ML20081A898
| ML20081A898 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/08/1995 |
| From: | Hannon J NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20081A901 | List: |
| References | |
| NUDOCS 9503150338 | |
| Download: ML20081A898 (10) | |
Text
{{#Wiki_filter:C 9% [ W B NUCLEAR REGULATORY COMMISSION ,2 T.
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WASHINGTON, D.C. 20066 4 001 .....,of Ci NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 115 License No. DPR-42 i 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Northern States Power Company (the licensee) dated January 13, 1995, 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 t the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-42 is hereby amended to read as follows: 9503150338 950308 PDR ADOCK 05000282 p
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Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 115, 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 issuance, with full implementation within 30 days. FOR THE NUCLEAR REGULATORY COMMISSION &a A John N. Hannon, Director Project Directorate III-I Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Spec.ifications Date of Issuance: Fbrch 8, 1995 I ) i j
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E~6 l ~ t c.- ATTACHMENT TO LICENSE AMENDMENT NO. 115 FACILITY OPERATING LICENSE NO. DPR-42 .l l DOCKET NO. 50-282 i ' Revise Appendix A Technical Specifications by removing.the pages identified below and inserting the attached pages..The revised pages are identified by amendment number and contain vertical lines indictting the area of change. REMOVE INSERT I TS 4.4-4 TS 4.4-4 l TS B.4.4-2 TS B.4.4-2 .l J 8 1 i ) q l -1 1
7 y y ? D TS 4.4-41 ^ 9 b.' Cold:DOP. testing shall be performed after each complete + {or partial _ replacement of a.HEPA filter bank.or after g .any structural-maintenance'on thersystem housing that Lcould affect-the HEPA bank bypass leakage, 4
- c. -Halogenated hydrocarbon testing shall be~ performed ~'
after each complete or partial replacement of a' chare 4 coal adsorber bank or after any structural maintenance on,the system housing that could affect the charcoal adsorber bank bypass, leakage. d. Each circuit shall be operated with the' heaters on at' least 10 hours every. month. l5. Perform an air distribution test on the HEPA filter bank' after any maintenance or testing that could affect the air. distribution within'the systems; 'Ihe; test shall be performed at rated flow-rate (1106). The results of the test shall b .show the' air distribution is uniform within 1204. y, C. Containment Vacuum Breakers The air-operated valve in each vent.line shall be tested at quarterly intervals to demonstrate that a simuisted contain-ment vacuus of 0.5 psi will open the valve and a simulated accident signal will close_the valve. The check. valves as. well as the butterfly valves will be-leak-tested during e'ach refueling shutdown in accordance with the requirements' of Speci- -fication 4.4.A.2.- j D. Residual Heat Removal System 1. Those portions of the residual heat removal. system external to the isolation valves at the containment, shall be hydro-j statically tested for leakage-during each refueling shutdown. l l 4 .i 2. Visual insp'ection shall be made for excessive leakage'from .l components of the system. Any visual leakage that cannot be stopped at test conditions shall be measured by collec-- .j tion and weighing or by another equivalent method'. 3. The acceptance criterion is that maximum allowable leakage from either train of the recirculation heat removal system .j components.(which includes valve stems; flanges and pump 1 seals) shall not exceed two gallons per hour when the system is at M0 psig. 4. Repairs shall be made as required to maintain leakage within the acceptance criterion _in Specification 4.4.D.3 5. If repairs are not completed within 7 days, the reactor shall be shut down and deprassurized until repairs are effected and1 the acceptance criterion in 3. above is satisfied. Prairie Island Unit 1 Amendment No. 17, 62, 115 Prairie Island Unit 2 Amendment No. 11, 56, 108 l i
B.4.4-2 4.4 CONTAINMENT SYSTEM TESTS Bases continued Several penetrations of the containment vessel and the shield building could, in the event of leakage past their isolation valves, result in leakage being conveyed across the annulus by the penetrations themselves, thus bypassing the function of the Shield Building Ventilation System (Reference 5). Such leakage is estimated not to exceed.025% per day. A special zone of the auxiliary building has minimum leakage construc-tion and controlled access, and is designated as a special ventilation zone where such leakage would be collected by either of two redundant trains of the Auxiliary Bui'. ding Special Ventilation System. This system, when activated, will supplant the normal ventilation and draw a vacuum throughout the zone such that all outleakage will be through particulate