ML20236W498
| ML20236W498 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 11/24/1987 |
| From: | Zech L NRC COMMISSION (OCM) |
| To: | Studds G HOUSE OF REP. |
| Shared Package | |
| ML20236W501 | List: |
| References | |
| NUDOCS 8712080059 | |
| Download: ML20236W498 (2) | |
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
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WASHINGTON, D. C. 20555 u
CHAIRMAN November 24, 1987 l
l The Honorable Gerry F. Studds 4
United States House of Representatives j
Washington,.D.C.
20515 i
Dear. Congressman Studds:
I am responding to your letter of September 18, 1987, in which you requested that NRC take two specific actions regarding the Pilgrim Nuclear. Power Station.
First, you' asked that we i
perform an environmental assessment of modifications made to J
the plant's. containment structure before the plant is allowed q
to restart.
Second, you asked that we defer judgment'on the plant's venting system until all containment venting issues,.
including environmental concerns, are thoroughly analyzed and resolved through the ongoing review process.
Vou also urged us-to defer judgment on the plant's venting system until revision 4 to the Boiling Water. Reactor Owners Group (BWROG) Emergency Procedure Guidelines (EPGs) has been completed.
As part of their Scfety Enhancement Program (SEP), Boston Edison Company (BECo), the Pilgrim licensee, proposed the installation of a Direct Torus Vent System (DTVS) as one of several SEP measures to improve containment performance at Pilgrim.
The other SEP modifications do not affect the contain-
. ment structure per se but are designed to mitigate the effects of abnormal conditioWs that would develop in containment during an accident scenario.
These modifications are in consonance with NRC goals to enhance containment performance under severe accident conditions but are not required for restart of Pilgrim.
We are, however, ensuring that these modifications do not constitute an unreviewed safety question as part of the I
restart conditions.
The NRC staff's eva'luation of the proposed SEP modifications will include the prospective operation of the DTVS and will address the need for an environmental assessment.
I have enclosed a copy of the staff's initial safety assessment of the SEP modifications in which they did not endorse the use of-the DTVS'at this time.
BECo will not be allowed to place the DTVS system into service until it is thcroughly evaluated and approved by the staff.
Concerning revision 4 of the BWROG EPGs., the staff is not j
expected to complete its review of that revision until early 1988.
Although the staff has previously approved a strategy l
8712080059 871124 DR ADOCK 05000293 PDR
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ifor)containmentiventingLfor Pilgrim?and"other'bo'iling water
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reactorsLin conjunctiontwith their. review of current!BWROG~
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m EPGs,<the proposedtrevision 4 guidelines l recommend 'a new'.
- approach,~with-containment? venting used:las,an anticipatory response';to elevatedKcontainment pressure,L.This newJapproach
/must:be~ thoro'ughly evaluated:and approvedLby the-?. staff beforez specific'. containment <fventingLmodifications'may be/found,
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acceptable at;. Pilgrim...
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.I'.Lapprecia'te!.yo'urliMerestl inl this ' matter and trus titSatithis in letter adequately' respondstoiyour request ~
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Enclosure:
'As stated-
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August 21. 1987
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Docket No. 50-293 Boston Edison Company ATTN: Ralph G. Bird Senior Vice President - Nuclear 800 Boylston Street Boston, Massachusetts 0?199
SUBJECT:
INITIAL ASSESSMENT OF PILGRIM SAFETY ENHANCEMENT PROGRAM
Dear Mr. Bird:
On July 8,1987, Boston Edison Company (BECo) submitted a detailed description of the Pilgrim Safety Enhancement Program (SEP) to the NRC. This letter transmits the staff's initial r.ssessment of this program (Enclosure).
The staff's initial assessment has been conducted to provide an understanding.
of the SEP modifications and assess the safety significance of those changes, when considered singularly or along with other changes.
