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* Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn.: Document Control Desk MONTHLY OPERATING REPORT SALEMNO. 1 DOCKET NO: 50-272 August 15, 1996 In compliance with Section 6.9.1.6, Reporting Requirements for the Salem Technical Specifications, the original copy of the monthly operating report for the month of July is being sent to you. RH:vls Enclosures C Mr. H.J. Miller Sincerely yours, David F. Garchow General Manager -Salem Operations Regional Administrator USNRC, Region I
PS~G
* 475 Allendale Road King of Prussia, PA 19046 --*-------------9608200121 960731 '\ PDR ADOCK 05000272 R PDR The power is in your hands. 95-2168 REV. 6/94
* Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit August 15, 1996 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn.: Document Control Desk MONTHLY OPERATING REPORT SALEMNO. 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 report for the month of July is being sent to you.
* OPERATING DATA REPORT e Docket No: 50-272 Date: 08/10/96 Completed by: Robert Phillips Telephone:
Sincerely yours, David F. Garchow General Manager -
339-2735 Operating Status 1. Unit Name Salem No. 1 Notes 2. Reporting Period July 1996 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
Salem Operations RH:vls Enclosures C       Mr. H.J. Miller Regional Administrator USNRC, Region I
: 9. Power Level to Which Restricted, if any (Net MWe) 10. Reasons for Restrictions, if any 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 N A This Month Year to Date 744 5111 0 0 0 0 0 0 0 0 0 0 0 0 -2600 -16221 0 0 0 0 0 0 0 0 100 100 N/A Cumulative 167328 104380.5 0 100388.3 0 318062229.2 105301000 100188692 60.0 60.0 54.1 53.7 24.2 24. Shutdowns scheduled over next 6 months (type, date and duration of each) Forced outage. 25. If shutdown at end of Report Period, Estimated Date of Startup: To be determined.
* 475 Allendale Road King of Prussia, PA 19046 IE~t- 1 11
8-1-7.R2 AVERA.DAILY UNIT POWER LEVEL Docket No. : 50-272 Unit Name: Salem #1 Date: 08/10/96 Completed by: Robert Phillips Telephone:
~ --              * - - --     ~~ -------- -     ~--
339-2735 Month July 1996 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 NO. DATE 4092 07-01-96 F 2 F: Forced S: Scheduled DURATION TYPE' (HOURS) REASON 1 744 F c Reason A-Equipment Failure (explain) 8-Maintenance or Test C-Refueling D-Requlatory Restriction 4 UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH June 1996 METHOD OF SHUTTING DOWN REACTOR LICENSE EVENT REPORT # ----------
9608200121 960731                               '\
HJ 3 Method: 1-Manual 2-Manual Scram SYSTEM CODE 4 E-Operator Training & License Examination F-Administrative 3-Automatic Scram 4-Continuation of Previous Outage 5-Load Reduction 9-0ther G-Operational Error (Explain)
PDR     ADOCK 05000272 R                         PDR The power is in your hands.
H-Other (Explain)
95-2168 REV. 6/94
COMPONENT CODE 5 DOCKET NO. UNIT NAME DATE COMPLETED BY TELEPHONE:
* OPERATING DATA REPORT   e Docket No:   50-272 Date:         08/10/96 Completed by:     Robert Phillips                 Telephone:   339-2735 Operating Status
50-272 Salem #1 08/10/96 Robert Phillips.
: 1. Unit Name                         Salem No. 1   Notes
609-339"*
: 2. Reporting Period             July       1996
2735 CAUSE AND CORRECTIVE ACTION TO PREVENT RECURRENCE HTEXCH Nuclear Steam Generator Systems. 4 Exhibit G -Instructions for Preparation of Data Entry Sheets for Licensee Event Report CLER) File (NUREG-0161) 5 Exhibit 1 -Same Source
