ML19209B233
| ML19209B233 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/21/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| NUDOCS 7910090359 | |
| Download: ML19209B233 (9) | |
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UNITED STATES y ; [3s-,(
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NUCLEAR REGULATORY COMMISSION
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. E WASHINGTON, D. C. 20555 o :7 f
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September 21, 1979 V
Docket No. 50-336 Mr. W. G. Counsil, Vice President Nuclear Engineering & Operations Northecst Nuclear Energy Company P. O. Box 270 Hartford, Connecticut 06101
Dear Mr. Counsil:
In the September 19, 1978 Safety Evaluation Report (SER) for Fire Prot-: tion supporting Amendmsnt No. 43 to Facility Operating License No. 00'd-65 fu the Millstone Nuclear Power Station, Unit No. 2, Alternate Safe Zhutdown Capability was identified as an incomplete item of review in Section 3.2.
To expedite the completion of our review of this item, we are enclosing our pot,ition entitled, " Safe Shutdown Capability". Section 8 of this document specifies the information we requird to complete our review.
In reviewing your submittals of April 11, 1979 and July 31, 1979, we find that the data requested by 8(a) has been supplied. Please provide the remaining responses (8(b) through 8(1)) within 90 days of the date of this letter.
If additional information is already available in a previous submittal, you need only reference that material.
Sincerely, 0
' uls / /p/l.t
.&A Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
Enclosure:
Staff Position -
Safe Shutdown Capability cc w/ enclosure:
See next page 1114 035 7910090Jf7
Northeast Nuclear Energy Company cc:
4 William H. Cuddy, Esquire l
Day, Berry & Howard Counselors at Lav One Constitution Plaza
{
Hartford, Connecticut 06103 Waterford Public Library l
Rope Ferry Road, Route 156 Waterford, Connecticut 06385 Northeast Nuclear Energy Cor pany ATIN:
Superintendent Millstone Plant Post Office Box 128 uaterford, Connecticut 06385 Northeast Utilities Service Company ATTN: Mr. James R. Hiernelwright Nuclear Engineering and Operations P. O. Box 270 Hartford, Connecticut 06101 1
Anthony Z. Roisman, Esq.
Natural Resources Defense Council hi g on
.' 'b005 Mr. John.'. Shedlosky l
Nuclear Regulatory Connission, Region I Office of Inspection and Enforcement 631 Park Avenue King of Prussia, Pennsylvania 19406 1114 036 i
STAFFPOSITION Enclosure SAFE SHUTDOWN CAPASILITY t
Staff Cencem i
I n Ulnfig.,
Duriel the staff's evaluation of fire protectien pregrams at c;e: :ing plants, one or :nore specific plant areas may be identified in which the staff does not have adequate assurance that a pcstulated fire will not damage both redundant divisions of shutdown systems.
This lack of assurance in safe shutdewn capability has resulted from ene or both of the follcwing situatiens:
- Case A: The licensee has not adaquately identified the systems and cer:penents required for safe shutdown and their location in specific fire areas.
Case 5: The licensee has net demeastrated that.he fire protecticn for specific plant areas wi".1 prevent damage to both redundant divisiens of safe shutdcwn e:m:cnents identified in these areas.
For Case A, the staff has required thet an adequate safe shutdewn analysis be perfe m ed. This evaluatien includes tre identification of the systems required for safe shutdcwn and the location of the system c:=;cnents in the plant. iihere it is determined by this evaluaticn that safe shutdcwn c mpene,:s of both redundant divisiens are located in the same fire area, the licensee js required to demenstrate chat a ;cstulated fire will not damage both divisions or provide alternate shutdewn capability as in Case B.
