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ENCLOSURE 1 EVALUATIONBYTHEOFFICEOFNUCLEARREACTORREGULATig TECHNICAL SPECIFICATION REOUIREMENTS FOR CONTROL R00 DRIVE SCRAM INSERTION TIME PILGRIM NUCLEAR POWER STATION, UNIT 1 BOSTON EDISON COMPANY DOCXET NO. 50-293
1.0 INTRODUCTION
By letter dated Nove ber 19,1985 (Ref.1), the licensee for Pilgrim Nuclear Power Station (PNPS), proposed a change to the Technical Specification (TS) 4.3.C.2.
This change would reduce the existing surveillance for scram insertion time.
The present TS requires scram insertion time testing of 50% of the control rod drives in 16 week intervals. This results in all operable control rods being tested every 32 weeks.
The licensee seeks to reduce the control rod system wear by reducing the frequency and number of control rods tested in a given cycle.
2.0 EVALUATION The proposed change (TS 4.3.C.2) would provide scram insertion time surveillance within each 120 days of operation for a minimum of 10% of the control rod drives (CRDs) on a rotating basis.
PHPS stated that incorporation of this change would provide assurance that the control rods are operable and not cause excessive wear on the control rod drive components.
In response to staff requests for further justification, PNPS stated (Ref. 2) that review of surveillance documentation from initial startup to May 1983 indicated
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2 that the probability of violating the TS for the limiting condition for operation for scram insertion speed is very low. Further, the raw data also showed that no TS violation with regard to scram speed occurred in that 13 year interval. This justification is acceptable to the staff.
I At the reques+ of the staff, PNPS, by references 3 and 4, proposed two additional changes which would bring the TS into reasonable conformance with the Standard TS. The first additional change provides a more frequent scram insertion time surveillance in the event three or more control rods fail to
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meet a maximum scram insertion of seven seconds.
The other change requir'es each operable CR0 to be scram time tested after a reactor shutdown greater than one hundred twenty days. As these changes are consistent with the Standard TS, we find these changes acceptable.
3.0 CONCLUSION
The purpose of the scram insertion time surveillance is to demonstrate operability of the control rod drives.
PNPS has proposed a reduction in the number of control rods tested and an increase in interval between the tests for the purpose of reducing CR0 wear. We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the Fealth and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Connission's regulations, and issuance of the amendment will not be inimical to the connon defense and security or to the health and safety of the public.
4.0 REFERENCES
i 1.
Letter from W. D. Harrington, Boston Edison Company, to D. B. Vassallo, NRC, dated November 19, 1985.
2.
Letter from James M. Lydon, Boston Edison Company to NRC, dated February 18, 1987, i
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3.
Letter from James M. Lydon, Boston Edison Company, to J. A. Zwolinski, I
NRC, dated December 5, 1986.
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Letter from R. G. Bird Boston Edison Company to NRC, dated July 24, 1987 t
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ENCLOSi1RE ?
Boston Edison Company Pilgrim Nuclear Power Station Systematic Assessment of Licensee Perfomrance*
Functional Areas 1.
Management Involvement in Assuring 'Juality.
Review generally timely, thorough and technically sound.
Rating:
Ca tegory 2 2.
Approach to Resolution of Technical Issues from a Safety Standpoint.
l Generally timely resolutirAs.
l Rating: Category ?
3.
Responsiveness to NRC Initiatives Generally timely responses.
Rating: Category 2 4.
Enforcement History N/A 5.
Operational and Construction Events i
N/A 6.
Staffing (includingManagement)
N/A Training and Qualification Effectiveness
Reference:
NRC Manual Appendix 0516 - Systematic Assessment of Licensee j
Performance
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MEMORANDllM FOR:
V. Nerses, Acting Project Director Project Directorate I-3 Division of Reactor Projects - I/II FROM:
M. W. Hodges, Chief Reactor Systems Branch Division of Engineering & Systems Technology
SUBJECT:
SAFETY EVALUATION REPORT RELATING TO TECHNICAL SPECIFICATION CHANGES FOR SCRAM INSERTION TIME Plant Name:
Pilgrim Nuclear Power Station Docket No:
50-293 TAC No:
60217 Project Directorate:
I-3 Project Manager:
R. Wessman Review Branch:
SRX8 Review Status:
Complete is the Reactor Systems Branch Safety Evaluation Report (SER) on Boston Edison Company's request to modify the Technical Specifications (TSs) for Pilgrim Nuclear Power Station with regard to scram insertion time.