Additionally the staff examined your evaluations of these changes and the BEco schedule for implementation of the modifications. The staff's review included a visit to BEco offices in Braintree on July 22, 1987, conversations with representatives of your staff over the past few weeks, and a meeting with BEco representatives in Bethesda on August 4, 1987.
The staff expects to continue its dialogue with BECo regarding the SEP program as part of its larger effort on severe accidents. The generic issue of containment venting has been under consideration by BWR owners and the NRC for several years.
It is a complex issue fraught with conflicting safety objectives. Because the severe accident effort is ongoing, the staff is not prepared to endorse the use of the Direct Torus Vent System (DTVS) at this time.
To assist the staff in its consideration of the DTVS, we request you provide the.
staff your written response to the concerns contained in the enclosure.
Installation of the DTVS under the provisions of 10 CFR 50.59 is precluded by the need for Technical Specifications on a containment isolation valve.
The staff still has questions regarding the proposed modification to the reactor core isolation cooling (RCIC) system.
Prior to implementing this modification the Staff requests that BEco condu::t an assessment of hydrodynamic loads on the RCIC piping and supports, based on the proposed exhaust pressure of 46 psig, and make the results of that assessment available to the staff.
The staff requests clarification regarding the function of one valve in the backup nitrogen supply system.
As described in the enclosure, valve A0-4356 appears to be a containment isolation valve and, consequently, would be appropriate for inclusion in the Technical Specifications.
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August 21, 1987 2
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The staff requests clarification regarding the modificc+, ion to the RHR system j
to provide additional-sources of water for RPV injection and containment spray. This modification may require a change to the Technical Specifications.
1 As described in the enclosure, the valves to be added to the RHR system become part of the' reactor coolant pressure boundary, during operation of the RHR system and, consequently, are subject to surveillance testing, h
We commend your efforts and leadership on this program. The quality of your July 8. 1987 submittal is impressive and the cooperation of your staff is
' appreciated.
As you are aware, the NRC will continue its inspection of SEP modifications, review'of affected plant procedures, and observation of related onsite activities. We will keep you informed, should we have additional concerns about this program.
Please contact the NRR Project Manager if you have any quest %ns.
Si
- erely, l
A Dire Division of Reactor P. jects I/II Office of Nuclear R ac or Regulation
Enclosure:
As stated cc w/ enclosure See next page
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il Mr. Ralph' G. Bird L
. Boston Edison Company'
.. Pilgrim Nuclear Power Station' a
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L Mr. K. P. Roberts, Nuclear Operations Boston Edison Company.
Pilgrim Nuclear Power Station.
ATTN:' Mr. Ralph G. Bird Boston Edison Company-Senior Vice President - Nuclear;
- RFD #1, Rocky Hill Road
- 800 Boylston Street
-Plymouth, Massachusetts 0?360-Boston, Massachusetts 02199:
Resident = Inspector's Office'-
Mr.' Richard N. Swanson, Manager
- U. S.- Nuclear Regulatory Comission.
Nuclear. Engineering Department a
Post Office Box 867
-Boston Edison Company Plymouth, Massachusetts ~ 02360 25 Braintree Hill Park-Braintree, Massachusetts 0218a-Chairman, Board of Selectmen
- 11. Lincoln Street Ms. Elaine D. Robinson Plymouth,-Massachusetts. 02360 Nuclear. Infonnation Manager Pilgrim Nuclear Power Station
' Office of the Commissioner RFD #1, Rocky Hill Road Massachusetts Department of.
. Plymouth, Massachusetts ' 02360 Environmental.- Quality Engineering One Winter; Street-Mr. Mike Ernst, Research Director Boston, Massachusetts 02108 Energy Committee.
Statehouse - Room 540.
Office of the' Attorney General Boston, Massachusetts 02133 l'Ashburton.P. lace.
19th Floor.