For Case B, the staff may have required that an alternate shutdewn ca;acility be ;revided with is incependent of the area of c:ncern or :ne licensee may have propesed such a capability in lieu of certain additienal fire protection modificatie-e in the area. The s;ecific mcdificatiens associated with the arc of concern along with otner systems and equipment already independent of the area form the altemate shutdewn capability. For each plant, the modifications needed and the combinations of systems which provide the shutdewn functions may be unique for each critical area; hewever, the sh0:dewn functions provided shculd main.ain plant para. meters within the bounds of the limiting safety consequences de2.med acceptable for the design basis event.
i Staff Positien
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!afe shu dewn ca: ability shculd be demcastrated (Case A) er alte-na:e shuticwn ca:a::ility previded (Case 5) in acc:rdance with the guiceif res pr0vided belew:
- 1. yes-n 3 asis Event he design basis even for considering te need for alternate snu ::wn is a ::stulatec ' ire in a s:ecFi: fire area ::ntaining ecuncan safe snu ::wn : ables /ecui: rent in ci:se pr:xi:-i y *nert i-.as :een :e er-inec -hat P re ;r: e::icn means cannet assure
- a: safe snu ::wn ca:c:ility wi'.1 :e ; reserve:. Twc cases sncui:
te ::nsi:ered:
(1) Offsite ;cwer is availa:Te; and (2) offsite
- wer f s a.
- : avaiia:ie.
l 7910090,36I 1II4.1.]
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. A.N -
- 2. Limitine Safety Consecuences and Recuired Shutdewn Functiens 2.1 No fissicn product boundary integrity shall be affected:
4.
No fut ~
..e /smage; b.
N: ruN a o',
iy primary coolant bcundary; c.
l._ ructurt af tie containment boundary.
2.2 Ths r tar coolant syr'em process variables shall be within the ndicted for a hss of nomal ac power.
2.3 The alternate shutdewn
.pability shall be able to achieve and maintain subcritical conditions in the raactor, maintain reactor c:olant inventory, achieve and -Tin: sin het standby
- c:nditions (hot shutdewn* for a 2*iR) for an extended period of time, achieve cold shutdown
- c:nditions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and maintain cold shutdown conditiens the,reafter.
As defined in the Standard Technical Specifications.
- 3. perfennance Geals 3.1 The reactivity centrol functicn shall be capable of achieving and maintaining cold snutdown reactivity conditiens.
3.2 The reactor coolant makeup function shall be capable of maintaining the reacter ceplant level above the top of the care for SWR's and in the pressuri er for PWR's.
3.3 The reactor heat removal functicn shall be espable of achieving and r.aintaining decay heat removal.
3.4 The process menitoring function shall be capable of providing direct readings of the process variables necessary to perform and centrol the above functions.
- 3. 5 The supporting function shall be capable of providing th'e process cccling, lubricatien, etc. necessary to permit t
the operatien of the equipment used for safe shutdown by the systers identified in 3.1 - 3.4 3.6 The equi;eene and systers used to achieve and maintain hot standby conditicas (hot shutdewn for a BWR) shculd be (1) free of fire damage; (2) capable of maintaining such conditiens for an extanded time period lenger than 72 hcurs if the ecui: ment recuired to achieve and maintain ecid shutdown is not available cue to fire damage; aad (3) capable Of being powered cy an ensite emergency power system.
3.7 The ecuipmen and systers usec t: achieve and maintain c:ld shutd:wn c:ncitions shculd be eit.9er free Of fire damage cr the fire damage to suc systems shcule :e limitec sucn that re: airs can be'made and colt snutd:wn c:ndi icns acnieved wi:rin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Ecuipment and systems used orier to 72 neurs after tne fire snculd be catacle of being cwered by an onsite emerg;acy :cwer system; ncse used after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> may be powered by 1114 038
P00R 0 mm j) ;
offsite power.
3.. B These systems need not be designed to (1) seismic categor/ I criteria; (2) single failure criteria; or (3) cepe with other plant accidents such as pipe bnaks or stuck valves
( A;;endix A BTP S.5-1), except these portions of these systems which interface with or impact existing safety systems.
