L-MT-03-010, Revised Pages for a Previously Submitted Technical Specification Change

Revised Pages for a Previously Submitted Technical Specification Change

ML030550135
Person / Time
Site:	Monticello
Issue date:	02/12/2003
From:	Denise Wilson Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-03-010, TAC MB4919
Download: ML030550135 (5)

Text

NMC Ej Committed to Nuc~learExe Monticello Nuclear Generating Plant Operated by Nuclear Management Company, LLC February 12, 2003 L-MT-03-01 0 10 CFR Part 50, Section 50.90 US Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Revised Pages for a Previously Submitted Technical Specification Change (TAC No. MB4919)

Reference 1: Nuclear Management Company, LLC submittal to the Nuclear Regulatory Commission, "License Amendment Request for Risk Informed Technical Specification Change Regarding Five Year Extension of Type A Test Interval," dated April 22, 2002.

Attached are revised Technical Specification (TS) pages reflecting revisions to the Monticello Technical Specifications. These pages were previously submitted to the Nuclear Regulatory Commission (NRC) from the Monticello Nuclear Generating Plant by letter dated April 22, 2002 (Reference 1).

Due to the approval of TS Amendment 132, proposed text in the referenced License Amendment Request to be added to TS page 159 was relocated to TS page 258. In addition, portions of TS page 185 were reworded.

Revisions to TS pages 185 and 258 are being submitted to reflect additional changes to the Monticello Technical Specifications that have occurred because of the issuance of TS Amendment Number 132.

The above referenced changes are administrative and editorial in nature and have no technical impact on pages or justifications previously submitted. These changes to Monticello TS are needed to correctly reflect the current page configuration of the TS due to the issuance of TS License Amendment Number 132. Therefore, the Determination of No Significant Hazards Consideration and Environmental Assessment submitted by the original letter dated April 22, 2002, are also applicable to this submittal.

2807 West County Road 75 e Monticello, Minnesota 55362-9637 0o\

Telephone: 763.295.5151

• Fax: 763.295.1454

USNRC Nuclear Management Company, LLC Page 2 The attachment cover page provides remove and insert instructions for the Reference 1 submittal.

If you have any questions relating to these revised pages or the previously submitted License Amendment Request, regarding the Five Year Extension of Type A Test Interval Technical Specification Change, please contact John Fields, Senior Licensing Engineer, at (763) 295-1663.

David L. Wilson Site Vice President Monticello Nuclear Generating Plant Subscribed to and sworn before me this ,,Pd'ay of _-_.__

.~~~~~ ... .........

Notar KATHRYN 1. KLEINE~

NOTARY PUBLIC - MINNESOTA t MY COMM. EX;.JaM31a, 2005

Attachment:

Revised Monticello Technical Specification Pages cc: Regional Administrator-Ill, NRC NRR Project Manager, NRC Sr. Resident Inspector, NRC Minnesota Department of Commerce J. Silberg, Esq.

ATTACHMENT Revised Monticello Technical Specification Pages Remove from Reference 1 Submittal Insert into Reference 1 submittal 159 185 185 258

Bases 4.7 (Continued):

On September 26, 1995, Regulatory Guide 1.163 became effective providing guidance on performance based testing to the requirements of 10 CFR 50, Appendix J, Option B. Monticello has adopted 10 CFR Part 50, Appendix J, Option B, and a Primary Containment Leakage Rate Testing Program. This program is modified by the following exception: NEI 94-01, Rev. 0, Section 9.2.3: The first Type A test performed after March 1993 Type A test shall be performed no later than March 2008.

Maintaining primary containment integrity requires compliance with the visual examinations and leakage rate test requirements of the Primary Containment Leakage Rate Testing Program. Failure to meet air lock leakage testing, primary containment purge valve resilient seal leakage testing or main steam isolation valve leakage does not necessarily result in a failure of this SR. The impact of the failure to meet these SRs must be evaluated against the Type A, B and C acceptance criteria of the Primary Containment Leakage Rate Testing Program.

Maintaining primary containment air locks requires compliance with the leakage rate testing requirements of the Primary Containment Leakage Rate Testing Program. This SR reflects the leakage rate testing requirements with respect to-air lock leakage (Type B leakage tests). The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall primary containment leakage rate. The frequency is required by the Primary Containment Leakage Rate Testing Program.

The SR has been modified by two footnotes. Note

• states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission barrier in the event of a DBA. Note ** has been added to this SR requiring the results to be evaluated against the acceptance criteria which is applicable to Primary Containment Integrity. This ensures that the air lock leakage is properly leakage.

accounted for in determining the combined Type B and C primary containment The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident primary containment pressure, closure of either door will support primary containment integrity. Thus, the interlock feature supports primary containment integrity while the air lock is being used for personnel transit in and out of the containment. Periodic testing of the interlock demonstrates that the interlock will function as designed and that the simultaneous inner and outer door opening will not inadvertently occur. Due to the purely mechanical nature of this interlock, and given that the interlock mechanism is not normally challenged when primary containment air lock door is used for entry and exit (procedures require strict adherence to single door opening), this test is only required to be performed every 24 months. 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 loss of primary containment integrity if the Surveillance were performed with the reactor at power. The 24 month frequency is based on engineeringjudgme'nt'and is considered adequate given that the interlock is not challenged during the use of the air lock.

4.7 BASES NEXT PAGE IS PAGE 188 185 Amendment No. 95, 400a, 132

I 44 6.8.J -'RESERVED K. Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

1. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

2. Changes to Bases may be made without prior NRC approval provided the changes do not involve either of the following:

a. a change in the TS incorporated in the license; or

b. a change to the USAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

3. The Bases Control Program shall contain provisions to ensure that the Bases, are maintained consistent with the USAR.

4. Proposed changes to the Bases that involve changes as described in a. or b. of Specification 6.8.K.2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71 (e).

6.8.L - RESERVED M. Primary Containment Leakage Rate Testing Program

1. This program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B, as modified by approved exemptions. 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 modified by the following exception: NEI 94-01, Rev. 0, "Industry Guideline for Implementing Pe'rformance-Based Option of 10 CFR 50, Appendix J."

Section 9.2.3: The first Type A test after the March 1993 Type A test shall be performed no later than March 2008.

2. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 42 psig. The containment design pressure is 56 psig.

6.8 258 Amendment No. 1 20, 22,1432 ,