View source for JAFP-92-0578, Clarifies 920713 Response to Violations & Deviations Noted in Insp Rept 50-333/92-81 on 920413-0501.Corrective Actions: Safety Function Sheet for Control & Relay Room Ventilation & Cooling Approved & QA Upgrade Initiated

{{Adams
| number = ML20099F219
| issue date = 08/04/1992
| title = Clarifies 920713 Response to Violations & Deviations Noted in Insp Rept 50-333/92-81 on 920413-0501.Corrective Actions: Safety Function Sheet for Control & Relay Room Ventilation & Cooling Approved & QA Upgrade Initiated
| author name = Salmon H
| author affiliation = POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000333
| license number = 
| contact person = 
| document report number = JAFP-92-0578, JAFP-92-578, NUDOCS 9208120184
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 11
}}

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2   *          [J;mes A. FitzPatrick Nucient Powst Plin?
P.o. Box 41 Lycoming. New York 13093 315 342 3840
        #W NewYorkPower W Authority                                                                                                        ""1"d*j"; "#
August-4, 1992 JAFP-92-0578 Director, Office of Enforcement U.S. Nuclear Regulatory Commission ATTN:. Document Control Desk Washington, D.C. 20555

==SUBJECT:==
James A. FitzPatrick Nuclear Power Plant Docaat No. 50-333 Revision to Notice of violation and Nptice__qf p_tiviation Inspection 50-133/92-81

==REFERENCES:==
: 1.                NYPA letter, H.P. Salmon to the NRC, dated July 13, 1992, (JAPP-92-0527), replies to a Notice of Violation and a Notice of Deviation, Inspection 50-333/92-81."
: 2.                 NRC letter, M.W. dodges to H.P. Salmon, dated June 11, 1992, "NRC Inspection Report 50-333/92-81."

==Dear Sir.:==

This letter provides a clarification to Reference 1, the Authority's written response to NRC Inspection Report 50-333/92-81 (Reference,2). The inspection report contained one Notice of Violation and one Notice of Deviation 16 .ified during the NRC's Safety System Functional Inspection of the Emergency Service Water System conducted April 13 through May 1, 1992 at the James A. FitzPatrick Nuclear Power Plant.
This letter supersedes Reference 1 with changes noted through the use of revision bars._ The Authority has determined additional information-is required to-correctly document the corrective steps that have been taken.in response to the violation, and the Edate when full compliance will be achiev'd.
If you have any questions, please contact Mr. M. Colomb.
          ;Very truly yours,
             . //4 ~ /
HARRY P. SALMON, m
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: JAMES:A..FITZPATRICK NUCLEAR' POWER PLANT:
: ATTACHMENT IlTOEJAFP-92-XXXX
_page 1 of-1'
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                                                             . NOTICE OF VIOLATLQM During7an NRC Emergency' Service Water (ESW) Safety _ System of::                       Functional Inspe' tion (SSFI)Lconducted April 13 through May-1,
                                ~1992 the-:following~ violation of NRC requirements was_identifiedi 110CFR50.59 L , a)? allows ~ the f hviders of- a license: to !make changes to 1the. facility assdescribed in'the safety analysis report (SAR).
:without~ prior: Commission. approval unless it involves an funreviewedlsafetyfquestion.         10CFR50.59 (b)' requires,.in part,-           '
,~              -
that:the: records;of ? a change to the< facility be1 maintained byLthe-y
                                ' licensee and'zmustlinclude ;a written ' safety evaluation-which~
Tprovides;the> basis for-the' determination'that-theHchange-does not involveJan1unreviewed safetyEquestion.-
:Contrarypto the above, the: safety evaluation JAF-SE-90-067, which (downgraded (the; control room chiller condensersLfrom.; safety related~to_non-safety.relatedLdid;not provide an adequate basis iforfthe determination thatLthe change;does not_ involve an unreviewod safety question. :The evaluation.didLnot: include;an
                                . evaluation;of  f flooding _andivas performed:basedLinlpart.on_a 1970 icontrol? room ~ heat generation analysis, which.did not; account for         G1
                                . changes madento theEcontrol. room since 1970.         An updated control-s room: heat 9 generation ratelanalysis indicated 1that the control-           q
                                ..roomitemperature{could exceed the; maximum design. temperature.1
                                -This11sta' Severity-Level IV Violation.
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                                  'N;w York P; war Auth3rity 0AMES A. FITZPATRICK NUCLEAR POWER PLANT ATTACHMENT II TO JAFP-92-XXXX page 1 of 4 EESPONSE T_Q_tiOTICE OF VIOLAT79H VIOLATION
         'The control room chiller condenser reclassification from safety related to non-safety related did not provide an adequate basis
         -for the determination that the change does not involve an unreviewed safety question.
ADMISSION - OR DENJM_97 THE ALI&QEQ VIOLATION
         .The Authority agrees with the violation, however the following clarification needs to considered.
Safety. Evaluation.JAF-SE-90-067, Revision 0, dated June 16, 1990 is not:the documentation basis that reclassified the control room and' relay: room chiller condensers. JAF-SE-90-067 consolidated existing emergency service water design basis information into a single l document. The chiller condensers were reclassified in accordance with; Modification Control Manual procedure MCM-6A,
          " Component. Classification and System Safety Function Control (JAF)"c May 22,-1990. .See-Attachment V for a description of MCM-   -
6A.
T_HE REASONS FOR THE VIOLATION Flooding Concern:
Personnel! error was the primary cause for.the inappropriate downgrade of a safety related pressure boundary. MCM 6A does not opecifically require interfacing systems be-identified and1 evaluated. -This procedural weakness contributed to the-violation.
MCM-6A provides a formaltprocess and step by step instructions necessary to determine:the correct QA classification of structures andJcomponents. . The procedure requires an evaluation
        ' to-determine if the component' functions as a pressure boundary Efor'any-. portion pf-the system being used to accomplish a safety 1 function.
The' Control Room and Relay Room Ventilation and Cooling. System is safety related.ard has a safety function to cool the rooms with emergencyEservice water (ESW) supplying the air. handling units
         '(ABUs) J The reclassification:of the control room and relay room
                                                              ~
         ' chiller' condensers' failed to conclude that~the service water piping supplying the chiller condensers provides a safety related pressure boundary.for'the ESW system supplying the AHUs. Had the
         ~ safety related pressure boundary been recognized, the reclassification of tdue chiller condensers would have been
         . limited to system function (heat removal capability) only and the pressure boundary would have remained QA Category I.
l

