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=Text=
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{{#Wiki_filter:' . , Public ::>ervice Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit AUG lJi 1997 LR-N970518 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn: Document Control Desk MONTHLY OPERATING REPORT SALEM UNIT NO. 2 DOCKET NO. 50-311 In compliance with Section 6.9.1.6, Reporting Requirements for the Salem Technical Specifications, the original monthly operating report for July, 1997, is attached.
{{#Wiki_filter:' .
F. Gare ow General Manager -Salem Operations RAR:tcp Enclosures c Mr. H. J. Miller Regional Administrator USNRC, Region 1 475 Allendale Road King of Prussia, PA 19046 -----
    , Public ::>ervice Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit AUG lJi 1997 LR-N970518 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn:         Document Control Desk MONTHLY OPERATING REPORT SALEM UNIT NO. 2 DOCKET NO. 50-311 In compliance with Section 6.9.1.6, Reporting Requirements for the Salem Technical Specifications, the original monthly operating report for July, 1997, is attached.
---------
I F. Gare ow General Manager -
.... ...-:.. __ n r:::_ ___ ------------..., 9708250118 970731 PDR ADOCK 05000311 R PDR The power is in your hands. I 1111111111111111111111111111111 11111111 . *lil.lil.5C17*
Salem Operations RAR:tcp Enclosures c       Mr. H. J. Miller Regional Administrator USNRC, Region 1 475 Allendale Road King of Prussia, PA 19046
I 95-2168 REV. 6/94 I I _J SALEM GENERATING STATION MONTHLY OPERATING  
        ----   - ---~- -- - --- --- .~~_£!_*-.... -*~- ...-:..__ n r:::_ ___ ---- ----- ---...,
9708250118 970731 PDR ADOCK 05000311 R                                           PDR                                             I1111111111111111111111111111111 11111111 .
                                                                                                          *lil.lil.5C17*
The power is in your hands.
I 95-2168 REV. 6/94 I
_J
 
SALEM GENERATING STATION MONTHLY OPERATING  


==SUMMARY==
==SUMMARY==
  -UNIT 2 JULY 1997 SALEM UNIT 2 The unit remained shutdown for the entire period. According to commitments from PSE&G and a subsequent confirmatory action letter from the NRC, the unit will remain shutdown pending completion of the following actions:
  - UNIT 2 JULY 1997 SALEM UNIT 2 The unit remained shutdown for the entire period. According to commitments from PSE&G and a subsequent confirmatory action letter from the NRC, the unit will remain shutdown pending completion of the following actions:
* Appropriately address long standing equipment reliability and operability issues.
* Appropriately address long standing equipment reliability and operability issues.
* After the work is completed, conduct a restart readiness review to determine for ourselves the ability of the unit to operate in a safe, event free manner.
* After the work is completed, conduct a restart readiness review to determine for ourselves the ability of the unit to operate in a safe, event free manner.
* After the restart review, meet with the NRC and communicate the results of that review.
* After the restart review, meet with the NRC and communicate the results of that review.
Refueling Information ,Month:, July, 1997 Month: July, 1997 Docket No. Unit Name: Contact: Telephone:
 
