SNRC-1326, Forwards Reactor Containment Bldg Integrated Leakage Rate Test,Types A,B & C Periodic Test & 1986 Local Leak Rate Test Summary Analysis

Forwards Reactor Containment Bldg Integrated Leakage Rate Test,Types A,B & C Periodic Test & 1986 Local Leak Rate Test Summary Analysis

ML20212Q934
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	04/21/1987
From:	Leonard J LONG ISLAND LIGHTING CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20212Q936	List: ML20212Q950 SNRC-1326, Forwards Reactor Containment Bldg Integrated Leakage Rate Test,Types A,B & C Periodic Test & 1986 Local Leak Rate Test Summary Analysis
References
SNRC-1326, NUDOCS 8704240152
Download: ML20212Q934 (5)

Text

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LONG ISLAND LIGHTING COMPANY SHOREHAM NUCLEAR POWER STATION P.O. BOX 618, NORTH COUNTRY ROAD e WADING RIVER, N.Y.11792 JOHN D. LEONARD, JR.

VICE PflE51 DENT . NUCLEAR OPE RATIONS APR 211987 SNRC-1326 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Requirements of 10 CFR 50 Appendix J Integrated Leak Rate Test

Shoreham Nuclear Power Station - Unit 1 l Docket No. 50-322 Gentlemen

In accordance with the requirements of 10 CFR 50 Appendix J, Paragraph V.B., attached are the results of the Shoreham Containment Building Integrated Leak Rate Test. Also attached is l the 1986 LLRT Summary Analysis.

If there are any questions or if additional information is required, please contact this office.

Very truly yours,

('W<d "I'o n b. Leonard, Jre -

Vi e President - N lear Operations a

JDG:ck Attachment l cc: R. Lo Region I Administrator C. Warren 0\'

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0704240152 870421 PDR ADOCK 05000322 p PDR l

ATTACHMENT 1986 LLRT

SUMMARY

ANALYSIS The pre-repair LLRT, the repair, and the post-repair LLRT for each boundary, or penetration, was reviewed. The net leakage contribution for each penetration was determined using the following criteria:

1. A leakage equivalent to the repair improvement achieved on each valve in the penetration is calculated.

2. The leakage equivalent is the difference between the pre-repair and the post-repair LLRT results.

3. If a repair was not performed, a zero leakage equivalent is assessed to the valve.

4. The leakage equivalent assessed to a penetration may be reduced due to the safety-related service of the system associated with the penetration (s). Justification for this reduction will be provided with the analysis.

5. The net equivalent leakage for the penetration is the lowest of the inside or outside valve grouping (e.g., simulates minimum pathway leakage).

6. If the "As-Left" leakage of a repaired valve is lower than the "As-Left" leakage of a valve that didn't require a repair, then the penetration net equivalent leakage is the difference between the "As-Left" leakages.

7. For series valves tested together, the penetration net equivalent leakage is half the total leakage when both valves are repaired at the same time (prior to performing another test).

8. When the summation of the leakage equivalent and the leakage measured during a successful Type A test is greater than L the penetration (s) ,

with excessive leakage (s) shall be analyzed unde 8,a corrective action ~

program.

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I. t ATTACHMENT- (Continued) 1986 LLRT

SUMMARY

ANALYSIS Penetration No. Inside Outside Net (sefh) Remarks X-2A Feedwater 0.0 0.0 0.0 See Note 2 X-2B Feedwater 0.0 0.0 0.0 See Note 2 X-5 RHR from 0.0 >72 0.0 RPV X-6B RHR Recire. > 72 0.0 0.0 Return X-8B RHR Pool Spray 0.0 0.0 1

1, X-9A RHR Suction > 72 0.0 See Note 3 X-9B RHR Suction 71.17 0.0 See Note 3 X-9C RHR Suction > 72 0.0 See Note 3 X-10A Radwaste, Fuel

• 0.0 See Note 3 Pool Cooling &4 and Cleanup i

X-12 HP_CI Turbine 0.0

• 0.0 See Note 4 Steam &5 X-13 HPCI Turbine 0.0 0.0 See Note 5 Exhaust X-16 RCIC Turbine

• 0.0 0.0 See Note 4 Steam e_

{ X-22B RBCLCW *

• See Note 4 X-25A RBCLCW to 0.0 14.95 0.0 i

Drywell Coolers

! X-31 Equipment 11.4 11.4 ,

Drains X-37A TIP 0.0 0.0 See Note 6 1

X-37B TIP 0.0 0.0 See Note 6 I

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4 ATTACHMENT (Continued) 1986 LLRT

SUMMARY

ANALYSIS Penetration No. Inside Outside Net (sefh) Remarks i

j X-37C TIP 0.0 0.0 See Note 6 i

l- X-37D TIP 0.0 0.0 See Note 6 1

XS-5 RHR 0.0 0.0 XS-21 Cont. Atmos. 8.03 0.0 0.0 Control F-10 Reactor Recire. 4.879 0.0 0.0

Pump Seal IT48*PNL-068A >72.0 >72.0 Hydrogen Recombiner 6.05 6.05

Loop A i Hydrogen Recombiner 10.04 10.04 Loop B

! 1B21*LT155A 0.05 0.05 (As-Found)

Total > 99.54 SCFH

, Total > 0.175970 %/ day

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ATTACHMENT (Continued) 1986 LLRT

SUMMARY

ANALYSIS NOTES:

1. The resulting net equivalent leakage of >99.54 SCFH or > 0.175970 percent / day indicates that the plant's maximum allowable leakage rate of 0.5 percent / day may have been exceeded.

2. Feedwater check valves were opened for inspection.

3. The objective of the emergency core cooling systems (ECCS), is to limit 'the release of radioactive materials should a loss of coolant accident (LOCA) occur. - The RER system restores and maintains the coolant inventory in the reactor vessel so that the core is adequately cooled after a LOCA, (LPCI mode of operation). The RHR system valves, piping, and components have been designed as essentially a leaktight system (seismic, safety-related). During plant operations periodic tests and inspections are performed. Reference SNPS FSAR Section 5.5.7, Residual Heat Removal (RHR) System and Section 6.3, Emergency Core Cooling Systems (ECCS).

4. Local leakage rate test data was not obtained for these valves prior to performing repairs.

5. The objective of the emergency core cooling systems (ECCS), is to limit the release of radioactive materials should a loss of coolant

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accident -(LOCA) occur. The HPCI system provides core cooling to ensure that the reactor is adequately cooled to meet the design bases in the event of a break in the reactor coolant pressure boundary (RCPB) and loss of coolant that does not result in' rapid depressurization of the reactor vessel. The HPCI system valves, piping, and components have been designed . as essentially a leaktight system (seismic, safety-related). During plant operations periodic u tests and inspections are performed. Reference SNPS FSAR Section 6.3, Emergency Core Cooling Systems (ECCS).

6. The explosive shear valves were not repaired.

7. Greater than 72 scfh represents the largest leakage measuring j equipment used for the 1986 LLRT program. ,

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