the proper: allowable values and surveillance. requirements lfor the steam generator level transmitters which had been installed in lieu of the level switches , in the facility steam and feedwater rupture. control system. ; Amendment No. 5 revised license NPF-3 by deleting license condition 2.C.(3)(c) which stipulated the time' allowed from date of issuance of NPF-3 for completing;the installation of a second oxygen monitor'in the gaseous radwaste system thereb'y providing red'undant oxygen monitors that alarm locally and in the control room at specified set points. . Amendment No. 6 revised license NPF-3 by deleting license condition . 2.C.(3)(b) which~ stipulated the' time allowed from date of issuance of NPF-3 for completing the installation of the modified seismic Category 1 emergency diesel fuel oil and storage transfer system. Amendment No. 7 revised the Technical Specifications as follows: (1) the allowable trip setpoint was revised to 7.0 + 1.5 seconds for the Sequence Logic Channel of the Essential Bus Feeder Trip, and (2) the surveillance '! l testing frequency for the Rosemont reactor coolant system pressure transmitters was changed from once every four months to once every l eighteen months. Amendment No. 7 revised license c odition 2.C.(3)(k) by removing the stipulations within condition 2.C.(J)(k) requiring that acceptable noise l test procedures be provided within four months from issuance of NPF-3. Also, license condition 2.C.(3)(1) was revised by removing the stipula-tions within-2.C.(3)(1) for providing large break spectrum analyses within l
-six (6) months from issuance of NPF-3. .
Amendment No. 7 deleted two license conditions 2.C.(3)(m) and 2.C.(3)(q) < from NPF-3 which stipulated the period of time allowed to provide additional analysis and/or modifications in the facility design. , P -, - ~, ,, ,4, ,,o., ..#.-m. . - , - ~ . - ,e ,,,,,L..e , ..E._.-,,m.- .,--omr- - - - e. .~, - - . - - . . - . . - ... . + me .* --.i---
License condition 2.C.(3)(m) required making changes to the existing low pressure and high pressure injection flow system which would provide a flow indication syster hich was seismically qualified and powered from essential power source. and with flow indication in the main control room. License condition 2.C.(3)(q) required the submittal of an evaluation and modifications proposed to assure that required Class lE equipment would operate properly in the event of offsite grid degradation. Amendment No. 8 deleted the requirements for an Annual Operating Report in , the. Technical Specifications in order to be consistent with recent . Commission guidance. Amendment No. 9 revised the Technical Specifications to incorporate limiting conditions for operation and surveillance requirements for existing fire . protection systems and administrative controls. The interim fire Technical l Specifications were stipulated.to become effective 30 days after issuance of Amendment No. 9. . Amendment No. 10 revised license NPF-3 by deleting license condition 2.C.(3)(n) which stipulated the time allowed from date of issuance of NPF-3 for completing the installation of flow measuring devices for boron dilution control. Amendment No. 11 revised the Technical Specifications to allow full power operation for the duration of Cycle No. 1 with all Burnable Poison Rod
, Assemblies and all but two of the Orifice Rod Assemblies removed from the core. Amendment No. 11 revised license NPF-3 by deleting license condition 2.C.(3)(i) which stipulated the penalties for rod bow effects on the departure from nucleate boiling ratio. Amendment No. 11 also revised the Technical Specification regarding the modification of alarm setpoints on quadrant tilt to accommodate a recently discovered increase in the measurement error associated with this quantity.
Amendment No. 12 revised the Administrative Control Section of the Technical Specifications to reflect changes in the corporate structure of the Toledo Edison Company. Amendment No. 13 revised the requirements for nonroutine environmental operating reports in order to make them consistent with Regulatory Guide 4.8, p
" Environmental Technical Specifications for Nuclear Power Plants, December 1975."
The current status of items requiring futher staff actions and the organi-zations responsible for completing these items are identified in Enclosure 1. Lists of Generic Concerns and Regulatory Guides used during the license review, with reference +o the locations where relevant information or evaluations.of records , e found, are provided in Enclosure 2 and 3, respectively.
5-By copy of this memorandum, Division of. Systems Safety, Division.of Site Safety and Environmental Analysis, Office of Inspection and Enforcement,. Office of Management Information and Program Control, Office of Executive Legal Director, Regulatory Files, Public Information and'Public Proceedings are being notified of-the following changes in management responsibilities which are effective per date of this memorandum. FROM TO Project Manager L. B. Engle G. Vissing Branch Chief J. F. Stolz R. Reid ~ Assistant Director D. B. Vassallo B. Grimes Licr.nsing Assistant E. H. Hylton - R. Ingram
~
Roger S. Boyd, Director uirector of. Project Management Office.of Nuclear Reactor Regulation
Enclosures:
- 1. Current Status of Items - -
Requiring Staff Action
- 2. Current Status of Generic Review Items
- 3. Regulatory Guides Used.
