ML20203H845

Proposed Tech Spec Pages 3.5-1b,4.5-1,2,3,3a,4,5a,7 & 8, Revising Acceptance Criteria for Containment Leakage Testing.Nshc Encl

ML20203H845
Person / Time
Site:	Oyster Creek
Issue date:	07/25/1986
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20203H841	List: ML20203H838 ML20203H845
References
NUDOCS 8608050109
Download: ML20203H845 (14)

Text

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7.

4 3.5-lb cold shutdown condition. The inoperable valve shall be returned to the operable status prior to placing the reactor in a condition where primary containment integrity is required,

b. If the primary containment air lock is inoperable, restore the inoperable air lock to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least shutdown condition within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. -

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8608050109 DR 860725 ADOCK 05000219 PDR a

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. 4.5-1 4.5 CONTAINMENT SYSTEM Applicability: Applies to containment system leakage rate, continuous leak rate monitor, functional testing of valves, standby gas treatment system operability, inerting surveillance, drywell coating surveillance, instrument line flow check valve surveillance, suppression chamber surveillance, and snubber surveillance.

Objectives: To verify that the condition of the containment system and the leakage from the containment sys. tem are maintained within specified values, as outlined in Appendix J of 10 CFR 50, and ANSI /ANS Standard 56.8-1981.

Specification: A. Type "A" Primary Containment Integrated Leak Rate Test (PCILRT).

1. PCILRT shall be performed at a pressure (Pt ) Of at least 20 psig but not greater than 35.0 psig (Pa)-

2.' Closure of the containment isolation valves shall be accomplished by their normal operation and without any adjustment, e.g. no tightening of the remote operated valves after closure.

3. The duration of the containment test stabilization period shall be at least four (4) hours prior to -

the start of the PCILRT.

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4. The verification test shall superimpose a

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calibrated leakage between 75% and 125% of the allowable leakage rate (L t ) for at least four l

(4) hours following the PCILRT.

5. A general inspection of the accessible interior and exterior surfaces of the containment structures and components will be performed prior to any PCILT. Any significant structural deterioration that could effect leak tightness will be repaired prior to the test.

B. Acceptance Criteria

1. The maximum allowable leakage rate (La) shall not exceed 1.0 weight percent of the contained air l adjusted to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at a test pressure adjusted to 35 psig (Pa).

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. 4.5-2

2. The allowable leakage rate (L t ), at the test pressure (Pt ) shall not exceed the following:

Lt = 1.0 w/o

• Pt i

_?u _

with Pt and Pa converted to absolute pressure

. for calculations.

3. The maximum allowable operational leakage rate (L tm ) Max., which shall be met prior to resumption of power operation following a PCILRT (either as measured or following repairs and retest), shall not exceed 0.75L t:

1.e., (Lt m) Max. {.75Lt

4. The difference between the verification test result and the Tyge "A" PCILRT test result (n'ot including Type "B and "C" test data) shall be within 0.25Lt-C. Corrective Action

1. If during the PCILRT Test, including the verification test, excessive leakage paths are identified which will interfere with meeting the allcwable leakage rate Lt i

. either: ,

,a. The Type "A" PCILRT will be terminated and the leakage through such paths shall be measured using Type "B" and "C" Local Leak Rate Test methods.

Repairs and/or adjustments to the equipment shall be made and a Type "A" PCILRT reperformed. If the containment was not depressurized below Pt , no additional stabiTTiation period is required prior to reperforming the PCILRT; or.

b. The leakage paths will be isolated and the Type "A" test continued until completion. At that time, local leakage tests shall be performed at a pressure of at least P t, before and af ter the -

repair of each isolated leakage path and the sum of the post repair local leakage rates will be added to the PCILRT result. This total shall be less than the allowable operational leakage rate (L tm) Max.

2. If repairs and/or adjustments are required.in order to meet the allowable operational leakage (L m) t Max.,

after having met the allowable leakage rate (Lt I*

a. The integrated leakage rate test need not be repeated, however;

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s 4.5-3

b. Local leak rate tests must be conducted at a pressure of Pt . The changes in local leak rates resulting from repairs or adjustments, when deducted from the PCILRT result, must yield a value not in excess of the allowable operational leakage rate (Ltm) Max. This situation demonstrates sufficient containment integrity for startup, but the PCILRT is to be considered as not meeting acceptance criteria as referenced in section 4.5.D.2. ,

D. Frequency

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1. Three Type "A" overall Integrated Containn.3nt .