50-311 Salem 2 D. Tisdel 609-339-1538
Refueling Information                       Docket No. 50-311
: 1. Refueling information has changed from last month: Yes: No: X 2. Scheduled date for next refueling:
,Month:, July, 1997                         Unit Name: Salem 2 Contact:   D. Tisdel Telephone:  609-339-1538 Month:    July, 1997
Currently in outage. Scheduled date for restart following refueling:
: 1. Refueling information has changed from last month: Yes:   No: X
To Be Determined
: 2. Scheduled date for next refueling:   Currently in outage.
Scheduled date for restart following refueling: To Be Determined
: 3. a. Will Technical Specification changes or other license amendments be required?
: 3. a. Will Technical Specification changes or other license amendments be required?
Yes: x No: Not Determined to Date: b. Has the reload fuel design been reviewed by the Station Operating Review Committee?
Yes: x   No:     Not Determined to Date:
Yes: X (for upcoming cycle) No: If no, when is it scheduled?
: b. Has the reload fuel design been reviewed by the Station Operating Review Committee?
: 4. Scheduled date (s) for submitting proposed licensing action: N/A -previously submitted
Yes: X (for upcoming cycle) No: If no, when is it scheduled?
: 4. Scheduled date (s) for submitting proposed licensing action:     N/A -
previously submitted
: 5. Important licensing considerations associated with refueling:
: 5. Important licensing considerations associated with refueling:
: 6. Number of Fuel Assemblies:
: 6. Number of Fuel Assemblies:
: a. Incore: 193 b. In Spent Fuel Storage: 584 7. Present Licensed spent fuel storage capacity:
: a. Incore:                                                   193
1632 Future spent fuel storage capacity:
: b. In Spent Fuel Storage:                                   584
1632 8. Date of last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity:
: 7. Present Licensed spent fuel storage capacity:                 1632 Future spent fuel storage capacity:                           1632
October, 2016 NO. DATE 4092 7-1-97 F 1 2 F: Forced S: Scheduled DURATION TYPE' (HOURS) REASON 2 744 F c Reason A-Equipment Failure (explain)
: 8. Date of last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity:       October, 2016
B"Maintenance or Test* C-Refueling D-Requlatory Restriction UNIT SHUTDOYN AND POYER REDUCTIONS REPORT MONTH July 1997 METHOD OF SHUTTING LICENSE DOIJN EVENT SYSTEM REACTOR REPORT # CODE 4 4 _ .. _________
 
2422 3 Method: 1-Manual 2-Manual Scram E-Operator Training & License Examination F-Administrative 3-Automatic Scram 4-Continuation of Previous Outage 5-Load Reduction 9-0ther G-Operational Error (Explain)
UNIT SHUTDOYN AND POYER REDUCTIONS REPORT MONTH July         1997                                           DOCKET NO.: ~5~0~-~31~1~--
H-Other (Explain) 4 DOCKET NO.:
UNIT NAME:   Salem #2 DATE: 08-10-97 COMPLETED BY:   Robert Phillips TELEPHONE:   609-339-2735 -
UNIT NAME: Salem #2 DATE: 08-10-97 COMPLETED BY: Robert Phillips TELEPHONE:
METHOD OF SHUTTING        LICENSE                                                                  -
609-339-2735  
DURATION                      DOIJN            EVENT          SYSTEM    COMPONENT             CAUSE AND CORRECTIVE ACTION     ..
--COMPONENT CAUSE AND CORRECTIVE ACTION .. CODE 6 TO PREVENT RECURRENCE Refueling Outage Extension Axhibit G -Instructions for Preparation of Data Entry Sheets for Licensee Event Report CLER) File (NUREG-0161) . 5 Exhibit 1 -Same Source -ca 
NO.          DATE      TYPE'      (HOURS)      REASON 2      REACTOR          REPORT #        CODE 4    CODE 6                   TO PREVENT RECURRENCE
. . -------------
_.. _________
------* DOCKET NO.: UNIT: DATE: COMPLETED BY: TELEPHONE:
4092      7-1-97    F          744            Fc            4                                2422                      Refueling Outage Extension
50-311 Salem 2 08/15/97 R. Ritzman (609) 339-1445  
                                                                                                                                                                      -ca 1              2                                                        3                        4                                  5 F:  Forced      Reason                                                    Method:                  Axhibit 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        (NUREG-0161) .
E-Operator Training & License Examination                  Previous Outage F-Administrative                                          5-Load Reduction G-Operational Error (Explain)                            9-0ther H-Other (Explain)
 
. .
* DOCKET NO.:
UNIT:
50-311 Salem 2 DATE:  08/15/97 COMPLETED BY:  R. Ritzman TELEPHONE:  (609) 339-1445


