ML20234E868

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Monthly Operating Rept for Nov 1987
ML20234E868
Person / Time
Site: Rancho Seco
Issue date: 11/30/1987
From: Bosakowski P, Karen Meyer
SACRAMENTO MUNICIPAL UTILITY DISTRICT
To: Martin J
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
NL-87-1584, NUDOCS 8801110216
Download: ML20234E868 (15)
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Text

NOVEMBER 1987

SUMMARY

OF PLANT OPERATIONS The plant was in cold shutdown for the entire month of November. The initial

- shutdown was due to the December 26, 1985, loss of Integrated Control System power event.

PERSONNEL CHANGES REQUIRING REPORT There were no personnel changes reportable pursuant to Technical Specification Figure 6.2-2 as revised in Proposed Amendment No.138.

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SUMMARY

OF CHANGES IN ACCORDANCE HITH 10 CFR 50.59

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The plant staff accepted documentation packages in November 1987 for the facility changes described below which required detailed safety analyses. In addition, procedure changes from the last month are included. These changes were reviewed in accordanc9 with the Technical Specifications by the Plant Review Committee (PRC) and the Management Safety Review Committee (MSRC).

There were no documentation packages compleied for tests or experiments during November 19C7.

1. Intermixing of power / control cables with instrumentation cables was not in agreement with the description in USAR Section 8.2.2.11.H.5.

The work performed under ECN R-1786, Revision 0 included:

  • Rerouting to power / control cable trays the power / control cables presently routed in instrumentation trays A28AAl and A28AB3.

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  • Separating from the instrumentation cables and installing a barrier to maintain that separation from the pcuer and control cables which are routed through instrumentation trays A28AN3 and A28AA3.

The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report was not created.

The margin of safety as defined in the basis for any Technical Specification was not reduced. Implementation of ECN R-1786 did not involve an unreviewed safety question. (Log No. 1022)

2. ECN R-1811, Revision 1 added a second spring and bottom blade reinforcement bar to seven horizontal fire dampers in the Nuclear Service Electrical Building (NSEB) to enable them to close against air flow.

Both the damper's manufacturer and ANI stated that the modification to the dampers was acceptable. A SMUD evaluation (calc. no. Z-FPP-M2250) concluded that the addition of a second spring and bottom blade reinforcement bar doe:; not degrade the existing 3-hour fire rating of the seven fire dampers in qusstion.

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The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report was not created since the functionality of the components and system are not affected by the modifications.

The margin of safety as defined in the basis for any Technical Specification was not reduced since the modifications increased the functional operation of the fire dampers within the HVAC System. An unreviewed safety question is not involved. (Log No. 1001)

3. The District provided an automatic trip of the main feedwater pumps on loss of ICS dc or ac power which yields an automatic response to help prevent overcooling of the reactor vessel. This modification (ECN R-0823, Revision 0) was an operational enhancement and will improve the safety of the plant following loss of ICS at or dc power, ,

This modification to the ICS will help prevent overcooling of the RCS by tripping the main feedwater pumps en loss of ICS ac or dc power. The +24 and -24V dc relays provide a two-out-of-two logic for tripping the pumps.

Provisions are made allowing cperators to restart the main feedwater pumps, if necessary, while a loss of ICS power condition exists. The trip logic for the main feedwater pumps requires that two redundant 118V ac undervoltage relays must both trip to cause an inadvertent trip. The failure analysis shows at worst a single failure could prevent the restart of one of two tripped main feedwater pumps. No single failure of the added equipment can cause the main feedwater pumps to trip.

Based on the above discussion, the changes do not adversely affect nuclear safety. No new failure modes are introduced. The failure analysis in the DBR and the effects on USAR Chapter 14 accident analysis are within the Licensing Design Basis. (Log No. 814)

4. This facility change (ECN R-0896, Revision 1) involved the design and-modifications required to transmit 22 cold shutdown parameter signals from various cabinets in the control room and computer room to the H1CO console. This required the installation of two new RTD signal conditioning modules and the installation of four new loops in parallel with four existing loops.

