ML18100A821

Monthly Operating Rept for Dec 1993 for Salem Nuclear Generating Station Unit 1.W/940112 Ltr

ML18100A821
Person / Time
Site:	Salem
Issue date:	12/31/1993
From:	Heller R, Shedlock M, Vondra C Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9401240190
Download: ML18100A821 (20)

Text

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.ps~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station January 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 December 1993 are being sent to you.

Average Daily Unit Power Level Operating Data Report Unit.Shutdowns and Power Reductions Safety Related Maintenance 10CFR50.59 Evaluations Operating Summary Refueling Information RH:pc

~Iii Salem Operations cc: Mr. Thomas T. Martin Regional Administrator USNRC Region I 631 Park Avenue King of Prussia, PA 19046 Enclosures 8-1-7.R4 The Energy People 940124

- -- --------r\

0190 931231

• 05000272 I ~

l'l j:&-:J..'-f 95-2189 (10M)

,,1~-!9 PDR AD0Cl1' PDR L R

~ERAGE DAILY UNIT POWER L~

Docket No.: 50-272 Unit Name: Salem #1 Date: 01/10/94 Completed by: Mark Shedlock Telephone: 339-2122 Month December 1993 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 16 0 P. 8.1-7 Rl

~ OPERATING DATA REPORT ~

Docket No: 50-272 Date: 01/10/94 Completed by: Mark Shedlock Telephone: 339-2122 Operating Status

1. Unit Name Salem No. 1 Notes

2. Reporting Period December 1993

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 8760 144697

12. No. of Hrs. Rx. was Critical 0 5949.99 95131.97

13. Reactor Reserve Shutdown Hrs. 0 0 0

14. Hours Generator On-Line 0 5747.43 91887.84

15. Unit Reserve Shutdown Hours 0 0 0

16. Gross Thermal Energy Generated (MWH) 0 18573705.6 290772314

17. Gross Elec. Energy Generated

{MWH) 0 6162980 96535970

18. Net Elec. Energy Gen. (MWH) -3654 5865894 91937553

19. Unit Service Factor 0 65.6 63.5

20. Unit*Availability Factor 0 65.6 63.5

21. Unit Capacity Factor (using MDC Net) 0 60.5 57.4

22. Unit Capacity Factor (using DER Net) 0 60.1 57.0

23. Unit Forced Outage Rate 0 12.6 21.0

24. Shutdowns scheduled over next 6 months (type, date and duration of: each)

The Unit is presently in a refueling outage.

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

January 16, 1994.

8-1-7.R2

UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH DECEMBER 1993 DOCKET NO. :_5~0~-=27~2~--

UNIT NAME: Salem #1 DATE: 01-10-94 COMPLETED BY: Mark Shedlock TELEPHONE: 339-2122 METHOD OF SHUTTING LICENSE DURATION DOWN EVENT SYSTEM COMPONENT CAUSE AND CORRECTIVE ACTION NO. DATE TYPE 1 (HOURS) REASON 2 REACTOR REPORT # CODE 4 CODE 5 TO PREVENT RECURRENCE 134 10-02-93 s 1824.90 c 1 ---------- RC FUELXX NUCLEAR NORMAL REFUELING 135 12-17-93 s 359.98 c 4 ---------- RC FUELXX NUCLEAR NORMAL REFUELING 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-Administrative 5-Load Reduction G-Operational Error (Explain) 9-0ther H-Other (Explain)

SAFETY RELATED MAINTE.CE DOCKE.O: 50-272

. MONTH: - *DECEMBER 1993 UNIT NAME: SALEM 1 DATE: JANUARY 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 WO NO UNIT EQUIPMENT IDENTIFICATION 901017137 1 lA DIESEL GENERATOR FAILURE DESCRIPTION: LUBE OIL PUMP LEAKS - REWORK 911015202 1 12 CONTAINMENT FAN COIL UNIT FAILURE DESCRIPTION: FAULTY CONTROL CIRCUIT CONNECTION