We have reviewed the licensee's submittals and find them acceptable.
Enclosure II is our SALP input.
M. W. Hodges, Chief Reactor Systems Branch Division of Engineering & Systems Technology
Enclosures:
As stated cc w/ enclosures:
S. Varga B. Boger A. Thadani i
R. Wessman SRXB Merrbers
Contact:
D. Katze,SRXB, x27588
'O MEMORANDUM FOR:
V. Nerses, Acting Project Director Project Directorate I-3 Division of Reactor Projects - I/II FROM:
M. W. Hodges, Chief l
Reactor Systems Branch Division of Engineering & Systems Tecnnology l
SUBJECT:
SAFETY EVALUATION REPORT RELATING TO TECHNICAL SPECIFICATION CHANGES FOR SCRAM INSERTION TIME Plant Name:
Pilgrim Nuclear Power Statd1n Docket No:
S0 293 TAC No:
60217 Project Directorate:
I.3 Project Manager:
R. Wessman Review Branch:
SRXB Review Status:
Complete is the Reactor Systems Branch Safety Evaluation Report (SER) on Boston Edison Company's request to modify the Technical Specifications (TSs) for Pilgrim Nuclear Power Station with regard to scram insertion time.
We have reviewed the licensee's submittals and find them acceptable.
Enclosure II is our SALP input.
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M. W. Hodges, Chief Reactor Systems Branch Division of Engineering & Systems Technology
Enclosures:
DISTRIBUTION As stated Docket File SRXB R/F cc w/ enclosures:
M. W. Hndges S. Varga R. Jones R. Boger D. Katze A. Thadani D. Katze R/F R. Wessmin Pilgrim P/F SRXB Memeers
Contact:
D. Katze,SRXB, x27588
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- 8/ /87 0FFICIAL RECORD COPY I
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August 21, 1957 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 P!LGRid 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 assessment 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 irmlementation 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 rneeting 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 conteinment isolation valve.
The staff still has questions regarding the proposed modification to the reactor core isolation cooling (RCIC) system. Prior to implernenting this modification the staff requests that BEco conduct an assessment of hydrodynamic loads on the RCIC piping and supports, based on the proposed exhaust pressure of 46 psig, and raake 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 ande consequently, would be appropriate for inclusion in the Technical Specifications.
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August 21, 1937 2
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 P.HR system and, consequently, are subject to surveillance testing.
We consnend your efforts and leadership on this program. The quality of your July 8,1987 suMittal is impressive and the cooperation of your staff is appreciated.
As you are awart, the NRC will continue its inspection of SEP modifications, review of affected plant procedures, and observation of related onsite activities. We will keep you infonned, should we have additional concerns about this program. Please contact the PRR Project Manager if you have any questions.
Si
- erely, 6, tre ar-Division of Reacto P jects !/II Office of Nuclear e :or Regulation
Enclosure:
As stated cc w/ enclosure See next page g
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Mr. Ralph G. Bird Boston Edison Compan.y Pi! grim Nuclear Power Station ec:
Mr. K. P. Roberts, Nuclear Operations Boston Edison Company Pilgrin 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 07360 Boston, Massachusetts 02190 Resident Inspector's Office Mr. Richard N. Swanson, M3nater U. S. Nuclear Regulatory Connission Nuclear Engineering Department Post Offics Box 867 Boston Edison Company Plymouth, Massachusetts 02360
?5 Braintree Hill Park Braintree, Massachusetts 02184 Chairman, Board of Selectmen 11 Lincoln Street Ms. Elaine D. Robinson Plymouth, Massachusetts 02350 Nuclear Inforretion Manager Pilgrim Nuclear Power Station Office of the Commissioner RF0 #1, Rocky Hill Road Massachusetts Departnent of Plyaouth, Massachusetts 02360 Environmental Quality Engineerirg One Winter Street Mr. Mike Ernst, Research Director Boston, Massachusetts 02108 Energy Committee Statehouse - Room 540 Office of the Attorney General Boston, Massacnusetts 02133 1 Ashburton Place 19th Fluer Boston, Massachusetts 02108 Mr. Robert M. Hallisey, Director R)diation Control Program Massachusetts Department of Public Health 150 Tremont Street, 2nd floor Boston, Massachusetts 02111 Regional Administ,*ator, Region !
