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public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station August 12, 1994 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555  
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i public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station August 12, 1994 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC               20555


==Dear Sir:==
==Dear Sir:==
MONTHLY OPERATING REPORT SALEM NO. 1 DOCKET NO. 50-272 In compliance with Section 6.9.1.6, Reporting Requirements for the Salem Technical Specifications, the original copy of the monthly operating reports for the month of July 1994 are being sent to you. RH:pc Average Daily Unit Power Level Operating Data Report Unit Shutdowns and Power Reductions 10CFR50.59 Evaluations Operating Summary Refueling Information Sincerely yours, cc: Mr. Thomas T. Martin Regional Administrator USNRC Region I 631 Park Avenue King of Prussia, PA 19046 Enclosures 8-1-7.R4 ThREnerav 9408190226 940731 PDR ADOCK 05000272 R PDR 1 I (10M) 12-89 e OPERATING DATA REPORT e Docket No: 50-272 Date: 08/10/94 Completed by: Mike Morroni Telephone:
 
339-2122 Operating Status 1. Unit Name Salem No. 1 Notes 2. Reporting Period July 1994 3. Licensed Thermal Power (MWt) 3411 4. Nameplate Rating (Gross MWe) 1170 5. Design Electrical Rating (Net MWe) 1115 6. Maximum Dependable Capacity(Gross MWe) 1149 7. Maximum Dependable Capacity (Net MWe) 1106 8. If Changes Occur in Capacity Ratings (items 3 through 7) since Last Report, Give Reason NA 9. Power Level to Which Restricted, if any (Net MWe) N/A 10. Reasons for Restrictions, if any
MONTHLY OPERATING REPORT SALEM NO. 1 DOCKET NO. 50-272 In compliance with Section 6.9.1.6, Reporting Requirements for the Salem Technical Specifications, the original copy of the monthly operating reports for the month of July 1994 are being sent to you.
: 12. Hours in Reporting Period 12. No. of Hrs. Rx. was Critical 13. Reactor Reserve Shutdown Hrs. 14. Hours Generator On-Line 15. Unit Reserve Shutdown Hours 16. Gross Thermal Energy Generated (MWH) 17. Gross Elec. Energy Generated (MWH) 18. Net Elec. Energy Gen. (MWH) 19. Unit Service Factor 20. Unit Availability Factor 21. Unit Capacity Factor (using MDC Net) 22. Unit Capacity Factor (using DER Net) 23. Unit Forced outage Rate This Month 744 678.35 0 648.58 0 2120853.6 697480 664768 87.2 87.2 80.8 80.1 12.8 Year to Date 5087 2914.58 0 2375.13 0 7949685.6 2276330 2108071 46.7 46.7 37.5 37.2 44.1 Cumulative 149784 98046.6 0 94262.97 0 298721999.6 98812300 94045624 62.9 62.9 56.8 56.3 21.8 24. Shutdowns scheduled over next 6 months (type, date and duration of each) None. 25. If shutdown at end of Report Period, Estimated Date of startup: N A. 8-1-7.R2
Average Daily Unit Power Level Operating Data Report Unit Shutdowns and Power Reductions 10CFR50.59 Evaluations Operating Summary Refueling Information Sincerely yours, RH:pc cc:         Mr. Thomas T. Martin Regional Administrator USNRC Region I 631 Park Avenue King of Prussia, PA               19046 Enclosures 8-1-7.R4
.ERAGE DAILY UNIT POWER Docket No.: 50-272 Unit Name: Salem #1 Date: 08/10/94 Completed by: Mike Morroni Telephone:
                                                                                              /(fl~
339-2122 Month July. 1994 Day Average Daily Power Level Day Average Daily Power Level (MWe-NET) (MWe-NET) 1 1050 17 0 2 1020 18 0 3 1055 19 587 4 1060 20 1028 5 1060 21 1048 6 1036 22 1073 7 1040 23 . 1036 8 1042 24 1060 9 1057 25 1056 10 1038 26 1044 11 1063 27 1075 12 1042 28 1057 13 1018 29 1060 14 982 30 1068 15 0 31 1060 16 0 P. 8.1-7 Rl NO. DATE 1440 07-14-94 1465 07-19-94 1 F: Forced S: Scheduled UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH JULY 1994 METHOD OF SHUTTING LICENSE DURATION DOWN EVENT SYSTEM TYPE 1 (HOURS) REASON 2 REACTOR REPORT # COOE 4 F 95.42 A 2 ----------
1 ThREnerav Peoole~-                                                                                      I 9408190226 940731 PDR ADOCK 05000272                                                             ~-2189 (10M) 12-89 R                        PDR
HF F 1.9 A 5 ----------
 