The facility change represents an operational enhancement to provide backup indication in the control room, independent of Non-Nuclear Instrumentation (NNI) indications, of the cold shutdown parameters of the plant. The signals were isolated from their circuits and are not dependent on NNI power supplies. This backup system is provided in console HlCO and used as backup monitoring in the event of the loss of NNI.

The Class 2 system is properly isolated so as not to affect the Class 1 interfaces. The system enhances the operator's ability to respond to transients by providing trending of key variables as well as providing backup to NNI. This change did not involve an unreviewed safety question.

The Technical Specifications were not affected.

The possibility for an accident or malfunction of a different type than any evaluated previously was not created. (Log No. 862, Revision 1)

5. .As part of the MOV Refurbishment Program performed in response to IE Notice 86-03, ECN R-1600, Revision 0 replaced the wiring and components within the Limitorque motor operator for PV-21520, Pressurizer Spray Bypass Valve. The internal wiring was replaced with qualified nuclear grade wire and motor splices were replaced with environmentally qualified Raychem splice kits or Okonite tape, installed per procedure.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased since the function of the valve is unchanged.

The margin of safety as defined in the basis for any Technical Specification is not reduced, because the modification of the valve is an operational upgrade and does not alter the function of the valve.

Implementation of this ECH does not involve an unreviewed safety question. (Log No. 696'

6. Auxiliary Building batteries had nearly reached the end of their design life. Visual examination of these batteries indicated a marked deterioration of portions of the positive plates. Batteries BA, BB, BC, and BD were replaced with nnw Class 1 batteries in accordance with ECN R-0E08, Revision 1.

New vital batteries BA, BB, BC, and BD meet all required design, operating and surveillance criteria as specified in the USAR and Technical Specifications. Replacement of the NS batteries affects nuclear safety because the NS batteries are a dual train, redundant, safety-grade system, and are designed to support SFAS and RPS and start the diesel generators during a LOOP. Replacement of the NS batteries does not change the faliure mode analysis or the result of any accident. The NS batteries are single failure proof. (Log No. 766)

7. To provide suitable low pressure operating characteristics for the PSV-21511 pilot operated relief valve, pilot valve and main valve springs were replaced with higher rate springs and machine reliefs on pilot seal bushings and pilot disc.

This facility change (ECN A-5426, Revision 0) does not affect operation or reliability of the valve at normal operating or maximum design temperatures and pressures and improves its operating characteristics at low pressure.

The change does not affect nuclear safety, does not alter the facility as described in the USAR, and does not involve an unreviewed safety question.

(Log No. 581)

8. ECN R-0904, Revision 0 modified the control of the CR/TSC Essential HVAC refrigeration system to provide increased cperating flexibility. ECN R-0904B, Revision 2 modified the Essential Refrigeration Pumpdown Control to provide automatic pumpdown control.

The controls for the refrigeration system artion of the CR/TSC Essential HVAC System did not permit sufficient fle) D illity for the system to respond to cooling load conditions. A.uton dc pumpdown control was required because without automatic pumpdown, liquid refrigerant collects '

in the oil separator when the system is shut down.

This facility change provides enhanced control of the CR/TSC Essential HVAC. This will allow for system flexi'ility o to respond to variations in cooling loads without shutting down and to maintain lower room temperatures.

The possibility for an accident or malfunction of a different type than l

any evaluated previously was not created because the change does not affect the design function of the CR/TSC Essential HVAC.

The margin of safety as defined in the basis for any Technical Specification was not decreased because the change improves the contro' but does not affect the design function of the CR/TSC Essential HVAC.

(Log No. 1046)

9. Thrcugh reanalysis required by NUREG-0737, Item II.D.1, SMUD determined that the supports on the safety and relief valve discharge piping required. I modifications in order to withstand the transients described in the NUREG.  !