: 3. Licensed Thermal Power (MWt)             3411
* 10CFR50.59 EVALUATIONS MONTH: JULY 1996 DOCKET NO: UNIT NAME: 50-272 SALEM 1 CONTACT: TELEPHONE:
: 4. Nameplate Rating (Gross MWe)             1170
R. HELLER 609-339-5162 The following items were evaluated in accordance with the provisions of the Code of Federal Regulations 10CFR50.59.
: 5. Design Electrical Rating (Net MWe)       1115
The Station Operations Review Committee has reviewed and concurs with these evaluations.
: 6. Maximum Dependable Capacity(Gross MWe) 1149
: 1. Design Change Packages (DCP) lEC-3505, Pkg. 2 2. Evaluations Radiological Safety "Control Area Emergency Air Conditioning System (EACS system) Condensate Drain Trap Seal" Rev. 0 -This DCP removed the existing Seismic Class 3 Condensate drain line from the Unit 2 EACS, located in the Auxiliary Building, elevation 122' -0", and installed a new Seismic Class 1 Condensate Drain piping with a fillable trap and a Seismic Class 3 site glass. Also this DCP removed two (2) 1" EACS housing drain lines located on the downstream side of the charcoal filter. There is no reduction in the margin of safety as defined in the basis for any Technical Specification because (1) there is no credible event analyzed in Chapter 15 of the UFSAR which can cause an unacceptable environment in the control room since core alterations are currently restricted; (2) fuel movement inside the Fuel Handling Building (FHB) is restricted unless FHB ventilation is operable, and (3) the beyond-design-basis event which might impact the control room (rupture of an ammonium hydroxide tanker) is precluded by procedure. (SORC 96-033) "Radiological Safety Organizational Reporting Structure" (Organizational Change) -The revision changes the structural reporting requirements for Radiological Safety, encompassing the functional areas of emergency preparedness, radiation protection, chemistry, dosimetry and radiological instruments, from Nuclear Operations Support to the Nuclear Training Center. The functions and responsibilities remain the same, although the groups will have different reporting requirements.
: 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
: 12. Hours in Reporting Period           744          5111          167328
: 12. No. of Hrs. Rx. was Critical           0              0          104380.5
: 13. Reactor Reserve Shutdown Hrs.         0              0              0
: 14. Hours Generator On-Line               0              0          100388.3
: 15. Unit Reserve Shutdown Hours           0              0              0
: 16. Gross Thermal Energy Generated (MWH)                         0              0      318062229.2
: 17. Gross Elec. Energy Generated (MWH)                         0              0        105301000
: 18. Net Elec. Energy Gen. (MWH)         -2600        -16221        100188692
: 19. Unit Service Factor                   0              0              60.0
: 20. Unit Availability Factor               0              0              60.0
: 21. Unit Capacity Factor (using MDC Net)                   0                0            54.1
: 22. Unit Capacity Factor (using DER Net)                   0                0            53.7
: 23. Unit Forced Outage Rate             100             100             24.2
: 24. Shutdowns scheduled over next 6 months (type, date and duration of each)
Forced outage.
: 25. If shutdown at end of Report Period, Estimated Date of Startup:
To be determined.
8-1-7.R2
 
