ML17156B455
| ML17156B455 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/03/1989 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17156B456 | List: |
| References | |
| NUDOCS 8911140292 | |
| Download: ML17156B455 (14) | |
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UNITED STATES NUGAE EAR R EG ULATORY COMMISSION WASHINGTON, O. C. 20555 PENNSYLVANIA POWER 8
LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.
DOCKET NO. 50-388 SUS UEHANNA STEAM ELECTRIC STATION, UNIT "
AMENDMENT TO I-ACTLITY OPERATING LICENSE Amendment No.
59 License No.
NPF-22 1.
The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The apolication for the amendment filed by the Pennsylvania Power 8
Light Company, dated June 16, 1989 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's requlations set forth in 10 CFP. Chapter I; B.
The facility wi 11 operate in conformity with the application, the provisions of the Act, and the requlations of the Commission; C.
There is reasonable assurance:
(i) that the activities authorized bv this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set foI.th in 10 CFR Chapter I; D.
The issuance of this amendment wi>1 not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 5]
of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No.
NPF-22 is herebv amended to read as follows:
(2)
Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 59 and the Environmental Protection Plan con-tained in Appendix B, are hereby incorporated in the license.
PP8L shall operate the <aci lity in accordance with the Technical Specifica-
- tions and the Environmental Protection Plan.
89 1 1 14p K pa'ppp388 292 8911P3 P
I
3.
This license amendment is effective as of its date of issuance FOR THE NUCLEAR REGULATORY COMMISSION
/S/
Mohan C. Thadani fo Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance:
November 3, 1989 PDI-2/LA
'g0'Brien PDI-2/PM MThadani Jt /L7V/89 PDI-2/D WButler Ei/5 /89 l( /+/89 yp,A OGC
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This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COb'YISSION
Attachment:
Changes
+o the TechnicaI Specifications Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II Date o~ Issuance:
Noveraber 3, 1989
ATTACHMENT TO LICENSE AMENDMENT N0.59 FACILITY OPERATING LICENSE NO.
NPF-22 DOCKET NO. 50-388 Replace the following pages of the Appendix A Technica> Soecifications with enclosed oages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
The overleaf oages are provided to maintain document completeness.*
REMOVE 3/4 6-21 3/4 6-22 3/4 8-31 3/4 8-32 3/4 8-33 3/4 8-34 B 3/4 6-3 8 3/4 6-4 INSERT 3/4 6-21 3/4 6-22*
3/4 8-3]*
3/4 8-32 3/4 8-33 3/4 8-34*
B 3/4 6-3*
B 3/4 6-4
TABLE 3. 6. 3-1 (Continued}
PRIMARY CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TIME Seconds)
VALVE FUNCTION ANO NUMBER Automatic Isolation Valves (Continued}
Containment Atmos here Sam le ISOLATION SIGNAL S}
SV"25734 A,B SV-25736 A
SV-25736 B
SV-25740 A,B SV"25742 A,B SV-25750 A,B SV-25752 A,B SV-25774 A,B SV-25776 A
SV-25776 B
SV-25780 A,B SV-25782 A,B ll SV-25737 SV-25738 SV-25767 SV-25789 Reactor Coolant Sam le HV-243F019 HV-243F020 Li uid Radwaste HV"26108 A1, A2 HV-26116 Al,A2 RHR -
Su ression Pool C~l/5 HV-251F028 A, B GS TBGG( )(
)
HV-252F015 A,B HPCI Suction ( }
'V-255F042 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A 15 15 90 60 90 B,Y B,Y B,Y B,Y B,Y B,Y B,Y B,Y B,Y B,Y B,Y B,Y B,Y,R B,Y,R B,Y,R B,Y,R B,C B,C B,Z B,Z X,Z X,Z L,LB SUSQUEHANNA - UNIT 2 3/4 6-21 Amendment No. 59
TABLE 3. 6. 3-1 (Continued)
PRIMARY CONTAINMENT ISOLATION VALVES vALvE PUNC ION ANO NUMBER Automatic Isolation Valves (Continued)
Su ression Pool Cleanu (0)
XV-25766 HV-25768 MAXIHUM ISOLATIOH TIME Seconds 35 30 ISOLATION (
SIGNALS S
A,Z A,Z HPCI Vacuum Breake~
HV-255F075 HV-255F079 15 15 LB,Z LB,Z RCIC Vacuum Bragger HV-249F062 HV 249F084 dl TIP Ball Valves
(
10 10 KB,Z KB,Z C51-J004 A,B,C,D, E A,Z
>anua>
- splat:on Valves MSIV-LCS Bleed Valve HV-239FQOl B, F, K, P
~
~IIa lair "
HV-241F032 A,b RWCU Return HV 24182 A,h RC I C In ection HY-249F013 2-49-020 SUSQUEHANNA - UNIT 2 3/4 6-22
~
Amendment Ho.
