ML18100A475

Monthly Operating Rept for June 1993 for Salem Unit 2.W/

ML18100A475
Person / Time
Site:	Salem
Issue date:	06/30/1993
From:	Shedlock M, Vondra C Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9307190097
Download: ML18100A475 (11)

Text

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

Dear Sir:

MONTHLY OPERATING REPORT SALEM NO. 2 DOCKET NO. 50-311 July 14, 1993 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 June 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 cc:

Mr. Thomas T. Martin Regional Administrator USNRC Region I 631 Park Avenue King of Prussia, PA 19046 Enclosures 8-1-7.R4

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The Energy People

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General Manager -

Salem Operations f Pf'f

'[I 95-2189 (10M) 12-89

~ERAGE DAILY UNIT POWER LE~

Docket No~:

50-311 Unit Name:

Salem #2 Date:

7/10/93 Completed by:

Mark Shedlock Telephone:

339-2122 Month June 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 0

16 0

P. 8.1-7 Rl

OPERATING DATA REPORT Docket No:

50-311 Date:

7/10/93 Completed by:

Mark Shedlock Telephone:

339-2122 Operating Status

1.

Unit Name Salem No. 2 Notes

2.

Reporting Period June 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-=--~~~~~~~~~~~~~~-

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)

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 720 52.4 0

0 0

8505.6 0

-22010 0

0 0

0 100 Year to Date Cumulative 4343 102696 1769.11 65534.71

~~~~o~-

o 1675.51 63233.97

