RS-03-031, Request for Amendment to Technical Specifications Surveillance Requirement 3.6.1.3.8

Request for Amendment to Technical Specifications Surveillance Requirement 3.6.1.3.8

ML031190554
Person / Time
Site:	LaSalle
Issue date:	04/18/2003
From:	Jury K Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-03-031
Download: ML031190554 (15)

Text

ExeNuclear Ike n M Exelon Generation wwwexeloncorpcorn 4300 Winfield Road Warrenville, IL 60555 10 CFR 50.90 RS-03-031 April 18, 2003 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555

Subject:

LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374 Request for Amendment to Technical Specifications Surveillance Requirement 3.6.1.3.8 LLC, hereby requests In accordance with 10 CFR 50.90, Exelon Generation Company (EGC),

of Facility Operating the following amendment to Appendix A, Technical Specifications (TS), modify TS change will License Nos. NPF-11 and NPF-18. Specifically, the proposed requirement is testing Surveillance Requirement (SR) 3.6.1.3.8 to identify that the specified consistent with the SR applicable to reactor instrumentation lines. The proposed change is General Electric Plants, wording specified in NUREG -1433, "Standard Technical Specifications BWR/4," Revision 2, dated June 2001.

as follows.

The information supporting the proposed TS change is subdivided Attachment 1 is the notarized affidavit.

Attachment 2 provides our evaluation supporting the proposed change.

Attachment 3 contains a copy of the marked up TS page.

Attachment 4 provides the retyped TS page and Bases pages for information only.

Station Plant Operations The proposed TS change has been reviewed by the LaSalle County Board (NSRB) in Safety Review Review Committee (PORC) and approved by the Nuclear accordance with the Quality Assurance Program.

by transmitting a copy of EGC is notifying the State of Illinois of this application for amendment this letter and its attachments to the designated State Official.

support the planning of We request approval of the proposed change by January 1, 2004 to testing during the scheduled Unit 1 refueling outage in February 2004.

pool

April 18, 2003 U. S. Nuclear Regulatory Commission Page 2 Should you have any questions concerning this submittal, please contact Mr. T. W. Simpkin at (630) 657-2821.

Sincerely, eith R. Jury Director-Licensing Mid-West Regional Operating Group Attachments: . Affidavit . Evaluation of Proposed Change . Markup of Proposed Technical Specification Page Change . Retyped Pages for Technical Specification Change and Bases Changes (for information only) cc: Regional Administrator - NRC Region IlIl NRC Project Manager - NRC NRR NRC Senior Resident Inspector - LaSalle County Station Office of Nuclear Facility Safety - Illinois Department of Nuclear Safety

ATTACHMENT I Affidavit STATE OF ILLINOIS  ! )

COUNTY OF DUPAGE )

IN THE MATTER OF: )

EXELON GENERATION COMPANY (EGC), LLC ) Docket Numbers LASALLE COUNTY STATION - UNIT 1 and UNIT 2 50-373 and 50-374

SUBJECT:

Request for Amendment to Technical Specifications Surveillance Requirement 3.6.1.3.8 AFFIDAVIT I affirm that the content of this transmittal is true and correct to the best of my knowledge, information, and belief.

T.W. Si W-n Manager-Licensing Mid-West Regional Operating Group Subscribed and sworn to before me, a Notary Public in and for the State above named, this /day of

, 2003 Notary Public A t OFFICIAL SEAL ANESE L. GRIGSBY NOTARY PUBLIC, STATE OF ILLINOIS MY COMMISSION EXPIRES 3-13-2005

ATTACHMENT 2 Evaluation of Proposed Change Page 1 of 6

2.0 DESCRIPTION

OF PROPOSED AMENDMENT

3.0 BACKGROUND

4.0 REGULATORY REQUIREMENTS & GUIDANCE

6.0 REGULATORY ANALYSIS

7.0 NO SIGNIFICANT HAZARDS CONSIDERATION (NSHC)

8.0 ENVIRONMENTAL CONSIDERATION

9.0 PRECEDENT

ATTACHMENT 2 Evaluation of Pr6posed Change Page 2 of 6

1.0 INTRODUCTION

In accordance with 10 CFR 50.90, Exelon Generation Company (EGC), LLC, hereby requests the following amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. NPF-11 and NPF-18. Specifically, the proposed change will modify TS Surveillance Requirement (SR) 3.6.1.3.8 to identify that the specified testing requirement is applicable to reactor instrumentation lines. The proposed change is consistent with the SR wording specified in NUREG -1433, 'Standard Technical Specifications General Electric Plants, BWR/4," Revision 2, dated June 2001.