- RESOLDER 920521149 1 VALVE 1CA572 FAILURE DESCRIPTION: VALVE CORRODED - REPLACE VALVE 921102094 1 VALVE 1CH156 FAILURE DESCRIPTION: VALVE LEAKS THROUGH & STEM SNAPPED - REPLACE 930118141 1 VALVE 1RC900 FAILURE DESCRIPTION: UPSTREAM FLANGE LEAKS - RELOCATE PER DCP 930810133 1 AFW PUMP ROOM COOLER FAN FAILURE DESCRIPTION: FOUND WORN SHAFT - REBUILD FAN 931004200 1 VALVE 11CA360 FAILURE DESCRIPTION: VALVE FAILED LOCAL LEAK RATE TEST

- TROUBLESHOOT 931005123 1 12 CONTAINMENT FAN COIL UNIT FAILURE DESCRIPTION: O.B. BEARING VIBRATION - OPEN AND INSPECT 931014111 1 VALVE 1CC283 FAILURE DESCRIPTION: VALVE HAS BROKEN STEM - REPLACE VALVE 931112063 1 lC DIESEL GENERATOR FAILURE DESCRIPTION: VOLTAGE TRANSDUCER FAILED RETEST -

INVESTIGATE & REPAIR

SAFETY RELATED MAINTE.CE DOCKE~"O: 50-272

. MONTH: - *DECEMBER 1993 UNIT NAME: SALEM 1 DATE: JANUARY 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

WO NO UNIT EQUIPMENT IDENTIFICATION 931115153 1 NUCLEAR INSTRUMENTATION FAILURE DESCRIPTION: GAMMAMETRICS SOURCE RANGE CHANNEL "C" FAILED LOW - INVESTIGATE 931130190 1 13 COLD LEG RTD FAILURE DESCRIPTION: 13 COLD LEG RTD IS BENT - REPLACE 931207147 1 VALVE 1CH4 DOWNSTREAM PIPING FAILURE DESCRIPTION: 1CH4 LEAK IN DOWNSTREAM PIPING -

REPLACE PIPING 931208227 1 12 CONTAINMENT FAN COIL UNIT FAILURE DESCRIPTION: HIGH SPEED BREAKER CLOSED IN WITHOUT 1B3X CLOSING IN -

INVESTIGATE 931208230 1 12 CONTAINMENT FAN COIL UNIT FAILURE DESCRIPTION: MOTOR WINDINGS AND BEARINGS RUNNING HOT - INVESTIGATE 931208241 1 12 CONTAINMENT FAN COIL UNIT FAILURE DESCRIPTION: REPLACE MOTOR 931209140 1 11 RHR PUMP FAILURE DESCRIPTION: CC FLOW SW ALARM UP - INVESTIGATE 931210142 1 VALVE 12AF21 FAILURE DESCRIPTION: 12 AUX FEEDWATER VALVE THROTTLING ERRATICALLY - INVESTIGATE 931213147 1 13 RCP LOOP FLOW TRANSMITTER FAILURE DESCRIPTION: RCP LOOP FLOW TRANSMITTER LEAK IN PANEL 447-lJ - REPAIR

SAFETY RELATED MAINTE.CE DOCKE-NO: 50-272

• MONTH: - *DECEMBER 1993 UNIT NAME: SALEM 1 DATE: JANUARY 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609) 339-5162 (cont'd)

WO NO UNIT EQUIPMENT IDENTIFICATION

---

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931214215 1 PRESSURIZER LEVEL CHANNEL II FAILURE DESCRIPTION: PRESSURIZER LEVEL CHANNEL II READS 5% LOW - INVESTIGATE 931215188 1 RVLIS TRAIN B FAILURE DESCRIPTION: RVLIS TRAIN B DIFFERS FROM TYGON TUBE - INVESTIGATE 931220075 1 1PR17 RC PRT GAS SAMPLE VALVE FAILURE DESCRIPTION: NO INDICATION ON CONSOLE -

INVESTIGATE

---

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931220109 1 VALVE 1CV35 FAILURE DESCRIPTION: 1CV35 HAS DUAL LIMITS, FAILED RETEST - INVESTIGATE 931221210 1 VALVE 12SW24 FAILURE DESCRIPTION: 12SW24 IS NOT RECEIVING A DEMAND TO STROKE WHEN STRAINER D/P IS HIGH - INVESTIGATE 931230117 1 lB DIESEL GENERATOR FAILURE DESCRIPTION: JACKET WATER LEAK BY 1DA46B -