~~~~o~~

o 5599250.4 150694940.2 1913950 66621438 1796757 63382733

~~~~3~8=-=-*=6-61.6

~~~~3~8~*~6-61.6

~~~~3~7'-'-.4-"-

55.8

~~~-3~7_.~1~

55.4

~~~~2~6=--=--.5=--

23.0

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:

NA 8-l-7.R2

NO.

DATE 0014 03-17-93 0021 06-01-93 0022 06-11-93 1

F:

Forced S:

Scheduled DURATION TYPE 1

(HOURS)

REASON2 s

1811.92 c

s 252.00 c

F 467.98 A

2 Reason A-Equipment Failure (explain)

B-Maintenance or Test C-Refueling D-Requlatory Restriction UNIT SHUTDOWN AND POWER REDUCTIONS REPORT MONTH JUNE 1993 METHOD OF SHUTTING DOWN REACTOR 4

4 4

LICENSE EVENT REPORT #

zz zz IB 3

Method:

1-Manual 2-Manual Scram SYSTEM CODE4 E-Operator Training & License Examination F-Administrative 3-Automatic Scram 4-Continuation of Previous outage 5-Load Reduction 9-0ther G-Operational Error (Explain)

H-Other (Explain)

COMPONENT CODE5 DOCKET NO.: ~5~0_-~27~2~--

UNIT NAME:

Salem #2 DATE:

07/10/93 COMPLETED BY:

Mark Shedlock TELEPHONE:

339-2122 CAUSE AND CORRECTIVE ACTION TO PREVENT RECURRENCE zzzzzzzz NUCLEAR NORMAL REFUELING zzzzzzzz NUCLEAR NORMAL REFUELING CONROD NUCLEAR CONTROL ROD INSTRUMENTATION 4

Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report CLER) File (NUREG-0161) 5 Exhibit 1 - Same Source

.SA.f'ETY 'RELATED MAINTEN.E MONTH: -

JUNE 1993 DOCKET,O:

UNIT NAME:

50-311 SALEM 2 WO NO UNIT 930519085 2

930602104 2

930605074 2

DATE:

COMPLETED BY:

TELEPHONE:

JULY 10, 1993 R. HELLER (609)339-2212 EQUIPMENT IDENTIFICATION 22 CONTAINMENT FAN COIL UNIT FAILURE DESCRIPTION:

22 CFCU HIGH VIBRATION -

INSPECT AND REPAIR 2A SAFEGUARDS EQUIPMENT CABINET FAILURE DESCRIPTION:

2A SEC SPURIOUS ALARMS -

TROUBLESHOOT "A" REACTOR TRIP BYPASS BREAKER FAILURE DESCRIPTION:

REPLACE THE UNDERVOLTAGE TRIP DEVICE

.10t:FR50. 59 EVALUATIONS e MONTH: -

JUNE 1993 DOCKET.O:

UNIT NAME:

DATE:

COMPLETED BY:

TELEPHONE:

50-311 SALEM 2 JULY 10, 1993 R. HELLER (609)339-2212 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 2SC-2267 Pkg 2 "Safeguards Equipment Controller -

ATI Restoration"

- Within each of the SECs is a Control Electronics Unit (CEU).

Each CEU contains an Automatic Test Insertion (ATI) card that was installed as part of DCP 2SC-2267 Pkg. 1.

These ATis weren't activated at the time, awaiting design verification from the manufacturer, Eaton Corporation.

(The design verification consequently arrived, but not in time for ATI activation.)

Wiring changes were also made during DCP 2SC-2267 Pkg. 1 to avoid nuisance alarms due to the ATis not being activated at that time, unless programming changes were made to the ATI EPROMs.

The purpose of the DCP is to activate the ATis.

The DCP will install new EPROMs in each of the ATis in the three SECs and the spare chassis.

For the SECs this modification will be accomplished by removing an entire CEU chassis from an SEC cabinet and bringing it up to a test rig.

The ATI EPROMs will then be changed out and then, the entire CEU will be tested in its SEC cabinet.

Output alarm relays will be reinstalled in existing XK90 and XK91 sockets.

Wires will be terminated, as required.

This will allow the Aux and Overhead Annunciators to receive inputs from the ATI.

Replacement EPROMs are qualified to the original specifications, and the modification will be thoroughly tested prior to declaring an SEC operable.

There is no requirement for an ATI, and therefore there is no requirement for the speed at which it operates.

The Master Time Response Procedure will be revised to include the postulated 400 millisecond delay for ATI disengagement to assure that Technical Specification limits are not violated.

There is no reduction in the margin of safety as defined in the basis for any Technical Specification.

(SORC 93-049)

.1o't:FRSO. 59 EVALUATIONS e MONTH: -

JUNE 1993 (cont'd)

ITEM 2EC-3179 Pkg 9 DOCKETttO:

UNIT NAME:

DATE:

COMPLETED BY:

TELEPHONE:

SUMMARY

50-311 SALEM 2 JULY 10, 1993 R. HELLER (609)339-2212 "Salem Fire Damper Upgrade" -

The design scope for this package involves the application of a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (2 layer) fire wrap on four dampers of the Diesel Fuel Oil Storage Area Ventilation (DFSAV) Subsystem of the Diesel Generator Area Ventilation (DGA) System and the application of a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (2 layer) fire wrap on eight dampers of the Control Area Ventilation (CAV) System Battery Rooms. The function, basic configuration and operation of the system will not be altered and the codes, standards, qualification and design criteria of the original system will apply.

In order to maintain the integrity of the fire barrier, the dampers and duct work between the dampers and the fire barrier wall must be wrapped with a fire wrap material to obtain a fire rating equivalent to the barrier.

The ductwork containing the dampers will be wrapped with two layers of 3M's E-54A Mat, a rated fire protective coating to obtain a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rating from the fire barrier up to and including the fire damper.