U. S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406 Mr. James D. Keyes Regulatory Affairs and Programs Group Leader Boston Edison Comocny 25 Braintree Hill Park Braintree, Massachusetts 02184
i Enclosure INITIAL ASSESSMENT OF PILGRIM 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 Venj System (DTVS)
The proposed design modification associated with the direct torus vent system (OTVS) 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 existing torus purge exhaust line between the containment isolation valves (outside containment) with a reconnection to the existing torus purge exhaust line downst Mam of the SGTS. The new torus vent line is also provided with its own containment isolation valve and a rupture df *.c. set to relieve at 30 psig.
The installation of an additional branch line and containment isolation valve would require a change to the plant Technical Specifications. Therefore, it is our view that installation of the DTVS cannot be implerented under the provisions of 10 CFR 50.59.
1 To assist the staff in its consideration of the proposed DTVS, we request a written response to the following conce ns:
1)
Provide comprehensive analyses of accident sequences, with their estimated frequency of occurrence, for which the vent would be called upon to operate.
2)
Provide estimate of the fraction of those sequences whare the vent would be operated but where the accident would hsve been teminated short of containment failure without vent operation. Consider the following situations in the accident sequences:
i (a) electric power returned to service (b) equipment returned to service (c) mis-diagnosed situation corrected by operators v --
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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 fr orrect 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 indyce 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.
2.
Sect. 3.3 - Containment Soray Header Nozzles The objective of installing new containment spray header nozzles in the drywell is to igrove the perfonnance 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 asseebly 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 1
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3-function is perfonned using the new backup diesel fire p.mp (750 gcr:1 Installation of the capped nozzle assemblies in conjunction with an RHR pump will reduce the drywell spray flow from the original design value of approximtely 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 through the RHR heat exchangers the licensee evaluated the consequences of tnis 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 teeperature.
In order to address this matter the licensee perfonned 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 detennined 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 781*F.
Bscause 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 ad*quately 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 t
appears acceptable.
3.
Sect. 3.4 - Additional Sources of Water for RPV Injection and Contain-Ent pray, 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
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4 between the Fire Protection System and the RHR system as well as the reinsta11ation of the RPV Head Spray line. The RPV Head Spray line was include <1 in the original design but was disconnected due to water harivner 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 try adding a piping connection from the fire protection system piping and the RHR Salt Service Water Injection Ifne. 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. Whan the removable pipe section is not i
installed the piping ends are caaed. 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 il protection system is provided by gate valve. The cate valves will be locked closed. The crosstie on the RHR sido of the removable pipe section is to be designed with ASME Section III, Class !! piping and ASME Section I!!, Class I valves (gate valve and chuk valve). On the fire protection j
side of the connection the crosstie is designed to ANSI and NFPA Standards andisdesignatedQualityClassFPQ(FireProtection).
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 the fire water storage tank and the city water supply line to provide I
water for core injection and containment sprays.
The 11cansee 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 initial assessment, that the
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design changes have no significant deleterious effects on the design or l
operation of the plant. However, the licensee should consider the need to propose Plant Technical Specifications regarding surveillance testing to l
-5 verify leak tightness of the RHR isolation valves to be added as part of this change.
This modification is expected to be completed af ter plant restart.
Installation of this 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 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 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 J
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 detemine if there were any j
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 detemined not to effect plant safety functions or systems. To the extent the design change effects the fire protection system the new components are designated 0 (fire protection).
4 This modification is expected to be completed after plant restart.
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installation of this modification under the provisions of 10 CFR 50.59 appears acceptable.
. 5.
Sec. 3.6 - Diesel Pumo Fire Pump Fuel Oil Transfer System 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 detemined to not effect plant safety functions or systems. The staff agrees with the licensee's evaluation.
Installation of this system is expected to be completed before plant l
res ta 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 N2 supply will provide a motive source for critical valves and 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 I
2 the additional cylinders is to provide a N2 supply in an interim period while the N trallar is being aligned. The nitrogen supply from the p
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 torms and drywell makeup.
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 desian 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 containrent 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 the 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.
1 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 Enhancerent Program Boston Edison Company will install a non-safety related Station Blackout ($80) 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 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 bachip power source is being installed to reduce the probability of a station blackout which could lead to core damage and/or containment failure. The i
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 perranent plant systems (except for a non-safety 480V feed from the plant for diesel generator maintenance loads when the unit is not 4
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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 existing emergency diesel generators for Pilgrim are ALCO engines. The new unit will be located south of the plant adjacent to the switchyard relay house.