==SUMMARY==
==SUMMARY==
OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION MONTH JULY 1997 The following items completed during July 1997 have been evaluated to determine:
OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION MONTH     JULY 1997 The following items completed during July 1997 have been evaluated to determine:
: 1. If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or 2. If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or 3. If the margin of safety as defined in the basis for any technical specification is reduced. The 10CFR50.59 Safety Evaluations showed that these items did not create a new safety hazard to the plant nor did they affect the safe shutdown of the reactor. These items did not change the plant effluent releases and did not alter the existing environmental impact. The 10CFR50.59 Safety Evaluations determined that no unreviewed safety or environmental questions are involved.
: 1. If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or
Design Changes Summary of Safety Evaluations 2EC-3341, Pkg. 1, Turbine Gland Sealing Steam and Leak Off System Modifications.
: 2. If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or
This design change modifies the Turbine Gland Sealing Steam and Leak-Off system. The modifications include a new control valve, a new gland seal steam header pressure controller, a strainer with a drain valves and blowdown line, new internals for the high pressure turbine cylinder warming header steam pressure control valves, new turbine gland local steam pressure gauges and root valves, new orifice couplings, and the relocation of free blow valve piping connections.
: 3. If the margin of safety as defined in the basis for any technical specification is reduced.
This design change does not negatively impact any accident response.
The 10CFR50.59 Safety Evaluations showed that these items did not create a new safety hazard to the plant nor did they affect the safe shutdown of the reactor. These items did not change the plant effluent releases and did not alter the existing environmental impact. The 10CFR50.59 Safety Evaluations determined that no unreviewed safety or environmental questions are involved.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
Design Changes     Summary of Safety Evaluations 2EC-3341, Pkg. 1, Turbine Gland Sealing Steam and Leak Off System Modifications. This design change modifies the Turbine Gland Sealing Steam and Leak-Off system. The modifications include a new control valve, a new gland seal steam header pressure controller, a strainer with a drain valves and blowdown line, new internals for the high pressure turbine cylinder warming header steam pressure control valves, new turbine gland local steam pressure gauges and root valves, new orifice couplings, and the relocation of free blow valve piping connections.
.. 2EC-3370, Pkg. 1, Replacement of the Main Steam Coil Drain Tank Dump Valves. This design change replaces the Main Steam Coil Drain Tank Dump Valves to improve leak tightness.
This design change does not negatively impact any accident response. This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
 
..
2EC-3370, Pkg. 1, Replacement of the Main Steam Coil Drain Tank Dump Valves. This design change replaces the Main Steam Coil Drain Tank Dump Valves to improve leak tightness.
This design change also installs "y" strainers before the dump valves and a manual isolation valve upstream of the strainer.
This design change also installs "y" strainers before the dump valves and a manual isolation valve upstream of the strainer.
This design change does not negatively impact any accident response.
This design change does not negatively impact any accident response. This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
2EC-3465, Pkg. 1, Circulating Water System Controls Modification. This design change modifies the Circulating Water system controls. This modification includes a revision to the circulator start logic, replacing the condenser waterbox vacuum switch, lowering the undervoltage time delay relay trip setpoint on the Circulating Water Pump Switchgear bus, and delaying the opening of the vacuum priming valves.
2EC-3465, Pkg. 1, Circulating Water System Controls Modification.
This design change does not negatively impact any accident response. This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
This design change modifies the Circulating Water system controls.
2EE-0046, Pkg. 1, Circulating Water Traveling Screen Modification. This design change modifies the 21A Circulating Water Traveling Screen. The modification includes replacing the fish baskets, adding spray nozzles, providing a shield over the top of the spray nozzles, and replacing the drive motor, gear reducer, and coupling.
This modification includes a revision to the circulator start logic, replacing the condenser waterbox vacuum switch, lowering the undervoltage time delay relay trip setpoint on the Circulating Water Pump Switchgear bus, and delaying the opening of the vacuum priming valves. This design change does not negatively impact any accident response.
This design change does not negatively impact any accident response. This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
2EE-0048, Pkg. 1, Circulating Water Traveling Screen Modification. This design change modifies the 22A Circulating Water Traveling Screen. The modification includes replacing the fish baskets, adding spray nozzles, providing a shield over the top of the spray nozzles, and replacing the drive motor, gear reducer, and coupling.
2EE-0046, Pkg. 1, Circulating Water Traveling Screen Modification.
 