and charcoal filters which exhaust to the shield building exhaust stack. The design basis loss-of-coolant accident was initially evaluated by the AEC staff (Reference 3) assuming primary containment leak rate of 0.5% per day at the peak accident pressure. Another conservative assumption in the calculation is that primary containment leakage directly to the ABSVZ is 0.1% per day and leakage directly to the environs is 0.01% per day. The resulting two hour doses at the nearest SITE BOUNDARY and 30-day doses at the low population zone radius of 14 miles are less than guidelines presented in 10CFR100. Initial leakage testing of the shield building and the ABSV resulted in a greater inleakage than the design basis. The staff has reevaluated doses for these higher inleakage rates and found that for a l primary containment leak rate of 0.25% per day at peak accident pres-sure, the offsite doses are about the same as those initially calculated for higher primary containment leakage and lower secondary containment in-leakage (Reference 6). l The Residual Heat Removal Systems functionally become a part of the l containment volume during the post-accident period when their operation is changed over from the injection phase to the recirculation phase. I Redundancy and independence of the systems permit a leaking system to be isolated from the containment during this period, and the possible consequences of leakage are minor relative to those of the Design Basis Accident (Reference 4); however, their partial role in containment l warrants surveillance of their leak-tightness. l l The limiting leakage rates from the recirculation heat removal system l are judgment values based primarily on assuring that the components could operate without mechanical failure for a period on the order of 200 days after a design basis accident. The test pressure, 350 psig, l gives an adequate margin over the highest pressure within the system after I a design basis accident. A recirculation heat removal system leakage of 2 gal /hr will limit off-site exposure due to leakage to insignificant levels relative to those calculated for leakage directly from the containment in the design basis accident. Prairie Island Unit 1 Amendment No. 91, 107, 115 j Prairie Island Unit 2 Amendment No. 84, 100, 108 i
./ 'o., ~ UNITED STATES [ t>( B NUCLEAR REGULATORY COMMISSION g * *v[p*s[ 4 4 WASHINGTON, D.C. 20seHX101 o I I . NORTHERN STATES POWER COMPANY ( g DOCKET NO. 50-306 i PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 2 i AMENDMENT TO FACILITY OPERATING LICENSE ~ t Amendment No. 108 License No. DPR-60 1. The Nuclear Regulatory Commission (the Commission) has found that: i A. 'me application for amendment by Northern States Power Company (the licensee) dated January 13, 1995, complies with the standards and. requirements of the Atomic Energy Act of 1954, as amended (the Act), l 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 l 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 i satisfied. i 2. Accordingly, the license is amended by changes to the' Technical Specifications as indicated in the attachment to this license amendment, i and paragraph 2.C.(2) of Facility Operating. License No. DPR-60 is hereby amended to read as follows: t l
( o; Technical Specifications' z i The Technical Specifications contained in Appendix A, as revised -{ through Amendment No. 108, are hereby incorporated in the license. i ~ The licensee shall operate the. facility in accordance with the-Technical Specifications. I 3.~ This license amendment is effective as of the date of. issuance' with full implementation within 30 days. FOR THE NUCLEAR REGULATORY COMMISSION t ' John N. Hannon, Director Project Directorate III-I ' Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation ' Attachment Changes to the Technical - i Specifications q . Date of Issuance: March 8, 1995 l i f i I I ? f i y w
.u i i 'f ATTACHMENT TO LICENSE AMENDMENT NO. 108-FACILITY OPERATING LICENSE NO. DPR-60 ? l-DOCKET NO. 50-306- [ ~ e- - i ~ Revise Appendix A Technical Specifications by removing the pages identified i below and inserting the attached pages. The revised pages~are identified by amendment number and contain vertical lines indicating the area of change. I REMOVE INSERT l f TS 4.4-4 TS 4.4-4 B.4.4-2 B.4.4-2 l i 5 [ t I l j iJ