The new diesel generator will be connected between the secondary side of the shutdown transformer (third source of power to the safety related buses) and emergency buses AS and A6 Pigure 1). The diesel generator and the existing SMVA shutdown transfonner 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 will be connected to the new switchgear A8 thru breaker #801 and the shutdown transformer will be j
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 transfonner was directly connected to breaker #600.
Breaker a802 which is connected to the shutdown transformer will be keot closed during noncal operation to supply power when required to safety buses AS and A6 thru breaker 600 (norrelly closed) and breakers 501 and 4
601 (nonnally open). This alignment of breakers is consistent with the present arrangement which maintains shutdown power transfonner 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
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transfonner (preferred source) and failure of the emergency diesel i
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 transfonner during plant operation. During this time breakers 802 (NO), 600 (NC), 501 (NO), 601 (NO) are maintained in their normal line up. The diesel generator will also be tested by
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energizing safety related loads when the reactor is shut down.
. The controls of breakers 801 and 802 are interlocked to prevent interconnection of the SB0 diesel generator with the shutdown transfonner 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 danage 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 transfonner breaker (802) trip / inoperative. The diesel generator has an independent sufficient fuel system with capacity to supply rated load for a minimum of one week.
The cabling for the diesel generator controls and new breakers 801 and l
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 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 ! criteria.
The staff has reviewed the information provided by the licensee or, 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 1
power sources are unavailable (Station Blackout). The new diesel generator is a batkup to the secondary offsite power source (shutdown transformer) and is manually started. The unit is fully self-contained and interfaces only with the shutdown transforver (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 t,reakers 600, 501, and 601 are not changed by this modification. Therefore, the shutdown transformers ability to supply power to buses A5 and A6 c.he design conditions will not be affected.
. There are no changes to the safety related portion of the emergency service buses as a result of this change.
The 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 1E wiring within the panels will be accordance with R.G.
1.75 (verbal agreement by the licensee). Thert. ore, 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 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 IE 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 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 BWR00 evaluation for item II.K.3.18 of NUREG 0737. The modification and proposed Technical Specification (BEco letter of May 20,1987) have been reviewed and approved by the staff. A license amendment is currently being processed.
. This modification is expected to be completed before plant restart.
9.
Sect. 3.10 - Addition of Enriched Boron to Standby Liquid Control System The use of enriched sodium pentaborate in the Standby Liquid Control System (SLCS) allow Pilgrim to meet the requirement of the Anticipated 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 Specifica' ion change which was approved by the staff on August 5.1987 (Amendment 102).
This modification is expected to be completed before plant restart.
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- 10. Sect. 3.11 - ATWS Feedwater Pumo Trio This change will provide an automatic trip to all feedwater pumps at 1400 psig reactor vessel pressure. This setDoint is selected so that feedwater pump trip occurs only when an ATWS event occurs following closure of Main Steam Isolation Yalves. It serves as a backup to the existing ATWS protection. The current ATWS design consists of trips of the recirculation pumps and initiation of the Automatic Rod Insertion (ARI) system on low water level or high reactor pressure.
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 associated with each reactor feed pump. The coils are "energized to trip" coils.
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 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 unde" the provisions of 10 CFR 50.59 appears acceptable.
- 11. Sec. 3.12 - Modification to Reactor Core Isolation Cooling System Turbine Exhaust Trio Setpoint During Station Blackout (SBO) events, the RCIC system is available to supply cooling water to the reactor and maintain the reactor water level. The RCIC pump is driven by a turbine using the primary system s
steam. The turbine exhaust is piped to the suppression pool. Continuous discharge of the steam to the suppression pool, hwever, will increase the suppression pool terrperature and the contlinment pressure. The existing RCIC exhaust trip pressure is 25 psig, which will be reached at 7,50ut 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, l
results in thennal-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 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 l
indicate that the dynamic load differs substantially from static load, l
i the licensee's present method based on static pressure is not acceptable.
Based on the above, we conclude that, prior to iglementing this l
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.
- 12. Sect. 3.13 - Additional ATWS Recirculation Pump Trip Trip of the recirculation purms is a feature for the mitigation of ATVS a-
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events. Pilgrim currently has the capability of tripping the recirculation pumps by opening the field breakers. Installation of a new trip coil within the breaker associated with each recirculation puep MG set drive motor will ir. crease 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" syste>i.
The licensee has analyzed this rodification 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 (10 CFR 50.62) is currently under staff review.
This modification is expected to be completed before plant restart.
Installation of this rodification under the provisions of 10 CFR 50.59 appears acceptable.
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