AVERA.DAILY UNIT POWER LEVEL   ~
Docket No. :   50-272 Unit Name:     Salem #1 Date:         08/10/96 Completed by:     Robert Phillips                 Telephone:     339-2735 Month       July       1996 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
 
UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH June       1996                                           DOCKET NO. 50-272 UNIT NAME    Salem #1 DATE  08/10/96 COMPLETED BY    Robert Phillips.
TELEPHONE:  609-339"* 2735 METHOD OF SHUTTING        LICENSE DURATION                      DOWN            EVENT            SYSTEM    COMPONENT              CAUSE AND CORRECTIVE ACTION NO.        DATE      TYPE'      (HOURS)      REASON 1 REACTOR          REPORT #        CODE 4    CODE 5                    TO PREVENT RECURRENCE 4092    07-01-96  F          744            F c            4              ----------      HJ        HTEXCH           Nuclear Steam Generator Systems.
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 8-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)
 
10CFR50.59 EVALUATIONS
* DOCKET NO:            50-272 MONTH: JULY 1996                                       UNIT NAME:             SALEM 1 CONTACT:           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.
: 1. Design Change Packages (DCP) lEC-3505, Pkg. 2             "Control Area Emergency Air Conditioning System (EACS system) Condensate Drain Trap Seal" Rev. 0 - This DCP removed the existing Seismic Class 3 Condensate drain line from the Unit 2 EACS, located in the Auxiliary Building, elevation 122' -0", and installed a new Seismic Class 1 Condensate Drain piping with a fillable trap and a Seismic Class 3 site glass. Also this DCP removed two (2) 1" EACS housing drain lines located on the downstream side of the charcoal filter.
There is no reduction in the margin of safety as defined in the basis for any Technical Specification because (1) there is no credible event analyzed in Chapter 15 of the UFSAR which can cause an unacceptable environment in the control room since core alterations are currently restricted; (2) fuel movement inside the Fuel Handling Building (FHB) is restricted unless FHB ventilation is operable, and (3) the beyond-design-basis event which might impact the control room (rupture of an ammonium hydroxide tanker) is precluded by procedure.
(SORC 96-033)
: 2. Evaluations Radiological Safety          "Radiological Safety Organizational Reporting Structure" (Organizational Change) - The revision changes the structural reporting requirements for Radiological Safety, encompassing the functional areas of emergency preparedness, radiation protection, chemistry, dosimetry and radiological instruments, from Nuclear Operations Support to the Nuclear Training Center. The functions and responsibilities remain the same, although the groups will have different reporting requirements.
The qualifications of the Director of Nuclear Training and the
The qualifications of the Director of Nuclear Training and the
* 10CFR50.59 EVALUATIONS MONTH: JULY 1996 DOCKET NO: UNIT NAME: 50-272 SALEM 1 R.HELLER 609-339-5162
* 10CFR50.59 EVALUATIONS                         DOCKET NO:            50-272 MONTH: JULY 1996                                 UNIT NAME:           SALEM 1 CONTACT:          R.HELLER TELEPHONE:            609-339-5162 General Manager of Nuclear Services are similar in nature, although there is a lesser requirement for time requirement for the director (8 versus I 0 years). This lesser requirement is sufficiently rigorous and exceeds the ANSI standard so the net effect will not diminish quality. The proposed changes do not relate to design criteria, specifications, or operations of systems or components relating to the fuel cladding, Reactor Coolant System (RCS) boundary, or containment, and do not address any margin of safety. Therefore, the proposed changes do not reduce the margin of safety as defined in the bases for any Technical Specification. (SORC 96-095)
: 3. UFSAR S96-086 S96-093 CONTACT: TELEPHONE:
: 3. UFSAR S96-086          "Minimum Neutron Count Rate on Source Range Instrumentation" - This change revises Section 4.2.3 .1 reducing the minimum required capability of the neutron sources from providing 2 counts per second to greater than I cps on the source range channels. The proposed changes will not impact the initial subcritical boration requirements or the capability of the instrumentation to detect changes in core multiplication.
General Manager of Nuclear Services are similar in nature, although there is a lesser requirement for time requirement for the director (8 versus I 0 years). This lesser requirement is sufficiently rigorous and exceeds the ANSI standard so the net effect will not diminish quality. The proposed changes do not relate to design criteria, specifications, or operations of systems or components relating to the fuel cladding, Reactor Coolant System (RCS) boundary, or containment, and do not address any margin of safety. Therefore, the proposed changes do not reduce the margin of safety as defined in the bases for any Technical Specification. (SORC 96-095) "Minimum Neutron Count Rate on Source Range Instrumentation" -This change revises Section 4.2.3 .1 reducing the minimum required capability of the neutron sources from providing 2 counts per second to greater than I cps on the source range channels.
The source range monitors' operability requirements for all affected Modes remain unchanged and are not based on minimal indicated neutron count rate but on the prescribed channel calibrations and functional test. The change does not reduce the requirement that both channels provide continuous visual indication of the neutron flux and core multiplication. The proposed change does not reduce the margin of safety as defined in the basis for any Technical Specification.
The proposed changes will not impact the initial subcritical boration requirements or the capability of the instrumentation to detect changes in core multiplication.
(SORC 96-096)
The source range monitors' operability requirements for all affected Modes remain unchanged and are not based on minimal indicated neutron count rate but on the prescribed channel calibrations and functional test. The change does not reduce the requirement that both channels provide continuous visual indication of the neutron flux and core multiplication.
S96-093          "SCN Service Water System Modifications" - This safety evaluation supports changes to sections of the UFSAR relative to the Salem service water system. These changes are required to address issues raised by problem reports. The changes include corrections to address discrepancies I clarifications determined to be necessary as a result of both internal and external system reviews. The external review
The proposed change does not reduce the margin of safety as defined in the basis for any Technical Specification. (SORC 96-096) "SCN Service Water System Modifications" -This safety evaluation supports changes to sections of the UFSAR relative to the Salem service water system. These changes are required to address issues raised by problem reports. The changes include corrections to address discrepancies I clarifications determined to be necessary as a result of both internal and external system reviews. The external review
* 10CFR50.59 EVALUATIONS                         DOCKET NO:          50-272 MONTH: JULY 1996                               UNIT NAME:           SALEM 1 CONTACT:          R.HELLER TELEPHONE:          609-339-5162 included comments from the NRC during the 5/96 licensing basis audit. The acceptance criteria for postulated design basis accidents affected by the Service Water system define the acceptable margin of safety. This proposal does not affect the design limits of the Service Water system, or design limits of components affected by this system. It is therefore concluded that this proposal will not reduce the margin of safety as defined in the basis for any Technical Specification. (SORC 96-098)
* 10CFR50.59 EVALUATIONS MONTH: JULY 1996 DOCKET NO: UNIT NAME: 50-272 SALEM 1 R.HELLER 609-339-5162 CONTACT: TELEPHONE:
 