Leakage Rate Tests shall be conducted at 40 + 10 month intervals during shutdown during each To-year service period. The third test of each set shall be conducted during the shutdown for the 10-year plant inservice inspection.

2. If any periodic Type A t.est fails to meet the acceptance criteria, the subsequent Type A test

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shall be performed at the next shutdown for

. refueling. Should this test also fail, the testing will be done at every refueling outage. This schedule will remain in effect until two consecutive

Type A tests meet the acceptance criteria, at which time the frequency of testing noted in 0.1 above may '

be resumeM.

E. Type "B" and "C' local Leak Rate Tests (LLRT)

1. Primary Containment testable penetrations (Type "B" Test) and isolation valves (Type "C" Test), except as stated below, shall be tested at a pressure of at least 35 psig (Pa) each refueling outage. ,

2. The main steam line isolation valves shall be tested at a pressure of at least 20 psig during each refueling outage to determine if corrective action is required.

3. Isolation valve, Type "C", tests shall have each valve closed by normal operation. (e.g. no tightening of valve after closures by valve motor).

4. bolted double gasketed seals shall 1:e tested whenever the seal is closed after being opened, and at least at each refueling outage.

5. The drywell airlock shall be demonstrated operable

, by performing the following tests:

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. 4-5-3a

a. The airlock must be tested at least once each 6 month interval at an internal pressure not less than Pa except:

(1) If there have been no airlock openings since the last successful test at Pa.

the test pressure may be reduced to 10 psig and the full pressure test interval may be extended to the next refueling outage or airlock opening. Opening of the airlock for the purpose of removing test equipment following any airlock test does not require further testing of the airlock.

(2) If the airlock is opened during'a period when Containment Integrity is required, it must be tested within 3 days. If the airlock is opened more frequently than once every 3 days, the airlock must be tested at least once every 3 days during the period of frequent openings. This intermediate testing may be accomplished at 10 psig. This reduced pressure testing may not be substituted for the full pressure testing requirement of the airlock in the previous paragraph.

(3) If the airlock is opened during a period 1 when Containment Integri.ty is not required, it need not be tested while Containment Integrity is not required, but must be tested at Pa prior to returning the reactor to an operating mode requiring containment integrity.

F. Acceptance Criteria

1. The combined leakage rate of all genetrations and isolation valves subject to Type B" and "C" tests shall be less than 0.60 of the maximum allohable limit (La) at 35 psig.

l 2. The MSIVs leakage rate shall not exceed 5% of the I

allowable operational leakage rate (L m) t Max. as ,

adjusted to or measured at 20 psig.

l G. Frequency Local Leak Rate Tests shall be perfonned as stated above but in no case may exceed intervals of 24 months.

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4.5-4

- H. Continuous Leak ' Rate Mon'itor

1. When the primary containment is inerted the containment shall be continuously monitored for gross leakage by review of the inerting system makeup requirements.

2. This monitoring system may be taken out of service for the purpose of maintenance or testing but shall be returned to service as soon as practical.

I. Report of Test Results -

Each integrated leakage rate test shall be the subject of a summary technical report, including results of the local leakage rate tests. The report shall include analysis' and interpretation of the results which demonstrate compliance in meeting the specified leakage rate limits.

J. Functional Test of Valves l

1. All containment isolation valves specified in Table 3.5.2 shall be tested for automatic closure by an isolation signal during each refueling outage. The following valves are required to close in the time specified below:

Main steam line isolation valves > 3 sec. and < 10 sec. '

Isolation condenser isolation valves < 60 sec.

Cleanup, system isolation valves 7 60 sec.

Cleanup auxiliary pumps system isolation valves 7 60.sec.

Shutdown system isolation . valves 160sec.

2. Each containment isolation valve shown in Table 3.5.2 shall be-demonstrated operable prior to returning the valve to service'after maintenance, repair or replacement work is performed on the- valve or its associated actuator by cycling the valve through at least one complete cycle of full travel and verifying the specified -

< isolating time. Following maintenance, repair or replacement work on the control or power circuit for the valves shown in Table ,

l 3.5.2, the affected component shall be tested to assure it will

• perform its intended function in the circuit.

3. During periods of sustained power operation each main steamline isolation valve shall be exercised in accordance with the following schedule,

a. Daily tests- Exercise valve (one at a time) to approximately 95 %open position with reactor at operation

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power level.

b. Quarterly tests - Trip valve (one at a time) and check full closure time, with reactor power not greater than 50%of rated power.