REPLACE NIPPLE

10CFR50.59 EVALUATIONSf/j

. MONTH: - 'DECEMBER 19 9 3 DOCKE~O: 50-272 UNIT NAME: SALEM 1 DATE: JANUARY 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 (DCPs) lEC-3278 Pkg 1 "Fuel Handling Building Ventilation Upgrade" -

This DCP entails modifications to the Unit 1 Fuel Handling Area Ventilation System which are summarized as follows: 1.) Replace backdraft dampers (1VHE304 & 1VHE305); 2.) Replace the inlet guide vanes and their operators on the Fuel Handling Area exhaust fans No. 11 & 12. 3.)

Remove the existing control loop instrumentation for the Unit 1 Fuel Handling Area supply fan (1VHE24) inlet guide vanes; 4.) Remove the Unit 1 Fuel Handling Area Truck Bay exhaust fan 1VHE23, damper 1VHE503; 5.) Install gravity type pressure relief damper 1VHE868 and fixed louver 1VHE503 in the truck bay areas; 6.) Replace the adjustable motor sheaves and V-belts for the Unit 1 Fuel Handling Area supply fan (1VHE24); 7.) Replace the adjustable motor sheaves and V-belts for the No.

11 & 12 Fuel Handling Area exhaust fans 1VHE20 &

1VHE21; 8.) Replace inlet and discharge flexible connections on the Unit No. 1 Fuel Handling Area supply fan (1VHE24) and the No. 11 & 12 Fuel Handling Area exhaust fans 1VHE20 & 1VHE21; 9.)

Install new flow and static pressure test ports on fan inlet and discharge ducts; 10.) Install discharge screens on the supply air duct openings; 11.) Replace Fuel Handling Area differential pressure switches; 12.) The Fuel Handling Area humidity monitoring system will be abandoned in place (This was previously evaluated in Safety Evaluation S-C-M941-MSE-234, Rev. O) and the applicable documentation will be revised under this DCP to reflect this abandonment. These modifications will not result in any changes to system operating parameters. Therefore, this proposal will not reduce the margin of safety as defined in the basis for any Technical Specification. (SORC 93-105)

10CFR50.59 EVALUATIONS~ DOCKE~O: 50-272

. MONTH: - 'DECEMBER 1993 UNIT NAME: SALEM 1 DATE: JANUARY 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

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ITEM

SUMMARY

lSC-2269 Pkg 6 "Modification to Salem Unit 1 lE Distribution System & Circ. Water Switchgear" Rev. 1 - This package will provide the final connections between the two (2) new offsite power sources between the offsite transmission network and the Salem Unit #1 lE distribution system (Vital buses lA, lB & lC) and Unit #1 new CW switchgear. This revision to the package increases the calibration tolerance on the degraded grid relays from -0 + .15 to+/- .5 VAC. This also requires a change in the trip value to 95.1%. The new values do not affect existing electrical minimum bus voltage (94.7%) or minimum bus relay voltage (97.0%). These modifications will improve the voltage profile on the vital and group busses. This has resulted in a change of the trip setpoint for the second level of undervoltage protection on the vital busses.

This will increase the margin of safety between the setpoint value and the minimum allowable value and ensure safe operation of the vital equipment.

(SORC 93-105) lEC-3220 Pkg 1 "AFll and AF21 Modifications (Trim Replacement)"

Rev. 2 - It is proposed to reduce the maximum auxiliary feedwater flow rates, under certain plant conditions, delivered by the motor driven and turbine driven Auxiliary Feedwater Pumps (AFWP), by modifying flow control valves 11 thru 14AF11, and 11 thru 14AF21. The proposed design change consists of replacing the existing valve trim with a reduced capacity trim (anti-cavitation design). In addition, 12AF21, currently installed in the reverse direction will be installed in the correct configuration (flow to close) by turning it 180°. The AFW flow reduction is required to limit the containment peak pressure and temperature following a steam line break inside the containment, and thus preclude the potential violation of the containment safety limits (particularly pressure limit) for fuel cycle 12 and beyond. The anti-cavitation design of the replacement trim is expected to improve the operation of these flow control valves,

10CFR50.59 EVALUATIONS~ DOCKE.O: 50-272

. MONTH: - 'DECEMBER 1993 UNIT NAME: SALEM 1 DATE: JANUARY 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

particularly AF21s, in the low flow range.