In accordance with ECN A-4615, Revision 2, and A-46158, Revision 3, ten I existing supports were deleted,13 existing supports were modified and 15 )

supports were added. l l

The coolant pump motor service platform was reinforced by adding bracing {

members and stiffening plates.  !

l The char.ges upgrade the existing support scheme to preclude pipe failure {

under SRV/PORV blowdown. A potential failure mode is being eliminated,  !

and thus nuclear safety is being improved. (Log No. 806)

10. The Limitorque motor valve operato-s specified in ECNs R-0699A, Revision 3, and Is-0699G, Revision 3, contained PVC insulatec diring and components which did not duplicate the configuration ustd d'. ring qualification testing. The aesign cnanges were made in iesponse to  ;

IE Notice 86-03. The changes upgraded the internal wiring and established l the qualification per 10 CFR 50.49 of Limitorque motor operator wiring and  !

components.

l The internal wires of Limitorque motor valve operators were replaced with qualified SIS nuclear grade wire. Space heaters md low point drain plugs were removed, and motor splices outside the li, nit switch ,

compartments were replaced with qualifier' splices. Existing terminal i blocks were removed and control wires laid dowa directly onto the limit switches. The modifications to the motor operators improve nuclear safety. (Log No. 784, Revision 1)

11. The following ECNs upgraded the discs of certain Velan gate valves or replaced these valves with ones capable of withstanding the thrusts required to close the valve at the maximum design basis differential pressure. These modifications were performed in response to IE Bulletin 85-03.

EC_N_.N_umb e r hhg Xumber/ App.)icableJyitsm_

R-1511G, Revision 0 SFV-22006/ Purification & Letdown R-1511H, Revision 0 HV-22007/ Purification & Letdown R-lS110, Revision 0 HV-26517/ Core Flood (Log No. 1015)

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12. ECH R-1993A, Revision 0, installed a jumper in panels H4FCP1, H4FCP2, and j H4FDC02A between the common alarm and trouble output circuits. Existing l windows on annunciator panel H3FPB in the Control Room were relabeled to i indicate both a trouble and an alarm condition. The windows were colored I red. I This facility change does not affect nuclear safety, does not alter the facility as described in the USAR, and does not involve an unreviewed l l safety question. (Log No. 1034) l 1
13. The District completed the refurbishment of the following MOVs as part of the MOV program performed in response to IE Bulletin 85-03.

ECN NLm.tter Valvp_NamAEr_LApplic3hl e SySigm R-0914AC, Revision 0 HV-31826/ Auxiliary Feedwater I R-0914AD, Revision 0 HV-31827/ Auxiliary Feedwater I R-0914AF, Revision 0 HV-20569/ Auxiliary Feedwater i R-0914HA, Revision 0 SFV-23809/High Pressure Injection i R-0914HC, Revision 0 SFV-23811/High Pressure Injection  ;

R-0914HL, Revision 0 SFV-23645/High Pressure Injection j R-0914HM, Revision 0 SFV-23646/High Pressure Injection '

R-0914HN, Revision 0 SFV-23604/High Pressure Injection R-0914HR, Revision 0 SFV-23616/High Pressure Injection R-0914HH, Revision 0 SFV-23004/High Pressure Injection ,

(Log No. 831, Revision 1) -

l R-0968Q, Revision 0 FV-20529/Feedwater 1 R-0968R, Revision 0 FV-20530/Feedwater I R-0968T, Revision 0 HV-20601/Feedwater R-0968U, Revision 0 HV-20611/Feedwater t.-0968V, Revision 0 HV-20560/ Main Steam i R-0968W, Revision 0 HV-20565/ Main Steam R-09682, Revision 0 HV-20597/ Main Steam l R-0968AA, Revision 0 HV-20598/ Main Steam i R-0968AB, Revision 0 HV-32243/ Main Steam j R-0968AE, Revision 0 HV-21505/ Reactor Coolant j R-0968AF, Revision 0 HV-21510/ Reactor Coolant i R-0968AJ, Revision 0 HV-21516/ Reactor Coolant l R-0968AK, Revision 0 HV-21517/ Reactor Coolant ,

R-0968AL, Revision 0 SFV-60001/ Reactor Coolant l R-09688G, Revision 0 HV-36051/ Auxiliary Steam R-0968BH, Revision 0 SFV-46203/ Component Cooling Water ,

R-0968BJ, Revision 0 SFV-46907/ Component Cooiing Hater  !