During Licensing Review
- 4. DSE Transfer Memorandum (November 21,1977) t
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__ _% __._.._,.a _- - e ., , ENCLOSURE 1 ; CURRENT STATUS OF ITEMS REQUIRING STAFF ACTION .t DAVIS-BESSE, UNIT 1 DOCKET NO. 50-346 FACILITY OPERATING LILL,4SE NPF-3 , The items requiring.further staff action are as follows: (Note - A letter. was sent.to the licensee on September 20, 1978 requesting schedule dates for required submittals of analyses, modifications, Land. Technical Speci_-- fications as. identified in Enclosure.1 which must be completed prior.to startup.following th~e first. scheduled refueling outage.).
- 1. Overpressurization Protection for the Reactor Coolant System License condition 2.C.(3)(d) of Facility Operating Licen'se'NPF-3' requires the li.censee to install a.long term'means of protection. ;
against reactor coolant-system overpressurization. The. installation ( shall. be completed prior to startup following the first scheduled > refueling. - By letter date April 7, 1977, the licensee responded with proposed-modifications to meet the long-term provisions for overpressure , protection. By. memorandum dated October 5, 1977 theLReactor S
. Branch provided their evaluation and approval.of:the licensee'ystem s
-proposal to install-pressurizer heater interlocks and to remove power to the decay heat removal isolation valves DH 11_and DH 12.during-decay heat removal operation. On. March 30, 1978, the licensee sub- .
mitted requested changes to the Technical Specifications regarding' ! the proposed modifications. The licensee stated on July 7, 1978 that i the proposed changes will be completed prior to startup following the first scheduled refueling outage. An augmented SER prepared by RSB, DSS approving the overpressure protection and Tech.nical Specification y changes is scheduled for issuance on October 30, 1978. :An amendment 2 9 is also in preparation-approving-the proposed. Technical Specifications which will be implemented when the' modifications are completed. Responsibility for issuance of- the amendment will be either Light , Water Reactors Group No. 1, DPM or Operating Reactors Branch No. 4,: ' whoever,has responsibility for the facility'when the amendment'is ready lfor issuance. (
Reference:
.SER Supplement No. 1, p. 5-2, and memorandum from D. F. Ross to D. Vassallo dated October 5,-1977) q
~ . - - - . _ . . . . - - . 'e I
- 2. Reactor Coolant System Flow Indication j License, condition 2.C.(3)(e) of NPF-3 req" ires the licensee to modify the flow indication of the reactor coolant system to meet the single failure criterion with regard to the pressure sensing lines to the flow differential pressure transmitters. This modification shall be completed prior to startup following the first scheduled refueling outage. Evaluation will be made by the Division of Operating Reactors and management respc..sibility will be carried out by Operating Reactors Branch No. 4. '(
Reference:
SER, p. 7-2)
- 3. - Diverse Source of Power for the Auxiliary Feedwater System License condition 2.C.(3)(f) of NPF-3 requires the licensee to modify -
the auxiliary feedwater by providing diverse direct current power to one of the redundant' auxiliary feedwater trains. This modification shall be completed prior to startup following the.first scheduled ' refueling outage. The licensee's proposed modifications have been approved by the' staff. Management responsibility will be carried out by Operating Reactors Branch No. 4. (
Reference:
SER, p. 9-7 and 9-8, and SER Supplement No. 1, p. 7-7)
- 4. Automatic Alignment of High Pressure Injection Pump Suction With the Low Pressure Injection Dischargi License condition 2.C.(3)(g) of NPF-3 requires the licensee to modify the emergency core cooling system by providing motor operated valves in lieu of the manually operated valves in each of the two crossover connection lines installed between the high pressure makeup pump suction and the low pressure injection discharge. The motor operated valves shall provide control and position' indication in the control room'. The modifications shall be completed prior to startup following the first scheduled fueling outage. Evaluation will be conducted by Division of Operating Reactors. Management responsibility will be carried out by Operating Reactors Branch No. 4. (
Reference:
SER,
- p. 6-8 and SER Supplement No. 1, p. 6-7)
- 5. Reevaluation of Fire Protection Program ,
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 protectisn as stipulated in item 2.C.(3)(h) shall be completed within three (3) years from the date of iseuance of NPF-3.