Furthermore, reliability of the turbine driven AFW pump wi-11 be enhanced by minimizing the cavitation potential for the secondary side depressurization events. since the AFll and AF21 valves are oversized, the installation of reduced capacity trim in these valves will still assure the minimum flow delivery capability of the system under design basis conditions. There will be no reduction in the margin of safety as defined in the basis for any Technical Specification.

( SORC 93-106) lEC-3288 Pkg 1 "Dose Assessment system Upgrade" - This DCP will upgrade the Dose Assessment System utilized IAW the Emergency Plan to monitor offsite radiological conditions during an emergency situation at Artificial Island. The Midas upgrade involves conversion of the current VAX/MicroVAX computer platform to a PC based platform. PC equipment will replace the current Midas terminals located at the Technical Support Center (TSC}, the Emergency Offsite Facility (EOF}, the Operations I Support Center (OSC), and the Emergency **

Preparedness Training Facility (Beach House). The Dose Assessment System will also be integrated to the ERDS network which will facilitate automatic operations and eliminate the need for manual data entry, custom interfaces and software. The Dose Assessment System Upgrade is being implemented to bring the Hope Creek and Salem plants into compliance with SPA-400-R-92-001 in accordance with New Jersey and Delaware state requirements and NRC guidelines. There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 93-108).

lEC-3300 Pkg. 1 "Service Water Pump Upgrade, No. 12 Pump" - The 1SWE2 replacement Service Water Pump wiil be compatible with brackish river water, able to withstand the high silt flows, and designed for a NPSHa at 76 feet river water level. The replacement pump will be designed to operate in

10CFR50.59 EVALUATIONS~ DOCKE~O:

UNIT NAME:

50-272

• MONTH: - 'DECEMBER 1993 SALEM 1 DATE: JANUARY 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

parallel with the existing pumps, and will use the existing motor and floor opening. The new pump will be designed to operate over a wider range of flows based on past plant operating experience.

Minor modifications to the Service Water Intake Structure are required. This includes the removal of the Seismic Restraint on Elevation 70 feet.

The installation of a vortex suppressor on elevation 70 feet. The decommissioning and removal of the existing Service Water Pump's line shaft bearing lubrication supply piping. The replacement pump does not require line shaft bearing lubrication. There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 93-108) lEA-1053 Pkg 1 "Update Documents for 1500 Gallon Nitrogen Package" - This DCP is an As-Built change document that provides a basis for as-built conditions and updates the associated documentation. Two low pressure bulk nitrogen packages (1WGE9 and lWGElO) are part of the nitrogen system and supplement the nitrogen bottled system (lWGEl). The bulk packages provide a large source of nitrogen for additional capacity and flexibility for plant operations and economics. The nitrogen system provides several functions to the Waste Gas System. The 1500 Gallon Package (lWGElO) has been tagged out since 1984. The 600 Gallon Package (1WGE9) is fully operational. The 1500 Gallon Bulk Liquid Nitrogen Package (lWGElO) is provided with a fill line or assembly used to accept liquid nitrogen from a vendor. The as-built conditions reflect only the drain valve 1NT144. The liquid nitrogen system is not discussed in the Salem Technical Specifications. Therefore, the as-buit conditions of the fill line of lWGElO will not reduce the margin of safety as defined in the Technical Specifications. (SORC 93-109)

