ML20072F412
| ML20072F412 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 08/12/1994 |
| From: | Quay T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20072F414 | List: |
| References | |
| NUDOCS 9408230326 | |
| Download: ML20072F412 (6) | |
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NUCLEAR REGULATORY COMMISSION k.4.,(' /
i WASHINGTON, D.C. 2055H001 Wolf CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482 AMENDMENT TO FACILITY 0PERATING LICENSE Amendment No. 76 License No. NPF-42 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the. Wolf Creek Generating Station (the facility) Facility Operating License No. NPF-42 filed by the Wolf Creek Nuclear Operating Corporation (the Corporation), dated October 27, 1993, 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, as amended, 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 license 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.
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4 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of facility Operating License No. NPF-42 is hereby amended to read as follows:
2.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 76, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license.
The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION h Asa,$ hv-cy Theodore R. Quay, Director Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance:
August 12, 1994
ATTACHMENT TO LICENSE AMENDMENT NO, 76 FACILITY OPERATING LICENSE NO. NPF-41 j
f DOCKET NO, 50-482 Replace the following pages of the Appendix A Technical Specifications with the attached pages, The revised pages are identified by Amendment number and contain marginal lines indicating the areas of change.
The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE INSERT 3/4 6-2a 3/4 6-2a B 3/4 6-1 B 3/4 6-1
CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria speci-fied in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI N45.4-1972:
a.
Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 10 month
- intervals during shutdown at I
a pressure not less than either P, 48 psig, or P,, 24 psig, during each 10-year service period.
The third test of each set shall be conducted during the shutdown for the 10-year plant inservice inspection;
- A one time extension of the test interval is allowed for the third Type A test of the first 10-year service period, provided that unit shutdown occurs no later than March 31, 1996, and performance of the Type A test occurs prior i
to unit restart following the eighth refueling outage.
WOLF CREEK - UNIT 1 3/4 6-2a Amendment No. 13,76
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3/4.6 CONTAINMENT SYSTEMS BASES J/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses.
This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.
3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, P. As an added conservatism, the measured overall integrated leakage rate,is further limited to less than or or 0.75 L as applicable, during performance of the periodic equal to 0,75 L,for possibi,e degradation of the containment leakage barriers test to account between leakage tests.
For reduced pressure tests, the leakage characteristics yielded by measurements L,, and L,, shall establish the maximum allowable test leakage to L, (P,/P,),n} L,,/L,, is greater than In the eve 0.7, L,, shall be specified,as equal).
rate L of not more than L ( L,,/ L, The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50
- l 3_/4.6.1.3 CONTAINMENT AIR LOCKS j
The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.
Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.
- A one time extension of the test interval is allowed for the third Type A test of the first 10-year service period, as required by Surveillance Requirement 4.6.1.2.a and by Section Ill.D.l.(a) of Appendix J of 10 CFR 50, provided unit shutdown occurs no later than March 31, 1996 and performance of the Type A test occurs prior to unit restart following the eighth refueling outage.
WOLF CREEK - UNIT 1 8 3/4 6-1 Amendment No. 76
CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 3.0 psig, and (2) the containment peak pressure does not exceed the design pressure of 60 psig during steam line break conditions.
The maximum peak pressure expected to be obtained from a steam line break event is 48.9 psig.
The limit of 1.5 psig for initial positive containment pressure will limit the total pressure to 50.4 psig, which is less than design pressure and is consistent with the safety analyses.
3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial temperature condition assumed in the safety analysis for a steam line break accident.
Measurements shall be made at all listed locations, whether by fixed or portable instruments, prior to determining the average air temperature.
3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained in accordance with safety analysis requirements for the life of the facility.
Structural integrity is required to ensbre that the contain-ment will withstand the maximum pressure of 50.4 psig in the event of a steam l
line break accident.
The measurement of containment tendon lift-off for'e, the tensile tests of the tendon wires or strands, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the contain-ment, and the Type A leakage test are sufficient to demonstrate this capability.
The Surveillance Requirements for demonstrating the containment's structural integrity are in compliance with the recommendations of proposed Regulatory Guide 1.35, " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures," April 1979, and proposed Regulatory Guide 1.35.1, " Determining Prestressing Forces for Inspection of l
Prestressed Concrete Containments," April 1979.
The required Special Reports from any engineering evaluation of containment abnormalities shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection
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procedure, the tolerance on cracking, the results of the engineering evaluation and the corrective actions taken.
WOLF CREEK - UNIT 1 B 3/4 6-2 Amendment No. 50
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