Boston, Massachusetts 02108:
Mr. Robert M.-Hallisey, Director Radiation Control Program
' Massachusetts.DepartmentLof Public Health 150 Tremont. Street, 2nd Floor Boston, Massachusetts. 02111 Regional; Administrator, Region I U. S. Nuclear Regulatory Comission 631' Park Avenue King of Prussia, Pennsylvania 19406 Mr. James D. Keyes g
Regulatory Affairs and Programs Group Leader Boston Edison rompany
- 25 Braintree Hill Park Braintree, Massachusetts- 02184
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Enclosure INITIAi. ASSESSMENT OF PlLGRIM SAFETY ENHANCEMENT PROGRAM Note: Section' numbers refer to section numbers in~ the BEco submittal of July.8, 1987.
1.
Sect. 3.2 - Installation of Direct Torus Vent. System (DTVS)
The proposed desion modification associated with the direct torus vent system (DTVS) provides a direct vent path from'the torus: air space to the-main stack, in parallel with and bypassing. the Standby Gas' Treatment System (SGTS). The DTVS provides a 'new 8".line branching off the existingLtorus purge exhaust line between the containment isolation valves -(outside containment) with a reconnection to the existing torus purge exhaust line I
' downstream of the SGTS. The new torus vent line is-also provided with.'
its own containment isolation valve and a rupture disc, set to relieve l-at 30 psig.
The installation of an additional branch 'line.and containment isolation valve would recuire a change to the plant Technical Specifications. Therefo'e, it is our view that installation of the DTVS cannot be-implemented unfu the provisions of 10 CFR 50.59.-
To assist the staff in its consideration of the proposed DTVS, we request
.a written response to the following concerns:
l 1)
Provide comprehensive analyses of accident sequences, with their.
j estimated frequency of occurrence, for which the vent would be called upon to operate.
2)
. Provide estimate of the fraction of those sequences where the vent would be operated but where the accident would have been terminated l
short of containment failure without vent operation. Consider the following situations in the accident sequences:
(a) electric power returned to service' l
(b) equipment returned to service (c) ~ mis-diagnosed situation corrected by operators y
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Provide comprehensive analysis of those accident sequences that:^
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~(a) could be improved by correct use of the vent, or (b): could be initiated or made worse by incorrect operation ~of the-vent.
4)
Provide. analysis of sequences that could lead to containment failure.
.by cperation of the vent followed by, excessive pressure differential
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5)
Provide, analysis of the probability of vent failure when called upon.-
'6)'
Provide' analysis.of maintenance or surveillance' errors ~on the vent-system that could' induce accidents.
7)
Provide an estimate of the radioactivity released for.all sequences'when the
. vent could be opened, including both correct usage according.to procedures and incorrect usage due to human error or' equipment -
4
. malfunction.
2.
Sect. 3.3-Containment'Soray Header Nozzles
'The objective of installing new containment. spray header nozzles in the -
v drywell is to improve the performance of drywell spray under severe accident conditions and to provide greater flexibility of use of the sprays under a variety of accident conditions. The replacement spray.
nozzles are' identical to the existing nozzles except that the replacement nozzle assenbly has 6 out of 7 nozzle outlets capped.while the original nozzle assemblies had all 7 nozzle outlets open. The effect of capping nozzles is to reduce drywell spray flow when the spray water is provided by the RHR pumps (5000 gpm) and preserve a basic spray pattern when the spray p
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. function.is performed using the new backup diesel fire pump (750 ppm).
Installation of the capped nozzle assemblies in conjunction with an RHR pump will. reduce' the drywell spray flow from the original design value'of approximately 5000 gpm to a calculated spray flow rate of. 543 gpm..
Because. installation of the new spray nozzles results in reduced drywell.
spray capacity and reduced flow t'nrough the RHR heat exchangers the' Ifeensee evaluated the consequences of this modification. With1 regard to drywell spray flow capacity, the design basis (and licensing basis) require use of the drywell sprays within roughly 30 minutes after the onset of a~ small break LOCA in the drywell in order to reduce the drywell atmosphere temperature.. In order to address this matter the licensee
. perfonned reanalysis of the containment response to steam line breaks for.