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~ 4 4
~
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s 4 'CU5
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I i HV-0 2224
~V-v 122 28 HV-01224Al
~v-0122481 r V-01224AZ HV-0122482 HV-21144A HV-211448 HV-~8693A
,"V-086338
~V-0120 '.Al hV-01201AZ HV-0120181 wv-0120182 HV-21210A HV-212108 SV-21215A HV-2.12158 RV-25766 HV-25768 HV-22603 ROARS'~
QHRSW RHRSW RHR5W RHRSW RHRSW ESW
~SW ESW ESW RHRSW RHR5W RHRSW RHRSW RHRSW RHRSM RHRSW RHR5W Cont..sc'.
Cont. I:
Cont Isol SUS(UEHANNA - UNIT 2 3/4 8-31 Amendment No 56
TABLE 3. 8. 4. 2. 1-1 (Continued)
MOTOR-OPERATEO VALVES THERMAL OVERLOAD PROTECTION CONTINUOUS VALVE NUMBER SYSTEM(S)
AFFECTED HV-21345 HV-21313 HV-21346 HY-21314 HV-E1 1-2F009 HV-Ell-2F040 HV-G33-2F001 HV-Ell-2F103A HY-E11-2F075A HV-E1 1-2F048A HV-E1 1-2F006C HV"Ell-2F004C HV-Ell-2F015A HV-Ell-2F024A HV"E21-2F015A HV"E41-2F002 HV-B21-2F016 HY-Ell"2F 022 HV-El1-2F010A HV"E 11-2F004A HV-Ell-2F006A HV-E1 1-2F027A HV-Ell"2F007A HV-Ell"2F 104A HV-E 11-2F026A HV-Ell-2F028A HY"Ell-2F047A HV-E 11-2F073A HV-Ell-2F003A HV-Ell-2F017A HV-E21-ZF001A HV-E21-2F031A HV-E21-2F004A HV-E21" 2F005A HV-E1 1-2F021A HY-Ell-2F016A HV-25112 HV-E51-2F007 HV-E51" 2F084 HV-Ell-2F027B HV-Ell-2F048B HV-Ell-2F015B HV-Ell-ZF006B Cont.
Cont.
Cont.
Cont.
RHR RHR RWCU RHR RHRSM RHR RHR RHR RHR RHR CS HPCI NSSS RHR RHR RHR RHR RHR RHR RHR RHR RHR RHR RHRSW RHR RHR CS CS CS CS RHR RHR RHR RCIC RCIC RHR RHR RHR RHR Isol.
Isol.
Isol.
Isol.
SUSQUEHANNA - UNIT 2 3/4 8-32 Amendment No. 59
TABLE 3.8.4.2. 1-1 (Continued}
MOTOR-OPERATEO VALVES THERMAL OVERLOAD PROTECTION CONTINUOUS VALVE NUMBER HV-Ell-2F0218 HV-Ell-2F0108 HV-E11-2F0048 HY-Ell-2FQ078 HY"Ell-2F1Q48 HY-Ell-2FQ268 HY-E11-2F0288 HY-E1 1-2F04? 8 HV-E 11" 2F 0168 HY"E1 1-2F0038 HV-Ell-2F0178 HV-E21-2F0318 HV-E21-2F0018 HV-E11" 2F 1038 HV-Ell-2F0758 HV-Ell-2F0738 HV-Ell-2FOOGD HV"Ell-2F004D HV-Ell-2F0248 HV-E21-2F0158 HV-E21-2F0048 HV-E21-2F0058 HV-E32-2F001K HV-E32-2F002K HV-E32-2F003K HV"E32-2FOOlp HV-E32-2F002P HV-E32"2F003P HV-E32-2F0018 HY-E32-2FOQ28 HY-E32-2F0038 HV-E32-2F001F HY-E32"2F002F HV-E32-2F003F HV"E32-2F006 HV-E32-2F007 HV"E32-2F008 HV-E32-2F009 HV-E51-2F 045 HV"E51-2F012 HV-E51-2F013 HV"25012 SYSTEM(5)
AFFECTED RHR RHR RHR RHR RHR RHR RHR RHR RHR RHR RHR CS CS RHR RHRSW RHRSW RHR RHR RHR CS CS CS MSIV MSIV MSIV MSIY MSIV MSIY MSIV MSIV MSIV MSIV MSIV MSIV MSIV MSIV MSIY MSIV RCIC RCIC RCIC RCIC SUSQUEHANNA - UNIT 2 3/4 8-33 Amendment No. 59
Tab1e 3.8.4.i 1-1 (Continued)
MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION CONTINUOUS VALVE NUMBER HV-E51-2F046 HV-E51-2F008 HV-E51-2F031 HV-E51-2F010 HY-E51-2F019 HV-E51-2F060 HV-E51-2F059 HY-E51-2F022 HV-E51-2F062 HV-E41-2F 012 HV-E41-2F001 HV-E41-2F011 HV-E41-2F 006 HV-E41-2F079 HV-E41-2F059 HV-E41-2F 004 HY-E41-2F003 HV-E41-2F042 HV-E41-2F075 HY-E41-2F008 HV-E41-2F007 HV-E41-2F066 HY-G33-2F004 HV-821-2F019 HV-E11-2F 008 HV-E11-2F023 HV-E11-2F049 HV-831-2F032A HY-B31-2F032B HV-B31-2F031A HV-B31-2F031B HV-24182A HV-24182B SYSTEM(S)
AFFECTED RCIC RCIC RCIC RCIC RCIC RCIC RCIC RCIC RCIC HPCI HPCI HPCI HPCI HPCI HPCI HPCI HPCI HPCI HPCI HPCI HPCI HPCI RWCU NSSS RHR RHR RHR Rx Recirc Rx Recirc Rx Recirc Rx Recirc RWCU RWCU SUS(UEHANNA - UNIT 2 3/4 8"34 Amendment No.