HB 2 3 Reason A-Equipment Failure (explain)
e   OPERATING DATA REPORT e Docket No:     50-272 Date:         08/10/94 Completed by:     Mike Morroni                     Telephone:     339-2122 Operating Status
B-Maintenance or Test C-Refueling D-Requlatory Restriction E-Operator Training & License Examination F-Aaninistrative G-Operational Error (Explain)
: 1. Unit Name                         Salem No. 1   Notes
H-Other (Explain)
: 2. Reporting Period             July       1994
Method: 1-Manual 2-Manual Scram 3-Automatic Scram 4-Continuation of Previous Outage 5-Load Reduction 9-0ther DOCKET NO.
: 3. Licensed Thermal Power (MWt)             3411
__ _ UNIT NAME: Salem #1 DATE: 08-10-94 COMPLETED BY: Mike Morroni TELEPHONE:
: 4. Nameplate Rating (Gross MWe)             1170
339-2"122 COMPONENT CAUSE AND CORRECTIVE ACTION COOE 5 TO PREVENT RECURRENCE CKTBRK LIGHTNING VAL VEX CONTROL VALVES TURBINE 4 Exhibit G -Instructions for Preparation of Data Entry Sheets for Licensee Event Report CLER) File CNUREG-0161) 5 Exhibit 1 -Same Source 
: 5. Design Electrical Rating (Net MWe)       1115
-e 10CFR50.59 EVALUATIONS MONTH: -JULY 1994 e DOCKET NO: UNIT NAME: DATE: COMPLETED BY: TELEPHONE:
: 6. Maximum Dependable Capacity(Gross MWe) 1149
50-272 SALEM 1 AUGUST 10, 1994 R. HELLER (609)339-5162  
: 7. Maximum Dependable Capacity (Net MWe)     1106
: 8. If Changes Occur in Capacity Ratings (items 3 through 7) since Last Report, Give Reason     NA
: 9. Power Level to Which Restricted, if any (Net MWe)           N/A
: 10. Reasons for Restrictions, if any ~~~~___,N..,,,.__,A=-~~~~~~~~~~~~~
This Month  Year to Date    Cumulative
: 12. Hours in Reporting Period           744          5087        149784
: 12. No. of Hrs. Rx. was Critical         678.35        2914.58      98046.6
: 13. Reactor Reserve Shutdown Hrs.         0              0            0
: 14. Hours Generator On-Line             648.58        2375.13      94262.97
: 15. Unit Reserve Shutdown Hours           0              0            0
: 16. Gross Thermal Energy Generated (MWH)                       2120853.6  7949685.6    298721999.6
: 17. Gross Elec. Energy Generated (MWH)                     697480        2276330      98812300
: 18. Net Elec. Energy Gen. (MWH)       664768        2108071      94045624
: 19. Unit Service Factor                   87.2          46.7          62.9
: 20. Unit Availability Factor               87.2          46.7          62.9
: 21. Unit Capacity Factor (using MDC Net)                   80.8          37.5          56.8
: 22. Unit Capacity Factor (using DER Net)                 80.1          37.2          56.3
: 23. Unit Forced outage Rate               12.8           44.1           21.8
: 24. Shutdowns scheduled over next 6 months (type, date and duration of each)
None.
: 25. If shutdown at end of Report Period, Estimated Date of startup:
N A.
8-1-7.R2
 
                          .ERAGE DAILY UNIT POWER LE~
Docket No.:   50-272 Unit Name:     Salem #1 Date:         08/10/94 Completed by:     Mike Morroni                       Telephone:     339-2122 Month     July.       1994 Day Average Daily Power Level             Day Average Daily Power Level (MWe-NET)                                 (MWe-NET) 1         1050                             17             0 2         1020                             18             0 3         1055                             19           587 4         1060                             20           1028 5         1060                             21           1048 6         1036                             22           1073 7         1040                             23         . 1036 8         1042                             24           1060 9         1057                             25           1056 10         1038                             26           1044 11         1063                             27           1075 12         1042                             28           1057 13         1018                             29           1060 14           982                             30           1068 15             0                             31           1060 16             0 P. 8.1-7 Rl
 
UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH JULY     1994                                             DOCKET NO. :_.5"""0..,..-2=7~2'="'"_ __
UNIT NAME: Salem #1 DATE: 08-10-94 COMPLETED BY: Mike Morroni TELEPHONE: 339-2"122 METHOD OF SHUTTING       LICENSE DURATION                       DOWN           EVENT           SYSTEM   COMPONENT              CAUSE AND CORRECTIVE ACTION NO.          DATE    TYPE 1 (HOURS)       REASON 2 REACTOR       REPORT #         COOE 4   COOE 5                    TO PREVENT RECURRENCE 1440      07-14-94  F         95.42         A               2             ----------       HF       CKTBRK          LIGHTNING 1465      07-19-94  F           1.9           A               5             ----------       HB       VAL VEX        CONTROL VALVES TURBINE 1              2                                                         3                         4                                  5 F:  Forced      Reason                                                   Method:                  Exhibit G - Instructions          Exhibit 1 - Same S:  Scheduled  A-Equipment Failure (explain)                             1-Manual                  for Preparation of Data          Source B-Maintenance or Test                                     2-Manual Scram            Entry Sheets for Licensee C-Refueling                                               3-Automatic Scram        Event Report CLER) File D-Requlatory Restriction                                 4-Continuation of        CNUREG-0161)
E-Operator Training & License Examination                   Previous Outage F-Aaninistrative                                         5-Load Reduction G-Operational Error (Explain)                             9-0ther H-Other (Explain)
 