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sizes ranging from 0.02.ft to 0.5 ft, as originally discussed in the FSAR. The licensee determined from the reanalyses that the reduced drywell spray flow was sufficient to reduce the drywell atmosphere temperature and maintain the drywell. liner temperature below the design temperature of 281*F.
Because total flow through the RHR heat exchanger would otherwise be-t dramatically reduced when operating the RHR system in the containment spray mode, the operator will be instructed to open the RHR suppression j
pool bypass valve so that rated flow may be maintained through the heat exchanger and decay heat adequately removed.
L Installation of this modification is expected to be completed before plant restart.
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. installation of this modification under the provisions of 10 CFR 50.59 appears acceptable.
3.-
Sect. 3.4 - Additional Sources of Water for RPV injection and Contain-
- ment Soray The basic objective of this design change is to provide additional sources of water that are not dependent on AC power and thus available for core cooling and containment spray during severe accidents, including station blackout. The design modification consists of a piping crosstie 1
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I.LreihstallationLof the RPVl Head 5 Spray line. The' Rky' Head Spray line' was m
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. concerns. - Reinsta11ation of the line;is ' accompanied by ' design ~ changes,
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rerouted" piping, and a' bypass line with restHetion orifices added -in order O
! to redu'ce th'eipotential for ' water hammer.1 W
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made_ by. adding a piping; connection from the fire protection' system piping
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.and the:RHR Salt Service Water Injection line. The design of, the
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connection leaves the path: interrupted; when the connection'is desired a
-removable pipe section,l16" in-length,imust be_ installed with ouick g ;
i connect Victaulic; couplings. LWhen the removable pipei section is not-
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- installed the piping' ends are capped.. Isolation of the RHR_ system is
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Lprovided by the a'ddition; bf a" gate valv'eL(local manual)'and checkLvalve.
1 During; operation'of: the RHR. system, these valves become. part Lof thi
' reactor coolant. pressure boundary. ' Isolation of the line!from tne fire-4
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protection system is provided by gate valve. The gate valves will:be-L'
. locked closed.1The crosstie on'the RHR' side of the removable pipe section' lisito,be. designed withLASME Section..III, Class' II; piping /and A' ME Section-S n e
'III Class I valves 1(gate' valve and check valve). On the fire protection a
f, sidef of the connectiorr the crosstie-is designed.to ANSI and_ NFPA Standards and is designated Quality Class'FPQ (Fire Protection).
hl The effect.of these changes will. be' to allow the use of diesel fire
-pumps, including a newly proposed _ diesel fire pump, which draw water from h(
she fire water storage' tank and. the city water supply line to provide p
- water for core injection and containment sprays, p.
m; The. licensee has evaluated the effect of the proposed design fj modif.ications 'and concluded that there is no adverse impact on the Performance of safety related systems or the fire protection system. The I%
staff has-similarly concluded, based on our..irwitial assessment, that the j
l design changes;have no significant deleterious effects on the design or
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1 verify. leak; tightness of the RHR' isolation valves to be added as part of
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thisfchange.-
This modification-is expected to be completed after' plant restart.
Installation of this -modification. under the provisions of 10 CFR 50.59 -
i-may not biacceptable and the If censee.should. provide clarification K
regarding the need to, include;RHR-isolation valve leak: testing in the:
. Plant Technical Specifications.
4.
Sect. 3.5 --Diesel Fire' Pumo for RPV Injection and Containment Spray This design change was prompted by-the licensee's desire to provide a redundant pumping capacity to the existing diesel fire pump and thus f provide additional protection for extended station blackout accidet sequences or other severe accident scenarios. The design change 1 E N des the; addition of a' new. di.esel. fire pump and. auxiliary equipment cons'isting of; piping, valves, and an enclosurt with foundation and lighting. The-new diesel fire. pump requires no AC. power to perform its function, however, enclosure lighting and HVAC, if needed, will. be powered by the newly proposed; station blackout diesel. The new diesel fire pump hasia-capacity ofL750 gpm at 125 psi which is compatible with the water supply-provided by the 6 inch city water. line. The. licensee has not provided analyses to justify the adequacy of the pump capacity to prevent the y
occurrence or mitigate.the consequences-of a severe accident.
