ML20210M515

Proposed Tech Specs,Removing Requirement in Plant TS to Perform Pbnp Unit 2 Containment Integrated Leak Rate 60-months from Previous Test

ML20210M515
Person / Time
Site:	Point Beach
Issue date:	08/14/1997
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20210M513	List: ML20210M515 NPL-97-0484, TS Change Request 199,consisting of Application for Amends to Licenses DPR-24 & DPR-27,removing Requirement in TS to Perform Containment Integrated Leak Rate 60-months from Previous Test
References
NUDOCS 9708210440
Download: ML20210M515 (4)

Text

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B. In Service Inspection and Testing of Safety Class Components Other than Steam

. Generator Tubes

1. Inservice inspection of AShiE Code Class 1, Class 2 and Class 3 components shall be performed in accordance with Section XI of the AShiE Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g) modified by Section 50.55a(b), except where specine written relief is granted by the NRC, pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

a. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

2. Contaimnent isolation valves will be tested in accordance with-Tw4mieal Sf wifieatiere45.1 i instead of Sectiorr!" 3120, Vse Leak nae Te : g ~--

CEnlainmenLLeakage Rate Testing Progrant

3. Inservice testing of ASME Code Class 1,2, and 3 pumps, valves, and snubbers shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a.

a. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

ILuis The steam generator tube inspection requirements are based on the guidance given in NRC Regulatory Guide 1.83, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes." ASME Section XI Appendix IV is being used for defining the basic requirements or the inspection method. However, at the present time, changes and improvements in steam generator eddy current inspection are occurring faster than the' code can be revised. Thus, in order to ensure that the best possible exam of the tubing and/or sleeves is being done, the technique utilized will, in general, be the latest industry-accepted technique. This means that complete word-for-word compliance with Appendix IV may not be possible. However, the basic requirements and intent will be met, to the extent practical.

Specification 15.4.2.B delineates programmatic requirements for establishing Inservice Inspection and Testing programs in accordance with the ASME Section XI Code and 10 CFR 50.55a requirements. The Code establishes criteria for system and component inspection and testing to ensure an appropriate level of reliability and detection of abnormal conditions.

Failure to meet Code requirements is evaluated on an individual system or component bases to determine operability. Appropriate LCOs are entered if a system or component is determined to be inoperable.

9708210440 970814 PDR ADOCK 05000266 P PDR Unit 1 - Amendment No.-440 15.4.2-5 August 2 5, "m '

Unit 2 - Amerdment NoA44 l

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l Safety analyses have been performed on the basis of a leakage rate of 0.40% by weight per 24

, hotirs at 60 psig. With this leakage rate and with minimum containment engineered safety systems for iodine removal in operation, i.e. one spray pump with sodium hydroxide addition, the public exposure would be well below 10 CFR 100 values in the event of the design basis-  ;

accident.* '

The safety analyses indicate that the containment leakage rates could be slightly in excess of 0.75% per day before a two-hour thyroid dose of 300R could be received at the site boundary.

The performance of periodic integrated leakage rate tests during plant life provide a current assessment of potential leakage from the containment in case of an accident that would pressurize the interior of the containment. These tests are performed in accordance with the Containment Leakage Rate Testing Program.

Periodic visual and physical inspection of the containment tendons is the method to be used to determine loss ofload-carrying capability because of wire breakage or deterioration. The tendon surveillance program specified in 15.4.4.11 is based on the recommendation of Regulatory Guide '

l.33 Rev. 3. Containment tendon structural integrity was demonstrated for both units at the end of one, three and eight years following the initial containment structural integrity test.

The pre-stress lift-off test provides a direct measure of the load-carrying capability of the tendon.

A deterioration of the corrosion preventive properties of the sheathing filler will be indicated by a change in the physical appearance of the filler, if the surveillance program indicates, by extensive wire breakage, tendon stress-strain relations, or other abnormal conditions, that the pre.

stressing tendons are not behaving as expected, the abnormal conditions will be subjected to an engineering analysis and evaluation in accordance with Specification 15.4.4.II.D to determine whether the condition could result in a significant adverse impact on the containment structural integrity. The specified acceptance critetia are such as to alert attention to the situation well before the tendon load carrying capability would deteriorate to a point that failure during a -

design basis accident might be possible. Thus, the cause of the incipient deterioration could be evaluated and corrective action studied without need to shut down the reactor. If the engineering evaluation determines that the abnormal condition could result in a significant adverse impact on the containment structural integrity, an abnormal degradation situation will be declared and a report submitted to the NRC in accordance with the specifications.

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Unit 1 - Amendment No.-49 15.4.4-7 October o,!n96 Unit 2 - Amendment No.4M

4 References

l. (1) FSAR Section 5.1.2.3 (2) FSAR Section 5.1.2 (3) FSAR Section 14.3.5 (4) FSAR Section 14.3.4 (5) FSAR S ::icr. 6.2.3 Deleied (6) FSAR pages 5.1-86 and 5.1-87 Unit 1 - Amendment No.-le 15.4.4-8 Oc:o'cc 9,1996 Unit 2 - Amendment No. F74

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! 15.6.12 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptient This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163,

" Performance Based Containment Leak-Test Program," dated September, 1995,-as-medified by the fol!cv in;; exceptiens+,

1. He intervabet".ren the 1992 Unk 2 Type A test ami-thet&4 2 Type A test sha!! be 60 months B. The peak design containmem internal accident pressure, P , is 60 psig.

C. The maximum allowable primary containment leakage rate, L., at P., shall be 0.4% of containment air weight per day.

D. Leakage rate acceptance criteria are:

1. The containment leakage rate acceptance criterion is sl.0 L,.

2. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 50.6 L, for the combined Type B and Type C tests and s0.75 L, for Type A tests.

E. The provisions of Specification 15.4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

F. The provisions of Specification 15.4.0.3 are applicable to the Containment Leakage Rate Testing Program.

Jnit 1 - Amendment No.460 15.6.12-1 Oetcher 9,1995

'Jnit 2 - Amendment No.4-74

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