  -
10CFR50.59 EVALUATIONS e                            e DOCKET NO:  50-272 MONTH: - JULY 1994                             UNIT NAME:   SALEM 1 DATE:   AUGUST 10, 1994 COMPLETED BY:  R. HELLER TELEPHONE:  (609)339-5162
-----------------------------------------------------------------------------
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The following items were evaluated in accordance with the provisions of the Code of Federal Regulations 10CFR50.59.
The following items were evaluated in accordance with the provisions of the Code of Federal Regulations 10CFR50.59. The Station Operations Review Committee has reviewed and concurs with these evaluations.
The Station Operations Review Committee has reviewed and concurs with these evaluations.
ITEM                              
ITEM A. Design Change Packages lEC-3252 Pkg 1 lEC-3340 Pkg 1


==SUMMARY==
==SUMMARY==
  "Reracking Of spent Fuel Pool For Increased Storage Capacity" -This DCP will remove nine existing Exxon racks and replace then with nine maximum density Holtec racks. Relocate three Exxon racks in the spent fuel pool. In the reracked pool racks are located closer to the wall. The hook of the FH crane will not reach the first row of cells near the North wall and the first two rows of cells near the North wall and first two rows of cells near the East wall. For this purpose a new off set tool will be procured and used. This DCP will also modify the skimmers and strainer and make them removable for future access to the cells directly under these items. New Holtec racks may require use of a funnel for inserting spent fuels. The spent fuel storage capacity will run out in 1998 if this change is not made. After this modification, the storage capacity in the spent fuel pool will be available until 2008 *. There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 94-056) "Waste Gas System Setpoint Modification" -1. The following setpoints will be verified and adjusted as necessary during DCP installation in order to assure optimum system operation:
 
1WG8 control, Auto Waste Gas Decay Tank (WGDT) swap-over, Compressor Auto Start and Compressor Auto Stop. 2. A Special Test Procedure (STP) will be created as part of this DCP which will include, but will not be limited to, provisions for optimum test conditions, test procedures, calibrations and adjustments.
A. Design Change Packages lEC-3252 Pkg 1        "Reracking Of spent Fuel Pool For Increased Storage Capacity" - This DCP will remove nine existing Exxon racks and replace then with nine maximum density Holtec racks. Relocate three Exxon racks in the spent fuel pool. In the reracked pool racks are located closer to the wall. The hook of the FH crane will not reach the first row of cells near the North wall and the first two rows of cells near the North wall and first two rows of cells near the East wall. For this purpose a new off set tool will be procured and used. This DCP will also modify the skimmers and strainer and make them removable for future access to the cells directly under these items.
Also, steps for adjustments of proportional band and reset for the 11WG22 and 12WG22 controllers will be provided in the STP in addition to the 1WG8 controller.
New Holtec racks may require use of a funnel for inserting spent fuels. The spent fuel storage capacity will run out in 1998 if this change is not made. After this modification, the storage capacity in the spent fuel pool will be available until 2008 * . There is no reduction in the margin of safety as defined in the basis for any Technical Specification.     (SORC 94-056) lEC-3340  Pkg 1        "Waste Gas System Setpoint Modification" - 1. The following setpoints will be verified and adjusted as necessary during DCP installation in order to assure optimum system operation: 1WG8 control, Auto Waste Gas Decay Tank (WGDT) swap-over, Compressor Auto Start and Compressor Auto Stop.
The goal of this DCP is to maintain a suction header pressure of 0.5 to 2.0 psig as described in UFSAR (Section 11. 3)
: 2. A Special Test Procedure (STP) will be created as part of this DCP which will include, but will not be limited to, provisions for optimum test conditions, test procedures, calibrations and adjustments. Also, steps for adjustments of proportional band and reset for the 11WG22 and 12WG22 controllers will be provided in the STP in addition to the 1WG8 controller. The goal of this DCP is to maintain a suction header pressure of 0.5 to 2.0 psig as described in UFSAR (Section
. e 10CFR50.59 EVALUATIONS MONTH: -JULY 1994 (cont'd) ITEM e DOCKET NO: UNIT NAME: DATE: COMPLETED BY: TELEPHONE:  
: 11. 3)
 
  .
10CFR50.59 EVALUATIONS e                            e DOCKET NO:  50-272 MONTH: - JULY 1994                           UNIT NAME:   SALEM 1 DATE:   AUGUST 10, 1994 COMPLETED BY:   R. HELLER TELEPHONE:   (609)339-5162 (cont'd)
ITEM                                 


==SUMMARY==
==SUMMARY==
50-272 SALEM 1 AUGUST 10, 1994 R. HELLER (609)339-5162 while operating in the automatic mode. Vendor recommended setpoint changes for the Waste Gas Compressors will be made so that the Waste Gas Compressors can be operated in the automatic mode, as described in the SAR. In addition, the vendor recommended setpoint changes will be used to improve the reliability of the Waste gas Compressors.
 