This design change modifies the 21A Circulating Water Traveling Screen. The modification includes replacing the fish baskets, adding spray nozzles, providing a shield over the top of the spray nozzles, and replacing the drive motor, gear reducer, and coupling.
This design change does not negatively impact any accident
This design change does not negatively impact any accident response.
    .response. This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
2EE-0049, Pkg. 1, Circulating Water Traveling Screen Modification. This design change modifies the 22B Circulating Water Traveling Screen. The modification includes replacing the fish baskets, adding spray nozzles, providing a shield over the top of the spray nozzles, and replacing the drive motor, gear reducer, and coupling.
2EE-0048, Pkg. 1, Circulating Water Traveling Screen Modification.
This design change does not negatively impact any accident response. This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
This design change modifies the 22A Circulating Water Traveling Screen. The modification includes replacing the fish baskets, adding spray nozzles, providing a shield over the top of the spray nozzles, and replacing the drive motor, gear reducer, and coupling.
2EE-0051, Pkg. 1, Circulating Water Traveling Screen Modification. This design change modifies the 23B Circulating Water Traveling Screen. The modification includes replacing the fish baskets, adding spray nozzles, providing a shield over the top of the spray nozzles, and replacing the drive motor, gear reducer, and coupling.
This design change does not negatively impact any accident .response.
This design change does not negatively impact any accident response. This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
Temporary Modifications     Sununary of Safety Evaluations TMR 97-024, 22 Reactor Coolant Pump High Vibration Alarm Setpoint Change. This temporary modification changes the 22 Reactor Coolant Pump High Frame Vibration Alarm setpoint to 8.5 mils to clear a nuisance Control Room annunciator alarm, permit condition monitoring on the Reactor Coolant Pumps, and allow the 22 Reactor Coolant Pump to remain in service at a higher Reactor Coolant System temperature.
2EE-0049, Pkg. 1, Circulating Water Traveling Screen Modification.
This temporary modification does not negatively impact any accident response. This temporary modification does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety.
This design change modifies the 22B Circulating Water Traveling Screen. The modification includes replacing the fish baskets, adding spray nozzles, providing a shield over the top of the spray nozzles, and replacing the drive motor, gear reducer, and coupling.
Therefore, this temporary modification does not involve an Unreviewed Safety Question.
This design change does not negatively impact any accident response.
 
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
Procedures   Summary of Safety Evaluations TS2.SE-SU.ZZ-000l(Q), Rev. 11, Startup and Power Ascension Sequencing Procedure. This procedure controls the sequence in which Startup and Power Ascension test activities are performed. This procedure also establishes hold points to demonstrate acceptable plant performance, test results, and management review of performance and emergent issues. The safety evaluation associated with this revision determined the operational modes in which each of the postulated accidents in UFSAR Chapter 15 is applicable and identified the startup teat procedures and other activities that correspond to those postulated accidents.
2EE-0051, Pkg. 1, Circulating Water Traveling Screen Modification.
The safety evaluation concluded that this procedure revision does not negatively impact any accident response. This procedure revision does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this procedure revision does not involve an Unreviewed Safety Question.
This design change modifies the 23B Circulating Water Traveling Screen. The modification includes replacing the fish baskets, adding spray nozzles, providing a shield over the top of the spray nozzles, and replacing the drive motor, gear reducer, and coupling.
UFSAR Change Notices     Summary of Safety Evaluations There were no changes in this category implemented during July, 1997.
This design change does not negatively impact any accident response.
Deficiency Reports     Summary of Safety Evaluations There were no changes in this category implemented during July, 1997.
This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.
Other    Summary of Safety Evaluation There were no changes in this category implemented during July, 1997.
Temporary Modifications Sununary of Safety Evaluations TMR 97-024, 22 Reactor Coolant Pump High Vibration Alarm Setpoint Change. This temporary modification changes the 22 Reactor Coolant Pump High Frame Vibration Alarm setpoint to 8.5 mils to clear a nuisance Control Room annunciator alarm, permit condition monitoring on the Reactor Coolant Pumps, and allow the 22 Reactor Coolant Pump to remain in service at a higher Reactor Coolant System temperature.
 