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4 license condition 2.C.(3)(h) also requires that the licensee implement Section B of Appendix A, " Administrative Procedures, Controls, and Fire Brigade," and Sectinn C of Appendix A, " Quality Assurance Programs," prior to startup following the first regularly scheduled fueling outage. By letter dated August 29, 1977, the licensee was provided a copy of NRC document, " Nuclear Plant Fire Protection Functional Responsibili-t:.s, 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 licensee where the staff addressed the inadequacies of the licensee's Fire Hazard Analysis Report for Davis-Besse, Unit 1 submitted on Feburary 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 where the licensee and the Auxiliary System Branch discussed the facility design for fire pro-tection in the cable spreading room. On January 11, 1978 the licensee submitted Revis hn No._1 to the Fire Hazard Analysis Report for Davis Besse, Unit 1. Revision No. I was found to be acceptable and is presently under detailed review by the staff and staff's consultants. A fire protection site visit was completed on May 23-25, 1978, and staff requests for information were issued to the licerisee on July 6, 1978. On August 1, 1978, a meeting was held with the :icensee to clarify certain staff requests for-information. Major milestones scheduled for completing the Davis-Besse, Unit 1, Fire Protection Review are: (1) licensee's response to staff requests for information received on or before September 27, 1978; (2) meeting with licensee to resolve any remaining open items - October 25, 1978; (3) DSS and DPM SER input to ORB No. 4 - December 1,1978; and (4) issuance of Davis-Besse, Unit 1, Fire Protection Evaluation Report - January 15, 1979. , Staff review and evaluation required to complete this task will be by Division of Systems Safety and Division of Project Management. DSS reviewer is V. Leung (ASB) and DPM reviewers are F. A11enspach and J. Conway (QAB) and J. Holman (0LB). (Note: The licensee states that they have committed to meeting BTP 9.5.1, Fire Brigade Training, in the Fire Hazards Analysis Report, Table 4.1, sheet 8.) Management responsibility will be carried out by Operating Reactors Branch No. 4. (
Reference:
SER, p. 9-14 and 9-15, SER Supplement No. 1,
- p. 9-1) i
- 6. Potential for an Inadvertent Closure of a Decay Heat Removal System Valve During Shutdown Operations License condition 2.C.(3)(j) of NPF-3, as amended by Amendment No. 3 to NPF-3, requires until such time as final resolution is obtained regarding the potential for and consequences of an inadvertent closure of a decay heat removal system valve during shutdown operations, the licensee is required to maintain power on decay heat removal valves DH ll and DH 12 and can only operate one decay heat removal train at -
a time. The condition does not preclude performance of specific surveillance or preoperational test requirements. For the period of times during which only one decay heat removal train is available or a standby heat removal train is being brought on line, an operator is required to be stationed in the centrol room to monitor flow rates and be available to immediately secure the reactor heat removal pumps should - loss of flow occur due to the inadvertent closure of DH 11 or DH 12. Condition 2.C.(3)(j) interfaces with Item 1, condition 2.C.(3)(d). Necessary actions required for resolution of these matters are the same as stated in Item 1 of this enclosure. Management responsibility will be carried out by Operating Reactors. Branch No. 4. (
Reference:
SER Supplement No. 1, p. 5-4; Amendment No. 4 to Facility Operating License NPF-3, Safety Evaluation)
- 7. Desion Modification Alternatives to the Present Key Lcck Control in Manual Bypass Valves DH 21 and DH 23 Licensee condition 2.C.(3)(p) 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 isolatien of the decay heat removal loop is required. The submitted analysis and installa-tion of approved design modifications shall be completed prior to startup following the first scheduled refueling outage. Evaluation will be made by the Division of Operating Reactors., Management responsibility will be carried out by Operating Reactors Branch No. 4. (
Reference:
SER Supplement No. 1, p. 5-5 and p. E-3)
- 8. Seismic Reanalysis License condition 2.C.(3)(r) requires the licensee to submit a seismic reanalysis and evaluation to the Commission for its review and approval of the adequacy of the facility systems needed to accomplish safe shutdown of the reactor and continued shutdown heat removal prior to startup following the first scheduled refueling outage.
5-In performing the reanalysis, a safe shutdown earthquake acceleration.
-of 0.~20g shall.be applied at.the. foundation. level of the plant and the response spectra shall be used as specified in Regulatory Guide 1.60.