- 4. FWR Ecui =ent Generally Necessarv For Het Standby _
(1) Reactivity Centrei Reactor trip capabil'ty (scram). Scratien :apability e.g.,
charging ; ump, makeup pu.:o er high pressura injection pump taking suction " rem cencentrated borated sater supplies.
and letdewn system if required.
(2) Reacter C:elant Makeue React:r coolant makeue capability, e.g., charging pum:s or the high pressure injection pumps.
Power coerated relief valves may be required to reduce pressure to allow use of the high pressure injection pumps.
(3) React:r Coelant System pressure Centrol_
Reac:cr pressure c:ntrol c3; ability, e.g., charging pumps or pressurizer heaters and use of the letdcwn systams if recuired.
(4) Ceca'v Heat Rereval Decay heat remcval capability, e.g., pcwer 8:erated relief valves.(steam generator) or safety relief var'es for heat removal with a water supply and eme gency or aux 4liary feedwater pum:s for makeup to the steam generator. S*rvica water or other pum::s may be recuired to provide water for au41. i..s feed pump sucticn if the cendensate st: rage tank capacity is not adequate for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
(5) precass Menit:rine Instrumentatien P ecess monit: ring capability e.g., pressuri:er pressure and level, steam generat:r level.
(6) Sue:cet.
~he ecui;nen: recui ed :: su::cr: ::eration Of the abcve ces:ri:ed snu acwn ecui: rent e.,.,............ ::ciing water servi:e a:er, et:. and :nsi e : wer s urces ( AC, CC) witn
- neir ass:ciated ele: rical :istributi:n system.
I114 039
- 5. FWR E:uiement Generally Necessary For Cold Shutdewn*
(1) Reactor Ceolant System Pressure Reduction to Residual Heat Removal System t RnR Cacacility Reactor coolant system pressure reduction by c:oldewn using steam generator power operated relief valves or atmcspheric dump valves.
(2)- Decay Heat Removal Decay heat removal capability e.g., residual heat removal system, ce=ponent cooling water system and service water system to removal heat and maintain cold shutdewn.
(3) Suceert Support capability e.g., ensite pcwer scurces (AC & DC) er offsite after 72 hcurs and the associated electrical distributien system to supply the above equipment.
Equipment necessary in additien to that alrearty provided to maintain het standby.
- 6. BWR Ecei: ment Generally Necessary For Het Shutdewn (1) Reactivity Cen tml Reactor trip capability (scram).
(2) Reactor Ccolant Makeue Reactor coolant inventory makeup capability e.g., reactor core system (RCIC) er the hign pressure coolant isolaticn eccling(HPCI).
injection system (3) Reacter Pressure Centrol and Decay Heat Removal Depressuri:atien system valves or safety relief valves for dump to the suporessien pool. The residual heat removal system in steam cendensing.cde, and service water system may also be 'Ised for heat remeval to the ultir. ate heat sink.
(4) Su: ressien Peel Coeline Residual heat remeval system (in su:cessien peci c: cling mece} service water system to maintain het shutd:wn.
(5) Process Menit:rine Pr: cess = nit: ring ca:a:ility e.;., rea:::r vessel level and Oressure and su::nssi:n 001 te=er atum.
I114 040
-5 (6) Suceert Support capability e.g., ensite pcwer source (AC & OC) and their associated distribution systems to provide for the shutdown equipment.
- 7. SWR Ecui: ment Generally Necessarv For Cold Shutdewn' At this point the equipment necessary fer het shutdewn has reduced the primary system pressure and temperature to where the RHR system may be placed in service in RHR ccoling mcde.
(1) Cecav Heat Removal Residual heat removal ystem in the RHR c: cling mede, service s
water system.
(2) Suceert Onsite scurces (AC & CC) er offsite after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and their associated distributien systems to prcvide for shutdown equipment.
Equipment provided in addition to that for achieving het shutdewn.
