ML20046B373

Proposed Tech Specs Eliminating Exemption Granted on 850510 for Air Lock Testing & Requiring Facility to Meet Paragraph III.D.2(b)(ii) of App J

ML20046B373
Person / Time
Site:	Millstone
Issue date:	07/27/1993
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20046B372	List: B14524, Application for Amend to License DPR-21,revising TS to Eliminate Exemption Granted on 850510 for Air Lock Testing & Require Facility to Meet Paragraph III.D.2(b)(ii) of App J ML20046B373
References
NUDOCS 9308040128
Download: ML20046B373 (12)

Text

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Docket No. 50-245 B14524 i

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Millstone Nuclear Power Station, Unit No. I-Proposed Revision:to Technica1' Specifications Appendix J Containment Air l-Lock Testing Markeo ' p Pages J

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PDR ADOCK 05000245 p-pop.

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February 26, 1992l I

1 (5) Intearated Imolementation Schedule

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Northeast Nuclear Energy Company shall implement and s.

maintain in effect the Integrated-Implementation j

Schedule Program Plan (the Program Plan) to be fol-j.

lowed for scheduling of plant modifications and engi-nearing studies. The Program Plan shall be followed from and after the effective date of this license l

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J condition.

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b.

This license condition shall be effective for three l

years from the date of issuance.

D.

The facility has been granted certain exemptions from the i

requirements of 10 CFR-Part 50 as set forth below:

j (1)Section III.G. of Appendix R to 10. CFR Part-50, " Fire _ Pro--

tection Program for Nuclear Power Facilities Operating g

Prior to January 1,1979."

This section relates to fire i

protection features for ensuring the systems and associated circuits used to achieve and maintain safe shutdown are i

free of fire damage. The staff safety evaluation, dated November 6, 1985, concluded that the licensee's existing i

fire-protection configuration with proposed modifications I

achieves an equivalent level _ of safety. _ Exemption granted November 6, 1985.

(2) Appendix J to 10 CFR Part 50, " Primary Reactor. containment l

Lankage Testing for Water-Cooled Power Reactors."

Appendix J relates to containment leakage test require-ments, specifically periodic verification by tests of the leak-tight integrity of the1 primary reactor containment.and

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-The-systemsandcomp{onentswhich_penetratecontainment.

threeexemptionsfEAntedonMay-10, 1985, on the basis of the staff safety evaluation,q relate to testing of expansion

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bellows at containment penet rations,Pand-low-pressure-te s ts -

ef-containment-sooass-air 4oska and main ~ steam isolation valves.

(;g rwo www ncmcw zN eracr (3) Section 50.71 of 10 CFR fart 50, " Maintenance _ of Records, Making of Reports." Section 50.71(e)(3) relates to the requirement for submittal of an. updated FSAR. An exemption for a schedular delay in the submittal of the updated FSAR was granted on November 22,'1985.

This_ exemption requires-that the FSAR update be completed and submitted by March 31 -1987.

9 23VR8X.081

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OCT 311986 SURVEILLANCE REQUIREMENT (Continued) 8 4.7 CONTAINMENT SYSTEMS 4.7.A.3.a.4.

All test leakage rates shall be calculated using observed data converted to absolute values.

Error analyses shall be performed to select a balanced integrated leakage measurements system.

b.

Acceptance Criteria for IPCLT 1.

The maximum allowable leak rate at P shall not exceed L, (1.2 weight percent of the contai8ed air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

The allowable operational leak rate, L which shall be 2.

metpriortoincreasingreactorcoolank0,ystemtemperature s

above 212*F following a test (either as measured or following repairs and retest), shall not exceed 0.75 L,.

c.

Corrective Action for IPCLT If leak repairs are necessary to meet the allowable operational leak rate, the integrated leak rate test need not be repeated provided local leak measurements are conducted and the leak rate differences prior to and after repairs, when corrected to P and deducted from the integrated leak rate measurements, yieldaleakageratevaluenotinexcessof.theallowable operational leak rate L to*

d.

Local Leak Rate Tests (LLRT)

(1) Primary containment testable penetrations and isolation valves shall be tested at a pressure of 43 psig, except the main steam line isolation valves shall be tested at a pressure of 25 psig each operating cycle.

Bolted dot.ble-gasketed seals shall be tested whenever the seal is closed after being opened and at least once during each operating

cycle, gtyPL /1 C C' Metsonnel air lock door seals.shafl be tested at aJresiU6 of 43'psiAleast once every 6 monthpIfdhe airlock Uy g is opened when prTman containment-Tntegrity is required f

A 77/Eh'Mc*b during the intervabbetweerrtheabove tests, the air. lock door sealsahalfbe tested at 10 pYg'vi% hours of p he n st of'a series of openings, Acceptance criteria and corrective action for LLRT:

e.