m                                   ,
N;w Y0rk P;w r Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT-ATTA'CHMENT II'TO JAFP-92-XXXX page 2 of 4 BE.SPO@E TO NOTICE OF VIOL &TLOj{
Control' Room Heat Loads:
The Authority's failure-to esta"11on and usintain as-built control room heat loads contributed to using 1970 design dat' to support calculation JAV-90-C50. At the time of the reclassification, May 1990, the control room design heat loads were the best available information to perform the heat load analysis.in calculation JAF-90-058. The. Authority recognized the need to'obtain as-built centrol' room heat loads as part of the Generic Letter 89-13 program design-review for the FitzPatrick Plant. Safety Evaluation JAF-90-067 Revision 0, dated June 16, 1990 acknowledged calculation JAF-90-058 assumed design control room heat loads and using' engineering judgement concluded the following:-
During August 1988, when lake water temperature reached 80*F, the actual. performance of the control room-chillers with normal operating loads (normal heat loads exceeded acciJent heat loads), was acceptable indicating the as-built loadt were within the capacity of the-chillers. As the design loads used to size the chillers are the same as were used to size the AHUs the ability of the Air Handling Units to naintain room temperatures with as-built heat loads was iconfirmad.
JThis qualitative analysis was considered appropriate based on the
               ;information available at the time, however to validate the above e             conclusions the-Authority' contacted Stone & Webster.in August 1990 to provide a. comprehensive studyLof as-built heat loads for the. control room. This analysis was completed during the'ESW 4     s Safety System Functional Inspection.
N             IHE ' CORRECTIVE STEPS THAT HAVE BEEN TAKEE._AND THE RESUIaIS ACHIEVED-Flooding Concern:
b           MCM-6A-Safety Function Sheet for control and relay room
            -  -ventilation.and cooling (system 70) has been-approved by the Plant'Oparating Review committee to identify the service water / emergency service water pressure boundary supplying the chiller condensers and the chiller room air handling units as a safety related, function.      The revision being implemented will
               . ensure the pressure boundary safety function is evaluated during
                ' future' system 70-component classifr :lons.
               ~The Authority has initiated a QA classification upgrade for the service water / emergency service water pressure boundary supplying
               .the controlHroom and relay room chiller condensers, the chiller room air handling units and associated valves. In the interim a preliminary flood analysis has been performed assuming a failure of the-pressure boundary and the results indicate that no safety related; equipment would be affected by the failure.
      .o.