t', TS 4.4-4 b. Cold DOP testing shall be performed after each complete or partial replacement of a HEPA filter bank or after any structural maintenance on the system housin5 that could affect the HEPA bank bypass leakage. c. Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of a char-coal adsorber bank or after any structural maintenance l on the system housing that could affect the charcoal adsorber bank bypass leakage. d. Each circuit shall be operated with.the heaters on at least 10 hours every month. 5. Perform an air distribution test on the HEPA filter bank after any maintenance or testing that could affect the air distribution within the systems. The test shall be performed i at rated flow rate (1106). The results of the test shall show the air distribution is uniform within 1204. C. Containment Vacuum Breakers The air-operated valve in each vent line shall be tested at I quarterly intervals to demonstrate that a simulated contain-ment vacuum of 0.5 psi will open the valve and a simulated accident signal will close the valve. The check valves as well as the butterfly valves will be leak-tested during each i refueling shutdown in accordance with the requirements of Speci-fication 4.4.A.2. D. Residual Heat Removal System i 1. Those portions of the residual heat removal system external to the isolation valves at the containment. shall be hydro-statically tested for leakage during each refueling shutdown, i 2. Visual inspection shall be made for excessive leakage from' components of the system. Any visual leakage that cannot i be stopped at test conditions shall be measured by collec-tion and weighing or by another equivalent method. 3. The acceptance criterion is that maximum allowable leakage from either train of the recirculation heat removal system ) components (which includes valve stems; flanges and pump seals) shall not exceed two gallons per hour when the system is at 350 psig. 4. Repairs shall be made as required to maintain leakage within the acceptance criterion in Specification 4.4.D.3 5. If repairs are not completed within 7 days, the reactor shall be shut down and depressurized until repairs are effected and the acceptance criterion in 3. above is satisfied. Prairie Island Unit 1 Amendment No. 17. 62. 115 Prairie Island Unit 2 Amendnent No. 11, 56, 108
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1 p ~ f4.4 CONTAINMENT SYSTEM TESTS j n AAAAA continued .j i Several penetrations of the containment vessel and the. shield building { could, in the event of leakage past.their isolation valves, result in a leakage being conveyed across,the annulus by the penetrations;themselves,- 1 'y ~ thus bypassing the function of the Shield Building Ventilation System. (Reference 5). Such leakage is estimated not to exceed.0254 per day. 1 A special zone of the auxiliary building has minimum-leakage construc - 'l tion and controlled access, and is_ designated as'a'special ventilation zone where. such leakage would be collected by either of two redundant M trains, of:the Auxiliary Bui'. ding Special Ventilation Systea.,. This system. .i when: activated, will supplant the normal ventilation and draw a vacuum 'I throushout the. zone such that all outleakage will be through particulate: 4 and charcoal filters which exhaust to the shield building. exhaust stack. { 6 The design basis loss-of-coolant accident was initially evaluated by the AEC staff (Reference 3) assuming primary containment leak rate of 0.54 j? e per day at the peak accident pressure. Another. conservative assumption in the calculation is that primary containment leakage.directly to the ABSVZ~ -l is 0.14 per day and leakage directly to the environs is 0.014 per day. The resulting two-hour doses at the nearest SITE BOUNDARY.and 30-day. doses at the low population zone radius of 14 miles are less.than guidelines 1 presented in 10CFR100. Initial leakage testing of the shield building and the ABSV resulted in a greater inleakage than the design basis. The staff has reevaluated doses for these higher inleakage rates and found that for a 'l primary containment leak rate of 0.256 per day at peak accident pres-- t sure, the offsite doses are about the same-as those initially calculated. j for higher primary containment leakage and lower secondary containment in-leakage (Reference 6). The Residual Heat Removal Systems functionally become a part of the containment volume during the post-accident period when their operation is changed over from the injection phase to the recirculation phase. - . Redundancy and independence of the' systems permit a leaking system to i be isolated from the containment during this period, and the.possible1 l 3 consequences of leakage are minor relative to those of the Design Basis j Accident (Reference 4); however, their partial role in containment i warrants surveillance of their leak-tightness. The limiting leakage rates " rom'the recirculation heat removal system are judgment values based primarily on assuring that the components .j could operate without mechanical failure for a period on the order of j 200 days after.a design basis accident.. The test pressure, 350 psig, l gives an' adequate margin over the highest pressure within tho' system after I a design basis accident. A recirculation heat removal system leakage of 2 gal /hr will limit off-site exposure due to leakage to insignificant levels relative to those calculated for leakage directly from the containment in the design basis accident. l Prairie Island Unit 1 Amendment No. 91, 107, 115 i Prairie Island Unit 2 Amendment No. H, 100, 108 l l 2. . =. -. - l}}