ML20206C204
| ML20206C204 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 10/17/1988 |
| From: | Grimsley D NRC |
| To: | Williams O JRA ASSOCIATES |
| References | |
| FOIA-88-505 NUDOCS 8811160159 | |
| Download: ML20206C204 (2) | |
Text
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RESPONSE TO FREEDOM OF (e )
INFORMATION ACT (FOIA) REQUEST OCT l'i neo xe.
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1 Lasovesna Ms. Ophelia G. Willians PART L-RECORDS RELEASED OR NOT LOCAftD (See checked ecseri No agency records ele to the roovest have teen located.
No edetional egency recoros eW to N request have been locatM.
are sheedy es nietes for pwtac irspection and cwytng h h NRC Pside Doeverent Rum, Agency recorde mAect to te reemet that are identred M AspeMis 1717 H Street. N.W., We#-gm DC.
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Agency records Mpc1 to te reest e o encioned Any opp 4catdo charge for cc5es of the records pmMed eM per Records subpct to te request bare been re erted to another FMerel egenepass) for re%iew org direct royores to you.
r In www of NRC's response to the tw no twrther octon is be.ng taken ca *W totter deted PART (1 A-lNFORMATION uTHHILD FROM PUBUC DISCLOSURE W records a be.ig.Ahheid from putse deetosu e pv ovem to the FOIA esemotens descrted M sad fer the resecre stated M P r
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Cormente The new PDR is located at 2120 L Street, N. W., Lower-l.evel, Washington, D. C., 20555.
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F01 A.88-505 APPEN0lX A-RECORDS MAINTAINED IN THE POR UNDER THE AB0VE REQUEST NUMBER MUM 8ER DATE DESCRIPi10N 1.
7/22/87 Letter from Hopkins to Nauman, subject:
Issuance of Amendment No. 67 to Facilit/
Operating License No. NPF-12 regarding Engineered Safety Features Response Times
- Virgil C. Summer Nuclear Power Station, Unit No. 1 (TAC No. 64150).
(13pages) e
July 22, 1987 Docket No. 50-395 Distribution: Docket File NRC PDR Local POR SYarga Glainas PAnderson JHopkins Mr. D. A. Nauman OGC-Bethesda DHagan Vice President, Nuclear Operations EJordan JPartlow South Carolina Electric & Gas Company TBarnhart 4 Wanda Jones P.O. Box 764 (Mail Code 167)
EButcher CLi Columbia, South Carolina 29218 ACRS 10 GPA/PA ARM /LFM6
Dear Mr. Nauman:
PD21 r/f
SUBJECT:
ISSUANCE OF AMENDMENT NO. 67 TO FACILITY OPERATING LICENSE NO. NPF-12 REGARDING ENGINEERED SAFETY FEATURES RESPONSE TIMES -
VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 (TAC NO. 64150)
The Nuclear R&gulatorr Ccmission has issued the enclosed Amendment No, to Facility Operating License No. N?F-12 for the Virgil C. Sumer Nuclear Station, Unit No.1. The amendment consists of changes to the Technical Specifications in response to your application dated December 12, 1986, as supplemented April 9, 1987.
The amendment clarifies the Service Water System and Reactor Building Cooling Unit response times.
A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Comission's next regular bi-weekly Federal Register notice.
Sincerely, h
Jon B. Hopkins, Project Manager Project Directorate 11-1 Division of Reactor Projects I/11
Enclosures:
1.
Amendment No. 67 to NPF-12 i
2.
Safety Evaluation cc w/ enclosures:
See next page
- See previous concurrence LA:P021:DRPR*
PM:P021:0RPR*
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EAdensam kins /dsf JHop/87 PAnderson MY ou/qq 74 4/87 7/1( o/
7/14 7/14/87
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Mr. D. A. Nauman South Carolina Electric & Gas Company Virgil C. Summer Nuclear Station Cc:
Mr. William A. Williams, Jr.
Technical Assistant - Nuclear Operations Santee Cooper P.O. Box 764 (Mail Code 167)
Columbia, South Carolina 29218 J. B. Knotts, Jr., Esq.
Bishop, Liberwan, Cook, Purcell and Reynolds 120017th Street, N.W.
Washington, D. C.