Guidelines for the seismic reanalysis shall be specified by the-staff. The responsibility for specifying the guidelines for.the seismic reanalysis to the licensee is the Division of Systems Safety The assignment for preparing these guidelines is specified in TACS, #4925 (May 23, 1978). Evaluation of the seismic reanalysis will be by the . Division of Systems Safety. Draft guidelines were issued.on September 9, 1978 and the draft guidelines were. discussed with the licensee on ' September 19, 1978. The schedule for completing this task is'(1) Issuance of guidelines to licensee - November 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,(5) Staff identification of' items requiring follow-up action - July '1,1979, (6). Resolution of follow-up . items - October 1,1979, and (6) Final Report to ACRS - November _l, 1979. Staff reviewer will be J. Rajan (MEB). Management responsibility will be carried out by Operating Reactors Branch No. 4. (
Reference:
SER Supplement No. 1, pgs. 2-3, 3-1, 18-1, 18-2, and E-2)
- 9. Instrument-Station Ground Grid System On August 25, 1977, the licensee presented and suosequently documented in letters on August 30, 1977 and September 16, 1977 their analysis for the effect of ground currents introduced into the station grounding grid from the worst postulated station _ electrical fault condition with inadvertent ties between the present instrument and station ground systems. The licensee's analysis showed that safety systems will still perform as intended.
The licensee committed to continue to analyze and test the instrumen-tation systems to identify and eliminate any inadvertent ties between the instrument and station ground grid and will provide to the NRC by no later than the end.of the first scheduled refueling outage the results of their analysis, testing and corrective actions that have been implemented to assure that the installation meets the design g criteria. Also, the licensee committed to carefully monitor the instrumentation systems for any spurious operations and degrading conditions requiring corrections. Normal instrumentation monitoring will meet the defini-
- tions of a channel check (every 12 hours), channel functional test (every 30 days), and daily heat balance check'as specified in the facility technical specifications for instrumentation systems. Also, W
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the licensee will maintain a record of their monitoring of these systems and will report in a timely manner any abnormalities to NRC (Inspection & Enforcement, Region 3). The staff concluded that the plant can be operated safely based on the information and coma.itments provided by the licensee. Evaluation of the analysis, testing, and corrective actions and operational reports will be by the Division of Operating Reactors. Management responsibility will be by Operating Reactors Branch No. 4. ' (
References:
Toledo Edison Company letters from L. Roe to D. Vassallo dated August 30, 1977 and September 16, 1977, and NRC letter from D. Vassallo to L. Roe dated September 26, 1977)
- 10. Inservice Inspection and Testing Programs By NRC letter dated April 22, 1977 the licensee was granted writter relief from the requirements of Section XI of the ASME Code for e ~
and valves testing program in accordance with Technical Specif* 4.0.5a from the date sf issuance of the facility operating lit the start of faci?4cy commercial operation. Relief was not, however, granted for the time after commercial ope. tion. NRC letter dated April 22, 1977 required the licensee to submit information which identifies any request for written relief according to the requirements of Technical Specification 4.05b and 10 CFR 50.55a(g)(6)(i). On June 29, 1977 the licensee submitted their Pump and Valve Test Program and submitted on November 22, 1977 the Inservice Inspection Program for compliance with 10 CFR 50.55a(g). A letter granting relief from the ASME Section XI Inservice Inspection (Testing) Requirements is being issued to the licensee in October i 1978. The relief is being granted until such time that the NRC staff's detailed review is complete. The licensee's submittal is presently j being evaluated by the MEB, Division of Systems Safety. Staff Q-l's are scheduled to be issued June 1,1979 and final resolution of these matters is scheduled for September 1, 1979. The DSS reviewer assigned to complete this task is J. Rajan (MEB). Management will be by Operating Reactors Branch No. 4.
- 11. Interim Fire Protection Technical Specifications The licensee was notified on November 21, 1977 by telephone and telecopy of our requirements for implementing interim fire prott Pion i Technical Specifications for plant systems and administrative pre.c-dures. A letter dated November 28, 1977 was sent to licensee protiding l the the sample standard Technical Specifications and requesting the licensee's response by December 7, 1977 for an application for license amendment with the plant specific interim technical specifications.
