ML20237K069
| ML20237K069 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 08/21/1987 |
| From: | Murley T Office of Nuclear Reactor Regulation |
| To: | Bird R BOSTON EDISON CO. |
| References | |
| NUDOCS 8708270038 | |
| Download: ML20237K069 (18) | |
Text
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August 21,$87; d4 j
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f docket No. 50-293
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y ' Bod on Edison Company
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p-s ATsN' Ralph G. Bird Senior Vice President - Nuclear iN ' '
-s; 800 Boylston Street q
Boston, Massachusetts 02193 1
SUBJECT:
INI % ASSESS'iENT OF PILGRIM SAFETY ENHANCEMENT PROGRAM a
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Dear Mr. Bird:
/Chduly'8,1987,,BestonEdbo'n.Comny(BEco)submittedadetaileddescription
?t of the Pilgrim Safety Enhancement Program.(SEP) to the NRC. This letter transmits th.Y staff's initial af,sessment of this program (Enclosure).
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s The daff's initial assessment has been conducted to provide an understanding
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'of the CEP modifications and assess the safety significance of those changes,
, when considered singularly or along with other changes.
Additionally the a
staff examired'your evaluations of these changes and the BEco schedule for impleinentation of the modifications. The staff's review included a visit to BEco office 5 in Braintree on July 122,1987, conversations with representatives q
af your sttff.over the past few weeks, and a meeting with BECo representatives 4
in Bethesda on August 4,1987. "
The, staff expects to continue its dialogue with BECo regarding the SEP program i,
as part of'its larger effort on severe accidents. The generic issue of j
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containment venting has been under consideration by BWR owners and the NRC for
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. It qis a complex issue fraught with cor.f>11cting safety
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severa) years.
obipc;ives. Eacause the severe accident effort is ongoing,. the staff is not
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preparedto;endorsetheuseoftheDirectTorusVentSystem(DTVS)atthistime.
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To assist ths staff in its consideration of the DTVS, de request you provide the staff your written response to the questions contained in the enclosure.
a Installation of the DTVS under the provisions of 10 Cn's50.59 is precluded by
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the need for Technical Specifications on a containment' isolation valve.
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i The staff still has questions regarding the proposed modification to the i
reactorcoreisolationcooling(RCIC) system Prior to implementing this 9 -
modification the staff requests that BEco conduct an assessment of hydrodynamic loads on the RCIC pthing and supports, based on the proposed exhaust pressure of 46 psig, and maki the results of that' assessment available to the staff.
The staff requests clarification regarding the function of one valve in the 4~
backup nitrogen supply system, As described in the enclosure, valve A0-4356 4.
appears.to be a containment isolation valve and, consequently, would ce appropriate for inclus_icir'in the Technical Specifications.
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August 21, 1987 2
The staf reouests clarification regarding the modification to the RHR system to provide additional sources of water for RPV injection and containment
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spray. Thi modification may require a change to the Technical Specifications.
As described the enclosure, the valves to be added to the RHR system become part of the rea r coolant pressure boundary during operation of the RHR system and, consequently, re subject to surveillance testing.
As you are aware, the 'RC will continue its inspection of SEP modifications, review of affecte' pan rocedures, and observation of related onsite j
activities. We will keep ou informed, should we have additional concerns about this program.
Please ontact the NRR Project Manager if you have any questions.
Sincerely, l
Thomas E. Murley, Director Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
See next page DISTRIBUTION:
PDI-3 R/F PDR C. Tinkler T. Collins N. Su V. Thomas
- 0. Chopra B. Clayton J. Wiggins, R:I S. Collins, R:1 W. Russell, R:I M. McBride, R:I FL
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NAME :RWess
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- 08/16/87
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- 08/W/87 0FFICIAL RECOR 0FC :
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- TMurley DATE :08//'//87 T/87
- 08/ /87
- 08/ /87 OFFICIAL RECORD COPY
r August 21, 1987 i
l The staff requests clarification regarding the modification to the RHR system to provide additional sources of water for RPV injection and containment spray. This modification may require a change to the Technical Specifications.
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.
l We comend your efforts and leadership on this program. The quality of your l
July 8,1987 submittal is impressive and the cooperation of your staff is l
appreciated.
1 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 questions.
