View source for JPN-87-038, Forwards Summary of Mods Completed During Reload 7 Refueling Outage.Dcrdr Control Room Enhancement Program Mods Not Completed.Program Rescheduled as Stated in

{{Adams
| number = ML20235U103
| issue date = 07/17/1987
| title = Forwards Summary of Mods Completed During Reload 7 Refueling Outage.Dcrdr Control Room Enhancement Program Mods Not Completed.Program Rescheduled as Stated in
| author name = Brons J
| author affiliation = POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
| addressee name = 
| addressee affiliation = NRC OFFICE OF ADMINISTRATION &amp; RESOURCES MANAGEMENT (ARM)
| docket = 05000333
| license number = 
| contact person = 
| document report number = JPN-87-038, JPN-87-38, NUDOCS 8707220356
| title reference date = 04-20-1987
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC
| page count = 9
}}

=Text=
{{#Wiki_filter:. , _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ .
i' 123 Mmn Street Wh!te P;emz. Nc.w Wuk 10601 M4 6810240
              #> NewWrkPbwer                                                                                           - c. ume.
W' Authority                                                                                             it: ="1:::l"*"'
July 17 , 1987 JPN-87-038 U. S. Nuclear Regulatory Commission Attn. Document Control Desk Washington, D. C. 20555

==Subject:==
James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Summary of Work                                                                                    1 Completed Durina Reload 7 Refuelina Outaae

==References:==
: 1. NYPA letter, J. C. Brons to the NRC, dated April 20, 1987 (JPN-87-022) regarding revised implementation schedule for Detailed Control Room Design Review (DCRDR).

==Dear Sir:==

The Authority recently completed the FitzPatrick Nuclear Power Plant Reload 7 refueling outage. During the outage, the Authority completed, or made significant progess on, a number of plant modifications. This includes modifications to satisfy Authority commitments to the NRC as well as modifications to improve plant performance.
All of the work scheduled for the Reload 7 refueling outage was completed on or ahead of schedule, except for one portion of the Detailed Control Room Design Review (DCRDR).
oh 8707220356 870717 PDR  ADOCK 05000333'            _
10 4
3
           -P                     pyg                                                                                              l 1l

This portion,.the Cer.trc_ Room enhancement program, was rescheduled in Reference 1. No other DCRDR work was rescheduled.
A summary of the more significant work completed during  .
j the Reload 7 refueling outage is attached.
Should you or your staff have any questions, please contact Mr. J. A. Gray, Jr. of my staff.
Very truly yours, d         a Tonn C. Brons                   ,
                                                        'xecutive Vice President fuclear Generation cc: U. S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406 Office of the Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 136 Lycoming, New York 13093 Mr. H. Abelson, Project Manager Project Directorate I-1 Division of Reactor Projects - I/II U. S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20014

n   ''
Attachment No. 1 to JPN-87-038 NEW YORK POWER-AUTHORITY James A..FitzPatrick Nuclear Power Plant Progress During Reload 7 Refueling Outage at FitsPatrick
          . Introduction The following paragraphs summarize progress on modifications at the Authority's James A. FitzPatrick Nuclear Power Plant during the Reload 7 refueling outage.
Except for the removal of the recirculation pump bypass valves and low-pressure turbine replacement, all work is associated with NRC commitments.
: 1. SPDS/ EPIC The FitzPatrick Safety Parameter Display System (SPDS)-
has been completed earlier than the Authority originally committed to the NRC. Originally scheduled for completion at the end of Reload 8/ Cycle 9-(Reference 13), the SPDS was                                    4 fully functional on May 22, 1987 - almost a full refueling cycle ahead of schedule. This accelerated schedule was possible because modifications associated with the ATWS rule (10 CFR 60.62) were delayed until the end of the Reload 8/ Cycle 9 refueling outage (Reference 12).
SPDS is one part of the Authority's new Emergency and Plant Information Computer (EPIC). Non-SPDS portions of the EPIC system will be gradually phased-in during this operating cycle to avoid a negative impact on daily operations. When fully integrated, EPIC will be more than just a replacement for the aging process computer or a display of critical safety functions. The EPIC advanced capabilities centralize much of the information required to operate the plant and should improve the ability of the plant staff to perform routine functions.
: 2. Regulatory Guide 1.97 During the Reload 7 refueling outage, instruments to monitor the following plant parameters were either upgraded or installed to meet the guidance of Regulatory Guide 1.97:

              - Suppression Pool Water Temperature
              - Drywell Atmosphere Temperature
              - Suppression Chamber Pressure
              - Radiation Level in Circulating Primary Coolant
              - Containment Effluent Radioactivity, Noble Gases
              - Main Steamline Isolation Valve Leakage Control        ;
System Pressure
              - RHR (Residual Heat Removal) Heat Exchanger Outlet Temperature                                  ;
              - RHR Service Water System Temperature                    1
              - Status of Standby Power and Other Energy              f Sources Important to Safety                          !
              - Turbine Building Vent, Inlet and Exhaust               i' Damper Position Indication
              - Radwaste Building Vent, Inlet and Exhaust               .
Damper Position Indication                           l
              - Turbine and Radwaste Buildings Ventilation            ]
Exhaust Flow Rates                                 ,
              - Cooling Water Temperature to ESF [ Engineered           ,
Safety Feature] Components                          i All remaining modifications necessary to comply with the regulatory guide will be completed during the next refueling outage (Reload 8/ Cycle 9). The completion of this final portion of Regulatory Guide 1.97 is the only Supplement 1 to NUREG-0737 issue that has not been fully      j implemented at FitzPatrick.                                    1
: 3. Appendix R Fire Protection Modifications                  !
The final modifications necessary to complete the Alternate Shutdown Capability System (ASCS) were installed during the Reload 7 refueling outage. Shortly before the       ,
ASCS wac scheduled to become operational in March 1985, the    l Authority reported (Reference 14) that redundant fuses were    l not part of the FitzPatrick ASCS.                              l The need for redundant fuses was identified in IE Information Notice 85-09 (Reference 15). This notice described a scenario in which a fire in the Control Room       l could result in blown fuses in equipment in the plant's alternate shutdown system. If redundant fuses were not installed, operability of alternate shutdown equipment could   j not be assured without first identifying and replacing any blown fuses.
1 In order to assure operability of the ASCS, the Authority developed a modification to add a second set of I
                                                                                                          )

fuses to preclude the type of event described in the notice.
           ~ The time necessary to engineer and install redundant fuses delayed completion of the ASCS by one operating cycle.
Compensatory measures were in effect in the interim.
: 4. Detailed Control Room Design Review The Authority rescheduled completion of one portion of the Detailed Control Room Design Review Human Engineering Discrepancies (DCRDR HED) resolution program due to unanticipated problems associated with modifications to upgrade control room instruments. This program was implemented to comply with Regulatory Guide 1.97 and the SPDS/ EPIC system. No other changes were made to the remaining portions of the DCRDR resolution schedule.
The control room enhancement program, originally scheduled for completion at the end of Reload 7, included seven HED resolution categories: demarcation, labeling, color coding, scale modifications, standards, procedures and miscellaneous. A new schedule was provided to the NRC in                                                                             ,
Reference 1. It divided many of these categories into two                                                                            l or three phases to clarify and simplify the schedule.
None of the 177 HEDs included in the DCRDR program were assigned to HED assessment category I (HEDs associated with documented error). Therefore, this new schedule did not delay the resolution of HEDs considered to be most significant to safety.
The new schedule extended the completion date by as little as one month. Most of the work will be completed in 1987. Some items were delayed longer so they could be efficiently integrated with existing programs.
: 5. Second Level of Undervoltage Protection Installation of a second level of undervoltage protection system was completed this outage. This new protection system assures that safety-related equipment on both 4160 volt (4kV) emergency electrical busses will not be damaged due to sustained operation at low voltage. The function of the new degraded voltage protection is very similar to the existing loss-of-voltage protection.
Technical Specifications changes were submitted in December of 1986 (Reference 11) to incorporate this new system. The changes proposed by Reference 11 were

r__
r.,      .'.                 .
essentially identical to those described in the Authority's {{letter dated|date=October 17, 1977|text=October 17, 1977 letter}} (Reference 19). The NRC subsequently approved and issued the proposed changes as Amendment No. 106 to the operating license (Reference 17).
: 6. Primary Containment Isolation Upgrade The Authority committed to upgrade the Reactor Water Clean-Up, Traveling Incore Probe (TIP) and Recirculation Pump Mini-Purge containment penetrations as the result of a study prepared in response to NUREG-0737 Item II.E.4.2,
                                                      " Containment Isolation Dependability." New isolation valves with diverse actuation signals were added to improve containment isolation dependability.
During the Reload 7 refueling outage, modifications were completed bringing the TIP and Recirculation Pump Mini-Purge systems into compliance with the requirements of Appendix A to 10 CFR 50, General Design Criteria 54, 55, 56 and 57. Problems procuring suitable isolation valves forced the Authority to reschedule the Reactor Water Clean-up System upgrade until the next outage of sufficient duration after December 31, 1987 (Reference 18).
: 7. Improved Long-term operability for ADS To improve the long-term operability of the Automatic Depressurization System (ADS), the ADS accumulators were provided with a second, alternate source of pneumatic power.
This modification upgraded ADS to comply with NUREG-0737,    l Item II.K.3.28, " Qualifications of ADS Accumulators."
A ring hea^ der inside the drywell is used to supply nitrogen (or air when containment integrity is not required) to the ADS accumulators. A second seismically qualified nitrogen supply line, redundant to the existing line, was added. This allows either line to supply nitrogen to the ring header.
Additional work on both pneumatic supply lines' containment isolation valves is required to complete this      ,
modification. This final phase is currently scheduled for    !
the next refueling outage (Reload 8).
: 8. Removal of Recirculation Pump Discharge Bypass Valves The recirculation pump discharge bypass valves and their associated piping were removed from the Recirculation    >
2 I