N;w York P;w r Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT ATTACHMENT II TO JAFP-92-XXXX page 3 of 4 RESPONSB TO NOTICE OF VLOlBTJ_OE control Room Heat Loads:
The as-built control room heat loade L-ve been determined through field testing and analysis. Maximum steady state control room temperatures have been established, based oli percent lighting energized,.and 82*F lake water supplying the control room AHUs.
The results of this analysis concluded that with all heat loads, including 100% of room lighting energized, the control room temperature would reach 102*F. This is 2*F greater than the control room temperature referenced in FSAR Section 9.9.3.11. To limit the control room temperature to less than 100 F, the following administrative controls have been established:
               *      ?.ppro41mately 40% of control room lighting is secured.
(The secured lights are not required to provide adequate lighting in the control room)
                *. Plant operating procedures have been revised to secure or verify secure these lights when EFli is supplying the control room AHUs.
THE CORRECTIVE STEPS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS i
71oiding concern:
The following short tern corrective actions will be taken to ensure interfacing systems are identified and evaluated during component classifications.
                *-    MCM-6A will be reviced to ensure the appropriate personnel (System Engineers and/or Nuclear Engineering Department) are assigned to the review.                         [Due date - 10/31/92)
* MCM-6A will be revised to provide additional guidance to ensure interfacing safety related systems are identified and evaluated during component classifications.                                           [Due date -
10/31/92]
                 *-    Training on the revised procedure will be provided.                                           [Due date - 12/31/92) l

N;w YGrk P0 wor Authority L                                                    JAMES.A. FITZPATRICK NUCLEAR POWER PLANT ATTACHMENT II TO JAFP-92-XXXX page 4 of 4 BraPONSE TO NOTICPuCF VIOLATIOE Control Room Heat Loads:
As part of the Design Basis Document (DBD) program a heating ventilation and air conditioning (HVAC) DBD is scheduled to be developed starting September 1992. This DBD will document heat loads and HVAC capabilities in the various buildings and the control / relay rooms. This document will ensure the effect of future modification heat loads will be evaluated.
TjiE DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED
                                      . Full compliance will be achieved when the service water / emergency service water pressure boundary supplying the control room and
: relay room chiller condensers and the chiller room air handling units are. reclassified as QA Category I. Included in the upgrade                               j is a revision to the Master Equipment List in accordance with procedure MCM-6A.
                                                                ~
An engineering evaluation verifying the acceptability of the installed components to meet QA Category I requirements is being performed in accordance with Engineering Design Procedure EDP-31, " Component QA Classification Upgrade Evaluation-Procedure."
                                      -The Authority will complete the upgrade prior to startup from the 1992 refuel outage.

u Ccw York pow 0r Authority l                             JAMES A. FITZPATRICK NUCLEAR POWER PLANT e                              ATTACHMENT III TO JAFP-92-XXXX page 1 of 1 NOTICE OF DEVIATION During an NRC Emergency Service Water (ESW) Safety System Functional Inspection (SSFI) conducted April 13 through May 1, 1992 tlut following deviation of the FitzPatrich Final Safety Analysis-Report (FSAR) was identified.
FSAR Table 9.7-1, sheet 1 of 3, Emergency Service Water Equipment
        - Flow Rates and Operating Modos, states that the minimum required flow to each crescent area unit cooler is 24 GPM.
Contary to the above, ductag performance of procedure ST-8Q, emergency service water flow rates to individual crescent area unit coolers were not adjusted to greater than the minimum value of 24-gallons per. minute that is specified by Table 9.7-1 of the Final Safety Analysis Report. For example, on July 28, 1991, the emergency service water flow rate to west crescent area unit cooler 66UC-22G was left at 21 gallons per minute and the emergency service water flow rate to east crescent area unit icooler 66UC-22F was left at 22.8 gallons per minute, on September 10, 1991.-
__________-_m._______m___ __