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4.5-Sa pressure of not less than 1.0 psi, the differential pressure decay rate shall not exceed the equivalent of air flow through a 2-inch orifice.

K. Reactor Building m 1. Secondary containment capability tests shall be conducted after isolating the reactor building and placing either Standby Gas Treatment System filter train in operation.

2. The tests shall be perfomed at least once per operating cycle and shall demonstrate the capability to maintain a 1/4 inch of water vacuum under calm wind conditions with a Standby Gas Treatment System Filter train flow rate of not more than 4000 cfm.

3. A. secondary containment capability test shall be conducted at each refueling outage ~ prior to refueling.

4. _ The results of the secondary containment capability tests shall be in the subject of a summary technical report which can be included in the reports specified in Section 6.

L. Standby Gas Treatment System

1. The capability of each Standby

• Gas Treatment System circuit shall be demonstrated by:

t a. At least once per 18 months, after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation, and following significant painting, fire, or chemical release in the reactor building during operation of the Standby Gas Treatment System by verifying that:

(1) The charcoal absorbers remove 2_99% of a halogenated

, hydrocarbon refrigerant test gas and the HEPA filters remove 199% of the DOP in a cold D0P test when tested in accordance with ANSI N510-1975.

(2) Results of laboratory carbon sample analysis show > -

90% radioactive methyl iodine removal efficiency when t tested in accordance with ASTM D 3803-79 (30*C, 95%

relative humidity).

b. At least once per 18 months by demonstrating:

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4.5-7 Although the dose calculations suggest that the allowable test leak rate could be allowed to increase to about-2.0%/ day before the guideline thyroid dose limit given in 10CFR100 wouTd be exceeded, establishing the limit at 1.0%/ day provides an adequate margin of safety to assure the health and safety of the general public. It is further considered that the allowable leak rate should not deviate significantly from the containment' design value to take advantage of the design leak-tightness capability of the structure over its' service lifetime. Additional margin to maintain the containment in the "as built" condition is achieved by establishing the allowable operational leak

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rate. The operational limit is derived by multiplying the allowable test leak rate by 0.75 thereby providing a 25%

margin to allow for leakage deterioration which may occur during the period between leak rate tests.

The primary containment leak rate test frequency is based on maintaining adequate assurance that the leak rate remains within the specification. The leak ~ rate test frequency is based on 10 CFR 50 Appendix J requirements for developing i leak rate testing and surveillance of reactor containment vessels.(9) Allowing the test intervals to be extended up l to 10 months pemits some flexibility needed to have the tests coincide with scheduled or unscheduled shutdown periods.

The penetration and air purge piping leakage test frequency, along with the containment leak rate tests, is adequate to allow detection of leakage trends. Whenever a double j gasketed penetration (primary. containment head equipment I hatches and the abs 6rption 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 35 psig is consistent with the accident analyses and the maximum prcoperational leak rate test pressure.

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. 4.5-8 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 practical minimum.

The containment integrity isolation valves are provided to maintain containment integrity following the design basis loss-of-coolant accident. The closure times of the isolation valves on the containment are not critical because it is on the order of minutes before significant fission product release to the containment atmosphere for the design basis loss of coolant. These valves are highly reliable',

see infrequent service and most of them are normally in the closed position. Therg outage is sufficient. Lggre, a test during each refueling J

Large lines connecting to the reactor coolant system, whose failure could result in uncovering the reactor core, are supplied with automatic isolation-valves (except containment cooling). The specified closure times are adequate to restrict the coolant loss from the circumferential rupture of any of these lines outside the containment to less than that for main steam line rupture. Therefore, this isolation valve Core. {}gsure time is sufficient to prevent uncovering the Since the main steam line isolation valves are normally in the open position, more frequent testing is specified.

l Daily exercising the valves to about the 95% open position j provides assurance of their operability and the quarterly full closure test provides assurance that the valves maintain l

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- . 0YSTER CREEK NUCLEAR GENERATING STATION PROVISIONAL OPERATING LICENSE N0. DPR-16 DOCKET NO. 50-219 TECHNICAL SPECIFICATION CHANGE REQUEST NO. 126 Applicant hereby requests the Commission to change Appendix A.to the above-captioned license as below and, pursuant to 10CFR50.91, an analysis concerning the determination of no significant hazards considerations is also presented:

1. Section to be changed:

Section 4.5 and the corresponding bases, and Section 3.5.A.3.