R-03688L, Revision 0 SFV-53603/HVAC R-0968BM, Revision 0 SFV-53 05/HVAC R-0968BS, Revision 0 SFV-60003/ Reactor Coolant Drain R-0963BN, Revision 0 SFV-72501/ Reactor Sample R-09688X, Revision 0 HV-26513/ Core Flood R-0968BZ, Revision 0 HV-26517/ Core Flood R-0968CA, Revision 0 HV-26518/ Core Flood R-0968CB, Revision 0 HV-26511/ Core Flood R-0968CC, Revision 0 HV-26512/ Core Flood R-0968CD, Revision 0 HV-26515/ Core Flood R-0968CE, Revision 0 HV-26516/ Core Flood R-0968CF, Revision 0 FV-95401/ Drainage & Sewerage (Log No. 849, Revision 2)

14. Procedure A.62 addresses the actions to be taken for the operation of the '

120V ac vital system. This includes startup, operation, and shutdown of the system's inverters.

Procedures OP-C.161, OP-C.162, OP-C.163 and OP-C.164 address the actions to be taken in the event of a loss of 120V ac vital buses S1C, S1C2-1, S10 and S102-1, respectively.

The procedures require a change to Technical Specifications since the procedures describe the vital buses as receiving power from inverters which are not contained in present Technical Specifications. The referenced inverters will replace the vital 120V ac inverters currently listed in Technical Specifications; the present inverters will be abandoned in place.

ECN R-0955 (Log No. 851) installed the new inverters and abandoned the present inverters in place. It also reassigned the power sources to 120V ac vital buses S1A, S18, SIC, and SID. Proposed Amendment 147 (Log 760) lists the new inverters in proposed Specification 3.7.1.I.

The changes covered by ECN R-0955 represent improvement over the existing arrangement regarding the reliability and availability of power supplies to vital 120V ac buses SIA, S18, SIC, and SID.

The calculations discussed in the DBR verify the following:

(1) NSEB inverters SIA2, S182, SIC 2, and SID2 can carry the additional 4 loads assigned to buses S1A, S1B, SIC, and SlD as the result of this design change.

(2) Batteries BA2, BB2, BC2 and BD2 have adequate capacity to supply the loads assigned to 120V at buses SIA, SIS, S1C, S1D, SlA2-1, S182-1, S1C2-1, and S1D2-1.

(D Circuit breakers of the vital 120V ac system affected by this design change, are properly coordinated.

(4) Cable sizes are satisfactory for ampacity and voltage regulation, and the sizing ensures that acceptable voltage levels will be maintained supplied from the inverter or the backup power supply.

l'SAR Section 8.2 is impacted by ECN R-0955. Sub-Section 8.2.2.7 and Figure 8.2-4 are to be modified to reflect this design change. The Associated Technical Specifications (Section 3.7) related to the design change will be modified in accordance with the operational requirements of the system as changed by ECN R-0955. The implementation of ECN R-0955 will not create an unreviewed safety question. (Log Nos. 1055, 105S, 1059, 1060, 1061) l l

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15. The District approved STP.961 (Loss of Offsite Power Test) which conducts a test of the emergency power system. The procedure uses guidance t provided in various Regulatory Guides dealing with emergency power requirements. These regulations, and Technical Specifications, require that the diesel generators start, load in sequence, and add and/or reject the largest single load for a loss of offsite power and SFAS signal. The tests will also demonstrate the independence of the diesel trains.  ;

Loss of offsite power is an analyzed accident in Chapter 14 of the USAR.

The accident analyzed in Chapter 14 occurs from full power, providing a high decay heat load. Interruption of offsite power (for test purposes),

in the present plant condition, is bounded by the analyzed accident and does not introduce a new failure mode.

