ML20235U422
| ML20235U422 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 09/21/1987 |
| From: | Salisbury J NEBRASKA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20235U383 | List: |
| References | |
| 3.3, CNSS876103, NUDOCS 8710140085 | |
| Download: ML20235U422 (8) | |
Text
{{#Wiki_filter:l CUUrtK buGLLAM d'1AA1US Ut'ttui1 AUhd v1M.4 U A L. ATTACEMbT"A"CN'SENGINEERINGPROCEDURE 3.3 STATION SAFETY EVALUATIONS CNSS,876103 Page 8 of 14 C ] 10CFR50.59 DEPORTABILITY ANALYSIS 1 i i Proposed Activity (Include Applicable Number And Description): PTM 87-j The PTM will require closure of the MS-HO-BV3 hydraulic oil isolation valve. This action will prevent MS-HOV-BV3 from opening under normal or emergency situations. The closure is required to isolate the hydraulic control unit to MS-HOV-BV3 which is currently leaking. i ) 1 I { i l (') i ( l 1 I. USAR 1. Does this proposed activity constitute 2. List the affected Section(s) of the l l a change in the facility or procedures USAR. as described in the Updated Safety Analysis Report? Page/ Volume Section Figure Yes [ No @ For purposes of determining report-ability, if the answer to Section II. or III. is YES, then the change is reportable under 10CFR50.90. If the answer.to Section I. is YES and answers to Section II. and III. are NO, then the change is reportable under 10CFR50.59b and a description .of.the change will be included in the Annual Report. All other changes are' .not reportable. I \\ 8710140085 871001 l PDR ADOCK 05000298 g PDR Procedure Number 3.3 Date Lf ~2_ - 3 7 Revision _5 Page. 1 Of 3 Pages
1 COOPER NUCLEAR STAT 10N UPt.KAT10M MANUAL ATTACW1NT "A" CNS" ENGINEERING PROCEDURE 3.3 STATION SAFETY EVALUATIONS ONSS876103 Page 9 of 14 10CFR50.59 DEPORTABILITY ANALYSIS C II. TECHNICAL SPECIFICATIONS i l l 1. Does this proposed activity involve a 2. List the affected Section(s) of the change in the Technical Specifications Technical Specifications. incorporated in the License? l Yes C No @ Section Page j i i l 1 I ~~UNREVIEVED SAFEkT QUESTION III. 1. Is this activity potentially an unreviewed safety question? Yes No @ j (A potential unreviewed safety question exists if the answer to 2.a., 2.b., or 2.c. (, below is Yes.) 2. Unreviewed safety question evaluation - answer the following questions with a Yes or 'No and'irov'ide specific Fe'asons' j'us~tifying 'the 7!ecision: ~ ~ Is the probability of occurrence or the consequences of an accident or a. malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report increased? Yes No @ Because: The USAR assumes complete BPV fniture in thn ninend pos'i t ion. This 'PTM fails only one valve'(BPV-3). ** b. Is the possibility of an accident or malfunction of a different type than any evaluated previously in the Updated Safety Analysis Report created? Yes C No @ Because: Load Reject without Bvpass Valve transient hns bnen fully evaluated and is more limitine than a Lond Reieet with nne Evnnss fniinve, Is the margin of safety, as defined, in the basis for any Technical Specification c. reduced? Yes O No X Because: The safety margin in Technical Specifications in based on complet e"BPV failure in the closed positinn, ** In the case of Feedwater Controller failure, General Electric has evaluated .this for two BPV operation and by limiting MCPR ?. 1.35 will not create the possibility for.an accident or malfunction of a dif ferent type than previously evaluated in the USAR. .ProcedureNu$ber 3.3-Date 84 3 ~7 Revisicn 5 .fage 2 Of 3 Pages O 1
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\\q - 1 j(. . f. g hUCIAAK ~ 5TA*110h.. J-S h s COOf t.K Urr nA11Und tumunt s 9 4 ~f i J 'ITTACl$ENT"A"CNSEh'CINEER'INGFROCEDUREl s 3. *3 - STATIOM"SAFECV EVALUJ? IONS w o. j 'CNSS876103 i 8 g - ,/ j \\ Pag 10 of 14' .5 , + r / 4 C .I 10CFR50.59 REPOR.TABI1?W ANALYFIS 71 3 ) q., IV. . PRIOR NRC AU"n10RTZATION 1. Is prior NRC authorization required? ~ .l. T, l ^* )' ] Yes C No ,'p i j X (Prior NRC authorizario g i,e1 required if the answer to II. or IIJ1,1. is Yes.) Y .s> t. change will be submitted by the Initiator ' If the answer to' IV.I. i's Yas, then ::%'g and Safety Department to obtain NRC autho 2. to the Manager of the Nuclear Licensin zation prior to imph mhntation af the dubject change.1, 3* ~ t 1 ~ y ) 3. NRC authorization received: Refe\\ i i rence: e .n l I 1 s g p -K t - ^ O J k, Date/ Licensing Departmentit s 4t t j i g R.{ ' ~ } \\ Performed By: 4%f?r h4ch. ~2/~b7 j Date: i G i, ~, s. i g ss t i g '. (' '{\\! (* y / i e '?. g s s, \\ i / l 4 / l t a s ~ \\ l Qt' 3 g ,4 ? e s 4 i -;o L. / 8 B i a ) j n i } j i z\\ f 'N I $. j \\ l. + 8 a ,/