Sincerely,
/.5/
Steven A. Varga, Director Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure See next page DISTRIBUTION:
PDI-3 R/F PDR C. Tinkler T. Collins N. Su V. Thomas
- 0. Chopra B. Clayton J. Wiggins, R:I S. Collins, R:I W. Russell, R:I M. McBride, R:1
- See previous concurrence OFC :PDI-3*
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- ACTDIR/PDI-3 NAME :RWessman:lm :JCraig
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- VNerses DATE :08/ /87
- 08/ /87
,:08/ /87
- 08/ /87
- 08/ /87
- 08/ /87
- 08/ /87
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OFC :AD/ORP*
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T.... __:....___....:..___.......:....._______:......______:........ __
NAME :BBoger e
DATE :08/ /87
- 08 87 OFFICIAL M CORD COPY
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DISTRIBUTION:
, Dockets File:.50-?93i t
^W. Rusis~,"RT" '"
Local PDR Murley/Sneizek R. Starostecki 1.. Shao-J. Craig i;
-W. Hodges F. Rosa-
' T. ~ Speis G. Hulman
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H. Li S. Varga B.'Boger
'V. Nerses-V. Rooney S. Schinki, 0GC W. Paton, OGC
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Enclosure INITIAL ASSESSMENT OF PILGRIM SAFETY ENHANCEMENT PROGRAM i
l 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 design modification associated with the direct torus vent system (DTVS) provides a direct tent 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 existing torus purge exhaust line between the containment isolation valves (outside containment) with a reconnection to the existing torus purge exhaust line 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 at 30 psig.
The installation of an additional branch line and containment isolatien valve would require a change to the plant Technical Specifications. Therefore, it is our view that installation of the DTVS cannot be implemented under 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:
1)
Provide comprehensive analyses of accident sequences, with their estimated frequency of occurrence, for which the vent would be called upon to operate.
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2)
Provide estimate of the fraction of those sequences where the vent would be operated but where the accident would have been terminated short of containment failure without vent operation. Consider the following situations in the accident sequences:
(a) electric power returned to service (b) equipment returned to service (c) mis-diagnosed situation corrected by operators
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3)
Provide comprehensive analysis of those accident sequences that:
(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 operation of the vent followed by excessive pressure differential (buckling).
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 malfunction.
1 2.
Sect. 3.3 - Containment Spray Header Nozzles The objective of installing new containment spray header nozzles in the 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 assembly has 6 out of 7 nozzle outlets capped while the original j
nozzle assemblies had all 7 nozzle outlets open. The effect of capping
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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
function is performed using the new backup diesel fire pump (750 gpm).
Installation of the capped nozzle assemblies in conjunction with an RHR
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l 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.
I Because installation of the new spray nozzles results in reduced drywell i
spray capacity and reduced flow through the RHR heat exchangers the licensee evaluated the consequences of this modification.
With 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 performed reanalysis of the containment response to steam line breaks for 2
2
' sizes ran,ging 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 dramatically reduced when operating the RHR system in the containment spray mode, the operator will be instructed to open the RHR suppression pool bypass valve so that rated flow may be maintained through the heat exchanger and decay heat adequately removed.
Installation of this modification is expected to be completed before plant restart.
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-l ment Spray 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 j
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between the Fire Protection System and the RHR system as well as the reinstallation of the RPV Head Spray line. The RPV Head Spray line was included in the original design but was disconnected due to water hammer concerns. Reinsta11ation of the line is accompanied by design changes, rerouted piping, and a bypass line with restriction orifices added in order to reduce the potential for water hammer.
The connection between the fire protection system and the RHR system is 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 connection leaves the path interrupted; when the connection is desired a removable pipe section,16" in length, must be installed with quick connect Victaulic couplings. When the removable pipe section is not installed the piping ends are capped.
Isolation of the RHR system is provided by the addition of a gate valve (local manual) and check valve.
During operation of the RHR system, these valves become part of the reactor coolant pressure boundary.
Isolation of the line from the fire protection system is provided by gate valve. The gate valves will be locked closed. The crosstie on the RhR side of the removable pipe section is to be designed with ASME Section III, Class II piping and ASME Section
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III, Class I valves (gate valve and check valve). On the fire protection side of the connection the crosstie is designed to ANSI and NFPA Standards
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and is designated Quality Class FPQ (Fire Protection).
The effect of these changes will be to allow the use of diesel fire
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pumps, including a newly proposed diesel fire pump, which draw water from the fire water storage tank and the city water supply line to provide water for core injection and containment sprays.
J The licensee has evaluated the effect of the proposed design modifications and concluded that there is no adverse impact on the performance of safety related systems or the fire protection system. The staff has similarly concluded, based on our initf al assessment, that the design changes have no significant deleterious effects on the design or operation of the plant. However, the licensee should consider the need to propose Plant Technical Specifications regarding surveillance testing to
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verify leak tightness of the RHR isolation valves to be added as part of this change.
This modification is expected to be completed after plant restart.
l Installation of thi.. modification under the provisions of 10 CFR 50.59 may not be acceptable and the licensee should provide clarification regarding the need to include RHR isolation valve leak testing in the Plant Technical Specifications.
4.
