3F0797-33, Provides Addl Info on Independent Assessment That Was Conducted on Certain Portions of Fire Protection Program & Listing of Issues & Tracking Status

Provides Addl Info on Independent Assessment That Was Conducted on Certain Portions of Fire Protection Program & Listing of Issues & Tracking Status

ML20148T783
Person / Time
Site:	Crystal River
Issue date:	07/03/1997
From:	Holden J FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
3F0797-33, 3F797-33, NUDOCS 9707090315
Download: ML20148T783 (8)

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Florida Power Unit 3 Dochat No.60-301 July 3,1997 3F0797-33 U.S. Nuclear Regulator Commission Attn.: Document Control Desk Washington, D. C. 20555-0001

Subject:

Florida Power Corporation (FPC) Meeting with the U.S. Nuclear Regulatory Commi.sion (NRC) staff on Crystal River Nuclear Plant, Unit 3 (CR3) Appendix R issues, June 11,1997 - Additional Information FPC met with the NRC staff at NRC Headquarters in Rockville on June 11,1997, and provided a presentation summarizing the status and actions on CR3 Appendix R issues. During this meeting, FPC presented an overview of an independent Assessment that was conducted on certain portions of CR3's Fire Protection Program. This letter provides additional information on the assessment and Attachment 2 listing the issues and tracking status.

Historv: During the development of CR3's Thermo-Lag Resolution Program, some -f i

discrepancies and inconsistencies were discovered in the CR310CFR50 Appendix R Fire Study. These problems were documented on Precursor Cards and entered into the CR3 Corrective Action System for identification, tracking and corrective action. As Thermo-Lag analysis work continued in 1996, more Appendix R inconsistencies were discovered and entered into the Corrective Action System.

This increasing trend of Appendix R related Precursor Cards was identified, and Problem Report 96-0401 was issued on September 27,1996.

This Problem Report questioned the adequacy of documentation and programmatic ,

controls that support the post-fire safe shutdown analysis, and included corrective actions for conducting a comprehensive examination of the post-fire safe shutdown 9707090315 970703 ADOCK 050003 2 l ll ll llll,lllll{lll l ll fDR Od O (} U O CRYSTAL RIVER ENERGY COMPLEX:16760 W Power une $t

• Crystal River, Florida 34428-6708 * (352) 795-6486 A Fkrida Progress Company

U. S. Nuclear Regulatory Commission 3 F0,797-33 Page 2 analysis and reconfiguring key documentation and administrative controls. An independent assessment was also recommended to obtain impartial confirmation of the Problem Report corrective actions.

The Independent Assessment was set up to obtain a critical / conservative review of CR3 Appendix R documentation and controls including:

Safe Shutdown Systems and Performance Goals Safe Shutdown / Alternative Shutdown Circuits Separation Analysis Appendix R compliance Manual Actions Procedures and Repairs Results: The Independent Assessment was a 10 man-week intensive review of the CR3 post-fire safe shutdown analysis. The assessment confirmed the need for a comprehensive re-evaluation of the post-fire safe shutdown analysis and the need to re-configure key Appendix R documentation and programmatic controls. Specific results from the assessment were provided in five categories summarized below:

Requirement /ssues (total of 4) - plant features or documentation that deviate from the requirem'.snts of Appendix R or othe NRC regulatory guidance. It was determined that three of those issues identified concerns with the remote shutdown capability and were reported in LER 96-022. All four have been placed in the Corrective Action System for identification, tracking, and corrective action. They are also covered under the CR3 restart program.

Commitment /ssues (total of 1) - plant features, analysis, or documentation that are inconsistent with NRC commitments. This issue was placed in the Corrective Action System and is included in the CR3 restart program.

/ndeterminate Comp //ance /ssues (total of 11) - unavailability of analysis documentation, unanalyzed conditions that could result in a deviation, and professional differences of opinion on interpretations of Appendix R or other NRC regulatory guidance documents. These issues have been placed in the Corrective Action System and are covered under the CR3 restart program. One Indeterminate

, issue concerning IN 92-18 " POTENTIAL FOR LOSS OF REMOTE SHUTDOWN l CAPABILITY DURING A CONTROL ROOM FIRE" is not included in the restart program. FPC's resolution program for this issue will be the subject of a separate

. NRC submittalin August 1997.