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7 -: The" licensee submitted proposed in'terim Technical Specifications on December 12, 1977 and on March 22, 1978'(Amendment No. 9) interim 1. fire technical specification was issued'for the presently installed 3' fire. protection equipment at the facility, as hassbeen:done for other: operating 1 facilities. The interim Technica1' Specifications deal-only-1with admin'istrative-surveillance and correctivejsteps1to; reduce,the Li likelihood of damaging fires pending ourifinal review of'the fira
-c
. protection of Davis-Besse, Unit 1. l
.1 o
- ResolutionLof this. matter will-be.by Division of. Systems Safety. DSS ;!
- reviewer is V. Leung (ASB), 'and the = schedule .for completing this task 1 ,
is the'same as:specified for completing.the Fire Protection' Review- . J (see Item.5, EnclosureL1); ' Management responsibility will1be: carried ~l out by Operating Reactors Branch No. 4. -
-{
- 12. Modification to Spent Fuel Storage-Capacity
~
By'1etter. dated December 19, 1977,.-the' licensee requeste'd'an: amendment -
'7' to License NPF-3 for augmenting the existing spent-fuel: storage
. capacity. JA Federal RegisterLNotice.wasLfiled on March 8,,1978' j
specifying a; time limit offApri13 14,1978 forl filing'a timely; petition to intervene. .No timely nor tardy petitions to intervene have been: ! received as of 0ctober 10r 1978. On' March 6,;1978,. staff requests (EEB/ DOR) for additional- information regar'ing d the. licensee's submittal ; of December 19, 1977 was sent'to the licensee LThe;1icensee's response ; to theLstaff's request.for information was received.on-April 4, 1978. ) Subsequent to theLlicensee's submittal dated April 4, 1978, problems-associated with the removal of the Burnable Poison. Rod Assemblies (BPRAs) and Orifice Rod Assemblies-(0 ras)'from the core after.87 effective full power. days of operation required transferring the :: BPRAs and.0 ras through.the. spent fuel pool.to the adjacent cask pit. 1 (see Amendment No.:11, June 16,.1978). During this transfer contami-nation occurred and'the licensee provided revised information on June 22, 1978 regarding their April.4, 1978' response to the staff's request for information dated March 8,.1978.. The Division of Operating Reactors schedule for completing ~the f.EB/ DOR questions and sections of the SER and Environmental impact' Appraisal for the spent fuel ~ modifications is approximately three weeks after. transfer :' to DOR.- (See Memorandum dated September 12, 1978 from L. Barrett to T. Carter - EEB/ DOR' schedule to complete spent fuel modifications.) i The Division of Systems Safety will evaluate the remaining' sections- !
- required.for issuance of.the SER for,the spent fuel. modifications. -l
=The DSS. reviewer is V. Leung (ASB) and the schedule for completing the 7
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i task is:- (1) Q-l's (if needed) to be issued on. November 15, 1978
. and (2) completing the SER is scheduled for January 5,.1979.
Management responsibility will be carried out.by Operating Reactors - Branch No. 4. 4
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- a 2 ENCLOSURE 2' CURRENT STATUS OF GENERIC REVIEW ITEMS DAVIS-BESSE, UNIT 1 A. Items which have been resolved and require no further'NRC action.
(For' Enclosure 2 all items are ider..ified with number which' refers to-generic review index, page 3-2, for the October 21,-1977 NUREG-0328,. ~, Vol. 4, No. 5-Pink Book.)- ITEM
REFERENCE:
SER, p. 6-5, TS p. 3/4 6-2'
~
- 1. Containment Leak Testing -
Appendix J (3-11) .through p. 3/4 6-4 '
- 2. ECCS FAC Evaluation (3-13) SER, Supplement No. 1, -
- p. 6-3 through p. 6-10:
- 3. Effluent Treatment Systems SER, p. 11-1 through (ALARA) (3-15) p. 11-2
- 4. Emergency Planning (3-16) SER, p. 13-3 through
- p. 13-5 ,
- 5. Filter Tech Specs'(3-17). TS p. 3/4 7-17 through.
, p. 3/4 7-19
- 6. Flood of Equipment Important SER, p. 10-3 and Supplement to Safety (3-19) No. 1, p. 6-8
- 7. Fracture Toughness Requirements - TS p. 3/4 4-24 through Appendix G (3-20) p. 3/4 4-29
' 8. Fuel Cask Orop Analysis (3-21) SER, p. 9-2
- 9. Fuel Handling Accident Inside Letter to applicant dated Containment (3-22) July 29, 1977 with NRR ~
Safety Evaluation . enclosed -
- 10. High Energy Line Break (3-24) SER, p. 3-7.through 3-10
- 11. Inservice Inspection PWR SER, p. 5-8 and TS 3/4 4-6 Steam Generator Tubes (3-26) .through p. 3/4 4-12 I
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. ~ .
