ML20127E326
| ML20127E326 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 01/07/1993 |
| From: | Larkins J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20127E330 | List: |
| References | |
| NUDOCS 9301190254 | |
| Download: ML20127E326 (11) | |
Text
. -..
'o UNITED STATES
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NUCLEAR REGULATORY COMMISSION 5
l WA$HINGTON, D C. 20555
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NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-713 (M PER NUCLEAR S1A110N 6Bl@HENT TO JACILITY OPERATING LICENSE Amendment No.157 License No. DPR-46 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Nebraska Public Power District (the lictnsee) dated May 4, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance:
(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that ruch activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this license amendment will not be ininical to the common defense and security or to the hehlth and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicahic requirements have been satisfied.
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i 9301190254 930107 PDR ADOCK 05000290:
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. 2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. DPR-46 is hereby amended to read as follows:
2.
Igchnical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.157, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
& lid W -
iGhJohnT.Larkins, Director Project Directorate IV-1 Division of Reactor Projects - Ill/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: January 7, 1992 l
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ATTACHMENT TO LICENSE AMENDMENT NO.157 FACILITY OPERATING LICENSE NO. DPR-46 QQCKET NO. 50-298 Replace the following pages of the Appendix A Technicai Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
REMOVE PAGES INSERT PAGES 162 162 162a 162a 167 167 168 168 169 169 171 171 172 173 174 175 178 178 183 183 i
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[]EITING CONDITIONS MR OPERATION
' SyRVEILIANCE REOUIREMENTS 15 3.7.A (Cont'd) 4.7.A.2.e (cont'd) repeated provided --locally ' measured -
leakage. reductions,_ achieved by:
repairs, reduce - the containment's overall me'asured -leakage ra t e --
sufficiently to meet the acceptance-
- criteria, f.
Local Leak Rate - Tests l'
With the exceptions specified below, local leak rate tests (LIRT's) shall be perforned on the primary contain '
ment testable penetrations ar.d isolation valves at.'a pressure' of
$8 psig during each reactor shutdown.-
for refueling, or other _' convenient-intervals, but in no case at; inter-vals greater than two _. years. The l
test duration of all valves and penetrations shall be of sufficient length-to determine repeatable results. The total _ acceptable leak-age for all valves and-penetrations other than the MSIV's'is 0.60 La.
2.
Bolted double-gasket seals shall be
_l tested after each opening and during -
each reactor shutdown for refueling, or other convenient-intervals but in no case at ' intervals - greater than two years.
3.
The - main steam isolation valves (MSIV's) shall be-tested at-.a pressure of 29 psig.
If-a - total 1eakage rate of 115 sef/hr for any ;
one MSIV ' is exceeded, repairs and retest shall be performed to correct-the condition. This is an exemption to Appendix J of-10CFR50.
P
-Amendment No.-2S, ;, 80, S2. 157
-162-
' LIMITING CONDITIONS FOR OPERATION EURVEILIANCE REOUfREMfiqE
-3.7.A (Cont'd) 4.7. A.2. f (cont'd) 4 Main = steam'line and feedwater-line expansion bellows shal1~ be _ tested _by
-l
- pressurizing between the laminations of the bellows at a pressure of 5-psig.- This is an exemption to:
Appendix J of-10CFR50.
5.
The personnel ~ airlock shall be-tested at 58 psig at intervals no -
longer than six months. This. test--
ing may be extended _ to the next refueling outage (not.to exceed 24
-months) provided - that there have=
been no airlock openings since the last successful _ test at 58 psig. In the event the personnel airlock is not opened between refueling out-ages, it shall be leak checked at 3 psig at intervals no longer than six months. Within three days-of open-ing- (or every three days during periods of frequent opening) when containment integrity is required, test. the personnel airlock at 3 psig.
This is : an exemption to Appendix-J of 10CFR50.
The maximum allowable leakage at a
test pressure of 58 psig is 12 scfh.
Leakage. measured at test pressure less than 58 psig is adjusted to the-equivalent value at 58 psig, g.
Deleted h.
Drvwell surfaces The interior surfaces of the drywell--
and to rus --
shall- -be visually inspected each operating cycle for evidence of torus corrosion or_
' leakage.
Amendment No. 82, SS. 91, 1107 157
-162a-
LIMITING CONDfTIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.7 D (cont'd.) _
.4.7.Di(cont'd.)
b.
At least once per quarter:
(1) All normally _ open power operated isolation ' valves (except- 'for the main steam line power-operated-isolation. valves)_ shall be_ fully closed and reopened.
(2) With the reactor power less than 75%,
trip main steam' isolation valves individually and verify closure time, c.
At least once per. operating : cycle the operability of. the reactor-coolant system-instrument line flow check valves shall be verified.
d.