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' ~ Court.K hucLean stastun urtna11Und nanunL ^ I ATTACHMEh4T"B"CNSENGINEERINGPROCEDURE 3.3 STATION SAFETY EVALUATIONS 1 Ci4SS876103 ) P ge 11 of 14 } FIRE PROTECTION / APPENDIX R DESIGN EVALUATION REVIEW CHECKLIST Document Number: PTM 87-Proposed Activity: Isolation of MS-HOV-BV3 Systems Affected: Turbine Bypass System A. Preliminary Fire Protection Evaluation (To Be Completed By The Responsible Engineer) (Place A Check Mark Next To Any Applicable Item). 1. Fire Protection Impact - Does the proposed modification impact any of the following fire protection features: l Yes X No Fire Detection System. Yes X No Suppression (fixed, portable extinguishers, hose stations). ~ Yes X No Fire area barrier (walls, ceilings, floors, doors, dampers). Yes X No Penetrations through fire area barriers (cable, piping, . conduit, duct work, hatches, etc.). i o 1 l l Yes X No Fire area barrier penetration seals (cable tray dividers, i ~ ~ silicone foam, knowool, 5 star grout, and 3M seals). ) _ Yes 1 No Fuel spread limiters (curbs, drains). Yes X No Combustibles (significant addition / deletion of oil, grease, I cabling, charcoal. PVC, Class A wood products, combustible i pipe insulation). l Yes X No Emergency lighting (8 hout battery packs, lamps). Yes X No Communication System (portable radio system, antennas, ~ ~ repeaters). l l Yes X No Plant access (locked doors, access to fire fighting equip-ment). If Yes to any of the above fire protection features, then complete Step B.l. of the Plant Fire Protection Review and a Fire Protection Engineer is I to complete Step B.2. 2. Plant Elcetrical System Impact - Does the proposed modification impact any of the following plant electrical features: """" Yes - No Cables associated with safe shutdown components (see X Attachment "C"). Yes X No Safe shutdown component control or logic. 1 1 l_ mn l_ l eamsmu m - m m
1 LUVt'En WULLLAK b1A1AVA Uf' Lad ! 1 Ud 3 n M usu, ATTACHMENT "B" CNS' ENGINEERIi4G PROCEDURE 3.3 STATION SAFETY EVALUATIONS CNSS876103 Pege 12 of 14 FIRE PROTECTION / APPENDIX R DESIGN EVALUATION REVIEW CHECKLIST Yes X No Components which support the operation of safe shutdown components. Yes X No Essential AC or DC Power Systems: 125 V DC/250 V DC. j j 120 V AC/440 V AC/4160 V AC. { Yes X No Emergency lighting. Yes X No Plant Communication System. If Yes, then complete Step C.1. of the Plant Electrical System Review and an Electrical Engineer is to complete Step C.2. j 1 l 3. Plant Mechanical System Impact - Does the proposed modification affect the operation of: Yes X No Safe shutdown system operational characteristics (flow rate, normal lineup) (see Attachment "C"). 1 Yes X No Safe shutdown component operation. l If Yes, then complete Step D.I. of the Plant Mechanical System Review and l a Mechanical Engineer is to complete Step D.2. l If all answers are No in Steps A.1., A.2., and A.3., then the review is i complete. I h ~ 2 f - b7> Responsible Engineer: e Date: If all answers in S eps A.l., A. n A.3. are No, en the remaining pages of Attachment "B" yb deleted. { \\ l l c) l l Procedure Number 3.3 Date y-2-87 Revision s Page 2, Of 9 Pages
GE.408 925-4091 TEL No. 408925 4091 Sep 21,87 12:30 P.02 CNSS875103 Edelosure 4 GENER AL h ELECTRIC Page 13 of 14 + COOPER NUCLEAR STATION TURBINE BYPASS VALVE OUT-OF-SERVICE ASSESSMENT September 21, 1987 Cooper Nuclear Station (CNS) is currently operating at approximately 75% of rated thermal power with a critical power ratio (CPR) of approximately 1.50. Early this. morning a hydraulic fluid leak associated with one of the three turbine bypass valves was discovered. The origin of the leak (high or low pressure hydraulic supply). is currently unknown. In order to determine both the source of the leak and the ascertain whether or not it is possible to isolate the nource (only the high pressure line is equipped. with an isolation valve) it is necessary to disablethe bypass valve. If it is possible to isolate the leak, it would be desirable to continue operating CNS until a time when the repairs can more conveniently be made. If-it is not possible to isolate the leak CNS may have to be shut down in order to make the necessary repairs. Currently, the limiting Cycle 11 fuel thermal limits (Option B) are dictated by the feedwater controller failure - maximum demand (WCF) event. The calculated minimum ' critical power ratio _ (MCPR) for this event at end-of-cycle (EOC) is 1,23 and the corresponding technical specification limit is 1.25 (i.e., a