ML14183A063
| ML14183A063 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 10/28/1986 |
| From: | Rubenstein L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML14183A064 | List: |
| References | |
| DPR-23-A-107 NUDOCS 8611140124 | |
| Download: ML14183A063 (5) | |
Text
QREG&
0R UNITED STATES 0
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 CAROLINA POWER AND LIGHT COMPANY DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.107 License No. DPR-23
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Carolina Power and Light Company (the'licensee) dated July 17, 1984, as supplemented February 5, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-23 is hereby amended to read as follows:
861114012486-1028 PDR ADOCK 05000261 P
-2 (B) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised throunh Amendment No. 107, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Lester S. Rubenstein, Director PWR Pro~iect Directorate P?
Division of PWR Licensing-A
Attachment:
Chanaes to the Technical Specifications Date of Issuance: October 28, 1986
ATTACPMENT TO LICENSE AMENDMENT AMENDMENT NO. 107 FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-461 Revise Appendix A as follows:
Remove Paaes Insert Pages 4.4-1 4.4-1 4.4-9 4.4-9
(RBR-10) 4.4 CONTAINMENT TESTS Applicability Applies to containment leakage and structural integrity.
Objective To verify that potential leakage from the containment and that pre-stressing tendon loads are maintained within acceptable values.
Specification 4.4.1 Operational Leakage Rate Tests 4.4.1.1 Integrated Leakage Rate Tests (ILRT)
- a.
Integrated leak rate tests shall be performed prior to initial plant operations at the containment design pressure (P ) of p
42 psig and.at a test pressure (Pt) of 21 psig to establish the respective measured leak rates, Lm (42) and Lm (21).
The minimum test temperature will be 50*F.
- b.
Subsequent integrated leakage rate tests shall be performed at intervals specified in 4.4.1.1.g at an initial pressure (beginning of test) at or above 21 psig (50% of design pressure).
The first integrated leak rate test shall be performed at 21 psig and 42 psig.
- c.
The test duration shall meet the requirements of 10CFR5O.
Appendix J, and ANSI N45.4 (1972) for leakage rate measurements, and shall be extended a sufficient period of time to verify, b'r superimposing a known leak rate on the containment, the validity and accuracy of the-leakage rate results.
4.4-1 Amendment No. 107
(KBR-10)
The performance of a periodic integrated leak rate test during plant life provides a current assessment of potential leakage from the containment in case of an accident that would pressurize the interior of the containment.
In order to provide a realistic appraisal of the integrity of the containment under accident conditions, this periodic leakage rate test is to be performed without preliminary leak detection surveys or leak repairs and containment isolation valves are to be closed in the normal.manner.
The test pressure of 21 psig for the periodic integrated leakage rate test is sufficiently high to provide an accurate measurement of the leakage rate and it duplicates the pre-operational leak rate test at 21 psig. The factor of 0.8 relates the measured leakage of air to the potential leakage of a steam-air mixture. The specification also allows for possible deterioration of the leakage rate between tests, by requiring that only 75% of the allowable leakage rates actually be measured. The basis for these deterioration allowances is arbitrary, but is believed to be conservative and will be confirmed or denied by periodic testing. If indicated to be necessary, the deterioration allowances will be altered based on experience.
As stated in ANSI N45.4 (1972), "The leakage-rate test period, for any method, shall extend to 24 h~ours] of retained internal pressure. If it can be demonstrated to the satisfaction of those responsible for the acceptance of the containment structure that the leakage rate can be accurately determined during a shorter test period, the agreed-upon shorter period may be used,"
If an ILRT of a duration less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is attempted, the provisions of the Bechtel Topical Report BN-TOP-1, Revision 1, will be met.
The specified frequency of periodic integrated leak rate tests is based on the following major considerations. First is the low probability of leaks in the liner, because of (a) the test of the leak tightness of the welds during erection; (b) conformance of the complete containment to a low leakage rate limit at 42 psig during preoperational testing which is consistent with 0.1% leakage at design basis accident (DBA) conditions; and (c) absence of any significant stresses in the liner during reactor operation.
4.4-9 Amendment No. 107