ML20113G824
| ML20113G824 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 05/04/1992 |
| From: | NEBRASKA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20113G822 | List: |
| References | |
| GL-91-08, GL-91-8, NUDOCS 9205120267 | |
| Download: ML20113G824 (8) | |
Text
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. LIMITING CONDIT10'E TOF OPERATIO!:
SUPVEILIA?;CE REOPIREME!!TS I
3.7.A (Cont'd)
- 4. 7./. 2.e (cont'd) repeated provided locally measured leakage redactions, achieved by repairs, reduce the containment's overall measured leakage rate sufficiently to meet the acceptance criteria.
f.
Loen1 Leak Rate Tests 1.
With the exceptions specified below, local leak rate tests (LUlT's) shall be performed on he primary containment testable penetrations and isolation valves at a pressure each reactor of 58 psig during_
shutdown for refueling, or other convenient intervals, but in no case at intervals greater than two years.
g The test duration of all valves and penetrations shall be of sufficient length to determine repeatable results.
1he total acceptable leakage for all valves and penetra tions other than the MSIV's is 0.60 La.
2.
Bolted double gasket reals shall be tested af ter each opening and during l
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 leakage rate of 11,5 scf/hr for any one MSIV is exceeded, 1, pairs and retest shall be performed to correct the condition. This is an exemption to Appendix J of 100FR50.
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LIMITING CONDIT10NF TOP OPERATIOJ SURVEILLANCE REOUIREMENTS l
3.7.A-(Cont'd) 4.7.A.2.f ' cont'd) 4.
Main steam line and feedwater line expansion belloes shall 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 ps'g at intervals no longer than six months.
This testing 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 tae persennel airlock is not opened between
. refueling outages. it shall be leak checked at 3 psig at intervals no longer than six months.
Within three days of opening-(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'10CFR$0.
i The maximum allowable leakage at a test pressure of 58 psig is 12 scfh.
Leakage _ measured lat _ test pressure less_ than 58 psig is adjusted to the equivalent value at 58 psig.
g.
Deleted.
h.
Drywell surfaces The interior surfaces of the drywell' and torus shall be visually inspected each operating cycle for evidence of torus corrosion or leakage.
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4.7.D (cont'd.)
b.
At least once per quarter:
(1) All normally open power operated isolation valves (except for the main steam line s wer-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 cycle, while shutdown, the devices that limit the maximum opening angle to 60* shall be verified functional for the folloving valves:
PC 230MV, PC-231MV, PC 232MV, and PC 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 ineperable, reactor power operation position of at least one other valve may-continua provided at least one in each line having ' an inoperable valve in. each. line having an valve shall be recorded daily.
-inoperable valve shall be in the mode corresponding to the isolated
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3.
If Specification 3.74 D.1 and 3.7.D.2
- cannot be met, an orderly shutdown-shall be initiated and the reactor shall be-in the Cold shutdown condition vithin 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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^ COOPER. NUCLEAR STATION TABLEL3.7.1'(Page 1)
PRIMARY CONTAINMENT IS0tATION VALVES 1
-Number of Power Maximum Action On Operated Valves Operating Normal l Initiating Valve & Steam Inboard Outboard Time (Sec) (1)
Position (2)
Sirnil O)
3<T<5 0
GC MS-AO A,B,C 6 D 4
'3<T<5 O
- MS-AO-86 "A,B,C, & D 2
15 0
CC Drywell Floor' Drain Iso. Valves RW-AO-82, RW-AO-83 2
15 0
CG Drywell Equipment Drain Iso. Valves RW-AO-94, RW-AO-95 Main Steam Line. Drain 1
1 30-0 CC lZ
' valves MS-MO-74, MS-MO-77 Reactor Vater Sample Valves 1
1 15 0
cc l
' RR-740AV, RR-741AV 1
1 60 0
GC Peactor Vater Cleanup System Iso. Valves RWCU-MO-15 -.RWCU-MO-18 1
1 40 C
SC l
RIIR Suction Cooling Iso.
l Valve RIIR-MO-17, RIIR-MO-18 2
20 C
SC RilR Discharge to Radwaste Iso. Valves RIIR-MO-57,' RilR-MO-67 2
15 C
SC Suppression C1mmber Purge &
Vent PC-245AV, PC-230HV 2
15 C
SC Supply Suppression Chamber N2 PC-237AV, PC-233MV i
COOPEP. NUCLEAR STATIO!i TABLE 3.7.1 (Page 2)
PRIMARY CONTAINMENT ISOLATION VALVES
~
Action On Number of Power Maximum Operated Valves Operating IIormal initiatin,.