At least once per operating cycae,-
while shutdown, the devices that limit the maximum opening angle to 60' shall be verified functional for -
the following valves: PC-230MV, PC-231MV, PC-232MV, and C 233MV.
2.
In the event any isolation valve 2.
Whenever an isolation valve listed specified in Table 3,7.1 becomes in Table 3.7.1 is inoperable, the inoperable, reactor power operation position of at least one other valve may continue provided at least one in each _ line having an inoperable valve in each line having an inoper-valve shall be recorded daily, able valve shall be in the mode corresponding to-the_
isolated l
condition.*
3.
If Specification 3.7.D 1 and 3,7.D.2 cannot be met, an orderly shutdown shall be iniciated and the reactor shall be in the Cold Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
- Isolation valves closed to satisfy these requirements may be reopened on an intermittent basis under administrative control.
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Amendment No. 4M r 157 167-
g COOPER NUCLEAR STATION
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E TABLE 3.7.1 (Page 1)
PRIMARY CONTAINMENT ISOIATION VALVES an Number of Power Maximum Action On ll Operated Valves Operating Normal Initiating Valve & Steam Inboard Outboard Time (Sec) (1)
Position (2)
Simm1 (3)
U Main Steam Isolation Valves u
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MS-AO A,B,C, & D 4
35T55 o
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MS-AO A,B,C, 6 D 4
35T55 0
GC Drywell Floor Drain Iso. Valves 2
15 0
GC j$
RV-AO-82, RW-AO-83 g.
Drvvell Equipment Drain 2
15 0
' Iso. Valves RW-AO-94, RW-AO-95 Main Steam Line Drain 1
1 30 0
GC Valves MS-MO-74, MS-MO-77 Reactor Water Sample Valves 1
1 15 0
GC RR-740AV, RR-741AV Reactor Water Cleanup System 1
1 60 0
GC Iso. Valves RWCU-MO-15, RWCU-MO-18 RHR Suction Cooling Iso.
1 1
40 C
SC Valve RHR-MO-17. RHR-MO-18 RHR Discharge to Radwaste 2
20 C
SC Iso. Valves RHR-MO-57, RHR-MO-67 Suppression Chamber Purge &
2 15 C
SC Vent PC-245AV, PC-230MV Suppression Chamber N2 Supply 2
15 C
SC PC-237AV, PC-233MV
COOPER NUCLEAR STATION TABLE 3.7.1 (Page 2)-
PRIMARY CONTAINMENT ISOLATION VALVES
.o S:
Number of Power Maximum Action On-Operated Valves Operating Normal Initiating
.E Valve & Steam Inboard Outboard Time (Sec) (1)
Position (2)
Sirnal (3)
Primary Containment Purge & Vent 2
15 C
SC PC-246AV, PC-231MV l$
Primary Containment & N2 Supply 2
15 C
SC PC-238AV, PC-232MV y
Suppression Chamber Purge & Vent 1
40 C
SC(4)
PC-230MV Bypass.(PC-305MV)
Primary Containment Purge & Vent 1
40 C
SC(4)
PC-231MV Bypass (PC-306MV)
Dilution Supply PC-1303MV, PC-1304MV 2
15 C
SC-PC-1305MV, PC-1306MV 2
15 C
SC ey bilution Supply PC-1301MV, PC-1302MV 2
15 0
GC PC-1311MV,_PC-1312MV 2
15 O
CC Suppression Chamber Purge and Vent Exhaust 1
15 C
SC PC-1308MV Primary Containment Purge and Vent Exhaust 1
15 C
SC PC-1310MV l
Pages 171, 172, 173, 174, and 175 have been deleted.
Amendment No. -W r-1407 157
-171-1 (Next page is 176) l
_3.7 A & 4.7.A EASES (cont'd.)
trends.
Whenever a bolted double-gasketed penetration is broken and remade, the space between the gaskets is pressurized to determine that-the seals are performing properly.
It is expected that the majority of the leakage from valves, penetrations and seals would be into the reactor building. However, it is possible that leakage into other parts _of the facility could occur.
Such leakage paths that may affect significantly -the consequences of accidents are to be minimized.
Certain isolation valves are tested by pressurizing the volume between the inboard and l
outboard isolation valves. This results in conservative test results since the inboa d-valve, if a globe valve, will be tested such that the test pressure is tending to lift the globe off its seat. Additionally, the measured leak rate for such a test is conservatively assigned to both of the valves equally and not divided between the two.
The main steam and feedwater testable penetrations consist of a double _ layered metal bellows. The inboard high pressure side of the bellows is subjected to drywell pressure.
Therefore, the hellows is tested in its entirety when the drywell is tested. The bellows layers are tested for the integrity of both layers by pressurizing the void between the layers to 5 psig.