20036 Resident Inspector / Summer NPS c/o U.S. Nucleer Ranu14 tory Commission Route 1, Box 64 Jenkinsville, South Carolina 29065 Regional Administrator, Region 11 U.S. Nuclear Regulatory Commission, 101 Marietta Street, N.W., Suite 2900 Atlanta, Georgia 30323 4
Chairman, Fairfield County Council P.O. Box 293 Winnsboro, South Carolina 29180 Attorney General Box 11549 Columbia, South Carolina 29211 Mr. Heyward G. Shealy, Chief Bureau of Radiolo91 cal Health South Carolina Department of Health and Environmental Control 1
2600 Bull Street Columbia, South Carolina 29201 l
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UNITED STATES
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'n NUCLEAR REGULATORY COMMISSION 5
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SOUTH CAROLINA ELECTRIC & GAS COMPANY SOUTH CAROLINA PUBLIC SERVICE AUTHORITY DOCKET NO. 50-395 VIRGIL C. SUMER NUCLEAR STATION, UNIT N0.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 67 l
License No. NPF-12 l
e gulatory Cornission (the Comissicn) has found that:
1.
The Nuclear e
A.
The application for amendment filed by South Carolina Electric & Gas Company and South Carolina Public Service Authority (the licensees),
dated December 12, 1986, as supplemented April 9, 1987, complies with the standards and req)uirements of the Atomic Energy Act of 1954, as amended (the Act, and the Comission's rules and regulations set forth in 10 CFR Chapter I; i
8.
The facility will operate in conformity with the application, the provisions of tie Act, and the rules and regulations of the Ccm ission; l
C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will t'e conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this arandment is in accordance with 10 CFR Part 51 I
of the Comission's regulations and all applicable requirements l
have been satis'ied.
2.
Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachnent to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-12 is hereby smended to read as follows:
1
O-2 (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.
67. and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
South Carolina Electric & Gas Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This amendment is effective as of its date of issuance, and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION I
.C.O6-l
%g Elinor G. Adensam. Director Project Directorate !!-1 Division of Reactor Projects 1/11
Attachment:
Changes to the rechnical Specifications g
' bate of Issuance: July 22, 1987 i
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ATTACHMENT TO LICENSE AMENDMENT F
AMENDMENT NO. 67 TO FACILITY OPERATING LICENSE NO. NPF-12 DOCKET NO. 50-395 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. Corresponding overleaf pages are also provided to maintain document completeness.
Remove Pages Insert Pages 3/4 3 30 3/4 3-30 t
3/4 3-31 3/4 3-31 B 3/4 3 1b B 3/4 3-lb B 3/4 3-1c B 3/4 6-4 8 3/4 6-4 3.
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NSTRUMENTAT!bN TABLE 3.3-5(Continued]
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECON05 e.
Reactor Building Purge and Exhaust Isolation Not Applicable f.
Emergency Feedwater Pumps Not Applicable g.
Service Water System 71.5(41/81.5(5) h.
Reactor Building Cooling Units 76.5(4)/86.5(5) 1.
Control Room Isolation Not Applicable 3.
Pressurizer Pressure-Low a.
SafetyInjection(ECCS) 1 12.0(2)/27.0(1) b.
Reactor Trip (from SI)
$ 3.0 c.
Feedwater Isolation
< 10.0 d.
Containment Isolation-Phase "A"
[45.0(4)/55.0(5) e.
Reactor Building Purge and
, I.
Exhaust Isolation Not Applicable f.
Emergency Feedwater Pumps Not Applicable g.
Service Water System
< 71.5(4)/81.5(5) h.
Reactor Building Cooling Units
[76.5(4)/86.5(5) 1.
Control Room Isolation Not Applicable 4.
Different__ir1 Pressure Between Steam Lines-High a.
SafetyInjection(ECCS) 5 12.0(2)/22.0(3) b.
Reactor Trip (from SI) 1 3.0 c.
Feedwater Isolation
< 10.0 d.
Containment Isolation-Phase "A"
[45.0(4)/55.0(5)
SUMER - UNIT 1 3/4 3-30 Amendment No. 67
N TABLE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TINES INITIATING SIGNAL AND FUNCTION RESPONSE TINE IN SECONOS e.
Reactor Building Purge and Exhaust Isolation Not Applicable f.
Emergency Feedwater Pumps Not Applicable g.