The changes performed by this modification to the WGS do not affect any credible failure modes. All plant modifications for this DCP are confined to the recalibration of selected Waste Gas instrumentation.
while operating in the automatic mode. Vendor recommended setpoint changes for the Waste Gas Compressors will be made so that the Waste Gas Compressors can be operated in the automatic mode, as described in the SAR. In addition, the vendor recommended setpoint changes will be used to improve the reliability of the Waste gas Compressors. The changes performed by this modification to the WGS do not affect any credible failure modes. All plant modifications for this DCP are confined to the recalibration of selected Waste Gas instrumentation. There no is no increased potential for WGDT rupture since the output of the compressors is reduced in regard to
There no is no increased potential for WGDT rupture since the output of the compressors is reduced in regard to _the design limits of the tanks. The modification involves non-safety related instrumentation and does not involve any Category 1E systems. No active or passive failure modes are associated with this change, and the electrical performance in regard to failure (e.g., hot shorts, grounded shorts, etc.) is not affected by this modification.
_the design limits of the tanks. The modification involves non-safety related instrumentation and does not involve any Category 1E systems. No active or passive failure modes are associated with this change, and the electrical performance in regard to failure (e.g., hot shorts, grounded shorts, etc.) is not affected by this modification. There is no reduction in the margin of safety as defined in the basis for any Technical Specification.   (SORC 94-060)
There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 94-060) B. Procedures and Revisions LOCA 3 & LOCA 4 "EOP & FSAR Changes for Recirculation Mode" -There have been several performance issues identified over the years related to the operation of the Residual Heat Removal System (RHRS) during the Recirculation Mode following a Loss Of Coolant Accident (LOCA). These issues include: 1.) Unintended flow paths and their consequences; 2.) Excessive RHRS pump flows during Hot Leg Recirculation; 3.) Excessive suction boost to the Charging/Safety Injection pumps and Intermediate Head Safety Injection pumps. PSE&G and Westinghouse have worked together to develop changes for Salem Units 1 and 2 to address these issues. The scope of this safety evaluation includes an assessment of the following changes: 1.) Modification of the RHR pump design basis for
B. Procedures and Revisions LOCA 3 & LOCA 4       "EOP & FSAR Changes for Recirculation Mode" -
. e 10CFR50.59 EVALUATIONS MONTH: -JULY 1994 (cont'd) ITEM NC.NA-AP.ZZ-0034(Z) e DOCKET NO: UNIT NAME: DATE: COMPLETED BY: TELEPHONE:  
There have been several performance issues identified over the years related to the operation of the Residual Heat Removal System (RHRS) during the Recirculation Mode following a Loss Of Coolant Accident (LOCA). These issues include: 1.)
Unintended flow paths and their consequences; 2.)
Excessive RHRS pump flows during Hot Leg Recirculation; 3.) Excessive suction boost to the Charging/Safety Injection pumps and Intermediate Head Safety Injection pumps. PSE&G and Westinghouse have worked together to develop changes for Salem Units 1 and 2 to address these issues. The scope of this safety evaluation includes an assessment of the following changes:
1.) Modification of the RHR pump design basis for
 
  .                 e 10CFR50.59 EVALUATIONS e
DOCKET NO:  50-272 MONTH: - JULY 1994                     UNIT NAME:   SALEM 1 DATE:   AUGUST 10, 1994 COMPLETED BY:   R. HELLER TELEPHONE:   (609)339-5162 (cont'd)
ITEM                           


==SUMMARY==
==SUMMARY==
50-272 SALEM 1 AUGUST 10, 1994 R. HELLER (609)339-5162 net positive suction head (NPSH) to take credit for Containment air pressure; and 2.) Selected Emergency Operating Procedure (EOP) step changes. The first is necessary to justify an increase in RHR pump "worst-case" maximum flow potential and address Containment Sump level setpoint accuracy.
 