If the total leakage rates listed below are exceeded, repairs' and retests shall be performed to correct the condition.

Millstone Unit 1 3/4 7-5 1

(2)

Personnel Air Lock Door Seals i)

The personnel air lock door seals shall be tested at a pressure of 43 psig at least once every six months.

ii)

Prior to entering a mode in which primary containment integrity is required, the personnel air lock door seals shall be tested at 43 psig if the air lock was opened during the period when primary containment integrity was not required.

iii)

The personnel air lock door seals shall be tested at 43 psig within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following any opening of the air lock during periods when primary containment integrity is required.

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4.7 CONTAINMENT SYSTEMS

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The penetration and air pur the containment leak testf,(ge piping leakage test frequency, along with j

is adequate to allow detection of leakage trends. Whenever a double-gasketed penetration (primary containment head, i

equipment hatches, and the suppression chamber access hatch)-is broken and remade, the space between the gaskets is pressurized to determine that the seals are performing properly. The test pressure ofA43 psig is consistent with the accident analyses and the maximum preoperational leak rate test]g g

4 pressure.

It is expected that the majority of the leakage from valves, i

penetrations and seals would be into the reactor building..However, it is possible that leakage into other parts of the facility could occur. Such-i

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1eakage paths that may affect significantly the consequences of accidents are to be minimized.

g mugem 2 l

Mcnitoring the nitrogen makeup requirements of the inerting system i

provides a method of observing leak rate trends and would detect gross i

leaks in a very short time.. This equipment must be periodically removed i

i from service for test and maintenance, but this out-of-service time will be kept to a practical minimum.

i Surveillance of the suppression chamber-drywell vacuum breakers consists i

of operability checks, calibration of instrumentation and inspection of the j

valves.

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The monthly operability tests are performed to check the capability of the disc to open and close and to functionally test the position indication

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system. This test frequency is justified based on previous experience and l

j the fact that these valves are normally closed and'are only open during tests or accident conditions.

The refueling outage surveillance tests are perfomed to check that the j

valve will perform properly during the accident condition and to verify the calibration of the position indication system. Measuring the force 1

required to lift the valve assures that the valve will function properly l

during an accident.

Inspection of a select number of valves during each refueling outage assures that deterioration of the valve internals or t

misalignment of the disc does not impair the proper operation of the valve.

l This test interval is based on equipment quality and previous equipment l

experience.

1-i B.

Standby Gas Treatment System and 1

i C.

Secondary Containment i

Pressure. drop across the combined HEPA filters and charcoal adsorbers of less than 7 inches' of water,- at the system design flow rate, will F

indicate that the filters and adsorbers are not clogged by excessive:

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amounts of foreign matter. Heater capability and pressure drop should be l'

determined at least once per operating cycle to-show. system performance-I capability.

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gil,stoneUnit1 B 3/4 7-9

Amendment No. 59

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1 Personnel air lock door seal testing is performed in accordance with 10CFR50, Appendix J requirements.

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4 Docket No. 50-245 B14524

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Millstone' Nuclear. Power Station, Unit No.1-i Proposed Revision to Technical: Specifications.

t Appendix J Containment Air _ Lock Testing _.

Retyped Pages.

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L July - 1993,.

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4 (5)

Intearated Imolementation Schedule a.

Northeast Nuclear Energy Company shall implement and maintain in effect the Integrated Implementation Schedule Program Plan (the Program Plan) to be followed for scheduling of plant modifications and engineering studies.

The Program Plar, shall be followed from and after the effective date of this license condition.

b.

This license condition shall be effective for three years from the date of issuance.

D.

The facility has been granted certain exemptions from the requirements of 10 CFR Part 50 as set forth below:

(1)Section III.G. of Appendix R to 10 CFR Part 50, " Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979." This section relates to fire protection features for ensuring the systems and associated circuits used to achieve and maintain safe shutdown are free of fire damage.

The staff safety evaluation, dated November 6,1985, concluded that the licensee's existing fire-protection configuration with proposed modifications achieves an equivalent level of safety.

Exemption granted November 6, 1985.

(2) Appendix J to 10 CFR Part 50, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." Appendix J relates to containment leakage test requirements, specifically periodic verification by tests of the leak-tight integrity of the primary reactor containment and systems and components which penetrate containment. Three exemptions were granted on May 10, 1985, on the basis of the staff safety evaluation. The two which remain in effect relate to testing of expansion bellows at containment penetrations and main steam isolation valves.

