ML20217D695
| ML20217D695 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  (DPR-24-A-181, DPR-27-A-185) |
| Issue date: | 09/29/1997 |
| From: | Gundrum L NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217D699 | List: |
| References | |
| NUDOCS 9710060073 | |
| Download: ML20217D695 (8) | |
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UNITED STATES t
P; NUCLEAR REGULATORY COMMISSION j'
WASHINGTON, D.C. 30 2 6-0001 5
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WLS_CONSIN ELECTRIC POWER COMPANY DOCKET NO.50-20g POINT BEACH NUCLEAR PLANT. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICEgig Amendment No.181 License No. DPR 24 1.
The Nuclear Regulatory Commission (the Commission) has found mat:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated August 14,1997, 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, 9710060073 970929 PDR ADOCK 05000266 P
A e
2-2.
Accordingly, the license is amended by changes to the Technical Specifications as 4
indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR 24 is hereby amended to read as follows:
B.
Technical Soecifications The Technical Specifications contain6d in Appendices A and B, as revised through Amendment No.181, are hereby incorporated in the license. The licensee shall operate the facility in accordance with Technical Specificatioris.
3.
This license amendment is effective immediately upon issuance. The Technical Specifications are to be implemented within 45 days from the date of issuanos.
FOR THE NUCLEAR REGULATORY COMMISSION
$h 1td' Lum Linda L. Gundrum, Project Manager Project Directorate 1111 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation e
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Attachment:
Changes to the Technical Specifications Date of issuance: September 29, 1997
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t UNITED STATES s
NUCLEA84 RECULATERY CSMMISSIEN WASHINGTON. D.c. 30646 4 001
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WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50 301 POINT BEACH NUCLEAR PLANT. UNIT NQJ AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.185 License No. DPR 27 1.
The Nuclear Regulatory Commission (the Commission) has found that A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated August 14,1997, complies with the standards and a
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.
t 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 27 is hereby amended to read as follows:
i B.
Technical Specifications l
The Technical Specifications contained in Appendices A and B, as revised I
through Amendment No.185, are hereby incorporated in the license. The licensee shall operate the facility in accordance with Technical Specifications.
3.
This license amendme'nt is effective immediately upon issuance. The Technical Specifications are to be implemented within 45 days from the date of issuo ce.
FOR THE NUCLEAR REGULATORY COMMISSION h
11clAunu Linda L. Gundrum, Project Manager Project Directorate lil 1 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: September 29, 1997 1
ATTACHMENT TO LICENSE AMENQNJjNT NG3.181 AND 185 TO FACILITY OPERATING LICENSE NOS. DPR-24 AND DPR-27 DOCKET NOS. 50 266 AND 50-301 Revise Appendix A Technical Specifications by remou.ig the pages identified below and inserting the enclosed pages. The revised pages are identified by amendinent number and contain verticallines indicating the area of change.
4 I
EEMOVE INSERT i
15.4.2 5 15.4.2-5 15.4.4 7 15.4.4-7 15.4.4-8 15.6.12-1 15.6.12-1
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B.~
- in-Service lnspection and Testing of Safety Class Components Other than Steam Generator Tubes -
1.
Inservice inspection of ASME Code Class 1, Class 2 and Class 3 components shall be performed in accordance with Section XI of the ASME Roller and
. Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g) modified by Section 53.55a(b), except where specific written relief is granted by the NRC, pursuant to 10 CFR 50, Section 50.55(g)(6)(i).
a.
Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
I 2.
- Containment isolation valves will be tested in accordance with the Containment IE?kage Rate Testing Program,-
3.
Inservice testing of ASME Cooe Class 1,2, and 3 pumps, valves, and snubbers shall be performsd in s:cordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicabee Addenda as required by 10 CFR 50.55a.
a.
Nothing in the ASME Boiler and Pressure Vessel Code shall be i
construed to supersede the requirements of any Technical Specification.
Ba8iL The steam generator tube inspection requirements are based on the guidance given in NRC Regulatory Guide 1.83, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes." ASME Section XI Appendix IV is being used for defining the basic requirements _or the inspection method, However, at the present tirn, changes and improvements in steam generator. eddy current inspection are occurring faster than the code can be revised. : Thus, in order to ensure that the best possible exam of the tubing and/or sleeves is being done, the technique utilized will, in general, be the latest industry-accepted technique. This means that complete word-for-word compliance with Appendix IV mcy not be possible, However, the basic requirements and intent will be met, to the extent practical.