The ductwork metal provides an additional 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rating for combined rating that is equal to or greater than the 1-1/2 hour rating required of the damper.

This DCP will also install two pairs of longitudinal braces, to minimize the possible bending stresses, on the Battery Room Ele. 100 1 exhaust air ductwork.

Longitudinal braces have been designed and will be installed as a modification to two existing supports, one on each end of the exhaust ductwork span, above the Battery Room ceiling.

The ductwork above the 64' Ele. Battery Room will have an access door installed as part of this modification to allow for easier access to reset the fire damper following periodic testing.

The basis for the Technical Specifications does not address the Fire Protection, Diesel Fuel Storage Area ventilation systems or the Battery Room ventilation system.

There is no reduction in the margin of safety as defined in the basis for any Technical Specification.

(SORC 93-049)

ldCFR50.59 EVALUATIONS~

MONTH: -

JUNE 1993 (cont'd)

DOCKET.O:

UNIT NAME:

DATE:

COMPLETED BY:

TELEPHONE:

50-311 SALEM 2 JULY 10, 1993 R. HELLER (609)339-2212 ITEM

SUMMARY

2SC-2254 Pkg 1 2EC-3197 Pkg 1 "HD9 Valves Replacement" -

The purpose of this change is as follows: 1) Replace and relocate 21, 22, and 23HD9 valves and actuators; 2) Modify heater drain tank level controllers 2LA1021C from proportional plus reset to proportional control only.

The reason for this design change is as follows: 1) Eliminate a source of maintenance problems; 2) Correct controller action after periods of prolonged standby operation.

There are no Technical Specifications associated with the HD9 valves, Heater Drain Tanks or level controllers which are affected by these valve replacements.

There is no reduction in the margin of safety as defined in the basis for any Technical Specification.

(SORC 93-051)

"Auxiliary Feedwater and Steam Generator Blowdown Valve Modification" - This DCP replaces the internals of the steam generator blowdown containment isolation valves (21-24GB4) with new plug and trim that is hardfaced with Stellite over the full surface of the plug and seat; it also replaces the GB4 actuators with equivalent components.

The DCP further replaces the AF23 valves with Edward 4 inch Univalve Y-Pattern Stop Check valves.

The modified and replacement valves will not impact the Technical Specifications.

The installation of this modification is subject to Technical Specification Section 3/4.7.2 and 3.9.4 for the GB4 valves and Section 3.7.1.2 for the AF23 valves.

The proposed change does not alter the original design intent or modes of operation or function for which both systems are currently analyzed.

Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

(SORC 93-057)

B.

Temporary Modifications TMOD # 93-061 "Restoration of the Blowdown Supply Line from 24 Steam Generator to Radiation Monitor Rl9D 11 A

temporary modification will be made to the 24 Steam Generator blowdown system which will allow sample flow to the 24 Steam Generator blowdown radiation

1o'cFR5E>. 59 EVALUATIONS e MONTH: -

JUNE 1993 (cont'd)

ITEM DOCKET-0:

UNIT NAME:

DATE:

COMPLETED BY:

TELEPHONE:

SUMMARY

50-311 SALEM 2 JULY 10, 1993 R. HELLER (609)339-2212 monitor (2Rl9D).

The temporary modification is needed because the normal sample path is plugged and cannot be restored until plant conditions (Steam Generator cooled down) allow.

By providing a continuous flow to the R19D monitor, this temporary modification maintains the present margin of safety.

(SORC 93-051)

SALEM UNIT NO. 2 SALEM GENERATING STATION MONTHLY OPERATING

SUMMARY

UNIT 2 JUNE 1993 The Unit remained shutdown until June 28, 1993, as the Seventh Refueling Outage was completed and the Rod Control System event investigations and corrective actions were addressed.

The reactor was taken critical at 1936 hours0.0224 days <br />0.538 hours <br />0.0032 weeks <br />7.36648e-4 months <br /> on June 28, 1993 and startup operations continued throughout the remainder of the period.

r

<,,REFUELl:NG INFORMATION DOCKET~O:

MONTH: -

JUNE 1993 UNIT NAME:

DATE:

COMPLETED BY:

TELEPHONE:

MONTH JUNE 1993

1.

Refueling information has changed from last month:

YES X

NO 50-311 SALEM 2 JULY 10, 1993 R. HELLER

{609)339-2212

2.

Scheduled date for next refueling:

SEPTEMBER 24. 1994

3.

Scheduled date for restart following refueling: NOVEMBER 22, 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 NO ~--=X~~

If no, when is it scheduled?:

5.

Scheduled date(s) for submitting proposed licensing action:

N/A

6.

Important licensing considerations associated with refueling:

7.

Number of Fuel Assemblies:

a.

Incore 193

b.

In Spent Fuel storage 464

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:

March 2003 8-1-7.R4