32
CONTAINMENT SYS.EMS BASES 3/4. 6. 2 OEPRESSURIZATION SYSTEMS The specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 53 psig during primary system blowdown from full operating pressure.
The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system.
The suppres-sion chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from 1055 psig.
Since all of the gases in the drywell are purged into the suppression chamber air space during a loss of coolant accident, the pressure of the liquid must not exceed 53 psig, the suppression chamber maximum pressure.
The design volume of the suppression
- chamber, water and air, was obtained by considering that the total volume of reactor coolant and to be considered is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.
Using the minimum or maximum water volumes given in this specification, containment pressure during the design basis accident is approximately 45.0 psig which is below the design pressure of 53 psig.
Maximum water volume of 133,540 ft results in a downcomer submergence of 12 feet and the minimum volume of 122,410 ft> results in a submergence approximately 24 inches less.
The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation.
Thus, with respect to the downcomer submergence, this specification is adequate.
The maximum temperature at the end of the blow-dOwn teSted during the HumbOldt Bay and BOdega Bay teStS waS 170 F and thiS iS conservatively taken to be the limit for complete condensation of the reactor
- coolant, although condensation would occur for temperatures above 170'F.
Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3,5.3.
Under full power operating conditions, blowdown from an initial suppression chamber water temperature of 90 F results in a water temperature of approx.
imately 1284F imtdiately following blowdown ~hich is below the 1704F used for complete condensation via T-quencher devices.
At this temperature and atmos-pheric pressure, the available NPSH exceeds that required by both the RHR and core spray pumps, thus there is no dependency on containment overpressure during the accident injection phase.
If both RHR loops are used for containment cooling, there is no dependency on containment overpressure for post-LOCA operations.
Experimental data indicate that excessive steam condensing loads can be avoided if the peak local temperature of the suppression pool is maintained below 200 F during any period of relief valve operation.
Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chaaber loadings.
SUS)UEHANNA - UNIT 2 8 3/4 6-3
CONTAINMENT SYSTEMS BASES OEPRESSURIZATION SYSTEMS Continued Because of the large volume and thermal capacity of the suppression
- pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.
By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken.
The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered.
Particular atten-tion should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.
In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a safety-relief valve inadvertently opens or sticks open.
As a minimum this action shall include:
(1) use of all available means to close the valve, (2) initiate suppres-sion pool water cooling, (3) initiate reactor
- shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety relief valve to assure mixing and uniformity of energy insertion to the pool.
3/4. 6. 3 PRIMARY CONTAINMENT ISOLATION VALVES The OPERABILITY of the primary containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GOC 54 through 57 of Appendix A to 10 CFR 50.
Containment isolation within the time limits specified for those isolation valves designed to close auto-matically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.
3/4.6.4 VACUUM RELIEF Vacuum relief breakers are provided to equalize the pressure between the suppression chamber and drywell.
This system will maintain the structural integrity of the primary containment under conditions of large differential pressures.
The vacuum breakers between the suppression chamber and the drywell must not be inoperable in the open position since this would allow bypassing of the suppression pool in case of an accident.
There are five pairs of valves to provide redundancy so that operation may continue for up to ?2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with no more than one pair of vacuum breakers inoperable in the closed position.
SUSQUEHANNA " UNIT 2 B 3/4 6-4 Amendment No. 59