Service Water System
< 71.5(4)/81.5(5) h.
Reactor Building Cooling Units
[76.5(4)/86.5(5) 1.
Control Room Isolation Nat Applicable 5.
Steam Line Pressure-Low a.
Safety Injection - ECCS 512.0(2)/22.0(3) b.
Reactor Trip (from SI)
$ 3.0 c.
Feedwater Isolation
< 10.0 d.
Containment Isolation - Phase "A"
[45.0(4)/55.0(5) e.
Reactor Building and Purge and
'a Exhaust Isolation Not Applicable
'f.
Emergency Feedwater Pumps Not Applicable g.
Service Water System
< 71.5(4)/81.5(5) h.
Reactot Building Cooling Units
[76.5(4)/86.5(5) 1.
Steam Line Isolation i 10.0 j.
Control Room Isolation Not Applicable I
j 6.
Steam Flow in Two Steam Lines - High Coincident with T,yg--Loc Low s
a.
Steam Line Isolation 1 12.0 7.
Reactor Buildina Pressure-High-2 a.
Steam Line Isolation d 9.0 SU M ER - UhlT 1 3/4 3 31 Amendment No.67
INSTRUNENTATION BASES REACTOR PROTECTION SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (continued)
Several automatic logic functions included in this specification are not necessary for Engineered Safety Feature System actuation but their functional capability at the specified setpoints enhances the overall reliability of the Engineered Safety Features functions.
These automatic actuation systems are l
purge and exhaust isolation from high containment radioactivity, turbine trip and feedwater isolation from steam generator high-high water level, initiation of emergency feedwater on a trip of the main feedwater pumps, automatic transfer of the suctions of the emergency feedwater pumps to service water on low suction pressure, and automatic opening of the containment recirculation sump suction valves for the RHR and spray pumps on low-low refueling water storage tank level.
The servics.;ater response time includas:
- 1) the start of the service water pumps and, 2) the service water pumps disc'iat ;e valves (3116A,B,C-SW) stroking to the fully opened position.
This corditi m of the valves assures that flow will become established through the coyoneret cooling water heat exchanger, diesel generator coolers, HVAC chP.ier, and to the suction of the service water booster pumps when these components are placed in-service.
Prior to this time, the flow is rapidly approaching required flow and i
sufficient pressure is developed as valves finish their stroke.
Each of the
'above-listed components will be starting to erform their accident mitigation function, either directly or indirectly depending upon the use of the compo-nent, and will be operatic 9al within the service water response time of 71.5/81.5 secondsE.
Only the service water booster pumps have a direct impact on the accident analysis via the RBCUs' heat removal capability as discussed below.
The RBCU response tire includes:
- 1) the start of the RBCU fan and the service water booster pumps and, 2) all the service water valves which must be driven to the fully opened or fully closed position.
This condition of the valves allows the flow to become fully established through the RetU.
Prior to this time, the flow is rapidly approaching required flow as the
(
1/ Total time is 1.5 second instrument response after setpoint is reached, plus 10 seconds diesel generator start plus 10 seconds to reach service water pumpstartandbegin3116-SWopeningviaEngineeredSafetyFeaturesLoading Sequencer, plus 60 seconds stroke time for 3116-SW.
During this total time, l
the service water pumps start and the service water pump discharge valve begins to open at 11.5 seconds and the pump discharge valve is fully open at 71.5 seconds without a diesel generator start required and 21.5 seconds and 81.5 seconds including a diesel generator start.
SUMER - UNIT 1 B 3/4 3-lb teendment No. 67
, -, ~. - -,
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INSTRUMENTATION BASES valves finish their stroke.
Although the RBCU would be removing heat through-out the Engineered Safety Features response time, the accident analysis does not assume heat removal capability from 0 to 71.5 seconds E ecause the b
industrial cooling water system is not completely isolated until 71.5 seconds.
A linear ramp increase from 95% fell heat removal capability to 100% full heat removal capability is issumed by the accident analysis to start at 71.5 seconds and end at 86.5 secondsE.
Full heat removal capability is assumed at 86.5 seconds based on the position of the valve 3107-SW.
,.i e
2/ Total tim'e"~is 1.5 second instrument response af ter setpoint is reached, plus
- 10 second diesel start plus 60 seconds" for valves to isolate industrial cooling water system.