This temporary modification does not negatively impact any accident response.
OPERATING DATA REPORT Docket No:   50-311¥ Date:         08/10/97 Compl*ated *by:   Robert Phillips                 Telephone:   339-2735 Operating Status
This temporary modification does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this temporary modification does not involve an Unreviewed Safety Question.
: 1. Unit Name                         Salem No. 2   Notes
Procedures Summary of Safety Evaluations TS2.SE-SU.ZZ-000l(Q), Rev. 11, Startup and Power Ascension Sequencing Procedure.
: 2. Reporting Period               July       1997
This procedure controls the sequence in which Startup and Power Ascension test activities are performed.
: 3. Licensed Thermal Power (MWt)               3411
This procedure also establishes hold points to demonstrate acceptable plant performance, test results, and management review of performance and emergent issues. The safety evaluation associated with this revision determined the operational modes in which each of the postulated accidents in UFSAR Chapter 15 is applicable and identified the startup teat procedures and other activities that correspond to those postulated accidents.
: 4. Nameplate Rating (Gross MWe)               1170
The safety evaluation concluded that this procedure revision does not negatively impact any accident response.
: 5. Design Electrical Rating (Net MWe)         1115
This procedure revision does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this procedure revision does not involve an Unreviewed Safety Question.
: 6. Maximum Dependable Capacity(Gross MWe) 1149
UFSAR Change Notices Summary of Safety Evaluations There were no changes in this category implemented during July, 1997. Deficiency Reports Summary of Safety Evaluations Other There were no changes in this category implemented during July, 1997. Summary of Safety Evaluation There were no changes in this category implemented during July, 1997.
: 7. Maximum Dependable Capacity (Net.MWe) 1106
OPERATING DATA REPORT Docket No: Date: Compl*ated  
: 8. If Changes Occur in Capacity Ratings (items 3 through 7) since Last Report, Give Reason.~~~~N~A-=-~~~~~~~~~~~~~~~~~~~~~--
*by: Robert Phillips Telephone:
: 9. Power Level to Which Restricted, if any (Net MWe)           N/A
Operating Status 1. Unit Name Salem No. 2 Notes 2. Reporting Period July 1997 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 50-311¥ 08/10/97 339-2735 8. If Changes Occur in Capacity Ratings (items 3 through 7) since Last Report, Give
: 10. Reasons for Restrictions, if any   ~~~~~~~~N~=A'--~~~~~~~~~~
: 9. Power Level to Which Restricted, if any (Net MWe) N/A 10. Reasons for Restrictions, if any
This Month  Year to Date    cumulative
: 11. 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 0 0 0 0 0 0 -22577 0 0 0 0 100 Year to Date 5087 0 0 0 0 0 0 -73517 o* 0 0 0 100 cumulative 150036 78083.6 0 75229.5 0 187781005 78648598 74622814 50.l 50.l 45.0 44.6 36.2 24. Shutdowns scheduled over next 6 months (type, date and duration of each) NA 25. If shutdown at end of Report Period, Estimated Date of startup: 3rd quarter 1997. 8-1-7.R2 
: 11. Hours in Reporting Period             744          5087          150036
.ERAGE DAILY UNIT POWER LE"' Docket No.: 50-311 Unit Name: Salem #2 Date: 08/10/97 Completed by: Robert Phillips Telephone:
: 12. No. of Hrs. Rx. was Critical           0              0          78083.6
339-2735 Month July 1997 Day Average Daily Power Level Day Average Daily Power Level (MWe-NET) (MWe-NET) 1 0 17 0 2 0 18 0 3 0 19 0 4 0 20 0 5 0 21 0 6 0 22 0 7 0 23 0 8 0 24 0 9 0 25 0 10 0 26 0 11 0 27 0 12 0 28 0 13 0 29 0 14 0 30 0 15 0 31 0 16 0 P. 8.1-7 Rl}}
: 13. Reactor Reserve Shutdown Hrs.           0              0              0
: 14. Hours Generator On-Line                 0              0          75229.5
: 15. Unit Reserve Shutdown Hours             0              0              0
: 16. Gross Thermal Energy Generated (MWH)                         0              0        187781005
: 17. Gross Elec. Energy Generated (MWH)                         0              0        78648598
: 18. Net Elec. Energy Gen. (MWH)         -22577        -73517        74622814
: 19. Unit Service Factor                     0              o*            50.l
: 20. Unit Availability Factor               0              0              50.l
: 21. Unit Capacity Factor (using MDC Net)                   0                0            45.0
: 22. Unit Capacity Factor (using DER Net)                   0                0            44.6
: 23. Unit Forced Outage Rate               100            100             36.2
: 24. Shutdowns scheduled over next 6 months (type, date and duration of each)
NA
: 25. If shutdown at end of Report Period, Estimated Date of startup:
3rd quarter 1997.
8-1-7.R2
 