included comments from the NRC during the 5/96 licensing basis audit. The acceptance criteria for postulated design basis accidents affected by the Service Water system define the acceptable margin of safety. This proposal does not affect the design limits of the Service Water system, or design limits of components affected by this system. It is therefore concluded that this proposal will not reduce the margin of safety as defined in the basis for any Technical Specification. (SORC 96-098)
REFUELING INFORMATION
* REFUELING INFORMATION MONTH: JULY 1996 MONTH : JULY 1996 DOCKET NO: UNIT NAME: CONTACT: TELEPHONE:
* DOCKET NO:         50-272 MONTH: JULY 1996                                       UNIT NAME:         SALEM 1 CONTACT:         R.HELLER TELEPHONE:          609-339-5162 MONTH : JULY 1996 Refueling information has changed from last month: YES_._.NO. X           .
50-272 SALEM 1 R.HELLER 609-339-5162 Refueling information has changed from last month: YES_. _.NO. X . Scheduled date for next refueling: (to be determined)
Scheduled date for next refueling: (to be determined)
Scheduled date for restart following refueling: (to be determined)
Scheduled date for restart following refueling: (to be determined)
: a. Will Technical Specification changes or other license amendments be required?
: a. Will Technical Specification changes or other license amendments be required?
YES.
YES.         .NO~.-~
NOT DETERMINED TO DATE . X . b. Has the reload fuel design been reviewed by the Station Operating Review Committee?
NOT DETERMINED TO DATE .             X   .
YES. . NO. X . If no, when is it scheduled? (to be determined ) Scheduled date(s) for submitting proposed licensing action: . n/a Important licensing considerations associated with refueling:
: b. Has the reload fuel design been reviewed by the Station Operating Review Committee?
YES.       . NO. X   .
If no, when is it scheduled? (to be determined )
Scheduled date(s) for submitting proposed licensing action: .     n/a Important licensing considerations associated with refueling:
Number ofFuel Assemblies:
Number ofFuel Assemblies:
: a. Incore b. In Spent Fuel Storage Present licensed spent fuel storage capacity:
: a. Incore                                                                   0
Future spent fuel storage capacity:
: b. In Spent Fuel Storage                                                   989 Present licensed spent fuel storage capacity:                             1632 Future spent fuel storage capacity: 1                                   1632 Date oflast refueling that can be discharged to the spent fuel pool assuming the present licensed capacity:             September 2008
1 Date oflast refueling that can be discharged to the spent fuel pool assuming the present licensed capacity:
0 989 1632 1632 September 2008
* SALEM GENERATING STATION MONTHLY OPERATING  
* SALEM GENERATING STATION MONTHLY OPERATING  