• ~OCFR50.59 EVALUATIONS~ DOCKE~O: 50-272

• MONTH: - *DECEMBER 19 9 3 UNIT NAME: SALEM 1 DATE: JANUARY 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

B. Procedures and Revisions NC.NA-AP.ZZ-0024(Q) "Radiation Protection Program" Rev. 3 - Specific changes are made to several policies and procedures, including those dealing with occupational and public dose limits, administrative dose limits, dose categories, requirements for monitoring and for wearing of dosimetry, dose calculations, dose records, requirements for bioassay, designation of areas for radiation protection purposes, fetal dose limitation, definitions of radiologically posted areas, high radiation area access and key control, methods for keeping doses ALARA, measurement units, special access controls, and radiation protection program reviews. This revision incorporates the requirements of the revision to 10CFR20. The changes being made by this revision do not relate to design criteria, specifications, or operation of the fuel cladding, RCS boundary, or containment and do not address*

any margin of safety as defined in the basis for any Technical Specification. (SORC 93-105)

SECURITY PLAN "Artificial Island Security Plan" Rev. 4 - One significant change has been made to the plan: The addition of a new gate on the protected area (PA}

perimeter as an egress point and a point of return to the PA for activities associated with a silt holding basin. The gate, which is used infrequently, provides the only vehicle access to the basin. The area encompassed by the protected area on three sides and by the Delaware River on the west. The second change incorporated by this revision involves a management title change to the General Manager - Nuclear Support & Services.

There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 93-107)

NC.NA-AP.ZZ-0007(Q) "ALARA Program" Rev. 2 - This revision is intended to implement the changes required by the revision to 10CFR20, Standards for Protection Against Radiation, and other outstanding revision

10CFR50.59 EVALUATIONS~ DOCKEtl~lO: 50-272

• MONTH: - "DECEMBER 19 9 3 UNIT NAME: SALEM 1 DATE: JANUARY 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

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requests. Specific changes are made to several policies and procedures including management responsibilities, revision of the scope of the procedure and clarification of ALARA review requirements. Also included are the addition of requirements for TEDE ALARA evaluation, revision of ALARA review dose trigger criteria, strengthening of evaluation for cobalt reduction.

The PAA form has been enlarged and definitions have been added. The revisions to this procedure do not relate to the design criteria, specifications, or operation of the fuel cladding, RCS boundary, or containment and do not address any margin of safety as described in Section B3/4.0. Therefore, the proposal does not reduce the margin of safety as defined in the basis for any Technical Specification. (SORC 93-108)

NC.NA-AP.ZZ-0045(Q) "Respiratory Protection Program" Rev 2 - This revision is intended to implement the changes required by the revision to 10CFR20, Standards for Protection Against Radiation, and other outstanding revision requests. Specific changes are made to several policies and procedures including management responsibilities, revision of the scope of the procedure, revision of hazard assessment requirements, revision of physical requirements, revision of WBC and fit test requirements, and additional definitions. The revisions to this procedure do not relate to the design criteria, specifications, or operation of the fuel cladding, RCS boundary, or containment and do not address any margin of safety as described in Section B3/4.0. Therefore, the proposal does not reduce the margin of safety as defined in the basis for any Technical Specification. (SORC 93-108)

NC.NA-AP.ZZ-0029(Q) "Radioactive Material Control Program" Rev. 2 -

This revision is intended to implement the changes required by the revision to 10CFR20, Standards for Protection Against Radiation, and other outstanding revision requests. Specific changes

10CFR50. 59 EVALUATIONS- DOCKE~O: 50-272

" MONTH: - .DECEMBER 19 9 3 UNIT NAME: SALEM 1 DATE: JANUARY 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

are made to several policies and procedures including management responsibilities, terminology for approval of radioactive material purchases, radioactive material release criteria, criteria for surveys upon receipt of radioactive material, evaluation of doses to non-RCA workers from storage of radioactive material, criteria for reporting theft or loss of radioactive material, and training requirements. The revisions to this procedure do not relate to the design criteria, specifications, or operation of the fuel cladding, RCS boundary, or containment and do not address any margin of safety as described in Section B3/4.0. Therefore, the proposal does not reduce the margin of safety as defined in the basis for any Technical Specification. (SORC 93-108)

NC.NA-AP~ZZ-0025(Q) "Operational Fire Protection Program" - Rev. 3 -

The proposal invoives a full revision to the Nuclear Department Operational Fire Protection Program procedure. This revision incorporates requests by the station to modify the section on hot work - control of ignition sources with regards to hot work performed over gratings and floor openings. The current method of control has been modified to include specific steps to take besides that which is spelled out on the hot work permit. With this revision, there is no increase in the potential for damage to safety related equipment from fire nor are there any adverse effects to the ability of the plant to achieve and maintain safe shutdown in the event of a fire.