The changes identified in Item 2 are being made to preclude undesirable RHR pump flow alignments and to enhance RHRS long term availability.
net positive suction head (NPSH) to take credit for Containment air pressure; and 2.) Selected Emergency Operating Procedure (EOP) step changes.
No Technical Specification flow limits are affected by these changes. There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 94-057) "Performance Indicator Program" -This proposal deletes Nuclear Administrative Procedure NC.NA-AP.ZZ-0034(Z), Performance Indicator Program. The Performance Indicator Program has been incorporated into the Business Plan. The Operational Quality Assurance Program does not apply to this procedure.
The first is necessary to justify an increase in RHR pump "worst-case" maximum flow potential and address Containment Sump level setpoint accuracy.
Although this procedure is briefly described in the UFSAR (Section 13.5.1 (Hope Creek), Plant Procedures, and Section 13.5.1.1 (Salem), Nuclear Administrative Procedures), it performs no function relative to the safe operation of either Hope Creek or Salem Stations.
The changes identified in Item 2 are being made to preclude undesirable RHR pump flow alignments and to enhance RHRS long term availability. No Technical Specification flow limits are affected by these changes. There is no reduction in the margin of safety as defined in the basis for any Technical Specification.   (SORC 94-057)
Therefore, deleting this procedure will have no affect on the safe operation of either Hope Creek or Salem Stations.
NC.NA-AP.ZZ-0034(Z)    "Performance Indicator Program" - This proposal deletes Nuclear Administrative Procedure NC.NA-AP.ZZ-0034(Z), Performance Indicator Program. The Performance Indicator Program has been incorporated into the Business Plan. The Operational Quality Assurance Program does not apply to this procedure. Although this procedure is briefly described in the UFSAR (Section 13.5.1 (Hope Creek), Plant Procedures, and Section 13.5.1.1 (Salem), Nuclear Administrative Procedures), it performs no function relative to the safe operation of either Hope Creek or Salem Stations. Therefore, deleting this procedure will have no affect on the safe operation of either Hope Creek or Salem Stations. UFSAR Change*
UFSAR Change* Requests for Hope Creek and Salem UFSARs are being concurrently submitted with this proposal to remove the description of this procedure from the UFSAR. This procedure does not affect the operation of any equipment, either safety related or non-safety related. Therefore, deleting this procedure will have no effect on any equipment important to safety. There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 94-060)
Requests for Hope Creek and Salem UFSARs are being concurrently submitted with this proposal to remove the description of this procedure from the UFSAR. This procedure does not affect the operation of any equipment, either safety related or non-safety related. Therefore, deleting this procedure will have no effect on any equipment important to safety. There is no reduction in the margin of safety as defined in the basis for any Technical Specification.   (SORC 94-060)
. e .10CFR50.59 EVALUATIONS  
 
-JULY 1994 (cont'd) ITEM c. SAR Changes UFSAR Section 17.2 e DOCKET NO: UNIT NAME: DATE: COMPLETED BY: TELEPHONE:  
    .
.10CFR50.59 EVALUATIONS e                      e DOCKET NO: 50-272 MON~H: - JULY 1994                      UNIT NAME: SALEM 1 DATE: AUGUST 10, 1994 COMPLETED BY: R. HELLER TELEPHONE:   (609)339-5162 (cont'd)
ITEM                           


==SUMMARY==
==SUMMARY==
50-272 SALEM 1 AUGUST 10, 1994 R. HELLER (609)339-5162 "Quality Assurance During the Operations Phase" -The proposed changes: 1.) Reflect current organizational structure; 2.) Make editorial enhancements, and 3.) Provide clarifications.
: c. SAR Changes UFSAR Section 17.2 "Quality Assurance During the Operations Phase" -
Additionally, the changes effect selected Quality Assurance (QA) oversight practices for monitoring implementation of the Nuclear Department Procedure System. Changes in QA practices will now utilize the assessment process to evaluate the effectiveness of Nuclear Department groups implementation of the Nuclear Department Procedure System. The assessment approach expands the scope of QA to evaluate process requirements for preparing, reviewing, revising and issuing departmental administrative and implementing procedures.
The proposed changes: 1.) Reflect current organizational structure; 2.) Make editorial enhancements, and 3.) Provide clarifications.
There are no credible failure modes associated with this change. There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 94-058}
Additionally, the changes effect selected Quality Assurance (QA) oversight practices for monitoring implementation of the Nuclear Department Procedure System. Changes in QA practices will now utilize the assessment process to evaluate the effectiveness of Nuclear Department groups implementation of the Nuclear Department Procedure System. The assessment approach expands the scope of QA to evaluate process requirements for preparing, reviewing, revising and issuing departmental administrative and implementing procedures. There are no credible failure modes associated with this change. There is no reduction in the margin of safety as defined in the basis for any Technical Specification.
SALEM UNIT NO. 1 SALEM GENERATING STATION MONTHLY OPERATING  
(SORC 94-058}
 
SALEM GENERATING STATION MONTHLY OPERATING  


==SUMMARY==
==SUMMARY==
  -UNIT 1 JULY 1994 The Unit began the period operating at full power and continued to operate at essentially 100% power until July 14, 1994, when operators manually tripped the reactor due to loss of all circulators.
  - UNIT 1 JULY 1994 SALEM UNIT NO. 1 The Unit began the period operating at full power and continued to operate at essentially 100% power until July 14, 1994, when operators manually tripped the reactor due to loss of all circulators. An electrical storm in the area caused a bus voltage perturbation which activated the protective relays on the bus feeding the Circulating Water Pumps. A time delay has been installed to prevent a future similar occurrence. The Unit was returned to 100% power on July 20, 1994, and continued to operate at essentially 100% power throughout the remainder of the period.
An electrical storm in the area caused a bus voltage perturbation which activated the protective relays on the bus feeding the Circulating Water Pumps. A time delay has been installed to prevent a future similar occurrence.
 