==SUMMARY==
==SUMMARY==
  -UNIT 1 JULY 1996 SALEMUNITNO.
  - UNIT 1 JULY 1996 SALEMUNITNO. 1 The Unit is in a refueling and Steam Generator replacement outage and 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:
1 The Unit is in a refueling and Steam Generator replacement outage and 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}}

Revision as of 09:38, 21 October 2019

Monthly Operating Rept for July 1996 for Salem Unit 1. W/960815 Ltr
ML18102A300
Person / Time
Site: Salem PSEG icon.png
Issue date: 07/31/1996
From: Garchow D, Phillips R
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9608200121
Download: ML18102A300 (9)
v • d • e


Text

PS~G

  • Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit August 15, 1996 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn.: Document Control Desk MONTHLY OPERATING REPORT SALEMNO. 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 report for the month of July is being sent to you.

Sincerely yours, David F. Garchow General Manager -

Salem Operations RH:vls Enclosures C Mr. H.J. Miller Regional Administrator USNRC, Region I

  • 475 Allendale Road King of Prussia, PA 19046 IE~t- 1 11

~ -- * - - -- ~~ -------- - ~--

9608200121 960731 '\

PDR ADOCK 05000272 R PDR The power is in your hands.

95-2168 REV. 6/94

  • OPERATING DATA REPORT e Docket No: 50-272 Date: 08/10/96 Completed by: Robert Phillips Telephone: 339-2735 Operating Status
1. Unit Name Salem No. 1 Notes
2. Reporting Period July 1996
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
12. Hours in Reporting Period 744 5111 167328
12. No. of Hrs. Rx. was Critical 0 0 104380.5
13. Reactor Reserve Shutdown Hrs. 0 0 0
14. Hours Generator On-Line 0 0 100388.3
15. Unit Reserve Shutdown Hours 0 0 0
16. Gross Thermal Energy Generated (MWH) 0 0 318062229.2
17. Gross Elec. Energy Generated (MWH) 0 0 105301000
18. Net Elec. Energy Gen. (MWH) -2600 -16221 100188692
19. Unit Service Factor 0 0 60.0
20. Unit Availability Factor 0 0 60.0
21. Unit Capacity Factor (using MDC Net) 0 0 54.1
22. Unit Capacity Factor (using DER Net) 0 0 53.7
23. Unit Forced Outage Rate 100 100 24.2
24. Shutdowns scheduled over next 6 months (type, date and duration of each)

Forced outage.

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

To be determined.

8-1-7.R2

AVERA.DAILY UNIT POWER LEVEL ~

Docket No. : 50-272 Unit Name: Salem #1 Date: 08/10/96 Completed by: Robert Phillips Telephone: 339-2735 Month July 1996 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

UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH June 1996 DOCKET NO. 50-272 UNIT NAME Salem #1 DATE 08/10/96 COMPLETED BY Robert Phillips.

TELEPHONE: 609-339"* 2735 METHOD OF SHUTTING LICENSE DURATION DOWN EVENT SYSTEM COMPONENT CAUSE AND CORRECTIVE ACTION NO. DATE TYPE' (HOURS) REASON 1 REACTOR REPORT # CODE 4 CODE 5 TO PREVENT RECURRENCE 4092 07-01-96 F 744 F c 4 ---------- HJ HTEXCH Nuclear Steam Generator Systems.

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 8-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)

10CFR50.59 EVALUATIONS

  • DOCKET NO: 50-272 MONTH: JULY 1996 UNIT NAME: SALEM 1 CONTACT: 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.
1. Design Change Packages (DCP) lEC-3505, Pkg. 2 "Control Area Emergency Air Conditioning System (EACS system) Condensate Drain Trap Seal" Rev. 0 - This DCP removed the existing Seismic Class 3 Condensate drain line from the Unit 2 EACS, located in the Auxiliary Building, elevation 122' -0", and installed a new Seismic Class 1 Condensate Drain piping with a fillable trap and a Seismic Class 3 site glass. Also this DCP removed two (2) 1" EACS housing drain lines located on the downstream side of the charcoal filter.