1 i ITEM REFERENCE
- 12. Potential Equipment Failure SER, p. 8-2, Supplement Associated with Degraded Grid No. 1, p. 8-1, and Amendment No. 7 Safety Evaluation
- 13. PWR Secondary Water SER p. 5-8 and TS 3/4 7-10 Chemistry Monitoring Requirements (3-37)
- 14. QA Program for Operations SER, p. 17-1 to 17-6 and (3-38) p. 14-1
- 15. Qualifications of Radiation Necessary action completed Protection Manager (3-39) prior to issuance of OL - ,
April 22, 1977
~
- 16. Hydraulic Snubbers (3-43) TS 3/4 7-20 through 7-35
- 17. Steam Generator Feedwater NRC generic letter on Flow Instability (SGFWFI) SGFWFI sent to licensee on (3-44) September 6, 1977 which stated B&W steam generators have not had this problem
- 18. Respiratory Protection Necessary action completed Program (3-42) prior to issuance of OL -
April 22, 1977
- 19. Deletion Technical Specification Amendment No. 8 issued on Requirements for Annual Operating February 28, 1977 deleted Report the Technical Specification Requirements for an Annual Operating Report
- 20. Interim Fire Technical Amendment No. 9 issued Specifications (except as on March 22, 1978 noted in Item 11, Enclosure 1) .
t i 21. Fuel Rod Bow Effects (3-22) Amendment No. 11 issued
- on June 16, 1978 l 22. PWR Reactor Vessel Seal By letter dated March 24, l Ring Missile 1978, the licensee stated that the cavity annulus seal ring is not left in place during normal operation
B. Items which have been evaluated in light of current NRC requirements / guidance and for which measures that will make the status acceptable have been initiated by the licensee. ITEM REFERENCE
- 1. Industrial Security Licensee's response to 10 CFR Part 73.55 submitted on May 25, 1977. NRC staff review for Phase I completed -
September 8, 1977. Licensee's ' Modified tmended Security 4 Plan submitted December 9, , 1977. Review completed April 21, 1978. Revised MASP submitted June 5, 1978 and in Review.
- 2. Anticipated Transients NRC letters to licensee Without Scram (3-6) dated October 10,.1973 and December 6, 1974. Licensee letters to NRC dated January 24.
. . 1974 and October 28, 1974.
(SER, p. 7-3 and SER Supplement i No. 1, p. 18-4)
- 3. Fire Protection (3-18) See Items 5 and 11 of 1 Enclosure 1
- 4. Inservice Inspection Program See Item 10 of Enclosure 1 !
(3-27) C. Items which have been evaluated in light of current NRC requirements / ! guidance and which are presently unresolved. Further action by NRC 1 and/or the licensee may be required in the future. I J I
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. i ITEM , REFERENCE. ,
-1. Reactor-Pressure Vessel . NRC letter.to: licensee
. Supports;(Asymetric LOCA ' dated November 21, 1975.
loads) (3-5) ' Licensee's letters to NRCL on December 119, 1975 and August 28, 1976. NRCLletthr1 w dated January 20,11978 re-questedLlicensee:toLproceedJ with an evaluation of' asym ; metric LOCA! loads. .By letter . ' dated May 5,:1978, the licensee provided a schedule for comple-ting. Phase 1-of the' evaluation 1 by December.31,fl978. .Sched-uling;was defined at the B&WL Owners Group and NRC/ DOR meeting held on March 31~,;1978. (SER,p.~3-14) 2 .' Anticipated Transients . NRC letters.to licensee
. Without Scram (3-6). ' dated October 10, 1973 and-
. . December.6, 1974. . Licensee-
- letters to NRC dated.
~ January 24, 1974 and October 28, 1974. .(SER,' ~
- p. 7-3 and SER Supplement No. 1, p. 18-4)
- 3. Fire Protection (3-18) See Items 5 &.11 of Enclosure 1-
- 14. Upgrading STS Bases Program. NRC letter to licensee August 19, 1977. Licensee-attended generic meeting on October 4, 1977 D.- Items which have'been evaluated in light of current NRC requirements /-
.guidarce and which are unacceptable.
e ITEM REFERENCE l None
.).
'l
5-I E. Items which have not been evaluated in light of current NRC requirements / l guidance. These items will be evaluated by DDR. ITEM REFERENCE
- 1. Diesel Generator Lockout (3-12) NRC l'etter to licensee on September 2, 1977.