Valve 6 Steam Inboard Outboard Time (Sec)_(1)
Position (2)
Siznal (3) 2 15 C
SC Prircary Containment Purge & Vent PC-246AV, PC-231tW 2
15 C
SC Supply Primary Containment & Ny PC-238AV, PC-232MV 1
40 C
SC ( 's )
Suppression Chamber Purge & Vent PC-230MV Bypass (PC-305tfV) l 1
40 C
SC(4)
Primary Containment Purge & Vent PC-231MV Bypass (PC-306MV) m
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Dilution Supply 2
15 C
SC PC-1303MV, PC-130itW 2
15 C
SC l
FC-1305MV, PC-1306tW Dilution Supply 2
15 0
GC PC-1301MV, PC-1302fW 2
15 0
GC PC-1311MV, PC-1312MV Suppression Chamber Purge and Ve-t Exhaust 1
15 C
SC PC-13087W 1
15 C
SC Primary Containment Purge and Vent Exhaust PC-1310MV t.
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3.7.A 6 4.7.A BASES (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. llowever, it is possibic that Icakage into other parts of the facility could occur.
Such 2 eakage paths that tray affect significantly the consequences of accidents are to be minimized.
l Certain isolation valves are tested by pressurizing the volume between the inboard and outboard isolation valves. This results in conservative test results since the inboard valve, if a globe valve, will be tested such that the test pressure is tending to lif t 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 f
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 bellows 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.
i Surveillance requirements for integrity of the personnel air lock ar# specifjad 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 Icakage at 58 psig.
Equation A 3 of Enclosure 3 (pranklin Research Center Technical Evaluation Report) to the letter, D. G. Eisenhut to J. M. pilant, September 3,1982, defines the method of adjustment.
The primary containment pre-operational test pressurc= 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 psig.
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 r.uppression 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 leek rate of 0.635%/ day at 58 psig. Calculations made by the NRC staff with leak rate and a sts.ndby 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 total thyroid dose is about 12 REM at 1100 meters from the stack over an exposure _ duration 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 'ssumption 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 the 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
i.e.,
it is not used for normal operation; therefore, a daily l
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I 4.7.B & 4.7 C BASES
'with an adsorbent qualified according to Table 5.1 of ANSI N509 1980.
The replacement tray for the adsorber tray removed for the test should meet the s e.me adsorbent quality. Tests of the HEPA filters with DOP eerosol shall be performed in accordance to ANSI N510 1980.
Any filters found defective shall be replaced with filters qualified pursuant to Regulatory Position C 3.d. of Regulatory cuide 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 heatera 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 respectivo door to check the gasket seal.
Any detection of DOP in the fan exhaust shall be considered an unacceptable tsst result and the gaskets repaired and test repeated.
If system drains are present in the filter /adsorber banks, loop seals must be used with adequate water level to prevent by pass leakage frorn the banks.
If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sarnple analysis shall be performed as required for operat.ional use. The determination of significance shall be made by the operator on duty at the titse 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 performance capability.
If one Standby Cas Treatment subsystem is inoperable, the operable subsystem's operability is verified daily.
This substantiates the availability of the operable subsystem and tnus reactor operation or refueling operation can continae for a liraited period of time.
3,7.D & 4.7.D BASES Primary Containment Isolation Valves 1
Double isolation valves are provided on lines penetrating the primary containment and open to the free' space of the containmant. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system.
Automatic initiation is.equired to minimite the potential leskage paths from the containment in the event of a loss of coolant accident.
The maximum closure times for the automatic isolation valves of the prirnary containment and reactor vessel isolation control = system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside _ the pritnary containment and the need to contain released fission products following pipe-breaks inside the primary containment, The USAR identifies those testable-primary contaitunent valves that perform an isolation function, and testable penetrations with Double 0 Ring Seals, and testable penetrations with n testable _ Bellows _ ensuring _ that any changes thereto receive. a
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10CFR$0.59 review.. In addition, plant procedures also identify containment isolation valves, and testab1_e penetrations 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
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