                          .ERAGE DAILY UNIT POWER LE"'
Docket No.:   50-311 Unit Name:     Salem #2 Date:         08/10/97 Completed by:     Robert Phillips                   Telephone:     339-2735 Month     July       1997 Day Average Daily Power Level             Day Average Daily Power Level (MWe-NET)                               (MWe-NET) 1             0                           17             0 2             0                           18             0 3             0                           19             0 4             0                           20             0 5             0                           21             0 6             0                           22             0 7             0                           23             0 8             0                           24             0 9             0                           25             0 10             0                           26             0 11             0                           27             0 12             0                           28             0 13             0                           29             0 14             0                           30             0 15             0                             31             0 16             0 P. 8.1-7 Rl}}

Revision as of 08:17, 21 October 2019

Monthly Operating Rept for Jul 1997 for Salem Generating Station,Unit 2.W/970815 Ltr
ML18102B525
Person / Time
Site: Salem PSEG icon.png
Issue date: 07/31/1997
From: Garchow D, Phillips R, Tisdel D
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LR-N970518, NUDOCS 9708250118
Download: ML18102B525 (10)
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, Public ::>ervice Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit AUG lJi 1997 LR-N970518 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn: Document Control Desk MONTHLY OPERATING REPORT SALEM UNIT NO. 2 DOCKET NO. 50-311 In compliance with Section 6.9.1.6, Reporting Requirements for the Salem Technical Specifications, the original monthly operating report for July, 1997, is attached.

I F. Gare ow General Manager -

Salem Operations RAR:tcp Enclosures c Mr. H. J. Miller Regional Administrator USNRC, Region 1 475 Allendale Road King of Prussia, PA 19046

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SALEM GENERATING STATION MONTHLY OPERATING

SUMMARY

- UNIT 2 JULY 1997 SALEM UNIT 2 The unit remained shutdown for the entire period. According to commitments from PSE&G and a subsequent confirmatory action letter from the NRC, the unit will remain shutdown pending completion of the following actions:

  • Appropriately address long standing equipment reliability and operability issues.
  • After the work is completed, conduct a restart readiness review to determine for ourselves the ability of the unit to operate in a safe, event free manner.
  • After the restart review, meet with the NRC and communicate the results of that review.

Refueling Information Docket No. 50-311

,Month:, July, 1997 Unit Name: Salem 2 Contact: D. Tisdel Telephone: 609-339-1538 Month: July, 1997

1. Refueling information has changed from last month: Yes: No: X
2. Scheduled date for next refueling: Currently in outage.

Scheduled date for restart following refueling: To Be Determined

3. a. Will Technical Specification changes or other license amendments be required?

Yes: x No: Not Determined to Date:

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

Yes: X (for upcoming cycle) No: If no, when is it scheduled?