(3) Section 50.71 of 10 CFR Part 50, " Maintenance of Records, Making of Reports." Section 50.71(e)(3) relates to the requirement for submittal of an updated FSAR. An exemption for a schedular delay in the submittal of the updated FSAR was granted on November 22, 1985. This exemption requires that the FSAR update be completed and submitted by March 31, 1987.

SURVEILLANCE RE0VIREMENTS (Continued) 4.7 CONTAINMENT SYSTEMS 4.7.A.3.a.4.

All test leakage rates shall be calculated using observed data converted to absolute values.

Error analyses shall be performed to select a balanced integrated leakage measurements system.

b.

Acceptance Criteria for IPCLT 1.

The maximum allowable leak rate at P shall not exceed L (1.2 weight percent of the contained air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) a a

2.

The allowable operational leak rate, L which shall be metpriortoincreasingreactorcoolanko,ystemtemperature s

above 212*F following a test (either as measured or.

following repairs and retest), shall not exceed 0.75 L,.

c.

Corrective Action for IPCLT If leak repairs are necessary to meet the allowable operational leak rate, the integrated leak rate test need not be repeated provided local leak measurements are conducted and the leak rate differences prior to and after repai s, when corrected to -

P and deducted from the integrated leak rate measurements, yield a leakage rate value not in excess of the allowable operational leak rate L to*

d.

Local Leak Rate Tests (LLRT)

(1) Primary containment testable penetrations and isolation valves shall be tested at a pressure of 43 psig, except the main steam line isolation valves shall be tested at a pressure of 25 psig each operating cycle. Bolted double-gasketed seals shall be tested whenever the seal is closed after being opened and at least once during each operating cycle.

(2) Personnel Air Lock Door Seals i)

The personnel air lock door seals 'shall be tested at a pressure of 43 psig at least once every six months.

ii)

Prior to entering a mode in which primary containment integrity is required, the personnel air lock door seals shall be tested at 43 psig if the air lock was opened during the period when primary containment integrity was not required.

iii) The personnel air lock door = seals _ shall. be tested at 43 psig within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following any opening of the air lock during periods when' primary containment i

integrity is required.

Millstone Unit 1 3/4 7-5 Amendment No.

0102

SURVEILLANCE REQUIREMENTS (Continued)

.e.

Acceptance. criteria and-corrective' action for LLRT:

If the total. leakage rates listed below are exceeded, repairs and retests shall be performed'to correct the. condition.

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-e Millstone Unit =1-3/4L.7-5a -

' Anendment No.

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' 0102-s

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4.

as 4.7 CONTAINMENT SYSTEMS BASES j

i The penetration and air purge piping leakage test frequency, along with the containment leak test, which is performed at a test pressure of at i

least 43 psig (P is adequate to allow detection of leakage trends.

)

Wheneveradoubl8-)g,asketedpenetration(primarycontainmenthead, equipment hatches, and the suppression chamber access hatch) is broken and remade, the space between the gaskets is pressurized to determine that the seals are performing properly. The test pressure of at least 43 l psig is consistent with the accident analyses and the maximum preoperational leak rate test pressure.

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 facil ity could occur.

Such leakage paths that may affect significantly the consequences of accidents are to be minimized.

Personnel air lock door seal testing is performed in accordance with 10CFR50, Appendix J requirements.

Monitoring the nitrogen makeup requirements of the inerting system provides a method of observing leak rate trends and would detect gross leaks in a very short time. This equipment must be periodically removed from service for test and maintenance, but this out-of-service time will be kept to a pr ctical minimum.

1 Surveillance of the suppression chamber-drywell vacuum breakers consists of operability checks, calibration of instrumentation and inspection of the valves.

The monthly operability tests are performed to check the capability of the disc to open and close and to functionally test the position indication i

system. This test frequency is justified based on previous experience and the fact that these valves are normally closed and are only open during tests or accident conditians.

The refueling outage sur eillance tests are performed to check that the valve will perform properly during the accident condition and to verify the calibration of the position indication system. Measuring the force required to lift the valve assures that the valve will function properly during an accident.

Inspection of a select number of valves during each refueling outage assures that deterioration of the valve internals or misalignment of the disc does not impair the proper operation of the valve.

This test interval is based on equipment quality' and previous equipment experience.

Millstone Unit 1 B 3/4 7-9 Amendment No. A9 0103

4.7 CONTAINMENT SYSTEMS BASES B.

Standby Gas Treatment System and C.

Secondary Containment Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 7 inches of water, at the system design flow rate, will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Heater capability and pressure drop should be determined at least once per operating cycle to show system performance capability.

l M,llstone Unit 1 B 3/4 7-9a Amendment No. Ef i

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