LR-N970795, Monthly Operating Rept for Nov 1997 for Salem,Unit 2
| ML18106A207 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 11/30/1997 |
| From: | Bakken A, Knieriem R, Todd F Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LR-N970795, NUDOCS 9712180332 | |
| Download: ML18106A207 (8) | |
Text
.
e
- O*ps~G 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 SALEM UNIT NO. 2 DOCKET NO. 50-311 Gentlemen:
IOEC 12 1997 LR-N970795 In compliance with Section 6.9.1.6, Reporting Requirements for the Salem Technical Specifications, the original Monthly Operating Report for November, 1997, is attached.
RBK/tcp Enclosures c
Mr. H. J. Miller Sincerely, A. C. Bakken III General Manager -
Salem Operations Regional Administrator USNRC, Region 1 475 Allendale Road King of Prussia, PA 19046 97i2180332 971130 PDR ADOCK 05000311 R
PDR ldG0.l8 The power is in your hands.
I llllll lllll lllll /llll /llll lllll 1111111/'
- QQ3376o I/
I If I
95-2168 REV. 6/94 I
SALEM GENERATING STATION DOCKET NO.:
MONTHLY OPERATING
SUMMARY
- UNIT 2 NOVEMBER 1997 UNIT:
DATE:
COMPLETED BY:
TELEPHONE:
50-311 Salem 2 12/15/97 R. Knieriem (609) 339-1782 Salem Unit 2 began the month of November operating at full power.
At 2216 November 20, load was reduced to 89.5% to perform a Steam Flow Differential Pressure Transmitter Test.
The unit returned to 100%
power at 0800 on November 21, and remained at full power for the remainder of the month.
DOCKET NO.: 50-311 UNIT: Salem 2 DATE: 12/10/97 COMPLETED BY: F. Todd TELEPHONE:
(609) 339-1316 OPERATING DATA REPORT OPERATING STATUS 1
Reporting Period NOVEMBER 1997 Hours in Report Period 2
Currently Authorized Power Level (MWt)
Max Dependable Capacity (MWe-Net)
Design Electrical Rating (MWe-Net) 3 Power level to which restricted (if any)
(MWe Net) 4 Reason For Restriction (if any)
This Month Yr To Date 5
No. of hours reactor was critical 6
Reactor reserve shutdown hours 7
Hours generator on line 8
Unit reserve shutdown hours 9
Gross thermal energy generated (MWH) 10 Gross electrical energy generated (MWH) 11 Net electrical energy generated (MWH) 12 Unit Service Factor 13 Unit Availability Factor 14 Unit Capacity Factor (MDC) 15 Unit Capacity Factor (DER) 720 0.0 720 0.0 2448156 825430 792965 100.0%
100.0%
99.6%
98.8%
2379 0.0 2125 0.0 5913921 1921763 1727434 26.5%
26.5%
19.5%
19.3%
16 Unit Forced Outage Rate 0.0%
73.5%
17 Shutdowns scheduled over next 6 months (type, date, duration):
18 If shutdown at end of report period, estimated date of Startup:
720 3411 1106
. 1115 None Cumulative 80463 0.0 77355 0.0 193694926 80570361 76430068 48.9%
48.9%
43.7%
43.4%
33.6%
DOCKET NO.:
UNIT:
DATE:
COMPLETED BY:
TELEPHONE:
OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS 50-311 Salem 2 12/10/97 F. Todd (609) 339-1316 MONTH NOVEMBER 1997 TYPE F=FORCED DURATION REASON NO.
DATE S=SCHEDULED (HOURS)
( 1)
(1) Reason A -
Equipment Failure (Explain)
B - Maintenance or Test C -
Refueling D -
Regulatory Restriction E - Operator Training/License Examination F - Administrative G - Operational Error (Explain)
H - Other METHOD OF SHUTTING DOWN THE REACTOR OR REDUCING CORRECTIVE POWER (2)
ACTION/COMMENT (2) Method 1 - Manual 2 - Manual Trip 3 - Automatic Trip/Scram 4 - Continuation 5 - Other (Explain)
DOCKET NO.:
UNIT:
DATE:
COMPLETED BY:
TELEPHONE:
AVERAGE DAILY UNIT POWER LEVEL MONTH NOVEMBER 1997 50-311 Salem 2 12/10/97 F. Todd (609) 339-1316 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MWe-Net) 1 1099 17 1102 2
1104 18 1107 3
1103 19 1109 4
1104 20 1110 5
1103 21 1087 6
1103 22 1109
,* ',*\\
I.
7 1100 23 1105 8
1100 24 1104 9
1101 25 1102 10 1095 26 1100 11 1098 27 1098 12 1102 28 1102 13 1104 29 1101 14 1103 30 1102 15 1104 16 1099
1-DOCKET NO.:
UNIT:
DATE:
COMPLETED BY:
TELEPHONE:
50-311 Salem 2 12/15/97 R. B. Knieriem (609) 339-1782
SUMMARY
OF CHANGES, TESTS, AND EXPERIMENTS FOR THE SALEM UNIT 2 GENERATING STATION MONTH NOVEMBER 1997 The following items completed during November 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 tha~*~h~s~-it~fu~ 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 2EE-0147, Pkg. 1, Control Valve 2CVSS Replacement.
This modification replaced the existing Centrifugal Charging pump Flow Control valve 2CV55 with a design that provides more reliable flow control during normal and depressurized Reactor Coolant system modes of operation.
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-3329, Pkg. 1, Condenser Hotwell Level Control Modifications.
This design change replaced the existing Hotwell Level instrumentation.
It also modified Control Room indication to provide level indication for all six
hotwells and provided trend recording for condensate overflow.
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-3319, Pkg. 1, Feedwater Flow Nozzle Replacement.
This design change involved the replacement of the Feedwater Flowmeter nozzles with ASME flow nozzles and added four new Chordal Type Leading Edge Ultrasonic flow meters.
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 Summary of Safety Evaluations There were no changes in this category implemented during November,. 1997.
Procedures Summary of Safety Evaluations NC.DE-AP.ZZ-0004(Q), Design Drawings.
The proposed change involves nomenclature and *:resp'ons'lbility changes related to the process for controlling engineering design drawings.
This UFSAR 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.
UFSAR Change Notices Summary of Safety Evaluations There were no changes in this category implemented during November, 1997.
Deficiency Reports Summary of Safety Evaluations There were no changes in this category implemented during November~ 1997.
Other Summary of Safety Evaluation Safety Evaluation -
100% Power Operation With Degraded Advanced Digital Feedwater Control System (ADFCS) Median Signal Select (MSS) Function.
This Safety Evaluation evaluated the proposal of "forcing" two of the three loop 2 steam flow channels to predetermined values for use in the non-safety related ADFCS during full power operation in response to the apparent failure of the 2FT523 and 2FA3472 channels.
This Safety Evaluation does not negatively impact any accident response.
It does not increase the probability or consequences of either an accident or a malfunction of equipment important to safety.
Therefore, this Technical Specification Bases change does not involve an Unreviewed Safety Question.