4. Scheduled date (s) for submitting proposed licensing action: N/A -

previously submitted

5. Important licensing considerations associated with refueling:
6. Number of Fuel Assemblies:
a. Incore: 193
b. In Spent Fuel Storage: 584
7. Present Licensed spent fuel storage capacity: 1632 Future spent fuel storage capacity: 1632
8. Date of last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: October, 2016

UNIT SHUTDOYN AND POYER REDUCTIONS REPORT MONTH July 1997 DOCKET NO.: ~5~0~-~31~1~--

UNIT NAME: Salem #2 DATE: 08-10-97 COMPLETED BY: Robert Phillips TELEPHONE: 609-339-2735 -

METHOD OF SHUTTING LICENSE -

DURATION DOIJN EVENT SYSTEM COMPONENT CAUSE AND CORRECTIVE ACTION ..

NO. DATE TYPE' (HOURS) REASON 2 REACTOR REPORT # CODE 4 CODE 6 TO PREVENT RECURRENCE

_.. _________

4092 7-1-97 F 744 Fc 4 2422 Refueling Outage Extension

-ca 1 2 3 4 5 F: Forced Reason Method: Axhibit 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 (NUREG-0161) .

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

. .

  • DOCKET NO.:

UNIT:

50-311 Salem 2 DATE: 08/15/97 COMPLETED BY: R. Ritzman TELEPHONE: (609) 339-1445

SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION MONTH JULY 1997 The following items completed during July 1997 have been evaluated to determine:

1. If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or
2. If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or
3. If the margin of safety as defined in the basis for any technical specification is reduced.

The 10CFR50.59 Safety Evaluations showed that these items did not create a new safety hazard to the plant nor did they affect the safe shutdown of the reactor. These items did not change the plant effluent releases and did not alter the existing environmental impact. The 10CFR50.59 Safety Evaluations determined that no unreviewed safety or environmental questions are involved.

Design Changes Summary of Safety Evaluations 2EC-3341, Pkg. 1, Turbine Gland Sealing Steam and Leak Off System Modifications. This design change modifies the Turbine Gland Sealing Steam and Leak-Off system. The modifications include a new control valve, a new gland seal steam header pressure controller, a strainer with a drain valves and blowdown line, new internals for the high pressure turbine cylinder warming header steam pressure control valves, new turbine gland local steam pressure gauges and root valves, new orifice couplings, and the relocation of free blow valve piping connections.

This design change does not negatively impact any accident response. This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.

..

2EC-3370, Pkg. 1, Replacement of the Main Steam Coil Drain Tank Dump Valves. This design change replaces the Main Steam Coil Drain Tank Dump Valves to improve leak tightness.

This design change also installs "y" strainers before the dump valves and a manual isolation valve upstream of the strainer.

This design change does not negatively impact any accident response. This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.

2EC-3465, Pkg. 1, Circulating Water System Controls Modification. This design change modifies the Circulating Water system controls. This modification includes a revision to the circulator start logic, replacing the condenser waterbox vacuum switch, lowering the undervoltage time delay relay trip setpoint on the Circulating Water Pump Switchgear bus, and delaying the opening of the vacuum priming valves.

This design change does not negatively impact any accident response. This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.

2EE-0046, Pkg. 1, Circulating Water Traveling Screen Modification. This design change modifies the 21A Circulating Water Traveling Screen. The modification includes replacing the fish baskets, adding spray nozzles, providing a shield over the top of the spray nozzles, and replacing the drive motor, gear reducer, and coupling.

This design change does not negatively impact any accident response. This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.

2EE-0048, Pkg. 1, Circulating Water Traveling Screen Modification. This design change modifies the 22A Circulating Water Traveling Screen. The modification includes replacing the fish baskets, adding spray nozzles, providing a shield over the top of the spray nozzles, and replacing the drive motor, gear reducer, and coupling.

This design change does not negatively impact any accident

.response. This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.

2EE-0049, Pkg. 1, Circulating Water Traveling Screen Modification. This design change modifies the 22B Circulating Water Traveling Screen. The modification includes replacing the fish baskets, adding spray nozzles, providing a shield over the top of the spray nozzles, and replacing the drive motor, gear reducer, and coupling.