The STP will only de-energize the Decay Heat Renoval (DHR) System and Reactor Coolant System momentarily duiing tests 9, 10, and 12 through

15. Due to the low decay teat loads, and the short length of time during the testing in which the DHR System and reactor coolant loops will be j de-energized, no safety concerns exist due to the performance of j STP.961. (Log No. 1023, Revision 4)
16. Decay heat pumps P-216A and B have a single boundary between the primary coolant and the Nuclear Service P.eN Water (NSRW) System. Contrary to l USAR 9.4.2.1, a single failure could lead to a direct transfer of i radioactive liquid to the NSRW. (NCR S-7082, Revision 1)

Based on the DHR pump stuffing box failure analysis, decay heat pump stuffing box failure is not a credible failure mode and, therefore, the current configuration is acceptable as-is. 4 l

USAR 9.4.2.1 will be revised to acknowledge that the potential leakage j through the pump casing to NSRW has been evaluated. The dose impact of j accidents will not be impacted because of the documented improbability of j the leak passing through the pump casing. (Log No. 1044, Revision 1) {

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HAJOR SAFETY-RELATED MAINTENANCE, TESTS AND MODIFICATIONS NOT REQUIRING i DETAILED SAFE 1Y ANALYSES

1. The District has completed Seal Injection and Makeup System functional testing.
2. The jacket water return piping has been Installed on'the TDI diesel generators. 1
3. Cold functional pre-operational testing of the Emergency Feedwater Initiation and Control (EFIC) system has been completed. '
4. ECN R-2324, Revision 0, repaired Auxiliary Boiler (E-360) casing welds.

The wolds were cracking frcm thermal expansion; more flexible fabricated I joints were installed. j

5. ECN R-1760, Revision 0, added a time delay relay in the HPI and makeup pump starting circuits to ensure a 3 second timc delay between starting I of DHS pump and HPI/MU pumps.

6 ECN A-5772, Revision 2, installed temperatyro measurement devices on the )

hot leg of the Decay Heat Line. This equipment will be used during the j temperature compensation of the hot leg level measuring system and will j be removed at a later date, l l

7. ECN .S1884, Revision 1, upgraded the HPI and makeup pump minimum flow lines from the stop check to the safety features valves. The piping upgrade and hanger installation ensure isolation of the lines during high  !

pressure injection and compliance with seismic analysis. i 8 ECN R-2141 Revision 0, corrected discrepancies in RPS signal and power grounding.

9. ECN R-0088, Revision 0, installed a hardwire jumper to bypass the auxiliary contact 2C305. This modification was necessary because the i

lower level pressurizer heaters failed, thus eliminating the indication l lights for all heaters.

10. ECN R-1894, Revision 0, replaced the personnel prote-tion insulation on line 32195-2"-GB with new microtherm insulation per disposition of NCR S-6599.
11. ECN A-5518, Revision 0, supplied permanent power to the Freon Decon unit from the local power pane .

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12. ECN R-1630. Revision 1, installed springs and blade locks for vertical I fire dampers in NSEB HVAC ductwork to enable the dampers to close against air flow.
13. ECH R-2151, Revision 0, adjusted the secondary load tap settings of )

station service transformers X43C1 and X43C2 from 4260/480V to 4160/480V.

This modification will eliminate undesired low voltage levels during degraded switchyard voltage conditions.

14. ECN R-2021, Revision 1, installed permanent hardeare to facilitate pitot tube traverses of significant duct mains and branches of supply and return ductwork. Required air flow data for Control Room normal HVAC system balance could not be obtained with the previous configuration.
15. ECN R-0197, Revision 2, revised and added drainage for the TSC Preaction Sprinkler System to meet the requirements of NFPA 13.
16. ECN R-2126, Revision 1, added fcur card readers and associated hardware to personnel access portal (PAP) building. The additional card readers allow Security to meet the 30 minute accountability commitment during an I

emergency as specified in NUREG-0654.

17. ECN R-2241, Revision 0, installed 1" socketweld coupling to line 1 20554-1"-HD (vent line to flash tank). This modification was necessary  !

to allow proper piping fit after RCS fill and vent.