Licensee letter to NRC on. October 29,'1977
- 2. Fracture Toughness.and Potential NRC letter to licensee on-for Lamellar Tearing of Steam September 8, 1977. Licensee Generator and Reactor Coolant letter to NRC on October 10, Pump Materials 1977 stating submittal to-NRC on November 11, 1977.
Applicant's submittal received on November 11, 1977. This review is . being accomplished under Category A, Task A-12.
- 3. PWR Moderator Dilution (3-36) NRC letter sent to licensee on September 16, 1977.
- Licensee's response was submitted on December 16, 1978.
- 4. PWR HPSI & LIPSI Flow NRC letter sent to licensee Resistance (3-35) on November 9, 1977.
Licensee's response raceived on January 13, 1978. This review is being accomplished under Pink Book Generic Issue
#053.
- 5. Implementatic, of Appendix I Letter with enclosed Standard Technical Specifications model specifications sent
- to applicant on 'July 10, 1978. -Licensee s submittal schedule is .
. 180 days from issuance of letter.
F. Items which are not applicable to Davis-Besse, Unit No. 1. 'i ITEM REFERENCE All. item identified as specific to GE, CE or Westinghouse plants. ,
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a- -- ENCLOSURE 3 REGULATORY GUIDES USED DURING THE-LICENSING REVIEW
'FOR DAVIS-BESSE, UNIT 1 REFERENCETOSER$5ER 2j REGULATORY GUIDE .. SUPPLEMENT-NO.-1, AND.
NUMBER REGULATORY GUIDE TITLE . APPLICABLE LICENSE AMENDEMENTS. 1.1 Net Positive Suction Head. SER p.-:6-3. ')
.for Emergency Core Cooling and Containment Heat.
Removal System Pumps- , _ , 1.4 Assumptions Used for SER p. 2-10 and . Evaluating the Potential Supplement No. 1 p. 2-3; " Radiological Consequences of.a Loss-of-Coolant-Accident for Pressurized . Water' Reactors 1.6 Independence Between .SER p. 8-l'and p. 8-3 Redundant Standby (Onsite) Power' Sources and Between Their Distribution. Systems 1.7 Control of Combustible Gas. SER p. 6-5 and 6-6 Concentrations 1.8 Personnel Selection and SER p. 13-3 Training 1.9 Selection of Diesel Generator SER p. 8-1 Set Capacity for Standby ~ Power Supplies 1.11 -Instrument Lines Penetrating- .SER p.-6-5' , Primary. Reactor Containment
- 1.12 Instrumentation for Earthquakes SER p. 3-17 1.13 Fuel Storage Facility Design SER p. 9-2, 9-3 and Bases 9-13
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2 r 1.17 Protection of Nuclear Power SER p. 13-16 Plants-Against Industrial Sabotage 1.20 .- Comprehensive Vibration SER p. 4-6 Assessment Program for ; Reactor Internals During i Preoperational and Initial Startup Testing . 1.21 Measuring, Evaluating, and SER p.'11i10 ] Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-
; Water-Cooled Nuclear.. - '
Power Plants. , SER p. 2-9 1.23 Onsite Meteorological Programs 1.27 Ultimate Heat Sink for Nuclear SER p.-2-8, 2-15 Power Plants . 1.28 -Quality Assurance Program SER p. 17-1 Requirements (Design.and Construction) , 1.29 Seismic Design Classi,fication SER p. 9-3 and 9-16 1.30 Quality Assurance Requirements SER p. 17-1 for the Installation, Inspection, and Testing of Instrumentation ' , and Equipment , 1.31 Control of Stainless Steel SER p. 4-6 and 5-5 We] ding 1.33 Quality Assurance Program SER p. 13-6 Requirements (Operation) 1.37 Quality. Assurance Requirements SER p. 17-1 1 for Cleaning of Fluid Systems and Associated Components of Light Water-Cooled Nuclear Power Plants , s -l
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3-1.38' Quality Assurance Requirements SER p. 17-1 , for Packaging, Shipping, Receiving,' Storage, and Handling of-Items for Water-Cooled Nuclear Power Plants 1.39L Housekeeping Requir'ements for SER p. 17-1 Water-Cooled Nuclear Power. Plants 1.44 Control of the use of SER p. 4-5 and - Sensitized Stainless Steel 4-6 - 1.45 Peactor Coolant Pressure Supplement ik). 1, Detectionssystems P. 5 j 1.46-Protection Against Pipe SER p. 3-7 Whip Inside Containment 1.47 Bypassed and Inoperable Supplement No.~1, Status Indication for p. 7-7 Nuclear Power Plant Safety Systems-
~
1.48 Design Limits'and SER p. 3-15 and 5-1 Loading Conditions. 1.52 . Design, TestingLand SER p. 9-11 ; Maintenance Criteria for Atmospheric Cleanup System Air Filtration and' Absorption Units of Light-Water-Cooled Nuclear Power Plants-1.54 Quality Assurance Requirements SER p. 17-1 for Protective Coatings Applied to Water-Cooled Nuclear Power Plants 1.58 Qualification of Nuclear Power - Plant Inspection, Examination, - and Testing Personnel 1 1.60- Design Response Spectra for SER p. 3-10 and
' Seismic Design of Nuclear Supplement No. 1,
- Power Plants p. 2-2, 3-1, and 18-1 .