The addition of a new c'esel fire pump to the plants fire protection system has been evaluated by the licensee to detemine if there were any.
concomitant' effects on plant safety functions.
In as much as the plant fire protection system is not a safety related system, addition of the new pump and its auxiliaries were determined not to effect plant. safety y
functions or systems. To the extent the design change effects the fire protection system the new components are designated Q (fire protection),
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This modification is expected to be completed after plant restart.
Installation of this modification under the provisions of 10 CFR 50.59 appears acceptable.-
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.Sec. 3.6 - Diesel Pump Fire Pumri Fuel 011 Transfer System -
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S This Ldesign change is.to provide a redundant (non-electric: power idependNt) diesel fuel oil) transfer pump for the dissel fire pump;P-140.
This, redundant. pump will. allow extended operation'of the diesel fire--pumpi asia water source for the RHR ' system during extended station blackout and;
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'other potential severe accident' scenarios beyond the.' design.. basis.- The-change: ados a hyroturbine driver (AC powerLindependent) fuel oil transfer;
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. pump in theLintakeLstructure, and associated auxiliaries andl piping,
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LThe. addition oflthis fuel oil transfer system.to the plant's fire protection system has been evaluated by:the licensee to determine effects
. onL plant safety functions. ' in that the plant fire protection ; system is :
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Y not' a safety related system. addition of this system was determined to.
y not effect plant safety functions or.' systems' : The staff' agrees with thel w'
licensee's evaluation.
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Installation of this : system is expected to be completed before plant pw 4
-restart.
Installation of this modification under the provisions of 10 CFR 50.59 appears; acceptable, 6.
Sect. 3.7 - Backup Nitrogen Supply System As:the title implies, this proposed design change involves the addition of' L
The backup N. supply.will provide a motive source for critical valves and 2
~ instruments and a source of N for torus and drywell atmosphere makeup.
2 The backup supply consists of an additional 20 cylinders of N with 2
. piping and valves' and a new liquid N / vaporizer trailer. The purpose of 2
the additional cylinders is to provide a N supply in an interim period.
g' while.the N trailer is being aligned. The nitrogen supply from the 1
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p cylinders will automatically, in the event of a loss of the existing.N2
. storage facility, provide makeup to drywell instrument supply piping.
M The' cylinders are capable of supplying N f a' minimum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> based 2
upon ass'uming' two cycles of the MSIV's, two cycles of the MSRV's and other leakage. The liquid N / vaporizer trailer will be sized for a minimum of l
2 20,000'sefh for 7 days or at less flow for extended periods. Nitrogen is j
' supplied L at 110-120 psig for instrument supply lines; nitrogen from the W.
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! trailer is provided at 70 psig for torus and drywell makeup.
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'In order to' improve the reliability of. nitrogen supply the licensee has inodified the design to alter the fail safe position of gate valve A0-4356 from fail closed to' fail open.
As' part of the design process the licensee has detennined that the design modification will not adversely affect the safety fu'nctions of the Inerting
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and Drywell Testing System nor adversely affect the safety function of the i
reactor ' building (mo'dified by an addf tional penetration through the reactor building wall).-
During' discussions with the licensee on July 22, 1987 the staff inquired
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about the effect of altering the fail safe position of valve A0-4356. At that' time the licensee indicated the valve in question was not a containment isolation valve, and thus a. change in fail safe position would not affect the containment isolation design. The staff, however,
'l during subsequent review, has determined that the valve is listed as a containment isolation valve (FSAR Table 5.2-5).
Therefore, the staff concludes that effects on the containment isolation function need to be-reassessed by the licensee. To the extent a change in the technical specifications is involved, this matter needs to be considered as pari. of the issue of 50.59 applicability.
t This modification is expected to be completed prior to plant restart.