          -                                    New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT ATTACHMENT IV TO JAFP-92-XXXX page 1 of 2 RESPONSE TO NOTLCE OF DEVIATION DEVIATION During performance testing of crescent area unit coolers, flow rates.wcre not adjusted to greater than the FSAR Table 9.7-1 specified 24 gallons per minute.
THE REASONS FOR THE DEVIATION The original design for crescent area cooling consisted of five unit coolers operating at a design ESW flow rate. In 1988 the Authority became aware of silting problems in the crescent area unit coolers and its effect on reducing ESW flows through the coolers. Recognizing ESW flow through individual coolers will vary over time the Authority performed analyses that defined operability requirements for crescent area coolera based on heat transfer capability.
Technical Specification 3.11.B, 4.11.B, and Technical Specification Interpretation No. 19 define the operability requirements for the crescent area coolers. Operability of the crescent area unit coolers is demonstrated through thermal performance testing in accordance with Surveillance Test ST-19C.
Operability of an-individual cooler is based on its ability to camove heat (UA-> 12,500 E2U/ hour.*F). Operability of the crescent area cooler train is based on its total heat removal            _
capability, Er_st Crescent > 672,750 BTU /hr and West Crescent >
588,655 BTU /hr. Only four out of the five unit. coolers need to be effective in removing heat for a train to be considered operable. This allows monitoring of cooler performance to effectively schedule removal from service one cooler in each
                   . train for cleaning.
Operability of the coolers is independent of the ST-8Q ESW flow rate acceptance criteria. Crescent cooler ESW' flow rates are measured and adjusted during surveillance test ST-8Q to monitor and maintain the ESW system hydraulic flow balance and to reduce / prevent silt build up. The ST-8Q acceptance criteri1 for the crescent area unit coolers is 120 gallons per minute per train which ensures ESW system flows are properly balanced.
While attempts were made to establish 24 gallons per minute to eachjcooler tha Authority recognized the increased time and exposure to achieve the design flow rate did not justify the incremental improvement in cooler heat remo'/al performance.
Operability of crescent area unit coolers was verified during 1991 with biweekly thermal performance testing.
I

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New York Power Authority JAMES A.-FITZPATRICK NUCLEAR POWER PLANT ATTACHMENT-IV TO JAFP-92-XXXX page 2 of 2 RESPONSE TO NOTI.Q_E OF DEVIATION THE CORRECTIVE STEPS THAT HAVE BEEN TAKEN TO AVQIR_fURTHER DEVIATIONS The following corrective actions will be taken to ensure current plant configuration and procedures are in agreement with the FSAR.
e     FSAR Section 9.7-1 will be revised to include both the design specifications and the operability requirements of y                       the crescent area unit coolers.     [Due date - 1993 FSAR I                       Update]
* The Authority has established a Nuclear Generation Business Plan Objective to review its internal procedures used to maintain and update the FSAR. Included in this review will be an assessment of the FSAE level of detail based on recommendations in-Reg Guide 1.70. .[Due date - 9/30/92)
* The Authority has established a Nuclear Generation Business Plan Objective to enhance the process for review and revision of the FSAR to reflect current plant configuration anw Design Basis Documents.    [Due date - 12/30/92)
* The Authority will formally document the FSAR deviation in accordance with Nuclear Generation Procedure NGP-38.

4 New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT M%CHMENT V TO JAFP-92-XXXX page 1 of 1 MCM-6A COMPONENT CLASSIFICATION PROCEDUR3 MCM-6A, " Component Classification and System Safety Function Control (JAF)", ensures the appropriate quality classification is assigned to systems, structures and components, rather than to document design changes to the faci;1ty. A reclassification does not establish new or revised design functions rather it reviews existing safety related system functions and evaluates the effect a component or structure failure would have on preventing performance of a safety related function.
Generic Letter 83-28, " Required Action Based on Generic                                                   !
Implications of Salem ATWS Events", required that FitzPatrick review and update its equipment safety classifications. JAF-SE-88-052 evaluated the methodology for this project (Master                                                  :   q Equipment List or MEL) and became a basis for FSAR Section 12.A,
                                                      " Safety Related Functional Analysis" and the development of a                                              y long term component classification control procedure, Pi'M- 6 A .
MCM-6A, provides the guidance and documentation to perform the following:
* Determine the correct QA classification of systems, structures and components.
t
* Maintain and-control System Safety Function Sheets for applicable plant systems, structures and components.                                                    _
e     Evaluate the effect of changing the System Safety Function Sheets or QA Classifications.
The procedure identifies the safety related functions for each system at FitzPatrick.          System Safety Function Sheets have been prepared based on the MEL program effort including the Safety Related Functional Analysis document in FSAR Section 12.A. Those sheets identify system safety related and non-safety related functions.
                                                      'A component classification is established by reviewing the System Safety Function Sheets and by answering specific questions for a given component type (mechanical, electrical, instrumentation, or structural). The questions are designed to determine if the component or structure supports a system safety function- Any affirmative response to the safety related questions requires the component or structure be classified as safet) related.
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