2.0 DESCRIPTION

OF PROPOSED AMENDMENT The proposed wording associated with the change is identified below in bold type.

SR 3.6.1.3.8 Verify each reactor instrumentation line EFCV actuates to the isolation position on an actual or simulated instrument line break signal.

3.0 BACKGROUND

Excess flow check valves (EFCVs) are used as a means of automatic isolation on all static instrument sensing lines that penetrate the drywell containment. EFCVs that are connected to the reactor coolant pressure boundary (RCPB) are classified as reactor instrumentation line EFCVs. EFCVs that are not connected to the reactor pressure boundary (i.e., containment atmosphere, and suppression pool water level) as classified as low pressure EFCVs.

4.0 REGULATORY REQUIREMENTS & GUIDANCE 10 CFR 50.36(c)(2)(ii)(c), "Criterion 3," requires that a structure, system or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier be included in the TS.

5.0 TECHNICAL ANALYSIS

5.1 Design Bases NUREG-1433, Rev. 2, provided licensees with the latest NRC recommended content and format for TS. The NUREG-1433 SR for testing EFCVs, SR 3.6.1.3.10, specifies that this testing is associated with reactor instrumentation line EFCVs.

ATTACHMENT 2 Evaluation of Proposed Change Page 3 of 6 The Bases to SR 3.6.1.3.10 in NUREG-1433, Rev. 2, provides a reference to NEDO-32977-A, "Excess Flow Check Valve Testing Relaxation," dated June 2000. NEDO-32977-A was approved for use by licensees in a NRC letter dated March 14, 2000. NEDO-32977-A states the following on the scope of TS testing associated with EFCVs.

"EFCVs in instrument lines which connect to the RCPB are normally tested during refueling outages to meet Technical Specification requirements. Instrument lines that connect to the containment atmosphere, such as those which measure drywell pressure, or monitor the containment atmosphere or suppression pool water level, are considered extensions of primary containment. A failure of one of these instrument lines during normal operation would not result in the closure of the associated EFCV, since normal operating containment pressure is not sufficient to operate the valve. Such EFCVs will only close with a downstream line break concurrent with a LOCA. Since these conditions are beyond the plant design basis, EFCV closure is not needed and containment atmospheric instrument line EFCVs need not be tested."

The proposed change will incorporate the wording from NUREG-1433 into LaSalle County Station SR 3.6.1.3.8 to limit the scope of TS required testing to EFCVs that are directly connected to the RCPB (i.e., reactor instrumentation line EFCVs). Additionally, this proposed change will allow LaSalle County Station not to test the low pressure EFCVs in the future and this will be specified in the bases to SR 3.6.1.3.8.

The proposed change is consistent with the wording and intent of NUREG-1433, NEDO-32977-A and the design bases for LaSalle County Station.

5.2 Risk Information This submittal is not based on risk informed decision making.

6.0 REGULATORY ANALYSIS

The proposed change to SR 3.6.1.3.8 will provide the testing requirements for EFCVs that are associated with instrument lines that are connected to the RCPB and penetrate the drywell. The function of reactor instrumentation line EFCVs, in combination with other accident mitigation features, is to limit fission product release. Therefore, the testing of reactor instrumentation line EFCVs must be included in LaSalle County Station TS in accordance with 10 CFR 50.36(c)(2)(ii)(c).