18. ECN A-5765, Revision 3, revised drawings to reflect as-built configuration, relocated terminal block, and renumbered cables ll2J118 and 112J11C in Panel H2AB and Auxiliary Boiler (per disposition of NCR S-4781).
19. ECN R-1948, Revision 1, provided a consistent method of measuring AFH pump bearing temperature.
20. ECN A-4369, Revision 1, changed configuration of auto stop oil pressure control valves to prevent plugging of valve orifices due to contaminants. This change will allow repair of PCVs while on line, if necessary.
21. ECN A-4353, Revision 0, replaced packed valves SFV-70003 and SFV-70001 with packless valves and new operators. The original valves had a history of poor operation, leakage and high maintenance.
22. ECN A-5098, Revision 1, provided 0TSG manway handling / tensioning device on permanently mounted rails for a more efficient alternative for removal and installation of the primary manways. This modification also replaced twelve Velan valves and rerouted existing drainlines around the manway handling device, thus ensuring a high degree of water-tight integrity at the manway joint and eliminating trip hazard around the manway handling device.
23. ECN R-1684, Revision 0, removed neoprene washers under retaining nuts on access doors on air handlers AH-A-545A and AH-A-5458. The removal of the resilient neoprene washers prevents a breach in the CR/TSC pressure boundary.
24. ECN R-2286, Revision 1, corrected a " Blue Ribbon" connector termination problem for cable IR2C210TA.
25. ECN R-1707, Revision 0, installed a travel stop on PV-36014A to limit the steam flow to less than the capacity of the two downstream relief valves PSV-36012A and B.

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26. ECN R-0746, Revision 1, modified the boot seals on heat traced lines (70423-2"-GD, 71125-3/4"-GD, 15113-1"-HD, 711?0-2"-HD, 15077-1"-HD),

reworked pipe insulation at penetration no. 4458 and replaced missing insulation, and modified penetration nos. 2233, 4560, 4561, 4562, 4563 and 4564 to correct problem of the boot clamp touching the heat traced cable. These changes were performed per disposition of NCR 5470.

27. ECN A-5415AG, Revision 0, provided access and support for new H1SS panel in the Control Room (EFIC) by installing Control Room floor penetration.
28. ECN A-5266A, Revision 1, added new resin discharge lines and dewatering skids running from the polishing demineralized beds to the condenser and regenerant holdup tank.
29. The District completed the refurbishment of the following M0Vs as part of the H0V program performed in response to IE Bulletin 85-03.

ECN Number Valve Number /Anolicable System R-0976C, Revision 0 HV-30123/ Gland Steam & Condensate R-0976E, Revision 0 HV-20585/Feedwater R-0976F, Revision 0 HV-20586/Feedwater R-0976J, Revision 0 HV-35108/ Main Condensate & Makeup R-0976AC, Revision 1 FV-33012J/Hain Condensate & Makeup R-0976AG, Revision 0 FV-24022/ Purification & Letdown R-0976AH, Revision 0 PV-21509/ Reactor Coolant R-0976AM, Revision 0 HV-35043/ Demineralized Hater R-0976AQ, Revision 0 HV-21909/ Pressurizer Relief Tank R-0976AT, Revision 0 FV-36543/ Auxiliary Steam R-0976AH, Revision 0 HV-20612/Feedwater R-0976AZ, Revision 0 HV-43010/ Site Reservoir R-0976BB, Revision 0 HV-43012/ Site Reservoir R-0976BC, Revision 0 HV-43401/ Site Reservoir R-0976BF, Revision 0 HV-43603/ Site Reservoir l

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REFUELING INFORMATION REQUEST

1. Name of Facility Rancho Seco Unit 1
2. Scheduled date for next refueling shutdown: March 1. 1989
3. Scheduled date for restart following refueling: June 27. 1989
4. Technical Specification change or other license amendment -required:

a) Change to Rod Index vs Power Level Curve (TS 3.5.2) b) Change to Core Imbalance vs Power Level Curve (TS 3.5.2) c) Tilt Limits (TS 3.5.2)

5. Scheduled date(s) for submitting proposed licensing action: March 15. 1988
6. Important licensing considerations associated with refueling: N/A
7. Number of fuel assemblies:

a) In the core: 177 b) In the Spent Fuel Pool: 316

8. Present licensed spent fuel capacity: 10H0

. Projected date of the last refueling that can be discharged to the Spent Fuel Pool: December 3, 2001 l

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AVERAGE DAILY UNIT POWER LEVEL DOCKET N0. 50-312 UNIT Rancho Seco Unit 1 DATE 11/30/87 COMPLETED BY P. Bosakawski TELEPHONE (916) 452-3211 MONTH November 1987 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 i

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 INSTRUCTIONS On this format, list the average daily unit power level in MWe-Net for each d;ty in the reporting month. Compute to the nearest whole megawatt.