l l i j
- . . - . - . - . - . . - - -- . .- - - - ..- - . . . - . _a...
F,h ig. Damping Valves in Seismic SER p. 3-10 1.61 .l Design of Nuclear Power Plants Quality Assurance Requirements SER p. 17-1 1.64 for the Design of Nuclear Power Plants Installation :,' Over Pressure SER p. 3-16 1.67 . Protection Devices Preoperational and initial SER p. 14-1 1.68 Startup Test Programs for Water-Cooled. Power Reactors Quality Assurance Terms SER p. 17-1 1.74 . and-Definitions
.i Assumptions Used for SER p. 15-4 i 1.77 ' Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors Preoperational Testing of SER p. 6-10 1,79 and 14-1 Emergency Core Cooling Systems for Pressurized Water Reactors Inservice Inspection SER p. 5-8 1.83 .
of Pressurized Water Reactor ! Steam Generator Tubes Code Case Acceptability SER p. 3-3 1.84 and p. 5-2 ASME Section III Design ; and Fabrication ! SER p. 3-3 l 1.85 Code Case Acceptability - and p. 5-2 ASME Section III Materials Quality Assurance Requirements SER p. 17-1 1.94 for Installation, Inspection, and Testing of Structural ,
- Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants f
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I 5-1.95 Protection of Nuclear Power SER p. 6-11 Plant Control Room Operators Against an Accidental Chlorine Release ,
. l.111 Methods for Estimating SER p. 2-11 Transport and Dispersion of Gaseous and Liquid Effluents in Routine ' '
Releases from Light-Water J. Cooled Reactors 8.8 Information Relevant to. SER p. 12-1 Maintaining Occupational and 12-3 Radiation Exposure as low as Practicable (Nuclear Reactor) . 8.10 Operating Philosophy for. SER p. 12-1 Maintaining Occupational
. Radiation Exposures as Low as' Practicable e
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. ENCLOSURE N0. 4
' DISTRIBUTION Docket File (Environ) LEngle
- 11/21/77 DSE Reading. '
.EP-1. Reading. .,
. PCota.
DocketIfo.: 50-346 DVassalio JStolz . I ItE!!0RAilDtM FOR: Roger S. Boyd, Director. Division of Project -
!!anagement FR0!i: liarold R.L Denten, Director, Division of Site Safety * '
and Environmntal Analysis. .i SU3 JECT: TRA!!SFER OF DAVIS-SESSE UNIT l10.1 TO OPERATIt:G P.EACTORS 3:W!CH NO. 3 He undcrstanIthat you' are planning 'to tiansfer Davis-Dessa Unit-1 from DPM to !T9.* this nenorandua should be attached as 4.n enclosure to your ! transfor nenorandra to effcct transfor of the environmntal revier responsibility fron Environ. mental Pro.iccts Franch "o. I to Operating l Reactors 3 ranch !!c. 3. .This transfer of environ-. ental responsibility - ' is to be effective as of the date of your transfer namo. - [ There are no unconpleted environmental review tasks associated with this [ unit. . At the tino of transfer of responsibility, I?.E. !!IPC, and others on the distribution for this ncno should be notified of the .followin; chcnges:
- Fmn To '
Environnental Project 11anager P. Cota- J]iannon.
.G. Vnighten 'G. Lear Branch Chief K. Goller-Assistant Director V. fiocre Licensing Assistant D. Slater C. Parrish Origi n:155ed by N. rt. Dn::en I!arold R. Denton, Director '
..Divisi on:of Site Safety
~ ' '
and Environ.catal Analysis Office of nuclear Reactor Aogulation cc: See next page I 4
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