The Unit was returned to 100% power on July 20, 1994, and continued to operate at essentially 100% power throughout the remainder of the period.
    -
-e e . REFUELING INFORMATION DOCKET NO: MONTH: -JULY 1994 ' . UNIT NAME: DATE: COMPLETED BY: TELEPHONE:
. REFUELING INFORMATION e                            e DOCKET NO: 50-272
MONTH JULY 1994 1. Refueling inf ormat.ion has changed from last month: YES NO X 2. Scheduled date for next refueling:
    ' .
APRIL 8, 1995 50-272 SALEM 1 AUGUST 10, 1994 R. HELLER (609)339-5162
MONTH:   - JULY 1994                             UNIT NAME:
: 3. Scheduled date for restart following refueling:
DATE:
JUNE 6, 1995 4. a) Will Technical Specification changes or other license amendments be required?:
SALEM 1 AUGUST 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 MONTH JULY 1994
YES NO NOT DETERMINED TO DATE b) Has the reload fuel design been reviewed by the Station Operating Review Committee?:
: 1. Refueling inf ormat.ion has changed from last month:
YES NO X If no, when is it scheduled?:
YES                 NO     X
MARCH 1995 5. Scheduled date(s) for submitting proposed licensing action: N/A 6. Important licensing considerations associated with refueling:
: 2. Scheduled date for next refueling:     APRIL 8, 1995
: 3. Scheduled date for restart following refueling:     JUNE 6, 1995
: 4. a)   Will Technical Specification changes or other license amendments be required?:
YES                 NO NOT DETERMINED TO DATE --=X~-
b)   Has the reload fuel design been reviewed by the Station Operating Review Committee?:
YES                 NO     X If no, when is it scheduled?:     MARCH 1995
: 5. Scheduled date(s) for submitting proposed licensing action:
N/A
: 6. Important licensing considerations associated with refueling:
: 7. Number of Fuel Assemblies:
: 7. Number of Fuel Assemblies:
: a. Incore 193 b. In Spent Fuel Storage 732 8. Present licensed spent fuel storage capacity:
: a. Incore                                                       193
1170 Future spent fuel storage capacity:
: b. In Spent Fuel Storage                                         732
1170 9. Date of last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity:
: 8. Present licensed spent fuel storage capacity:                   1170 Future spent fuel storage capacity:                             1170
September 2001 8-l-7.R4}}
: 9. Date of last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity:                                       September 2001 8-l-7.R4}}

Revision as of 10:22, 21 October 2019

Monthly Operating Rept for Jul 1994 for Salem Generating Station,Unit 1.W/940812 Ltr
ML18101A184
Person / Time
Site: Salem PSEG icon.png
Issue date: 07/31/1994
From: Hagan J, Morroni M
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9408190226
Download: ML18101A184 (10)


Text

,.,_

PS~G*

i public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station August 12, 1994 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Dear Sir:

MONTHLY OPERATING REPORT SALEM NO. 1 DOCKET NO. 50-272 In compliance with Section 6.9.1.6, Reporting Requirements for the Salem Technical Specifications, the original copy of the monthly operating reports for the month of July 1994 are being sent to you.

Average Daily Unit Power Level Operating Data Report Unit Shutdowns and Power Reductions 10CFR50.59 Evaluations Operating Summary Refueling Information Sincerely yours, RH:pc cc: Mr. Thomas T. Martin Regional Administrator USNRC Region I 631 Park Avenue King of Prussia, PA 19046 Enclosures 8-1-7.R4

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1 ThREnerav Peoole~- I 9408190226 940731 PDR ADOCK 05000272 ~-2189 (10M) 12-89 R PDR

e OPERATING DATA REPORT e Docket No: 50-272 Date: 08/10/94 Completed by: Mike Morroni Telephone: 339-2122 Operating Status

1. Unit Name Salem No. 1 Notes
2. Reporting Period July 1994
3. Licensed Thermal Power (MWt) 3411
4. Nameplate Rating (Gross MWe) 1170
5. Design Electrical Rating (Net MWe) 1115
6. Maximum Dependable Capacity(Gross MWe) 1149
7. Maximum Dependable Capacity (Net MWe) 1106
8. If Changes Occur in Capacity Ratings (items 3 through 7) since Last Report, Give Reason NA
9. Power Level to Which Restricted, if any (Net MWe) N/A
10. Reasons for Restrictions, if any ~~~~___,N..,,,.__,A=-~~~~~~~~~~~~~

This Month Year to Date Cumulative

12. Hours in Reporting Period 744 5087 149784
12. No. of Hrs. Rx. was Critical 678.35 2914.58 98046.6
13. Reactor Reserve Shutdown Hrs. 0 0 0
14. Hours Generator On-Line 648.58 2375.13 94262.97
15. Unit Reserve Shutdown Hours 0 0 0
16. Gross Thermal Energy Generated (MWH) 2120853.6 7949685.6 298721999.6
17. Gross Elec. Energy Generated (MWH) 697480 2276330 98812300
18. Net Elec. Energy Gen. (MWH) 664768 2108071 94045624
19. Unit Service Factor 87.2 46.7 62.9
20. Unit Availability Factor 87.2 46.7 62.9
21. Unit Capacity Factor (using MDC Net) 80.8 37.5 56.8
22. Unit Capacity Factor (using DER Net) 80.1 37.2 56.3
23. Unit Forced outage Rate 12.8 44.1 21.8
24. Shutdowns scheduled over next 6 months (type, date and duration of each)

None.