I t ATTACHMENT 2 Evaluation of Proposed Change Page 4 of 6 7.0 NO SIGNIFICANT HAZARDS CONSIDERATION EGC has evaluated the proposed change to the TS for LaSalle County Station, Unit I and Unit 2, and has determined that the proposed change does not involve a significant hazards consideration and is providing the following information to support a finding of no significant hazards consideration.

Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change to the Technical Specifications (TS) Surveillance Requirement (SR) 3.6.1.3.8 will incorporate into the SR, wording specified in NUREG -1433, "Standard Technical Specifications General Electric Plants, BWRI4," Revision 2, dated June 2001. The proposed change will specify that the testing required by SR 3.6.1.3.8 is applicable to reactor instrumentation line excess flow check valves (EFCVs). The performance of TS surveillance testing is not a precursor to any accident previously evaluated. Thus, the proposed change does not have any affect on the probability of an accident previously evaluated.

The function of reactor instrumentation line EFCVs, in combination with other accident mitigation features, is to limit fission product release. The surveillance testing specified in SR 3.6.1.3.8 will provide assurance that the reactor instrumentation line EFCVs will perform as designed. Thus, the radiological consequences of any accident previously evaluated are not increased.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change does not affect the control parameters governing unit operation or the response of plant equipment to transient conditions. The proposed change does not introduce any new equipment, modes of system operation or failure mechanisms.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

ATTACHMENT 2 Evaluation of Proposed Change Page 5 of 6 Does the change involve a significant reduction in a margin of safety?

Response: No NUREG-1433, Rev. 2, provided licensees with the latest NRC recommended content and format for TS. The NUREG-1433 SR for testing EFCVs, SR 3.6.1.3.10, specifies that this testing is associated with reactor instrumentation line EFCVs.

The Bases to SR 3.6.1.3.10 in NUREG-1433, Rev. 2, provides a reference to NEDO-32977-A, "Excess Flow Check Valve Testing Relaxation," dated June 2000. NEDO-32977-A was approved for use by licensees in a NRC letter dated March 14, 2000. NEDO-32977-A states the following on the scope of TS testing associated with EFCVs.

"EFCVs in instrument lines which connect to the reactor coolant pressure boundary (RCPB) are normally tested during refueling outages to meet Technical Specification requirements. Instrument lines that connect to the containment atmosphere, such as those which measure drywell pressure, or monitor the containment atmosphere or suppression pool water level, are considered extensions of primary containment. A failure of one of these instrument lines during normal operation would not result in the closure of the associated EFCV, since normal operating containment pressure is not sufficient to operate the valve. Such EFCVs will only close with a downstream line break concurrent with a Loss of Coolant Accident (LOCA). Since these conditions are beyond the plant design basis, EFCV closure is not needed and containment atmospheric instrument line EFCVs need not be tested."

The proposed change will incorporate the wording from NUREG-1433 into LaSalle County Station SR 3.6.1.3.8 to limit the scope of TS required testing to EFCVs that are directly connected to the RCPB.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based upon the above, EGC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

ATTACHMENT 2 Evaluation of Proposed Change Page 6 of 6

8.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

9.0 PRECEDENT The proposed amendment incorporates into the LaSalle County Station a change to SR 3.6.1.3.8 that is consistent with the wording and intent of NUREG-1433, Rev. 2, and NEDO-32977-A.

ATTACHMENT 3 MARKUP OF PROPOSED TECHNICAL SPECIFICATION PAGE CHANGE Revised TS Page

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify the isolation time of each MSIV is In accordance

> 3 seconds and < 5 seconds. with the Inservice Testing Program SR 3.6.1.3.7 Verify each automatic PCIV actuates to 24 months the isolation position on an actual or simulated isolation signal.

SR 3.6.1.3.8 Verify each EFCV actuates to the 24 months isolation position on an actual or simulated instrument line break signal.