2. Extent of change:

The revision to section 3.5.A.3 is actually the addition of 3.5.A.3.b which is a Limiting Condition for Operation (LCO) concerning plant operations if the drywell airlock is not operable. The revisions made to Section 4.5 reflect the requirements of Appendix J of 10CFR50 as detailed in ANSI /ANS Standard 56.8-1981. This revision also incorporates a change to the paragraph numbers as necessary to correct inconsistencies caused by this and previous revisions. The specific changes requested are as follows:'

(1) Specification 3.5.A.3 is modified as follows: Step 3.5.A.3.b is added to create an additional LC0 concerning drywell airlock operability.

(2) Specification 4.5, " Applicability", is modified .as follows: This section now lists the major system surveillances and tests described in this section.

(3) Specification 4.5, " Objectives", is modified as follows: This section now refers to Appendix J of 10CFR50 and the corresponding

ANSI /ANS Standard 56.8-1981.

l (4) Specification 4.5.A is modified as follows:

a) Step 1 concerning the pre-operational testing is deleted as well as step 5 concerning test duration. Step 1 is no longer relevant, as it applies only to initial (pre-startup) testing of the containment. Step 5 is covered in the referenced standards, and appropriate procedures. ,

b) Steps 2 and 3 are modified to reflect the referenced standards.

Part of step 2 has become the new step 1, and the rest of step 2 along with step 3 are moved to 4.5.C.

c) Step 4 has remained essentially intact and renumbered Step 2.

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d) Steps 3 and 4 are added to reflect the requirements of Appendix J. Step 3 establishes a stabilization period prior to beginning l

the PCILRT and step 4 establishes a verification test to confirm calibration of instruments.

e) Step 5 is added to reflect the requirements of 10CFR50, Appendix J.V.A.

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.(5) Specification 4.5.B is modified as follows: l a) Step 1 remains essentially unchanged with minor subscript i changes to parallel variables used in Appendix J.  !

b) Steps 2 and 3 are modified to reflect the applicable standards.

c) Step 4 is added to established an acceptance criteria for the verification test in accordance with section III.A.3(b) of Appendix J.

(6) Specification 4.5.C is modified as follows:

a) This section is the largest change and adds more restrictions than previously existed. These additions reflect compliance with Appendix J.

(7) Specification 4.5.D is modified as follows:

a) This first section concerning the first refueing outage is deleted, b) The remainder of this section is modified to more closely reflect the testing frequency limits as imposed by Appendix J.

(8) Specification 4.5.E is modified as follows:

a) Steps 1 through 4 are taken apart and rearranged, but technically are still steps 1 through 4 with the addition of a requirement to use normal valve closures.

b) Step 5 is added to define testing of the largest containment penetration, the airlock.

(9) Specification 4.5.F is modified as follows:

a) The heading is changed from " Corrective Action" to " Acceptance Criteria".

b) Step 4.5.F.1 establishes tne acceptance limits as presented in Appendix J.

l c) Step 4.5.F.2 maintains the special case limits established for j MSIYs at Oyster Creek.

(10) Specification 4.5.G is modified as follows:

a) The original specification 4.5.G is moved to specification 4.5.H. This is the beginning of the paragraph numbering change.

b) The new specification 4.5.G is added to establish a local leak l rate testing interval limit in accordance with the referenced standards.

(11) Specifications 4.5.H, I, J, K, and L in the proposed change correspond respectively to 4.5.G, H, I, J, and K in the present Technical Specifications. This change is merely a change to the paragraph numbering system.

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, (12) Specificiation L in the present technical sptcificaticn is deleted.

Th2 proposed chang 2 to the Technical Sp:cifications will utilize paragraph L. The rest of specification 4.5. is unchanged with the exception of the corresponding bases.

3. Changes requested:

Replace the old pages 4.5-1, 2, 3, 4, Sa, 7, 8, and 3.5-1b, with the attached new pages 4.5-1, 2, 3, 3a, 4, Sa , 7, 8, and 3.5-1 b.

4. Discussion:

Appendix J to 10 CFR Pa'rt 50 was published on February 14, 1973. On August 7,1975, the NRC requested Jersey Central Power and Light (JCP&L)

Company to review its containment leakage testing program for Oyster Creek and the associated technical specifications, for compliance with the requirements of Appendix J.

JCP&L responded by letter dated December 24, 1975, which was supplemented by letters dated August 12, 1976, November 22, 1978 and June 27, 1980.

NRC letter dated March 4,1982 transmitted their Safety Evaluation of the Appendix J review for the Oyster Creek Nuclear Generating Station.