There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 93-108)

NC.NA-AP.ZZ-OOOl(Q) "Nuclear Department Procedure System" Rev. 5 -

Section 6.13 of NC.NA-AP.ZZ-OOOl(Q),"Nuclear Department Procedure System" was revised to read:

shall, should, may and will. These words are used in procedure as follows: 1.) "shall" denotes a regulatory requirement or commitment to a regulatory agency and is to be adhered to without

10CFR50.59 EVALUATIONSf/j DOCKE~O: 50-272 "MONTH: - 'DECEMBER 1993 UNIT NAME: SALEM 1 DATE: JANUARY 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

deviation (i.e. "shall is a commitment to regulators such as the NRC or the State of New Jersey); 2.) "should" denotes an expected action that is to be adhered to unless supervision or management determines that compliance is not mandatory (i.e. "should is a Nuclear Department management requirement to perform an action.

Therefore, supervision or management may authorize deviation from the action as they deem necessary);

3.) "may" is used to denote permission. The Nuclear Department is committed to ANSI NlS.7-1976/ANS 3.2, Administrative Controls and Quality Assurance for the operational phase of Nuclear Power Plants, that defines "shall" to denote a requirement, "should" to denote a recommendation and "may" to denote permission, neither a requirement nor a recommendation. The proposed changes provide a stronger management position with respect to implementation of activities associated with the term "should" while maintaining the term "shall" for activities traceable to regulatory requirements or commitments. There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 93-108)

NC.NA-AP~ZZ-0032(Q) "Preparation, Review and Approval of Procedures" Rev. 3 - Section 6.15 of NC.NA-AP.ZZ-0032(Q),

"Preparation, Review and Approval of Procedures" were revised to read: shall, should, may and will. These words are used in procedures as follows: 1.) "shall" denotes a regulatory requirement or commitment to a regulatory agency and is to be adhered to without deviation (i.e.

"shall is a commitment to regulators such as the NRC or the State of New Jersey); 2.) "should" denotes an expected action that is to be adhered to unless supervision or management determines that compliance is not mandatory (i.e. "should is a Nuclear Department management requirement to perform an action. Therefore, supervision or management may authorize deviation from the action as they deem necessary); 3.) "may" is used to

10CFR50.59 EVALUATIONS~ DOCKE.O: 50-272 "MONTH: - 'DECEMBER 1993 UNIT NAME: SALEM 1 DATE: JANUARY 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

denote permission. The Nuclear Department is committed to ANSI NlS.7-1976/ANS 3.2, Administrative Controls and Quality Assurance for the operational phase of Nuclear Power Plants, that defines "shall" to denote a requirement, "should" to denote a recommendation and "may" to denote permission, neither a requirement nor a recommendation.. The proposed changes provide a stronger management position with respect to implementation of activities associated with the term "should" while maintaining the term "shall" for activities traceable to regulatory requirements or commitments. There is no reduction in the margin of safety as defined in the basis for any Technical Specification.

(SORC 93-108)

c. Safety Evaluations (S'/E)

TSl.IC-TI.RCS-OOOl(Q) "Evaluation of the ABB Combustion Engineering Rod Position Indication System" - The purpose of this proposal is to evaluate the ABB Combustion Engineering Rod Position Indication (CE RPI) system by installing it on several Salem Control Bank A and Control Bank D rods. This CE RPI evaluation will take place in Mode 3. RPI is not required in Mode 3. Rod movement is allowed, according to Tech Spec, in Mode 3 with only Rod Control's group demand counters operational. This proposal does not affect Rod Control in any way.

The Technical Specification bases for movable control assemblies defines operability for RPI and indicates that if the rods are maintained above the insertion limits, all accident analysis assumptions concerning rod positioning are valid.