25. If shutdown at end of Report Period, Estimated Date of startup:

N A.

8-1-7.R2

.ERAGE DAILY UNIT POWER LE~

Docket No.: 50-272 Unit Name: Salem #1 Date: 08/10/94 Completed by: Mike Morroni Telephone: 339-2122 Month July. 1994 Day Average Daily Power Level Day Average Daily Power Level (MWe-NET) (MWe-NET) 1 1050 17 0 2 1020 18 0 3 1055 19 587 4 1060 20 1028 5 1060 21 1048 6 1036 22 1073 7 1040 23 . 1036 8 1042 24 1060 9 1057 25 1056 10 1038 26 1044 11 1063 27 1075 12 1042 28 1057 13 1018 29 1060 14 982 30 1068 15 0 31 1060 16 0 P. 8.1-7 Rl

UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH JULY 1994 DOCKET NO. :_.5"""0..,..-2=7~2'="'"_ __

UNIT NAME: Salem #1 DATE: 08-10-94 COMPLETED BY: Mike Morroni TELEPHONE: 339-2"122 METHOD OF SHUTTING LICENSE DURATION DOWN EVENT SYSTEM COMPONENT CAUSE AND CORRECTIVE ACTION NO. DATE TYPE 1 (HOURS) REASON 2 REACTOR REPORT # COOE 4 COOE 5 TO PREVENT RECURRENCE 1440 07-14-94 F 95.42 A 2 ---------- HF CKTBRK LIGHTNING 1465 07-19-94 F 1.9 A 5 ---------- HB VAL VEX CONTROL VALVES TURBINE 1 2 3 4 5 F: Forced Reason Method: Exhibit G - Instructions Exhibit 1 - Same S: Scheduled A-Equipment Failure (explain) 1-Manual for Preparation of Data Source B-Maintenance or Test 2-Manual Scram Entry Sheets for Licensee C-Refueling 3-Automatic Scram Event Report CLER) File D-Requlatory Restriction 4-Continuation of CNUREG-0161)

E-Operator Training & License Examination Previous Outage F-Aaninistrative 5-Load Reduction G-Operational Error (Explain) 9-0ther H-Other (Explain)

-

10CFR50.59 EVALUATIONS e e DOCKET NO: 50-272 MONTH: - JULY 1994 UNIT NAME: SALEM 1 DATE: AUGUST 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162


The following items were evaluated in accordance with the provisions of the Code of Federal Regulations 10CFR50.59. The Station Operations Review Committee has reviewed and concurs with these evaluations.

ITEM

SUMMARY

A. Design Change Packages lEC-3252 Pkg 1 "Reracking Of spent Fuel Pool For Increased Storage Capacity" - This DCP will remove nine existing Exxon racks and replace then with nine maximum density Holtec racks. Relocate three Exxon racks in the spent fuel pool. In the reracked pool racks are located closer to the wall. The hook of the FH crane will not reach the first row of cells near the North wall and the first two rows of cells near the North wall and first two rows of cells near the East wall. For this purpose a new off set tool will be procured and used. This DCP will also modify the skimmers and strainer and make them removable for future access to the cells directly under these items.

New Holtec racks may require use of a funnel for inserting spent fuels. The spent fuel storage capacity will run out in 1998 if this change is not made. After this modification, the storage capacity in the spent fuel pool will be available until 2008 * . There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 94-056) lEC-3340 Pkg 1 "Waste Gas System Setpoint Modification" - 1. The following setpoints will be verified and adjusted as necessary during DCP installation in order to assure optimum system operation: 1WG8 control, Auto Waste Gas Decay Tank (WGDT) swap-over, Compressor Auto Start and Compressor Auto Stop.

2. A Special Test Procedure (STP) will be created as part of this DCP which will include, but will not be limited to, provisions for optimum test conditions, test procedures, calibrations and adjustments. Also, steps for adjustments of proportional band and reset for the 11WG22 and 12WG22 controllers will be provided in the STP in addition to the 1WG8 controller. The goal of this DCP is to maintain a suction header pressure of 0.5 to 2.0 psig as described in UFSAR (Section
11. 3)

.

10CFR50.59 EVALUATIONS e e DOCKET NO: 50-272 MONTH: - JULY 1994 UNIT NAME: SALEM 1 DATE: AUGUST 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

while operating in the automatic mode. Vendor recommended setpoint changes for the Waste Gas Compressors will be made so that the Waste Gas Compressors can be operated in the automatic mode, as described in the SAR. In addition, the vendor recommended setpoint changes will be used to improve the reliability of the Waste gas Compressors. The changes performed by this modification to the WGS do not affect any credible failure modes. All plant modifications for this DCP are confined to the recalibration of selected Waste Gas instrumentation. There no is no increased potential for WGDT rupture since the output of the compressors is reduced in regard to

_the design limits of the tanks. The modification involves non-safety related instrumentation and does not involve any Category 1E systems. No active or passive failure modes are associated with this change, and the electrical performance in regard to failure (e.g., hot shorts, grounded shorts, etc.) is not affected by this modification. There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 94-060)

B. Procedures and Revisions LOCA 3 & LOCA 4 "EOP & FSAR Changes for Recirculation Mode" -

There have been several performance issues identified over the years related to the operation of the Residual Heat Removal System (RHRS) during the Recirculation Mode following a Loss Of Coolant Accident (LOCA). These issues include: 1.)

