ML18101A819

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Monthly Operating Rept for June 1995 for Salem Nuclear Generating Station Unit 2
ML18101A819
Person / Time
Site: Salem PSEG icon.png
Issue date: 06/30/1995
From: Heller R, Phillips R, Summers J
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9507190337
Download: ML18101A819 (8)
v • d • e


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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. 2 DOCKET NO: 50-311 July 14, 1995
  • 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 are being sent to you.

RH:vls Enclosures C

Mr. Thomas T. Martin General Manager -

Salem Operations Regional Administrator USNRC, Region I 631 Park Avenue King of Prussia, PA 19046 8-1-7.R4 Theoower_isin_vour_h~nrk__ _ _____ --- --------

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95-2168 REV. 6/94 9507190337 950630 PDR ADDCK 05000311 R

PDR

AV~GE DAILY UNIT POWER LEVE.

Completed by:

Robert Phillips

  • Month June 1995 Day Average Daily Power Level (MWe-NET) 1 1072 2

1087 3

740 4

1037 5

1076 6

1073 7

903 8

0 9

0 10 0

11 0

12 0

13 0

14 0

15 0

16 0

P. 8.1-7 R1 Day Docket No.:

50-311 Unit Name:

Date:

Telephone:

Average Daily (MWe-NET) 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31

Salem #2 07-10-95 339-2735 Power Level 0

0 0

0 0

0 0

0 0

0 0

0 0

0

.OPERATING DATA REPORT Docket No:

50-311 Date:

07/10/95 Completed by:

Robert Phillips Telephone:

339-2735 Operating Status

1.

Unit Name Salem No. 2 Notes

2.

Reporting Period JUNE 1995

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=-~~~~~~~~~~~~~~~
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 720 167 0

166.97 0

528849.6 174980 157494 23.2 23.2 19.8 19.6 76.8 Year to Date Cumulative 4343 120216 2468.40 78083.62 0

0 2261.60 75229.52 0

0 6807220.8 187781005.0 2198890 78648898 2053875 74777419 52.1 62.6 52.1 62.6 42.8 56.2 42.4 55.8 35.9 22.3

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

To be determined.

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

To be determined.

8-l-7.R2

NO.

DATE 2681 6-3-95 s

2692 6-7-95 F

1 2

F:

Forced S:

Scheduled REPORT MONTH ~JU=N=E~~1~9~9~5-DURATION TYPE 1 (HOURS)

REASON 2 7.0 B

553.0 A

Reason A-Equipment Failure (explain)

B*Maintenance or Test C-Refueling D-Requlatory Restriction METHOD OF SHUTTING DOWN REACTOR 5

3 E*Operator Training & License Examination F*Administrative G-Operational Error (Explain)

H-Other (Explain)

LICENSE EVENT SYSTEM REPORT #

CODE4 HA SH 3

Method:

1-Manual 2-Manual Scram 3-Automatic Scram 4-Continuation of Previous Outage 5-Load Reduction 9-0ther COMPONENT CODE 5 DOCKET NO.

UNIT NAME DATE COMPLETED BY TELEPHONE 50-311 Salem #2 7-10-95 Robert Phillips 609-339-2735 CAUSE AND CORRECTIVE ACTION TO PREVENT RECURRENCE VAL VEX TURBINE VALVE TEST CKTBRK BREAKER FAILURE 1-9 500KV 4

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

10CFR50.59 EVALUATIONS MONTH: JUNE 1995 DOCKET NO:

UNIT NAME:

50-311 SALEM2 07/10/95 R.HELLER 609-339-5162 DATE:

COMPLETED BY:

TELEPHONE:

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)

The reporting ofDCP related 10CFR50.59 evaluations is being modified to address these items only after they have been implemented and turned over to Operations. For the next several months, during this transition phase, we anticipate few new DCPs to report.

2. Procedures TS2.0P-PT.RHR-0001(Q)

"RH29 Testing and Cross Checking of Flow Devices" Rev.

0 - The purpose of this special test for the RHR system is in support of the operability determination for 21 & 22 RHR trains. The test essentially comprises the following two parts: 1) To determine the function response of the RHR pump miniflow valves 21RH29 and 22RH29 by conducting a dynamic flow test; and 2.) to obtain/record the flow data for the existing permanent flow orifices at the pump discharge and the downstream orifices in the safety injection headers. In Mode 5, two RHR loops are required to be operable for single failure considerations. Two steam generators with the secondary side water level greater than 5% narrow range will be substituted for the loop being tested. Therefore, the redundancy requirements will be met.

Only one loop will be tested at a time and it will not affect the operation of the other train. The operation of one RHR pump will provide adequate flow to ensure mixing, prevent boron stratification, and produce gradual reactivity changes during any boron concentration reductions. These gradual reactivity changes will be within the capability of operator recognition and response consistent with the

10CFR50.59 EVALUATIONS MONTH: JUNE 1995 DOCKET NO:

UNIT NAME:

50-311 SALEM2 07/10/95 R.HELLER 609-339-5162 DATE:

COMPLETED BY:

TELEPHONE:

Technical Specification bases for the RHR system operation. There is no reduction in the margin of safety as defined in the bases for any Technical Specifications.

(SORC 95-067)

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REFUELING INFORMATION MONrn,~ JUNE l995, :----J.

MONTH: JUNE 1995* -.c DOCKET NO:

UNIT NAME:

DATE:

COMPLETED BY:

TELEPHONE:

50-311

.SALEM2 07/10/95 RHELLER 609-339-5162

1. Refueling information has changed from last month: YES X

NO ~~

2. Scheduled date for next refueling: (to be determined)

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

3. 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)

  • 5. Scheduled date(s) for submitting proposed licensing action: ---~NIA. __ _
6. Important licensing considerations associated with refueling:
7. Number of Fuel Assemblies:
a. Incore
b. In Spent Fuel Storage
8. Present licensed spent fuel storage capacity:

Future spent fuel storage capacity:

9. Date of last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity:

8-1-7.R4

_193 __

_556 __

1632 1632 March2012

SALEM GENERATING STATION MONTHLY OPERATiNG

SUMMARY

- UNIT 2 JUNE 1995 SALEM UNIT NO. 2 The Unit began the period operating at 100% power and continued to operate at that level until 6/3 when load was reduced to 50% to perform turbine valve testing. The Unit returned to 100% power following the test. A Unit shutdown was initiated on 6/7 to comply with Technical Specification LCO 3.0.3 due to both residual heat removal systems

500KV circuit breaker failed, causing the loss of two 4KV group busses (power to the two reactor coolant pumps). According to commitments from PSE&G and a subsequent confirmatory action letter from the NRC, both Units will remain shutdown pending completion of the following actions:

Complete the Significant Event Response Team (SERT) review of the reactor trip Appropriately address long standing equipment reliability and operability issues After defining the work scope and approach to the outages, meet with the NRC to discuss these plans After the work is completed, conduct a restart readiness review to determine for ourselves the ability of each 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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