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7.
Sect. 3.8 - Blackout Diesel Generator including Protected Installation Facilities
.i As part of the Safety Enhancement Program Boston Edison Company will install a nor;-safety related Station Blackout (SBO) diesel generator rated at 2000KW to provide a non-safety related source of onsite ac power to the 4.16kV safety buses. This unit will be utilized to operate one ECCS pump and all other assNiated loads from one safety train required for reactor snutdown, without LOCA, when all other sources of ac power are unavailable. Boston Edison states that this unit can be made available (manually) from the control room within an hour. This backup power source is being installed to reduce the' probability of a station blackout which could lead to core damage and/or containment failure. The l
unit is skid mounted and housed in a pre-engineered enclosure to protect it from the environment.
The unit is fully self-contained, not dependent on-any permanent plant systems (except for a non-safety 480V feed from mm m ma _-
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- diesel) generator fuel supplies), and a cooling radiator.. The new^ diesel-generator andithe two existing emergency diesel generators 'for' Pilgrim are; h
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/, ALCOl engines. The new unit will be located southfof the. plant adjacent, t
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to'the' switchyard relay house..
q The new' diesel generator will be connected between the-secondary side of
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the' shutdown-transfomer (third source of power to the safety related buses) Land emergency buses AS?and A6.(Figure 1).. TL, diesel generator.
J and the existingiSMVA shutdown' transformer will be connected to the existing safety-related 4.16kV buses A5 and A6 through a new two-breaker 4.16kV? bus A8.
- The diesel generator wil1 be connected to the new-f
- sWitchgear. A8 thru breaker #801 and the shutdown transformer will bei w
connected;to switchgear~ A8 thru breaker #802..The outgoing: feed from the A
switchgear A8 willibe connected to the' existing 4.16kV breaker #600 which.
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.is in turn connected to breakars'#501:and #601 of the safety buses. AS and 1
A6.
Inf the original design the secondary of the; shutdown transformer was-(
directly connected to' breaker #600.
1 Breaker #802,which is connected to the' shutdown-transformer will be kept r
closed duri.ngLnomal operation.to supply. power when required to safety 1
buses AS and: A6 thru breaker 600 (normallyfclosed) and breakers 501 and 601L(noma 11y open). This alignment of breakers is consistent with the present arrangement which maintains shutdown: power transfomer power available for automatic connection to the emergency buses (via automatic closing of 501 and _601) upon a unit trip', loss of the start-up transfomer (preferred source) and failure of the emergency diesel generator. The blackout diesel generator output breaker 801 will be maintained open during' normal operation and will be closed to the safety
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=related buses only-during station blackout (loss of all ac power) or test. The diesel generator will be tested at regular intervals, when the plant is operating,.for its ability to start gnd assume load by'
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- L synchronizing. to the shutdown transformer during plant operation. During 1
athis' time breakers 802 (NC), 600 (NC), 501 (NO), 601 (NO) are maintained
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in their nomal line up. The diesel generator will also be tested by
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. energizing? safety related ~1oads when the reactor is shut'down, w 1,
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Thel controls of breakers 801 and 802 ~are' interlocked to prevent H
. interconnection of the SB0 diesel generator with the shutdown transfonner Lexc.ept for testing 'of the diesel generator.. The diesel generator and.the
- j 4.16kV breakers of. switchgear' A8 are controlled manually either from the'
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. main control' room'or locally from the. diesel generator enclosure. Protective relaying is provided to prevent damage-to the diesel generator. - An 'in' dependent 125.de system attery and charger)-is provided to' supply control power to diesel generator unit controls and associated 4.16kV switchgear A8 (breakers:
801 and 802). Loss of de power will be annunciated in the control room'..
j In addition, annunciation will be.provided in the main control room for l
diesel generator trouble, diesel generator breaker (801) trip / inoperative 3
and' shutdown transformer breaker (802) trip / inoperative. The diesel generator has an independent sufficient fuel' system with capacity to
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supply, rated load for a minimum of one week.