Sect. 3.5 - Diesel Fire Pump for RPV Injection and Containment Spray This design change was prompted by the licensee's desire to prov(de a redundant pumping capacity to the existing diesel fire pump and thus
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provide additional protection for extended station blackout accident sequences or other severe accident scenarios. The design change includes the addition of a new diesel fire pump and auxiliary equipment consisting of piping, valves, and an enclosure 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 has a capacity of 750 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 occurrence or mitigate the consequences of a severe accident.
The addition of a new diesel fire pump to the plants fire protection system has been evaluated by the licensee to determine if there were any concomitant effects on plant safety functions.
In as muen 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 functions or systems. To the extent the design change effects the fire protection system the new components are designated Q (fire protection).
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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5.
Sec. 3.6 - Diesel-Pump Fire Pump Fuel Oil Transfer System j
This design change is to provide a redundant (non-electric power dependent) diesel fuel oil transfer pump for the diesel fire pump P-140.
This redundant pump will allow extended operation of the diesel fire pump as a water source for the RHR system during extended station blackout and other potential severe accident scenarios beyond the design basis. The change adds a hyroturbine driver (AC power independent) fuel oil transfer pump in the intake structure, and associated auxiliaries and piping.
The addition of this fuel oil transfer system to the plant's fire protection system has been evaluated by the licensee to determine effects on plant safety functions.
In that the plant fire protection system is not a safety related system, addition of this system was determined to l
not effect plant safety functions or systems. The staff agrees with the licensee's evaluation, i
Installation of this system is expected to be completed before plant resta rt.
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 a backup nitrogen supply to provide nitrogen during a station blackout.
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 edditional 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 2
while the N trailer is being aligned. The nitrogen supply from the 2
cylinders will automatically, in the event of a loss of the existing N 2
storage facility, provide makeup to drywell instrument supply piping.
The cylinders are capable of supplying N for 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 assuming 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 2
20,000 scfh for 7 days or at less flow for extended periods. Nitrogen is supplied at 110-120 psig for instrument supply lines; nitrogen from the trailer is provided at 70 psig for torus and drywell makeup.
, f In order to improve the reliability of nitrogen supply the licensee has modified 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 determined that the design modification will not adversely affect the safety functions of the Inerting and Drywell Testing System nor adversely affect the safety function of the reactor building (modified by an additional penetration through the reactor building wall).
During discussions with the licensee on July 22, 1987 the staff inquired 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 i
containment isolation' valve, and thus a change in fail safe position would not affect the containment isolation design.
The staff, however, 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 tha licensee. To the extent a change in the technical specifications is involved, this matter needs to be considered as part of the issue of 50.59 applicability.
This modification is expected to be completed prior to plant restart.
7.
Sect. 3.8 - Blackout Diesel Generator Including Protected Installation Facilities As part of the Safety Enhancement Program Boston Edison Company will install a non-safety related Station Blackout (SB0) 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 I
ECCS pump and all other associated loads from one safety train required for reactor shutdown, 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
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blackout which could lead to core damage and/or containment failure. The 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 the plant for diesel generator maintenance loads when the unit is not
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running) and has a independent fuel tank (no connection with emergency diesel generator fuel supplies), and a cooling radiator. The new diesel generator and the two entsting emergency diesel generators for Pilgrim are j
ALC0 engines. The new unit will be located south of the plant adjacent i
to the switchyard relay house.
The new diesel generatcr will be connected between the secondary side of i
the shutdown transfonner (third source of power to the safety related buses) and emergency buses A5 and A6 (Figure 1). The diesel generator and the existing SMVA shutdown transformer will be connected to the 1
existing safety-related 4.16kV buses A5 and A6 through a new two-breaker 4.16kV bus A8. The diesel generator will be connected to the new switchgear A8 thru breaker #801 and the shutdown transformer will be connected to switchgear A8 thru breaker #802. The outgoing feed from the switchgear A8 will be connected to the existing 4.16kV breaker #600 which is in turn connected to breakers #501 and #601 of the safety buses A5 and A6.
In the original design the secondary of the shutdown transformer was directly connected to breaker #600.
t Breaker #802 wnich is connected to the shutdown transformer will be kept closed during normal operation to supply power when required to safety buses A5 and A6 thru breaker 600 (normally closed) and breakers 501 and 601 (normally open). This alignment of breakers is consistent with the present arrangement which maintains shutdown power transformer 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 transformer (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 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 and assume load by synchronizing to the shutdown transformer during plant operation. During this time breakers 802 (NC), 600 (NC), 501 (NO), 601 (NO) are maintained in their nonnal line up. The diesel generator will also be tested by energizing safety related loads when the reactor is shut down.
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The controls of breakers 801 and 802 are interlocked to prevent interconnection of the SB0 diesel generator with the shutdown transformer except for testing of the diesel generator. The' diesel generator and the 4.16kV breakers of switchgear A8 are controlled manually either from the main control room or locally from the diesel generator enclosure.