3/ Total time is 1.5 second instrument response af ter setpoint is reached, plus
~ 10 second diesel generator start plus 75 seconds to stroke valves 3107A,B-SW.
"During this time period, the Engineered Safety Features loading Sequencer starts the RBCV fans at 25 seconds and service water booster pumps at 30 seconds after the valves begin to stroke.
SLM4ER - UNIT 1 B 3/4 3 Ic bendment No. 67
CONTAINMENT SYSTEMS BASES 3/4.6.2.2 SPRAY ADDITIVE SYSTEM The OPERABILITY of the spray additive system ensures that sufficient NaOH is added to the reactor building spray in the event of a LOCA.
The limits on NaOH volume and concentration ensure a pH value of between 7.8 and 11.0 for the solution recirculated within containment after a LOCA.
This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.
The contained solution volume limit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics.
These assumptions are con-sistent with the iodine removal efficiency assumed in the accident analyses.
3/4.6.2.3 REACTOR BUILDING COOLING SYSTEM The OPERABILITY of the reactor building cooling system ensures that
- 1) the reactor building air tem during normal operation, and 2)perature will be maintained within limits adequate heat removal capacity is available when operated in co.1 junction with the reactor building spray systems during post-LOCA conditions.
}.
.The reactor building cooling system and.the rwactor building spray system are redundant to each other in providing post accident cooling of the reactor building atmosphere.
As a result of this redundancy in cooling capability, the allowable out of service time requirements for the reactor building cooling systemhavebeenappropriatelyadjusted.
However, the allowable est of service time requireme e s for the reactor building spray system have been ma*ntained consistent with that assigned other inoperable ESF equipment since ths reactor building spray system also provides a mechanism for removing iodine frot the reactor building atmosphere.
The accident analysis requires the service water booster pump to bi pass-ing 4,000 gpa to both RBCV's within 86.5 seconds.
This time encompasses the driving of all necessar service wSter valves to the correct positions, i.e.,
i fully opened or fully c.osed.
Reference Technical Specification Bases B3/4.3.1 and B3/4.3.2 for additional details.
3/4.6.3 PARTICULATE IODINE CLEANUP SYSTEM The OPERABILITY of the containment filter trains ensures that sufficient iodine removal capability will be available in the event of a LOCA.
The i
reduction in containment iodine inventory reduces 15e resulting site boundary
{
radiation doses associated with containment leakage.
The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analyses.
SLM4ER UNIT 1 B 3/4 6-4 Amendment No. H 67
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UNITEO STATES A
NUCLEAR REGULATORY COMMISSION i
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REAC70R REGULATION SUPPORTING AMEN 0 MENT NO. 67 TO FACILITY OPERATING LICENSE NO. NPF.12 SOUTHCAROQNAELECTRIC&GASCOMPANY SOUTH CAROL!kA PUBLIC SERVICE AUTHOR 11Y Vl'tGIL C. SUH4ER NUCLEAR STATION, UN!! NO.1 DOCKET NO. 50-395
1.0 INTRODUCTION
By letter dated December 12, 1986 South Carolina Electric & Gas Company submitted a request for changes to the Ytrgil C. Sunner Nuclear Station Technical Specifications (TS).
The amendment would revise Table 3.3-5 "Engineered Safety Features Response Times " and its associated Basis. This change would clarify the Service Water System and Reactor Building Cooling Unit response times.
8 A Notice of Consideration of Issuance of hnendment to License and Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing related to the requested action was published in the Federal Register on January 14,1987(52FR1557). No connents or requests for hearing were received.
By letter dated April 9, 1987, the licensee submitted additional inforination which clarified certain aspects of the December 12, 1986 application.
Since the additional information did not change the requested TS revision or affect the staff's initial determination, the amen h nt was not renoticed in the Federal, Register.
L 2.0 EVA:.UATION The proposed changes will increase the required response time for the service water system (SWS) and the reactor building cooling units (RBCU) to 'as-built" values. The proposed SWS response time is 81.5 seconds (including diesel generator delay) or 71.5 seconds (not including diesel generatordelay).
The response time of 81.5 secoria.ncludet 1.5 seconds fc. instrur,ent res sonse.10 seconds to start the diesel generator,10 i
se.onds to start tie service water pump and 60 seconds to open the SWS pump discharge valves (3116A, B, C-SW).