This design change does not negatively impact any accident response. This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.

2EE-0051, Pkg. 1, Circulating Water Traveling Screen Modification. This design change modifies the 23B Circulating Water Traveling Screen. The modification includes replacing the fish baskets, adding spray nozzles, providing a shield over the top of the spray nozzles, and replacing the drive motor, gear reducer, and coupling.

This design change does not negatively impact any accident response. This design change does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this design change does not involve an Unreviewed Safety Question.

Temporary Modifications Sununary of Safety Evaluations TMR 97-024, 22 Reactor Coolant Pump High Vibration Alarm Setpoint Change. This temporary modification changes the 22 Reactor Coolant Pump High Frame Vibration Alarm setpoint to 8.5 mils to clear a nuisance Control Room annunciator alarm, permit condition monitoring on the Reactor Coolant Pumps, and allow the 22 Reactor Coolant Pump to remain in service at a higher Reactor Coolant System temperature.

This temporary modification does not negatively impact any accident response. This temporary modification does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety.

Therefore, this temporary modification does not involve an Unreviewed Safety Question.

Procedures Summary of Safety Evaluations TS2.SE-SU.ZZ-000l(Q), Rev. 11, Startup and Power Ascension Sequencing Procedure. This procedure controls the sequence in which Startup and Power Ascension test activities are performed. This procedure also establishes hold points to demonstrate acceptable plant performance, test results, and management review of performance and emergent issues. The safety evaluation associated with this revision determined the operational modes in which each of the postulated accidents in UFSAR Chapter 15 is applicable and identified the startup teat procedures and other activities that correspond to those postulated accidents.

The safety evaluation concluded that this procedure revision does not negatively impact any accident response. This procedure revision does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety. Therefore, this procedure revision does not involve an Unreviewed Safety Question.

UFSAR Change Notices Summary of Safety Evaluations There were no changes in this category implemented during July, 1997.

Deficiency Reports Summary of Safety Evaluations There were no changes in this category implemented during July, 1997.

Other Summary of Safety Evaluation There were no changes in this category implemented during July, 1997.

OPERATING DATA REPORT Docket No: 50-311¥ Date: 08/10/97 Compl*ated *by: Robert Phillips Telephone: 339-2735 Operating Status

1. Unit Name Salem No. 2 Notes
2. Reporting Period July 1997
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.~~~~N~A-=-~~~~~~~~~~~~~~~~~~~~~--
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

11. Hours in Reporting Period 744 5087 150036
12. No. of Hrs. Rx. was Critical 0 0 78083.6
13. Reactor Reserve Shutdown Hrs. 0 0 0
14. Hours Generator On-Line 0 0 75229.5
15. Unit Reserve Shutdown Hours 0 0 0
16. Gross Thermal Energy Generated (MWH) 0 0 187781005
17. Gross Elec. Energy Generated (MWH) 0 0 78648598
18. Net Elec. Energy Gen. (MWH) -22577 -73517 74622814
19. Unit Service Factor 0 o* 50.l
20. Unit Availability Factor 0 0 50.l
21. Unit Capacity Factor (using MDC Net) 0 0 45.0
22. Unit Capacity Factor (using DER Net) 0 0 44.6
23. Unit Forced Outage Rate 100 100 36.2
24. Shutdowns scheduled over next 6 months (type, date and duration of each)

NA

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

3rd quarter 1997.

8-1-7.R2

.ERAGE DAILY UNIT POWER LE"'

Docket No.: 50-311 Unit Name: Salem #2 Date: 08/10/97 Completed by: Robert Phillips Telephone: 339-2735 Month July 1997 Day Average Daily Power Level Day Average Daily Power Level (MWe-NET) (MWe-NET) 1 0 17 0 2 0 18 0 3 0 19 0 4 0 20 0 5 0 21 0 6 0 22 0 7 0 23 0 8 0 24 0 9 0 25 0 10 0 26 0 11 0 27 0 12 0 28 0 13 0 29 0 14 0 30 0 15 0 31 0 16 0 P. 8.1-7 Rl

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