In addition, shutdown margin is maintained through boration such that even with two control banks fully withdrawn, Keff .remains less than .95. In that shutdown margin is assured and operability of RPI is not required, the margin of safety is not affected. (SORC 93-104)

10CFR50.59 EVALUATION- DOCKE.O: 50-272

.. MONTH: - ,DECEMBER 19 9 3 UNIT NAME: SALEM 1 DATE: JANUARY 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

D. SAR Change SCN 93-52 "SAR Change Notice" - This change is an UFSAR documentation change only, to revise Section *9.0 of the UFSAR to reflect the existing Diesel Generator Day Tank level alarm and transfer pump control setpoints. The UFSAR presently states that one of the two fuel transfer pumps is started when oil level in any day tank drops to one-third full. This in an incorrect statement since the existing setpoint as per field settings and other design basis documentation, for the primary transfer pump start is 33 inches. The total span of the tank is 48 inches therefore the actual pump start occurs when the tank is over two-thirds full. The UFSAR also presently states that should the primary pump fail and level drops to "one fourth capacity", that another level switch will give a "FUEL OIL DAY TANK LEVEL LOW" alarm and start the backup fuel transfer pumps. The setpoint for the low alarm and back up transfer pump start is presently set to 18 inches per field settings and other design basis documentation, which is more than one third tank capacity. A technical specification minimum tank level exists and is located at 11.45 inches. The low alarm and back-up transfer pump setpoints should be set at a level at which the technical specification limit would not be exceeded given a negative instrument error. Setpoint calculation SC-DG009-0l has been performed and has acceptably analyzed both the existing primary pump start setpoint of 33 inches and the back-up pump start and low alarm setpoints of 18 inches. The proposed change will revise the UFSAR section to reflect the actual setpoints of 33 inches for the primary pump start and 18 inches for the low alarm and backup pump start. There is no reduction in the margin of safety as defined in the basis for any Technical Specification.

(SORC 93-109)

• ~OCFR50.59 EVALUATION- DOCKE.O: 50-272

" MONT.H: -'DECEMBER 1993 UNIT NAME: SALEM 1 DATE: JANUARY 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 (cont'd)

ITEM

SUMMARY

E. Temporary Modifications (TMOD}

93-143 Rev. O "lB EDG Jacket Water Heaters - Replace With Plugs"

- The purpose of this modification is to allow lB Emergency Diesel Generator (EDG) to be declared operable with pipe plugs in lieu of 2 jacket water heaters. Jacket water heaters were burnt out and no spares are in FOLIO. The associated 10CFR50.59 evaluation addresses operability of the lB EOG without jacket water heaters pending arrival and installation of the proper heaters. The EOG will be operated in the interim with the jacket water heater holes plugged with pipe plugs and the heater breaker tagged. A Technical Specification Interpretation regarding operability of the EOG states that operation without jacket water heaters is not a concern provided lube oil is at the proper temperature. This has been confirmed via a letter from the vendor which states the cold start capability would not be reduced providing the lube oil was at temperature and the aftercooler jacket water heater was inservice. Thus the acceptance criteria that the EOG start and load within 10 seconds would not be affected during the period of the TMOD. There is no reduction in the margin of safety as defined in the basis for any Technical Specification. (SORC 93-109).

SALEM GENERATING STATION MONTHLY OPERATING

SUMMARY

- UNIT 1 DECEMBER 1993 SALEM UNIT NO. 1 The Unit remained shutdown throughout the entire period for the eleventh refueling outage.

• ~

REF;ELING INFORMATION MONTH: -*DECEMBER 1993 e DOCKE.O:

UNIT NAME:

50-272 SALEM 1 DATE: JANUARY 10, 1994 COMPLETED BY: R. HELLER TELEPHONE: (609)339-5162 MONTH DECEMBER 1993

1. Refueling information has changed from last month:

YES X NO ____

2. Scheduled date for next refueling: OCTOBER 2, 1993

3. Scheduled date for restart following refueling: JANUARY 16, 1994

4. a) Will Technical Specification changes or other license amendments be required?:

YES NO X NOT DETERMINED TO DATE b) Has the reload fuel design been reviewed by the Station Operating Review Committee?:

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

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

N/A

6. Important licensing considerations associated with refueling:

i I

• 1

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 8-1-7.R4