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OPERATING DATA REPORT DOCKET NO. 50-312 DATE 11/30/87 COMPLETED BY P. Bos7kowski TELEPHONE (916) 452-3211 OPERATING _ STATUS

1. Unit Name: Rancho Seco Unit 1 Notes:
2. Reporting Period: November 1987 _
3. Licensed Thermal Power (MHt): 2.772
4. Nameplate Rating (Gross M7ie): 963
5. Design Ele (irical Rating (Net MWe): 918
6. Maximum Dependable Capacity (Gross MWe): 917
7. Maximum Dependable Capacity (Net MWe): 873
8. If changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report, .

Give Reasons: N/A )

9. Power Level to Which Restricted, If Any (Net HHe): 0
10. Reasons For Restrictions, If Any: NRC letter dated 12/26/85 This Month Yr-to-Date Cumulative I
11. Hours in Reporting Period 720 8.016 110.640
12. Number of Hours Reactor Has Critical 0 0 54.322
13. Reactor Reserve Shutdown Hours 0_ 0 10.300.2
14. Hours Generator On-Line 0 0 50.363.8
15. Unit Reserve Shutdown Hours 0 0 1.210.2_
16. Gross Thermal Energy Generated (HWH) 0 0 127.861.688
17. Gross Electrical Energy Generated (MHH) 0 0 41.123.197
18. Net Electrical Energy Generated (HWH) -3.675 -46.422 38.242.363
19. Unit Service Factor 0.0% 0.0% 45.5% i
20. Unit Availability Factor 0.0% 0.0% 46.6%
21. Unit Capacity Factor (Using HDC Net) 0.0% __

0.0% 39.6%

22. Unit Capacity Factor (Using DER Net) 0.0% 0.0% 37.6%
23. Unit Forced Outage Rate 100.0% _.

100.0% 43.0%

24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):

N/A

25. If Shut Down At End Of Report Period, Estimated Date of Startup: 1/15/88 1
26. Units In Test Status (Prior to Commercial Operation): Forecast Achieved I l

INITIAL CRITICALITY N/A N/A i INITIAL ELECTRICITY N/A N/A i COMMERCIAL OPERATION N/A N/A 1

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Osuu . e<Cemee SACRAMENTO MUNICIPAL UTILITY DISTRICT O P O. Box 15830, Sacramento CA 9585Qt .,( 16) 452 3211 AN ELECTRIC SYSTEM SERVIN THE HE CALIFORNIA BIDEC 16 g y ,

NL 87- 1584 December 15, 1987 U. S. Nuclear Regulatory Commission Attn: J. B. Martin, Regional Administrator Region V Office of Inspection and Enforcement 1450 Maria Lane, Suite 210 Walnut Creek, CA 94596 I DOCKET NO. 50-312  !

RANCHO SECO NUCLEAR GENERATING STATION )

q LICENSE NO. DPR-54 OPERATING PLANT STATUS REPORT l

Dear Mr. Martin.

Enclosed is the November 1987 Monthly Operating Plant Status Report for the Rancho Seco Nuclear Generating Station. The District submits this report pursuant to Technical Specification 6.9.3.

Due to an apparent discrepancy in the statistical results of our Eddy Current Examination, the District is deferring the reporting of the examination results. This issue will be addressed independently pending verification of the Eddy Current Examination statistics. i Sincerely, 1

Karl A. Meyer Manager, Nucl r Licensing Enci (5) cc: I&E Hash (12)

F. J. Miraglia, NRR, Bethesda MIPC (2)

INP0 G. Kalman R. Twilley, Jr.

! I ffAY RANCHO SECO NUCLEAR GENERATING STATION O 14440 Twin Cities Road, Herald, CA 95638-9799;(209) 333 2935

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