Consistent with that safety evaluation, and by a letter dated September 25, 1984, General Public Utilities (GPU) Nuclear (now the licensee) submitted Technical Specification Change Request No.130 to change paragraph 4.5.F.1.b. After the NRC staff June / July Progress Review meeting .with GPUN on July 31 and August 1,1985, the licensee agreed' to withdraw Technical Specification Change Request No.130. The withdrawal

! was confirmed by NRC letter dated August 26, 1985.

GPUN is now submitting Technical ' Specification Change Request No.126.

l Change No.126 addresses the program which verifies that the leakage from l .

the Primary Containment, both integrated and local, is maintained within specific values as outlined in Appendix J of 10CFR50, and as detailed in ANSI /ANS Standard 56.8-1981. The major modifications incorporated in the Integrated Leak Rate Testing Program are the establishment of a '

stabilization period for internal containment pressure, and a verification '

test to help check the accuracy of leakage detection methods. The leakage limits are also more closely defined in this proposed revision. The new ~ ,

section on " Corrective Action" gives detailed options on what may be done

• to limit leakage during the PCILRT. This specification allows for repairs and local testing of the repairs. It also allows for tne re-commencement of the PCILRT without the required stabilization period if containment was not depressurized. The testing frequency of three times in ten years, or approximately every 40 months is established and the reference to doing the pre-operational test is eliminated.

The major modification to the LLRT program is the modification to the drywell airlock test. The 35 psig peak pressure airlock test required by Appendix J is established, but because of concerns described in NUREG/CR-4398 the frequency of airlock tests at 35 psig will be limited.

When permissible a 10 psig test will utilized. The acceptance criteria for the LLRT program is established as well as a testing frequency. for it. The change also adds an LC0 in section 3.5. The LC0 limits plant operation when the airlock is not operable.

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Th:re is no plant configuraticn change involved with this technical specification change request. The testing described here is merely a surveillance program designed to verify primary containment integrity.

The program outlined here is designed to bring the current program in line with the requirements of Appendix J to 10CFR50 as detailed in ANSI /ANS 56.8-1981.

5. Determination The Commission has provided guidance concerning the application of the standards of 10CFR50.92 for detennining whether a significant hazards consideration exists by providing certain examples as discussed in the Federal Register on April 6,1983 (48 FR 14870) under the heading

" Examples of Amendments That Are Considered Not Likely to Involve Significant Hazards Considerations". Example (1) relates to a purely administrative change to Technical Specifications: 1.e., a change to

! achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclature. Example (11) relates to a change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications; i.e., a more stringent surveillance requirement. Example (vii) relates to a change to make a license conform to changes in the regulations, when the license change results in very minor changes to facility operations clearly in keeping with.the regulations.

In this case, the proposed change described above is similar to at least one of the three examples. The change in the numbering scheme is clearly an administrative change as described in example (1). The addition of i

Specification 3.5.A.3.b is consistent with both examples (ii) and (vii). '

The modifications and additions made to Specifications 4.5.A through 4.5.G

also ~ relate easily to example (ii) in that a more stringent and comprehensive surveillance requirement is established. Example (vii) also

! relates in that the surveillance program, in the form presented in this proposal, is defined by a regulaticn to which the license is conforming.

The proposed modification to the Technical Specifications will not involve

, a significant hazards cons'ideration because operation of Oyster Creek l Nuclear Generating Station in accordance with this change would not:

1 i (1) involve a significant increase in the probability or consequences of l r.n accident previously evaluated. This change merely re-defines .the leak rate testing program for Primary Containment. This program is designed to ensure that the Primary Containment is able to perform its design function. That function is to contain the energy and the radioactive release of the design basis loss of coolant accident.

Therefore, this change cannot increase the probability or consequences of an accident previously evaluated.

(2) create the possibility of a new or different kind of accident from any previously analyzed. It has been detennined that, because this revision more clearly establishes the requirements and methods of testing the Primary Containment Integrity and does not involve a change to the containment configuration, this change will not create the possibility of a new or different kind of accident from any previously evaluated.

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(3) in' valve a significant reducticn in a margin of safety. This changa has increased the requirements as established in Appendix J that the primary containment must meet to be considered operable. Therefore, this change will not reduce the margin of safety.

This change reflects the requirments of Appendix J to 10CFR50 as described in ANSI /ANS Standard 56.8-1981. No changes proposed in this TSCR are outside the scope of those two documents.

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