U. S. Nuclear Regulatory Commission 1 3F0797-33 l Page 3 l

Enhancement /ssues (total of 14) - plant feature or documentation that is not critical to regulatory compliance but hinders implementation and long-term ,

maintenance of the Appendix R program. These issues have been reviewed and are '

included in the Precursor System. Some of these issues are covered under the CR3 restart program, some are not applicable, and the remainder will be completed post-restart. ,

i Draft Phase // Re-analysis Observations (total of 12) - The Draft Phase ll Re-analysis is support documentation for the CR3 Thermo-Lag resolution program. The assesant made observations recommending additional calculations, clarifications, or information to include in this re-analysis. These Observations will be included, as appropriate, in the Thermo-Lag resolution program.

A tabulation of the Requirement, Commitment, indeterminate, and Enhancement issues is provided in Attachment 2. '

It is further noted that the FPC CR3 Quality Assurance (QA) Department has recently completed an audit which included the Fire Protection Program and implementing procedures including the performance of Fire Area inspections. The Audit also addressed the Independent Assessment and the current resolution status. Precursor Card 97-3386 was issued by QA to document recommendations regarding the Independent Assessment. These recommendations are currently under review by management and will be accommodated during the resolution of the Independent Assessment " issues".

Attachment 1 is a listing of commitments made in this letter. If you have any questions regarding this letter, please contact D. F. Kunsemiller Manager, Nuclear Licensing at 352-563-4566.

Sincerely,

, , km o n J. Holden Director Nuclear Engineering & Projects JJH/ jnb Attachment xc: Regional Administrator, Region il Senior Resident inspector NRR Project Manager l

t

U. S. Nuclear Regulatory Commission 3F0797-33 Page 4 Attachment 1 List of Regulatory Commitments The following table identifies those actions committed to by Florida Power Corporation in this document. Any other actions discussed in the submittal represents intended or planned actions by Florida Power Corporation. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Manager, Nuclear Licensing of any questions regarding this document or any associated regulatory commitments.

ID NUMBER COMMITMENT COMMITMENT DATE FPC will resolve the Regulatory, Prior to Restart Commitment, indeterminate, and and Post-Restart 3F0797-33-1 Enhancement issues identified in the as indicated in

" Independent Assessment of FPC's Attachment 2 to Appendix R Safe Shutdown Analysis" 3F0797-33 using current plant corrective action system and restart resolution processes, as indicated in Attachment 2 to 3F0797-33.

3F0797-33-2 FPC will incorporate the Observations of the " Independent Assessment of Post-Restart FPC's Appendix R Safe Shutdown Analysis", as appropriate, in the Thermo-Lag resolution program.

U. S. Nuclear Regulatory Comrrdss'on Attachment 2 Page 1 of 4 3F2797 33

SUMMARY

STATUS AND TRACKING OF INDEPENDENT ASSESSMENT ISSUES ,

Independent Precursor Card No. Restart

• " Assessment issue Description issue Status and/or Problem Resolution Number Report No. Task Classification Ememency Diesel Generator (EDG) Govemor A design review of the Remote Shutdown Capability is on eqwenst Control A portion of the EDG circuits were not going. This and other discovered isolation problems will be PC 96-5253 D-11D 3.1.1 lssw completed prior to restart isolated from the control room Information Notice 85-09 Assessment: Single Re Wh @ily b m s scoM Mah Ws d M N M 252 MD i s could use fuses and loss of power to shutdown equipment.

Circuit Seoaration in Remote Shutdown Panet Room (RSP): A postulated fire in the remote A design review of the Remote Shutdown Capability is on

• • • shutdown panel could disab;e control power fuses going. This and other discovered isolation probiems will be PC 96-5252 D-11D 3.1.3 issw and cause loss of control room control of shutdown completed prior to restart equipment.

Ememency Lichtina (EL) in Three Areas: 1) EDG monitoring does not require 84mur EL 2) MSIV 3'1'4 eq Ins g ncy ng r mn n m ms aM M M wahn b on PC 97-0497 ash O Issue EDGs 2) verificaton of MSIV closure, and 3) the RSP and both have EL in addition. EL will be verified & 44 and OP-19A manual operation of atmospheric dump valves during examination of post-fire safe shutdown analysis.

The one example cited, letdown isolation, is on the RSP and Emeraency Liahtino in Remote Shutdown has EL However procedural reliance on EL will be re-verified Commitment Procedure: Reliance on EL in AP-990 " Shutdown D-11 Subtasks 43 exannahon of posMm sak SWown anaW and N M97 issue From Outside Control Room" inconsistent with & 44 and OP-19A verification of procedure AP-990 " Shutdown From Outside Control Room".