- 8. Informatien Recuired Fer Staff Review (a) Descri : ion of the syste :s er per 1ons thereof used to pr vide the shutdown ca: ability and modificaticns required
, := achieve the alternate shutdcwn capability if recuired.
(b) System design by drawings which shew nemal and alternate snutdewn centrol and pcwer circuits, location of c:meenents, and
. that wiring which is in the arsa and the wiring which is out of ::.e area that re:uired the alternate system.
(c)
Demonstrate that changes to safety systems will not degrade safety systems.
(e.g., new isolation swit:hes anc centrol switees shculd meet design criteria and standards in FSAR for electrical ecui: rent in the system that the switch is to be installed; cabinets that ce switchas are to be mounted in should also meet ce same criteri; (FSAR) as other safety related cabinets and panels; to avoid inadvertent isolation from the c:ntrol recm, the isola:icn switees shculd bt Seylocxed, or alamed in :ne centr:1 recm if in ce " local *
- isclared" positi:n ;
- eri:cic : necks snculd be mace to vers
. witch is in the
=rt:er :csitien for nemal coeration; anc a single transfer swite. Or ::ner new cevice sncuic nc: be a scur:e f:r a single failure :: cause 1:ss of necuncant safety systems).
')
Demenstrate that wiring, inclu:ing =cwer sources f:r the
- ntr:1 circui anc equi: ment :: era:i:n for -he al te-: ate shutscwn :e ned, is ince:encen: :f ecui: men; wiring in
- ne area :: :e avciced.
1114 041
P00R ORGINA
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(e)
Demonstrate that alternate shutdcwn power sources, including all breakers, have isolation devices en centrol circuits
~
that arc routed througn the area to be avoided, even if the breaker is to be cperated manually.
(f)
Demonstrate that licensee procedure (s) have been develcped which describe the tasks to be parfor-ted to effect the shutdown me thod. A sum.ary of these procedures should be submitted.
(.g)
Demonstrate that spare fuses are available'for centrol circuits where these fuses,may be required in supplying pcwer to centrol circuits used for the shutdown method and may be blcwn by the effects of a table spreading room fire. The spare fuses should be locat2d convenient to the existing fuses. The shutdc*n pr:,ce:ure should inferm the cperator to check these fuses.
(h)
Demonstrate that the man;cwer required to perform the shutdown functions using the crocedures of (f) as well as to provide fire brigade members to fight the fire is avr.ilable as required by the fire brigade technical s;ecifications.
(i) Demonstrate that adequate acceptance tests are performed.
These should verify that: equipment operates from the local control station wnen the transfer or isolation switch is placed in the "lecal" nesition and that the equipment cannot be operated frem the centrol reem; and that equip-ment operates frem the centrol reem but cannot be operated at the local c:ntrol statien when the transfer or isolatien switch is in the " remote" positicn.
(j) Technical Specifica*iens of the surveillance requirements and limiting c:nditiens for cperatien for that equipment not a'iready covered by existing Tech. Specs. Fcr example, if new isolation and c:ntrol switches are added to a service water system, the existing Tech. Spec. surveillance require-ments en the service water system shculd add a statement similar to the follcwing-
"Every third pump test shculd also verify that the pu=p star.s frem the alternate shutdewn statien after moving all service water system isolatien switches to tne local centrol pcsitien."
(k) Demonstrate that the systems available are adecuate to :erf:rm
=e necessary shut::0wn functions. ~he functions recuired sneuld be based On previcus analyses, if ;cssible (e.g.,
in ne :5AR), sucn as a less of nor al a.c. Ocwer er snuticwn
- n a Grou: I isciati:n (SWR).
he ecui: ment recuired for tne al:ema e ca:a::ility sneuld :e me same r epuivalent t:
=a relied :n in ne a:cve analysis.
1114 042 e
7 (1)
Demonstrate that npair procedures for cold shutdewn systems are developed and material for npairs is maintained en site.
e e
O d
e 1114 043