L The cabling' for the diesel generator controls and new breakers 801 and.
802 will be routed in separate conduit and duct banks from the' diesel generator enclosure and switchgear A8 to the control room. The physical'
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separation within the control panels between non-safety related diesel
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generator control wiring and existing class. IE wiring will be in j
accordance with R.G. 1.75. All conduit and cable installed by this design change located within safety related areas will be supported in-3 accordance with seismic I criteria.
The staff has reviewed the information provided by the licensee on its proposed modification to add a new diesel generator at Pilgrim which will power required loads for safe shutdown without a LOCA when all other ac j
power sources are unavailable (Statiun Blackout). The new diesel generator 1
is a backup to the secondary offsite power source (shutdown transformer) and ~
is manually started. The unit is fully self-contained and interfaces only with the shutdown transformer (which is the third power source to the safety i"
buses) and no-other system except for a 480 volt ac feeo from a. non-safety
.related load center.
The diesel generator breaker 801 is normally closed p
and the present alignment of breakers 600, 501, and 601 are not changed by
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h this modification. Therefore, the shutdown transformers ability to supply power to buses A5 and A6 under design conditions will not be affected.
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LThere' are no changes,to :the' safety; related portion of the' emergency service i buses as;a result of this change'. '.
iThe: control l cabling _ of. diesel.generacor and breakers 801 and 802 are routed in a separate conduit and duct banks' from ths diesel generatorf N
enclosure and switchgear A8;to the control panels C3 and C5;i.n'the i
control room. The' physical-separation between new non-safety wiNng and existing class IE wiring within the panels will: be' accordance with R.G.
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I 11.75 (verbal-agreement = by th'e licensee).- Therefore, although' the
. licensee has' not specifically addressed conformance to R.G.1.75, the acceptance of this design.is'. based upon our understanding that the proposed modification willisatisfy R.G.' 1.75, 4
dBased on.'the above, the staff concludes that the addition of the.-
non-safety-related' diesel generator at Pilgrim will reduce the..
protiabi11ty __of station: blackout 'and-have 'no ' adverse effect 'on the offsite
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power systems..the Class IE emergency diesel generators or.the shutdown transfonners and is, therefore, acceptable.. It:is also concluded.that.
this modification does not require any Tech. Spec.. changes or result in
'an unrev'iewed safety question 'per 10 CFR 50.59.. The implementation of-the design:will be verified by Region I,' with support from NRR as requested by'the Region.
This modification is expected to be completed after plant restart.
Installation of this modification under the provisions of 10 CFR 50.59 appears acceptable.
'8..
Sect. 3.9 - Automatic Depressurization System Logic Modifications This modification provides a timed bypass of the high drywell-(.
pressure initiation signal and a manual inhibit of existing ADS actuation logic. This modification responds to the BWROG evaluation for Item II.K.3.18 of NUREG 0737. The modifi, cation and proposed
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Technical Specification (BECo letter of May 20, 1987)-have been reviewed and' approved by the staff. A license amendment is currently being g
' processed..
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J This-modification _ :is expected to be completed before; plant restart.
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Sect.!3.10'-- Addition of Enriched Boron to StandbfLiquid' Control' System
.. :The use of enriched sodium;pentaborate in the Standby Liquid:Contro11 C
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- System (SLCS) allow Pilgrim to. meet the requirement of.-the: Anticipated
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> Transient Without Scram (ATWS) 'Ru1e. ('10. CFR. 50;62) withlone' pump
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i operable, thereby retainirfg the. redundancy'of the SLCS design.
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- The licensee submitted a proposed TechnicaliSpecification change which.
wasapprovedby;the.staffonAugust5,:1987.(Amendmen?.102).
This. modification!is expected to be completed before plant restart..