Fuse / Breaker Coordination; No breaker / fuse a s han WaW WMm May aM Indeterminate coordination study for reley and instrument power ns p wer ination s M mhed pN D-11 M ng e caWahns ahss coodeon. N 97 6 issue and three 480V loads found that were not Subtask 37 Better documentation of breaker coordinaten is being nM addressed.

The original analysis is being re-venfied during the S uri us Oeeration: Inadequate documentattun of Subtasks 40 & 41 e e pg 3.3.2 iire induced spurious signal ana!ysts was available g PC 97-04% (Documentation to conduct proper associated circuits assessment Enhancements are resM Post-Restart)

Hiah/ Low Pressure Boundarv Valves: An analysis This assessment recommendaten confirmed an already Indeterminate of a three phase hot short of power cables on PC 97-4675 and PR-0401 3.3.3 identified condition and that the work in progress for DHV4 issue Decay Heat pump suction valves had not been CAP ftems 10 through 14 and DHV-4 was appropriate.

performed

U. S. Nuclear Regulatory Commission Attachment 2 Page 2 cf 4 3F0707-33

SUMMARY

STATUS AND TRACKING OF INDEPENDENT ASSESSMENT ISSUES ,

, independent Precursor Card No. Restart .

Section issue Description issue Status and/or Problem Resolution Assessment

• Report No. Task Classification Sinale Sounous Ooeratign: The single spunous TM m M gh N M @ W m W indetenmnate opwats mNologW W N Study b clan 5 cations of GL 81-12 and GL 86-10 and is in agreement PC 97-0496 3.3.4 Subtasks 40 & 41 issue insumcientty consuative to meet NRC staff g exceptions.

• P'" *

• 9 #*

Extema! Hot Short Appendix R analysis did not consWed W N Ny aM mM W K Hot D-11 Indetermmate consider single fire-induced externaf hot shorts in 1 3.33 s s wen uctors @ a caw um consh n. MN issue all areas which is considered non-conservative and Subtask 34 all areas and shorts between cables were considered n not n agreement with GL 86-10.

congested cable areas.

IN 92-18 Asse==nmni Fire inouced hot shorts in FPC is currently reevaluating this IN 92-18 issue and a

• '"*"* submittal to NRC is planned for August '97 that will describe PC 97-3963 Post-Restart 3.36 DC valves could bypass limit switches as tssue desenbed in IN 92-18 our resolution program.

Reactor Ooeration in Solid Mode: Neither the Fire M R deem Study nor AP990 provide guidance on 1) reactor however, during the evaluation of the post-fire safe shutdown PC 97-0496 (related D-11 33.7 coolant system pressure control to assure sub-analysis, consolidation of this guidance into the Fire study Subtasks 30 & 31 issue PC 96-4534) cooling margin or 2) methods for operating solid or AP 990 M N MaM with pressurizer level.

Decav Heat System Oceration: Process monitoring.The Fire Study describes the use of RCS hot and cold leg for the Decay Heat removal system is not listed m indetenmnate wide range instrumentation and decay heat system inlet and 338 the Fire Study, hence the thoroughness of the PC 97-0496 issue outlet temperature is on the RSP as desenbed in the Design Subtasks 32 & 33 analysis of the decay beat removal funchon is in Basis Document.

question.

Borated Water Storace Tank (BWST) Levet: The F S id@ h BW n a @ mh

• """*'* Y ** '" *** ~

33.9 a emah wwweit and the BWST levelis on the RSP as devribed PC 97-0496 Subtasks 32 & 33

'*** "9 " """*" sa in the Design Basis Document.

shutdown Source Ranoe Monitonna: Following NRC review etennmate a ms nn s u ce mnge NW msMon oms use Re SWy &

3'3 10 PC 964990 D-11E Issue monitoring for a reactor building fire, the Fire Study revised as appropriate.

should be updated as appropriate.

72-Hour Cold Shutdown Caoability: The calculabon for a!!emative shutdown to cold shutdown The calculation for cooling down to 200 degrees is being conditions in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> does not include a delay revised. The new calculation willinclude the appropriate post PC M96 Indetennmate 3.3~11 PC 97-1522 Issue time prior to initiating the cooldown or Decay Heat fire delay and time to cool from 280 to 200 degrees on decay Subtasks 30 & 31 PC 97-2360 removal system operation from 280 to 200 heat removal.

degrees.