1 J
The rain function of the SWS is to deliver water to the following components:
component cooling water /SW heat exchanger; diesel generator coolers; heating, ventilation, and air conditioning chiller; and the suction of the SW booster pumps.
Each of the above listed components wi)1 be operational within the SWS retponse time. The licensee identified that only the service water booster pumps have a direct safety impact on the containment analysis via the RBCU's heat removal capability.
This impact was evaluated in conjunction with the effect of the RBCU response time as discussed below.
The proposed RBCU response time is 86.5 seconds (including generator delay) or 76.5 seconds (not including diesel generator delay). The response time of 86.5 seconds includes 1.! seconds for instrument response, 10 seconds to start the diesel generator and 75 seconds to open the slowest valves (3707A, B-SW) in the SW line. The Itcensee performed two contain.w nt response analysis cases, i.e., the worst pressure case and the worst temperature case.
In the analysis, no flow to the RBCU was assumed until 71.5 seconds because valves 3110A, B-SW would remain completely closed throughout this time period to isolate the SWS from the non-safety relsted, non-seismic industrial cooling water (Cl) system.
At 71.5 seconds, the RBCU discharge valvet 3107A, B-SW would still be o>ening and would be passing 3800 gpm of service water. At 86.5 seconds,
.1 tie valves 317A, B SW would be fully open and passing 4000 gps.
In tht-analysis, the licensee assuned a linear ramp increase of RBCU beat removal capability from 951 at 71.5 seconds to 100% at 86.5 seconds.
As indicated in the licenste's letter of April 9,1987, the analyses used the mass and energy release data that were calculated by the Westinghouse LOFTRAM code April 1986 revision.
This revision of the LOFTRAN code was found acceptable by the staff 1: its safety evaluation transmitted to Westinghouse by letter dated May 27, 1986. The computer code CONTEMPT LT/26 was used by the licensee to calculate contairment pressure and temperature responses. This is consistent with the staff's recomendation in SRP Section 6.2.1.1.A.
A peak pressure of 46 psig was calculated which is less than the design pressure of 57 psig with some margin. The peak calculated temperature of 322'F is bounded by the equipment qualification temperature profiles specified in FSAR Figures 3.11-8, 3.11-9 and 3.1*-10.
The staff has reviewed the licensee's analyses, including the assumptions, methodologies, and results. Based on the above review, the staff finds that the proposed TS changes to increase the response time of the SWS and RBCU to the specified values have only a minor impact on the containment pressure and temperature responses.
The resulting containment pressure and temperature were determined in accordance with SRP Section 6.2.1.1.A and are within the design values specified in the FSAR. The staff, therefore, concludes that the proposed TS changes are acceptable.
3
3.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change in the surveillance requirement of a facility component located within the restricted area, as defined in 10 CFR Part 20. The ste ff has determined that these amendments involve no significant increase in the amounts, and no significant change in the types of any effluents that may be released offsite; and that there is no significant increase in individual or cumulative occupational radiation exposure. The Connission has previously issued a proposed finding that these amendments involve no significant hazard censideration, and there has been no public connent on such finding. Accordingly, this amendments meets the eligibility criteria for categorica! exclusion set forth in 10 CFR 51.22(c)(9).
Pursuantto10CFR51.22(b)noenvironmentalimpact statement or environmental assessment need be prepared in connection with the issuance of this amensnent.
d0 CONCLUSION The Connission made a proposed determination that this amendment involves no significant hazards consideration which was published in the Federal Register (52 FR 1557) on January 14, 1987, and consulted with the state of Mabama.
No pubite comments or requests for hearing were received and The Connission has issued a "Notice of Consideration of Issuance of knendments to facility Operating ucenses and Opportunity for Prior pas Hearing" which was published in the FEDERAL REGISTER on January 14, 1987 (52 FR 1557) and consulted with the state of South Carolina. No public connents or requests for hearing were received, and the state of South Carolina did not have any connents.
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will nct be endangered by operation in the proposed manner, and (2) such activities will be conducted in ompliance with the Connission's regulations and the issuance of this ament:ent will not be inimical to the connon defense and security or to the health and safety of the public.
Principal Contributors:
J. Hopkins, Project Directorate !!-1 C. Li, Plant Systems Branch Dated: July 22, 1987