margin of 0.02 exists to the technical specification limit to allow for 10CFR50.59 evaluations of' the cycle specific limits). The WCF transient is analy:cd assuming all ' ( three bypass valves are operable. Disabling one of the bypass. valves may increase the calculated CPR for the WCF event. In order to allow-Nebraska Public Power District (NPPD) to investigate the source of the leak and determine if the leak can be isolated, General Electric has been requested to perform an assessment of the potential increase in the CPR for the WCF with one bypass valvo inoperable. Based on a review of j the current CNS response to the WCF and load rejection without bypass t events, the maximum increase in the CPR (over the current value) fpr the I WCF event. with one bypass valve inoperable' is expected to be significantly less than 0.10 Consequently, administratively limiting the operatinS CPRs above 1.33 (0.10 over the Cycle 11 calculated value at EOC, Option D) will conservatively ensure that the safety limit MCPR ~ of 1.0/ will not be violated while operating with one bypass valve out of service considering any of the abnormal transient events l previously analy::ed for CNS. Prepared by: mt egv// K.F. Cornwell, Senior Engineer I Application Engineering Services Reviewed by d Gk/htb J.F. Klapproth4 Principal Engineer g* l Licensin5 and Consulting Services F ,i a Approved by: N! o ht4 s G.L.Sozzi/,Mangg6r' Application Enginocring Services j l.
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CNSS876103 -g riffFlosure 5 NElbASKA Pusuc POWER LhdTRICT - g' f %n 1 nr 1 CNSS870476 Date September 22, 1987 FOR INTER-DISTRICT SORC Members To I1USINESS ONLY From G. R. Horn Subject SORC Meeting S87-097, September 21, 1987 Attendees: G. R. Horn E. M. Mace J. M. Meacham P. L. Ballinger D. M. Norvell R. L. Gibson R. Brungardt K. C. Walden SORC convened at 1500 to review a Plant Temporary Modification (PTM) to isolate HP hydraulic oil to Bypass Valve No. 3 due to an oil leak. The PTM was approved as submitted. Additionally, SORC requested issuance of a Special Order to administratively control MCPR greater than or equal to 1.35 until such time that General Electric provides the formal computer analysis to re-define MCPR limits with one bypass valve inoperable. A temporary procedure change to 6.2.4.1 to reflect the conservative MCPR change will also be processed. Paul Ballinger was directed to change MCPR limits in the process computer to reflect the administratively controlled MCPR limit of greater than.or equal to 1.35. SORC contacted General Electric, San Jose, to confirm that the turbine stop valve closure 25% scram block would not be affected by the single inoperable bypass valve. General Electric concurred with this interpretation and after discussing the fact that at less than 30% the high flux and high pressure scram adequately protect the reactor, the telecon was terminated. /% G. lorn Divi on Manager of Nuclear Operations GRH:EMM:lb cc: D. E. Schaufelberger H. G. Parris L. G. Kuncl P. V. Thomason K. C. Walden C. M. Kuta 1 NRC Resident Inspector 1
'CNSS876103 7 R$ N; 6 NE uASKA Punuc Powen d ornicT v CNSS870482 Date September 24, 1987 FOlt INTElt-DISTIllCT To SORC Members BUSINESS ONLY Frorn G. R. Horn Subject SORC Meeting S87-098, September 23, 1987 Attendees: G. R. Horn E. M. Mace J. M. Meacham P. L. Ballinger R. Brungardt R. L. Gibson SORC convened at 1600 on September 23, 1987, to review the turbine bypass l valve out-of-service evaluation summary received from General Electric. The j summary establishes a HCPR limit of 1.26 for the remaining Cycle 11 reload or until the bypass valve is repaired. SORC reviewed this change and concurred with the evaluation. Additionally, SORC requested revision of the Special Order to administratively control MCPR greater than 1.26. A temporary procedure change to 6.2.4.1 to reflect the MCPR change will also be processed. Paul Ballinger was directed to change MCPR limits in the process computer to reflect a value of greater than 1.26. l Additionally, based on input from General Office Licensing, SORC is of the understanding that through discussions with the CNS NRC Project Manager, there will not be a need for a formal Tech Spec change. However, per request, a letter will be sent to the NRC Project Manager providing the details of this issue and the revised MCPR limit. 1 0% ) G. . Horn ] Division Manager of I Nuclear Operations GRH:EMM:lb 1 Attach. I cc: D. E. Schaufelberger H. G. Parris L. G. Kuncl P. V. Thomason K. C. Walden C. M. Kuta NRC Resident Inspector l i ._-_______a}}