system during the Reload 7 refueling outage. The removal of these two valves was recommended by the reactor manufacturer, General Electric, to reduce the potential for pipe failure due to intergranular stress corrosion cracking (IGSCC). Technical Specification changes were submitted as Reference 7. The NRC approved these changes and issued them as Amendment No. 108 to the Operating License (Reference 8).
: 9. Low-Pressure Turbine Replacement FitzPatrick's "B" low-pressure turbine rotor was replaced during this refueling outage. Ultrasonic testing conducted during the 3983 outage revee. led crack indications on both the "A" and "B" low-pressure turbines. Replacement of the "A" rotor is scheduled for the next refueling outage.
The replacement rotors are of the integral design which elminates the crack-susceptible wheel-to-shaft keyway.
1
   . 5,6 ',
L
            -Notes and References
: 1. NYPA letter, J. C. Brons to NRC, dated April 20, 1987 (JPN-87-022) regarding revised implementation schedule for-Detailed Control Room Design Review (DCRDR).         I
: 2. NYPA letter, J. P. Bayne to D. B. Vassallo, dated April L                 1,:1985 (JPN-85-024) regarding NUREG-0737 Item l                 II.K.3.28, Qualification of Accumulators on Automatic    l Depressurization System Valves.                          {
l
: l.             3. NYPA; letter, J. P. Bayne to D. B. Vassallo, dated l                 September 4, 1984 (JPN-84-058) regarding qualification   y h                 of ADS accumulators. Responds to NRC July.16, 1984       {
l request for information.
i
: 4. NRC letter, D. B. Vassallo to J. C. Brons, dated July    )
24, 1985 regarding NUREG-0737 Item II.K.3.28. This       i letter concludes that "the requirements of Item II.K.3.28 have been satisfactorily addressed and that qualification of ADS accumulators at FitzPatrick has been verified." Includes a copy of NRC Safety Evaluation dated July 24, 1985.
l              5. NYPA letter, J. C. Brons to D. R. Muller, dated l
December 19, 1986 (JPN-86-061) regarding JPTS-86-017, proposed changes to the technical specifications regarding containment isolation valves. Revises Table 3.7-1.to include new valves.                             l l
: 6. NYPA letter, J. C. Brons to NRC, dated March 13, 1987    $
l (JPN-87-009) regarding revised proposed changes to the technical specifications regarding containment isolation valves (JPTS-86-017). Revises December 19, 1986 submittal to reflect discussions with NRC staff.
: 7. NYPA letter, J. C. Brons to D. R. Muller, dated October  ;
8, 1986 (JPN-86-044) regarding proposed changes to th~
technical specifications regarding the recirculation pump discharge bypass valves (JPTS-86-006).              !
: 8. NRC letter, H. I. Abelson to J. C. Brons dated February 9,  1987 (JAF-87-030) issues Amendment No. 104 to the      i FitzPatrick Technical Spec.4fications regarding the      i removal of recirculation' pump bypass valve removal. I l '
: 9. NYPA letter, J. C. Brons to NRC, dated April 20, 1987 (JPN-87-02 2 ) regarding revised implementation schedule for Detailed Control Room Design Review (DCRDR).
I i
      . O.
c
: 10. NYPA letter, J. C. Brons to NRC, dated March 13, 1987 (J PN-87-009 ) regarding revised proposed' changes to the technical specifications regarding containment isolation valves (JPTS-86-017) .
: 11. NYPA letter, J. C.'Brons to D. R. Muller, dated December 31, 1986 (JPN-86-064) regarding proposed changes to the technical specifications regarding a second level of undervoltage protection.
: 12. NYPA letter, J. C. Brons to D. B. Vassallo, dated June 14, 1985 (JPN-8 5-047 ) regarding Integrated Implementation Schedule of SPDS and~ATWS Rule             ;
Modifications.
: 13. PASNY letter, J. P. Bayne to D. B. Vassallo, dated April 15, 1983 (JPN-83-33) regarding Safety Parameter Display System Implementation Plan.
: 14. NYPA letter, C. A. McNeill, Jr. to D. B. Vassallo, dated March 15, 1985 (JPN-85-021) regarding Appendix R to 10 CFR 50, Request for Schedular Exemption.
: 15. IE Information Notice 85-09, " Isolation Transfer Switches and Post-fire Shutdown Capability", dated January 31, 1985.
: 16. NYPA letter, J. P.Bayne to T. A. Ippolito, dated January 7, 1982 (JPN-82-005) regarding Containment Isolation Dependability, NUREG-0737, Item II.E.4.2.
Submits containment isolation study.
: 17. NRC letter, H. I. Abelson to J. C. Brons, dated March     j 20, 1987 (JAF-87-059) issues Amendment No. 106 to FitzPatrick Operating License.
: 18. NYPA letter, J. C. Brons to NRC, dated January 3, 1987 (JPN-87-004 ) regarding revised schedule for installation of new reactor water cleanup system          }
containment isolation valves.
: 19. PASNY letter, G. T. Berry to R. W. Reid, dated October 17, 1977 (JNRC-77-002) regarding second level of undervoltage protection.
_ _    _ - _ - - _ _ - _}}