SR 3.6.1.3.9 Remove and test the explosive squib from 24 months on a each shear isolation valve of the TIP STAGGERED TEST System. BASIS SR 3.6.1.3.10 Verify leakage rate through any one main In accordance steam line is < 100 scfh and through all with the four main steam lines is < 400 scfh when Primary tested at > 25.0 psig. Containment Leakage Rate Testing Program SR 3.6.1.3.11 Verify combined leakage rate through In accordance hydrostatically tested lines that with the penetrate the primary containment is Primary within limits. Containment Leakage Rate Testing Program LaSalle 1 and 2 3.6.1.3 -8 Amendment No. 147/133

ATTACHMENT 4 RETYPED PAGES FOR TECHNICAL SPECIFICATION CHANGE AND BASES CHANGES (FOR INFORMATION ONLY)

Retyped TS Page Retyped Bases Pages

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify the isolation time of each MSIV is In accordance 2 3 seconds and

• 5 seconds. with the Inservice Testing Program SR 3.6.1.3.7 Verify each automatic PCIV actuates to 24 months the isolation position on an actual or simulated isolation signal.

SR 3.6.1.3.8 Verify each reactor instrument line EFCV 24 months I actuates to the isolation position on an actual or simulated instrument line break signal.

SR 3.6.1.3.9 Remove and test the explosive squib from 24 months on a each shear isolation valve of the TIP STAGGERED TEST System. BASIS SR 3.6.1.3.10 Verify leakage rate through any one main In accordance steam line is

• 100 scfh and through all with the four main steam lines is

• 400 scfh when Primary tested at 2 25.0 psig. Containment Leakage Rate Testing Program SR 3.6.1.3.11 Verify combined leakage rate through In accordance hydrostatically tested lines that with the penetrate the primary containment is Primary within limits. Containment Leakage Rate Testing Program LaSalle 1 and 2 3.6. 1.3 -8 Amendment No.

PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.8 REQUIREMENTS (continued) This SR requires a demonstration that each reactor instrument line EFCV is OPERABLE by verifying that the valve I actuates to the isolation position on an actual or simulated instrument line break condition. This SR provides assurance that the reactor instrumentation line EFCVs will perform as I designed. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass this Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Instrument lines that connect to the containment atmosphere, such as those which measure drywell pressure, or monitor the containment atmosphere or suppression pool water level, are considered extensions of primary containment. A failure of one of these instrument lines during normal operation would not result in the closure of the associated EFCV, since normal operating containment pressure is not sufficient to operate the valve. Such EFCVs will only close with a downstream line break concurrent with a LOCA. Since these conditions are beyond the plant design basis, EFCV closure is not needed and containment atmospheric instrument line EFCVs need not be tested (Ref. 6).

SR 3.6.1.3.9 The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. .The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. Other administrative controls, such as those that limit the shelf life and operating life, as applicable, of the explosive charges, must be followed. The Frequency of 24 months on a STAGGERED TEST BASIS is considered adequate given the administrative controls on replacement charges and the frequency checks of circuit continuity (SR 3.6.1.3.4).

(continued)

LaSalle 1 and 2 B 3.6.1.3-14 Revi sion

PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.10 REQUIREMENTS (continued) The analyses in Reference 2 are based on leakage that is less than the specified leakage rate. Leakage through any one main steam line must be < 100 scfh and through all four main steam lines must be s 400 scfh when tested at Pt (25.0 psig). This ensures that MSIV leakage is properly accounted for in determining the overall primary containment leakage rate. The Frequency is required by the Primary Containment Leakage Rate Testing Program.

SR 3.6.1.3.11 Surveillance of hydrostatically tested lines provides assurance that the calculation assumptions of Reference 2 are met. The acceptance criteria for the combined leakage of all hydrostatically tested lines is 1 gpm times the total number of hydrostatically tested PCIVs when tested at 2 1.1 Pd. The combined leakage rates must be demonstrated in accordance with the leakage test Frequency required by the Primary Containment Leakage Rate Testing Program.

REFERENCES 1. Technical Requirements Manual.

2. UFSAR, Section 15.6.5.

3. UFSAR, Section 15.6.4.

4. UFSAR, Section 15.2.4.

5. UFSAR, Section 6.2.4.2.3.

6. NEDO-32977-A, "Excess Flow Check Valve Testing Relaxation," June 2000 LaSalle 1 and 2 B 3.6.1.3-15 Revi sion