Protective relaying is provided to prevent damage to the diesel generator. An' independent 125 de system (battery 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.
In addition, annunciation will be provided in the main control room for diesel generator trouble, diesel generator breaker (801) trip / inoperative and shutdown transformer breaker (802) trip / inoperative.
The diesel generator has an independent sufficient fuel system with capacity to y
supply rated load for a minimum of one week.
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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 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 accordance with R.G. 1.75. All conduit and cable installed by this design change located within safety related areas will be supported in accordance with seismic I criteria.
I The staff has reviewed the information provided by the Heensee 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 power sources are unavailable (Station Blackout). The new diesel generator 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 buses) and no other system except for a 480 volt ac feed from a non-safety related load center.
The diesel generator breaker 801 is normally closed and the present alignment of breakers 600, 501, and 601 are not changed by this modification.
Therefore, the shutdown transformers ability to supply power to buses AS and A6 under design conditions will not be affected.
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There are no changes to the safety related portion of the emergency service buses as a result of this change.
1 Tha control cabling of. diesel generator and breakers 801 and 802 are routed in a' separate conduit and duct banks from the diesel generator enclosure and switchgear A8 to the control panels C3 and C5 in the control room. The physical separation between new non-safety wiring and existing class IE wiring within the panels will be accordance with R.G.
1.75 (verbal agreement by the licensee). Therefore, although the i
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 will satisfy R.G.1.75.
Based on the above, the staff concludes that the addition of the non-safety-related diesel generator at Pilgrim will reduce the probability of station blackout and have no adverse effect on the offsite power systems, the Class 1E emergency diesel generators or the shutdown transformers and is, therefore, acceptable.
It is also concluded that this modification does not require any Tech. Spec. changes or result in an unreviewed safety question per 10 CFR 50.59. The implementation of t
the design will be verified by Region.I, with support from NRR as requested by the Region, j
i This modification is expected to be completed after plant restart.
Installation of this modification under the provisions of 10 CFR 50.59 l
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 j
for Item II.K.3.18 of NUREG 0737. The modification and proposed l
Technical Specification (BECo letter of May 20,1987) have been reviewed i
and approved by the staff. A license amendment is currently being processed.
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This modification is expected to be completed before plant restart.
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9..
Sect. 3.10 - Addition of Enriched Boron to Standby Liquid Control System 1
-The use of enriched sodium pentaborcte in the Standby Liquid Control System (SLCS) allow Pilgrim to meet the requirement of the Anticipated
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Transient Without Scram (ATWS) Rule (10 CFR 50.62) with one pump operable, thereby retaining the redundancy of the SLCS design.
The licensee submitted a proposed Technical Specification change which was approved by the staff on August 5, 1987 (Amendment 102).
This modification is expected to be completed before plant restart.
- 10. Sect. 3.11 - ATWS Feedwater Pump Trip This change will provide an automatic trip to all feedwater pumps at 1400 psig reactor vessel pressure. This setpoint is selected so that feedwater 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 the recirculation pumps arc initiation of the Automatic Rod Insertion (ARI) system on low water ievel or high reactor pressure.
I The existing reactor feedwater pump trip logic will be modified to accept an additional trip signal from ATWS.
A new trip coil (in addition to the existing trip coil) will be installed in the breaker associr ted with each reactor feed pump. The coils are " energized to trip" co113.
I The licensee 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 j
staff agrees with the licensee's evaluation.
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.
- 11. Sec. 3.12 - Modification' to Reactor Core Isolation Cooling System Turbine j
Exhaust Trip Setpoint During Station Blackout (SB0) events, the RCIC system is available to supply cooling water to the reactor and maintain the reactor water I
level. The RCIC pump is driven by a turbine using the primary system
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steam..The turbine exhaust is piped to the suppression pool. Continuous l
discharge of the steam to the suppression pool, however, will increase the suppression pool temperature and the containment pressure. The existing RCIC exhaust trip pressure is 25 psig, which will be reached at about 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 of trip pressure will allow the RCIC system to operate until about 15.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the event.
Steam discharge into the suppression pool, where the steam is condensed, results in thermal-hydraulic loads both on the containment structures and the discharge pipe. These loads will be increased significantly with increasing exhaust back pressure. Assessment of the :nage.itude of these i
loads is required in order to ensure that the RCIC exhaust pipe will not l
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 assessment of hydrodynamic loads on the RCIC piping and supports based on the proposed exhaust 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 ATWS Recirculation Pump Trip Trip of the recirculation pumps is a feature for the mitigation of ATWS
14.
-events.
Pilgrim currently has the capability of tripping the recirculation pumps by opening the. field breakers.
Installation of a t.ew 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 breakers 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 steff agrees with the licensee's evaluation. The overall compliance of Pilgrim with ATWS Rule (10 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 of 10 CFR 50.59 appears acceptable.
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