Any higher pressure could cause permanent deformation, damage and possible ruptures of the bellows.
Surveillance requirements for integrity of the personnel air lock are specified in (Exemption) to the letter, D. G. Eisenhut to J. M. Pilant, September 3, 1982.
When the Personnel Air Lock Leakage Test is performed at a test pressure less than 58 psig, the measured leakage must be adjusted to reflect the expected leakage at 58 psig.
Equation A-3 of Enclosure 3 (Franklin Research Center Technical Evaluation Report) to the
- letter, D.
G.
Eisenhut to J. M.
Pilant, September 3, 1982, defines the method of adj us tment.
The primary containment pre-operational test pressures are based upon the calculated primary containment pressure response in the event of a loss-of-coolant accident.
The peak drywell pressure would be about 58 psig which would rapidly reduce to 29 psig following the pipe break.
Following the pipe break, the suppression chamber pressure rises to 27 psig, equalizes with drywell pressure and therefore rapidly decays with the drywell pressure decay.
The design pressure of the drywell and suppression chamber is 56..ig.
Based on the calculated containment pressure response discussed above, the primary containment preoperational test pressure was chosen. Also, based on the primary containment pressure response and the fact that the drywell and suppression chamber function as a unit, the primary containment will be tested as a unit rather than the individual components separately.
The design basis loss-of-coolant accident was evaluated at the primary containment maximum allowable accident leak rate of 0.635%/ day at 58 psig. Calculations made by the NRC staff with leak rate and a standby gas treatment system filter efficiency of 90% for halogens and assuming the fission product release fractions stated in NRC Regulatory Guide 1.3, show that the maximum total whole body passing cloud dose is about 1.0 REM and the maximum i
total thyroid dose is about 12 REM at 1100 meters from the stack over an exposure duration l
of two hours. The resultant doses reported-are the maximum that would be expected in the unlikely event of a design basis loss-of-coolant accident.
These doses are'also based on the assumption of no holdup in the secondary containment resulting in a direct release of_ fission products from the primary containment through the filters and stack to che i
environs. Therefore, the specified primary containment leak rate and filter efficiency are conservative and provide margin between expected off-site-doses and 10 CFR 100 guidelines.
The water in the suppression chamber is used for cooling in the event of an accident;-
i.e.,
it is not used for normal operation; therefore, a daily Amendment No. &2, 91, 110,- 157
-178-
. _ =.
4.7.B & 4.7.C BASES Table 5.1 of ANSI N509-1980. The replacement tray for the adsorber. tray removed for the-test should meet the same adsorbent quality. Tests of the HEPA filters with _DOP aerosol shall be performed in accordance to ANS1 N510-1980.
Any filters found defective shall be replaced with filters qualified pursuant to Regulatory Position C.34 d. of Regulatory Guide 1.52, Revision 2, March, 1978.
All elements of the heater should be demonstrated to be functional and operable during the test of heater capacity. Operation of the heaters will prevent moisture buildup in the filters and adsorber system.
With doors closed and fan in operation, DOP aerosol shall be sprayed externally along the full linear periphery of each respective door to check the gasket seal.
Any detection of DOP in the fan exhaust shall be considered an unacceptable test result and the gaskets repaired and test repeated.
It system drains are present in the filter /adsorber banks, loop-seals must be used with adequate watet level to prevent by-pass leakage from the banks.
If significant painting, fire or chemical releas'e occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for operational use.
The determination of significance shall be made by the operator on duty at the time of the incident.
Knowledgeable staff members should be consulted prior to making this determination.
Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system perfo rmance capability.
If one Standby Cas Treatment subsystem is inoperable, the operable subsystem's operability is verified daily..This substantiates *he availability of the operable subsystem and thus reactor operation or refueling operation can continue for a limited period of time.
3.7.D 6 4.7.D
- DASES, Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment.
Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system.
Automatic initiation is_ required _to minimize the potential leakage paths from the containment in.
the event of a loss-of-coolant accident.
The maximum closure times for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncoveringL following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.
The USAR identifies those testable primary containment valves that perform an isolation function, and testable penetrations with Double 0 Ring Seals, and testable penetrations with testable Bellows ensuring that any changes thereto receive a 10CFR50.59 review. In addition, plant procedures also identify containment isolation valves, and testable pene-trations with Double 0-Ring Seals, and testable penetrations with testable Bellows changes to these procedures and the USAR are controlled by Technical Specification 6.2.1. A.4 (Administrative Controls).
.These valves are highly reliable,-have a low service requirement, and most are normally closed.
The initiating sensors and associated trip channels are also checked to demonstrate the capability for automatic isolation.
The test interval of once per operating cycle for automatic initiation 2
i LAmendment No. 82, 102, 13S, I 'd, 152, 157-183-
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