U. S. Nuclear Regulatory Commission Attachment 2 Page 3 ef 4 3F0717-33

SUMMARY

STATUS AND TRACKING OF INDEPENDENT ASSESSMENT ISSUES -

Int',ependent Precursor Card No. Restart .

Aasessment issue Description issue Status and/or Problem Resolution Numbe'

• Report No. Task Classification EDG Load Calculations: Decay Heat Removat This is not a discrepancy and is in agreement unth the actual 34.1 Pump is included in the load for EDG *B" and not *

, loading of the EDGs in the event of an Appendix R fire.

EDG*A Accendix R Documentation- A consolidation of 3.4.2 Appendix R documentation and review for ema a a WegW Wer mstad isse M PC 97-04%

'*** consistency and data integrity are recommended.

ete a e nsWah d M compW post E n - .s am restart Post-Restart)

Conficuration Control of Accendix R Documentatro_n; in order to better assure a s u a Ms a sWestad s ue inee P be re sed to reference appropriate Appendix R e'ectncal design criteria Cable EGE86 for DGB: Based on the similarity of Enhancement functions between this cable and EGE85 it is EGE86 is a spare cable and should not be listed in the Fire 344 PC 9N96 -

Issue recommended that tNs cable also be listed in the Study.

Fire Study

" " " TNs be d s during Appendix R dW 34.5 9 N 92-82 s Id be ad t PC 97-0496 Post-Restart Appendix R Topical DBD IN 92-82 Assessment: It is recommended that ,y gg ,,, g g ,

3.4.6 Thermo-Lag be used to ehminate intervenmg PC 97-0496 -

combustibles in the CR3 ThermM.ag resolution program.

combustibles per IN 92-82.

Exemotion Bases: It is recommended that previous Enhancement 3.4.7 assure that exemptmbases remain intact during deletions and modifications have been considered in our PC 97-0496 -

and following the implementation of new planned activites.

compliance strategies Surveillance of Attemative Shutdown Eouicment:

The review determened that certain instrumentation , ,, g g Enhancement was not in Swvemance Rocedwe SW338 aM 3.4.8 SP161 A and SP-398A. Administrative controls, includog PC 97-0496 Post-Restart issue recommended that a review of all Altemative Han pWmA M elidated Mmth Shutdown equipment be conducted and SP-338 be revised accordingty.

4

. ...r

- U. S. Nuclear Reguletory Commission Attachment 2 Page 4 of 4 :

3F0797-33

SUMMARY

STATUS AND TRACKING OF INDEPENDENT ASSESSMENT ISSUES .

Independent Precursor Card No. . Restart ..

Assessment issue Description issue Status , and/or Problem Resolution Classification Report No. Task g-g_, gg Teneimes: A twnehne study was recommended t if it is determoed during revision of AP 990 that a tune line is 3.4.9 . g, assure that sufficient manpower as available to PC 97-0496 - OP-19A perform the shutdown funcbons Multele High impedance Faults: A more Safe shutdown power sources are correctly evaluated for the 3'4.10 effects of mulbpie high impedance faults, and the existing PC 97-0496 -

Issue impedance faults is recommended to be consistent calculabon is consestent with GL B6-10 and industry pracha with GL 86-10 EDG Local Control: Local EDG start and stop Enhancement capabihty is rewn.nended in the event that a - This potential enhancement will be evaluated during our 3 4'11 PC 97-0496 issue control room fire causes the air start soler'oid valve review of post-fire safe shutdown w,n, v.en;s. Subtask 32 to remain open.

Fuse / Breaker Cuvid;..nm. It is recommended Breaker coordmabon calculations were conduded in 3.4.12 that additional information be included in several PC 97-0496 -

" " {'" breaker coordmation calculations wMh our nuclear engineenng procedures.

D uiiv6.= instrun=ni i;vn. It is recommended Enhancement that diagnostic instrumentation for safe and Diagnostic instrumentation is identified and maintamed in the 3A13 9 ~

issue' afternative shutdown equipment be idenbfied in the DBD for the Remote Shutdown Panet.

Fire Study.

Instrument Sensino Lines: It is icwn... coded that A drawmg review indicated that appropriate separation exists "C"

3.4.14 analysis for separation of redundant instruments - for the sensing lines. However this will be further evaluated PC 97-0496 I'8"'

be revised toinclude sensmg lines. during the post 4re safe shutdown analysas review.

i

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