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c 10.c Sect. 3.11 -- ATWS Feedwater. PumpHTrip This change will provide!an1 automatic tHp. tola11 feedwater pumps"at 1400 1
psig reactor vessel pressure 4. This setpoint is' selected so?that ifeedwater pump trip occurs only when an; ATWS event occurs:following closure of Main Steam Isolation Valves. 'It serves as a backup to the existing ATWS. protection. The current ATWS design consists of trips of -
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the recirculation' pumps and ; initiation of the Automatic Rod Insertion m
'(ARI) system on low water level'or high rea'ctor. pressure..
Thel existing reactor feedwater pump trip logic. will be modified to. accept
' an additional trip signal;from ATWS. - A new trip' cofi (in addition to the
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existing trip coil) will be installed in the breaker associated with each reactor feed pump..The coils are'" energized to trip" coils.-
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The 1.icensee has analyzed this modification and concluded that the modifications to the feedwater pump, trip breakers, ATWS system, and
_ safety related power supplies do not have an adverse safety impact. The i
staff agrees with the licensee's evaluation.
This ' modification is expected to be completed before plant restart.
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k installation of this modification under the provisions of 10 CFR 50.59 l appears acceptable.
til.' Sec. 3.12 Modification to Reactor Core Isolation Cooling ~ System Turbine
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iExh'aust: Trip Setpoint Du' ring. Station' Blackout (SBO) events, the.RCIC system is available to-u supply cooling water to the reactor and maintain the reactor 'wateri
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level.t JThe RCIC pump is' driven by a turbine using the primary system h.
steam.. LThe: turbine: exhaust is piped to'the suppression pool. Continuous
- discharge'of the steam.to the suppression pool, however,'will increase
.the1 suppression pool temperature and the containment' pressure. The.-
existing RCIC exhaust trip-pressurt is 25 psig, which will be reached at:
- 'o Labout 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into'the SB0 event. To extend the use of the RCIC-system, the. licensee proposed to increase, the trip pressure ~ to 46 psig.
.This increase off trip pressure will allow the _RCIC' system to operate until about 15;5-hours into the event.
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. Steam discharge into the suppression pool, where the steam is condensed, results in thermal-hydraulic loads both on the containment structures and g
the discharge pipe. ~ These loads will_ be increased significantly with increasing exhaust back pressure. Assessment'of the magnitude of these loads is required in order;to ensure-that the RCIC exhaust: pipe will'not fail during the increased trip setpoint. Discussions with the licensee's.
technical; staff indicated that the licensee has assessed the loads on.the
- basis of static pressure. 'Since experiments and analytical methods indicate that the dynamic load differs substantially from static load, the licensee's present method based on static pressure is not acceptable.
Based on the above, we conclude that, prior to implementing this modification, the licensee should conduct an i ssessmant of hydrodynamic loads on the RCIC piping and supports based on the proposed exhaust b
. pressure of 46 psig.
It should be noted that the analysis should consider both air clearing loads and steam condensation' loads.
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'12.. Sect. 3.13 -' Additional A1WS Recirculation ' Pump Trip Y
- . Trip of'the recirculation pumps'is a feature for the mitigation of ATWS in~'
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Installation of a new
-trip coil within the breaker associated with each recirculation pump MG set drive motor will _ increase the pump trip reliability.
The design change will add an. ATWS initiated trip signal to the 4160 volt
' drive motor-breDers of the recirculation pump' motor generator sets A and' B. = The trip will be_ at either high reactor pressure (1175 psig) or low reactor water level.(-46 inches indicated level). Signals will be taken.
from existing sensors.
The system will be an " energize to trip" system.
The licensee has analyzed this modification and concluded that it does not degrade.the existing recirculation system. ATWS system or safety related power supplies. The. staff agrees with the licensee's.
evaluation. 'The overall compliance of Pilgrim with ATWS Rule (20 CFR 50.62) is currently under staff review.'
This _ modification is expected to be completed before plant restart.
Installation of this modification under the provisions pf 10 CFR 50.59 appears acceptable.
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