Unintended flow paths and their consequences; 2.)

Excessive RHRS pump flows during Hot Leg Recirculation; 3.) Excessive suction boost to the Charging/Safety Injection pumps and Intermediate Head Safety Injection pumps. PSE&G and Westinghouse have worked together to develop changes for Salem Units 1 and 2 to address these issues. The scope of this safety evaluation includes an assessment of the following changes:

1.) Modification of the RHR pump design basis for

. e 10CFR50.59 EVALUATIONS e

DOCKET NO: 50-272 MONTH: - JULY 1994 UNIT NAME: SALEM 1 DATE: AUGUST 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

net positive suction head (NPSH) to take credit for Containment air pressure; and 2.) Selected Emergency Operating Procedure (EOP) step changes.

The first is necessary to justify an increase in RHR pump "worst-case" maximum flow potential and address Containment Sump level setpoint accuracy.

The changes identified in Item 2 are being made to preclude undesirable RHR pump flow alignments and to enhance RHRS long term availability. No Technical Specification flow limits are affected by these changes. There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 94-057)

NC.NA-AP.ZZ-0034(Z) "Performance Indicator Program" - This proposal deletes Nuclear Administrative Procedure NC.NA-AP.ZZ-0034(Z), Performance Indicator Program. The Performance Indicator Program has been incorporated into the Business Plan. The Operational Quality Assurance Program does not apply to this procedure. Although this procedure is briefly described in the UFSAR (Section 13.5.1 (Hope Creek), Plant Procedures, and Section 13.5.1.1 (Salem), Nuclear Administrative Procedures), it performs no function relative to the safe operation of either Hope Creek or Salem Stations. Therefore, deleting this procedure will have no affect on the safe operation of either Hope Creek or Salem Stations. UFSAR Change*

Requests for Hope Creek and Salem UFSARs are being concurrently submitted with this proposal to remove the description of this procedure from the UFSAR. This procedure does not affect the operation of any equipment, either safety related or non-safety related. Therefore, deleting this procedure will have no effect on any equipment important to safety. There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 94-060)

.

.10CFR50.59 EVALUATIONS e e DOCKET NO: 50-272 MON~H: - JULY 1994 UNIT NAME: SALEM 1 DATE: AUGUST 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

c. SAR Changes UFSAR Section 17.2 "Quality Assurance During the Operations Phase" -

The proposed changes: 1.) Reflect current organizational structure; 2.) Make editorial enhancements, and 3.) Provide clarifications.

Additionally, the changes effect selected Quality Assurance (QA) oversight practices for monitoring implementation of the Nuclear Department Procedure System. Changes in QA practices will now utilize the assessment process to evaluate the effectiveness of Nuclear Department groups implementation of the Nuclear Department Procedure System. The assessment approach expands the scope of QA to evaluate process requirements for preparing, reviewing, revising and issuing departmental administrative and implementing procedures. There are no credible failure modes associated with this change. There is no reduction in the margin of safety as defined in the basis for any Technical Specification.

(SORC 94-058}

SALEM GENERATING STATION MONTHLY OPERATING

SUMMARY

- UNIT 1 JULY 1994 SALEM UNIT NO. 1 The Unit began the period operating at full power and continued to operate at essentially 100% power until July 14, 1994, when operators manually tripped the reactor due to loss of all circulators. An electrical storm in the area caused a bus voltage perturbation which activated the protective relays on the bus feeding the Circulating Water Pumps. A time delay has been installed to prevent a future similar occurrence. The Unit was returned to 100% power on July 20, 1994, and continued to operate at essentially 100% power throughout the remainder of the period.

-

. REFUELING INFORMATION e e DOCKET NO: 50-272

' .

MONTH: - JULY 1994 UNIT NAME:

DATE:

SALEM 1 AUGUST 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 MONTH JULY 1994

1. Refueling inf ormat.ion has changed from last month:

YES NO X

2. Scheduled date for next refueling: APRIL 8, 1995
3. Scheduled date for restart following refueling: JUNE 6, 1995
4. a) Will Technical Specification changes or other license amendments be required?:

YES NO NOT DETERMINED TO DATE --=X~-

b) Has the reload fuel design been reviewed by the Station Operating Review Committee?:

YES NO X If no, when is it scheduled?: MARCH 1995

5. Scheduled date(s) for submitting proposed licensing action:

N/A

6. Important licensing considerations associated with refueling:
7. Number of Fuel Assemblies:
a. Incore 193
b. In Spent Fuel Storage 732
8. Present licensed spent fuel storage capacity: 1170 Future spent fuel storage capacity: 1170
9. Date of last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: September 2001 8-l-7.R4