There is no reduction in the margin of safety as defined in the basis for any Technical Specification because (1) there is no credible event analyzed in Chapter 15 of the UFSAR which can cause an unacceptable environment in the control room since core alterations are currently restricted; (2) fuel movement inside the Fuel Handling Building (FHB) is restricted unless FHB ventilation is operable, and (3) the beyond-design-basis event which might impact the control room (rupture of an ammonium hydroxide tanker) is precluded by procedure.

(SORC 96-033)

2. Evaluations Radiological Safety "Radiological Safety Organizational Reporting Structure" (Organizational Change) - The revision changes the structural reporting requirements for Radiological Safety, encompassing the functional areas of emergency preparedness, radiation protection, chemistry, dosimetry and radiological instruments, from Nuclear Operations Support to the Nuclear Training Center. The functions and responsibilities remain the same, although the groups will have different reporting requirements.

The qualifications of the Director of Nuclear Training and the

  • 10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH: JULY 1996 UNIT NAME: SALEM 1 CONTACT: R.HELLER TELEPHONE: 609-339-5162 General Manager of Nuclear Services are similar in nature, although there is a lesser requirement for time requirement for the director (8 versus I 0 years). This lesser requirement is sufficiently rigorous and exceeds the ANSI standard so the net effect will not diminish quality. The proposed changes do not relate to design criteria, specifications, or operations of systems or components relating to the fuel cladding, Reactor Coolant System (RCS) boundary, or containment, and do not address any margin of safety. Therefore, the proposed changes do not reduce the margin of safety as defined in the bases for any Technical Specification. (SORC 96-095)
3. UFSAR S96-086 "Minimum Neutron Count Rate on Source Range Instrumentation" - This change revises Section 4.2.3 .1 reducing the minimum required capability of the neutron sources from providing 2 counts per second to greater than I cps on the source range channels. The proposed changes will not impact the initial subcritical boration requirements or the capability of the instrumentation to detect changes in core multiplication.

The source range monitors' operability requirements for all affected Modes remain unchanged and are not based on minimal indicated neutron count rate but on the prescribed channel calibrations and functional test. The change does not reduce the requirement that both channels provide continuous visual indication of the neutron flux and core multiplication. The proposed change does not reduce the margin of safety as defined in the basis for any Technical Specification.

(SORC 96-096)

S96-093 "SCN Service Water System Modifications" - This safety evaluation supports changes to sections of the UFSAR relative to the Salem service water system. These changes are required to address issues raised by problem reports. The changes include corrections to address discrepancies I clarifications determined to be necessary as a result of both internal and external system reviews. The external review

  • 10CFR50.59 EVALUATIONS DOCKET NO: 50-272 MONTH: JULY 1996 UNIT NAME: SALEM 1 CONTACT: R.HELLER TELEPHONE: 609-339-5162 included comments from the NRC during the 5/96 licensing basis audit. The acceptance criteria for postulated design basis accidents affected by the Service Water system define the acceptable margin of safety. This proposal does not affect the design limits of the Service Water system, or design limits of components affected by this system. It is therefore concluded that this proposal will not reduce the margin of safety as defined in the basis for any Technical Specification. (SORC 96-098)

REFUELING INFORMATION

  • DOCKET NO: 50-272 MONTH: JULY 1996 UNIT NAME: SALEM 1 CONTACT: R.HELLER TELEPHONE: 609-339-5162 MONTH : JULY 1996 Refueling information has changed from last month: YES_._.NO. X .

Scheduled date for next refueling: (to be determined)

Scheduled date for restart following refueling: (to be determined)

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? (to be determined )

Scheduled date(s) for submitting proposed licensing action: . n/a Important licensing considerations associated with refueling:

Number ofFuel Assemblies:

a. Incore 0
b. In Spent Fuel Storage 989 Present licensed spent fuel storage capacity: 1632 Future spent fuel storage capacity: 1 1632 Date oflast refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: September 2008
  • SALEM GENERATING STATION MONTHLY OPERATING

SUMMARY

- UNIT 1 JULY 1996 SALEMUNITNO. 1 The Unit is in a refueling and Steam Generator replacement outage and 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
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