Specification 15.4.2.B delirmates programmatic requirements for establishing inservice inspection and Testing programs in accordance with the ASME Section XI Code and 10 CFR 50.55a requirements. The Code establishes criteria for system and component inspection and testing to ensure an appropriate level of reliability and detection of abnormal conditions. Failure to meet Code requirements is evaluated on an individual system or component bases to determine operability. ' Appropriate LCOs are entered if a system or component is determined to be inoperable.
' As stated in 15.4.2.B.1, safety class components, other than the steam generator tubing, will be inspectad in accordance with ASME Section XI. The code edition / addenda utilized for the inspection interval will be as defined in Unit 1 - Amendment No. 83,95,440,181 Unit 2 - Amendment No. SS,90,464,185 15.4.2-5
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Safety analyses have been performed on the basis of a leakage rate of 0.40% by weight per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60 psig. With this leakinge rate and with minimum containment engineered safety systems for iodine removal in operation, i.e. one spray pump with sodium hydroxide addition, the public exposure would be well below 10 CFR 100 values in the event of the design basis accident.* '
The safety analyses indicate that the containment leakage rates could be slightly in excess of 0.75% per day before a two-hour thyroid dose of 300R could be received at the site boundary.
The performance of periodic integrated leakage rate tests during plant life provide a current assessment of potentialleakage from the containment in case of an accident that would pressurize the interior of the containment. These tests are performed in accordance with the Containment Leakage Rate Testing Program.
Periodic visual and physical inspection of the containment tendons is the method to be used to detoimine loss of load-carrying capability because of wire breakage or deterioration. The tendon surveillance program specified in 15.4,4.ll Is based on the recommendation of Regulato,y Guide 1.35 Rev. 3. Containment tendon structural integrity was demonstrated for both units at the end of one, three and eight years following the initial containment structural integrity test.
The pre stres: lift-off test provides a direct measure of the load-carrying capability of the tendon. A deterioration of the corrosion preventive properties of the sheathing filler will be.
indicated by a change in the physical appearance of the filler, if the surveillance program indicates, by extensive wire breakage, tendon stress-strain relations, or other abnormal conditioris, that the pre-stressing tendons are not behaving as expected, the abnormal conditions will be subjected to an engineering analysis and evaluation in accordance with Specification 15.4.4.ll.D to determine whether the condition could result in a significant adverse impact on the containment structural integrity. The specified acceptance criteria are such as to alert attention to the situation well before the tendon load-carrying capability would deteriorate to a point that failure during a design basis accident might be possible. Thus, the cause of the incipient deterioration could be evaluated and corrective action studied without
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need to shut down the reactor. If the engineering evaluation determines that the abnormal condition could result in a significant adverse impact on the containment stru.,tural integrity, an abnormal degradation situation will be declared and a report submitted to the NRC in accordance with the specifications.
Referengga (1) FSAR Section 5.1.2.3 (2) FSAR Section. 5.1.2 (3) FSAR Section 14.3.5 (4) FSAR Section 14.3.4 (5) Deleted l
(6) FSAR pages 5.1-86 and 5.1-87 Unit 1 - Amendment No. 440,181 Unit 2 - Amendment No. 474,18.5 15.4.4-7
1 15.6,12 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A.
A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 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.
B.
The peak design containment intemal accident pressure, P, is 60 psig.
1 C.
The maximum allowable primary containment leakage rate, L, at P,,
shall be 0.4% of containment air weight per day.
D.
Leakage rate acceptance criteria are:
1.
The containment leakage acceptance criterion is s1.0 L.,
2.
During the first unit startup following testing in s.ccordance with t
this program, the leakage rate acceptance criteria are s 0.6 L, for the combined Type B and Type C tests and s 0.75 L, for Type A tests.
E.
The provisions of Specification 15.4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.
F.
The provisions of Specification 15.4.0.3 are applicable to the Containment Leakage Rate Testing Program.
Unit 1 - Amendment No. 449,181 Unit 2 - Amendment No. 474,185 15.6.12-1
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