NL-88-223, Monthly Operating Rept for Jan 1988

Monthly Operating Rept for Jan 1988

ML20149H971
Person / Time
Site:	Rancho Seco
Issue date:	01/31/1988
From:	Bosakowski P, Crunk S SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
NL-88-223, NUDOCS 8802220222
Download: ML20149H971 (15)

Text

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JANUARY.1988 SUM 4ARY OF PLANT OPERATIONS The plant was in cold shutdown for the entire month of January. The initial shutdown was due to the December 26, 1985, loss of Integrated Control System power event.

PERSONNEL CHANGES REQUIRING REPORT Two personnel changes which require reporting pursuant to Figure 6.2-2 as revised in Proposed Amendment No. 138 were made during January 1988.

Bob G. Croley is the acting Assistant General Manager, Technical and Administrative Services and Steve L. Crunk is the acting Manager, Nuclear Licensing Department.

Hr. Croley, acting AGH, Technical and Administrative Services, has over twenty years of engineering, licensing, project management, training, and plant operations experience in commercial nuclear power generation with Babcock &

Hilcox, Westinghouse, South Carolina Electric and Gas Company, and Sacramento Municipal Utility District. He has a Bachelor of Science and a Master of Science in Nuclear Engineering from University of Tennessee and Kansas State University, respectively. He was certified at the SR0 Level by Westinghouse on the Zion Station.

Mr. Crunk, acting Manager, Nuclear Licensing, has over 15 years of nuclear experience with Commonwealth Edison Company and the Sacramento Municipal Utility District. His experience includes engineering management, licensing, power plant startup and testing, construction, planning and scheduling, and Quality Assurance in the area of nuclear fuels. He has a Bachelor of Science degree in Physics /Hath from Missouri Valley College and a Master of Science degree in Nuclear Engineering from University of Missouri-Columbia. He is a Registered Professional Engineer in Nuclear Engineering and was certified at the SR0 level for a Westinghouse PHR.

SUW4ARY OF CHANGES IN ACCORDANCE HITH 10 CFR 50.59 The plant staff accepted documentation packages in January 1988 for the facility changes, tests and procedures described below which required detailed safety analyses. These changes were reviewed in accordance with the Technical Specifications by the Plant Review Committee (PRC) and the Management Safety Review Committee (HSRC). There were no documentation packages completed for experiments during January 1988.

1. ECN R-1656, Revision 2, installed three drain sumps with pumps at three different locations within the Tank Farm area. Potentially contaminated condensate from various Auxiliary Steam System steam traps is routed to the three new sumps. Water pumped from the three new sumps is routed to the Regenerant Hold-Up Tanks (RHUTs) for disposal.

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J All equiprent installed under this ECN is Quality Class II. All porer requirements are non-Class 1. The amount of water to be pumped into the RHUTs as a result of these modifications is negligible, and the quantity of radioactive material to be added to the RHUTs will not exceed Technical Specification requirements (Section 3.17.3).

ECN R-1656 did not alter radioactive liquid effluent instrumentation but added HI-HI level annunciation on the Regenerative Haste Transfer panel H4RH. It affected control room instrumentation in that it added one input to the RHUT trouble alarm on panel H2YSA.

These modifications did not meet the definition of "Hajor Change to Radioactive Haste Treatment System" as given in Technical Specification Section 6.17.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously eva M ted in the safety analysis report was not increased because the consequenc u of any spill are bounded by a rupture of the RHUTs, as given in tue Baus for Technical Specification Section 3.17.3 Liquid Holdup Tanks.

The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report was not increased since the modification redirects the flow of potentially contaminated liquid from an unmonitored, off-site release pathway to a known point which can be sampled prior to release to the environment The margin of safety as defined in the basis for any Technical Specification was not reduced because any failure postulated as a result of this modification is bounded by the Bases for Technical Specification Section 3.17.3, Liquid Holdup Tanks.

The implementation of ECN R-1656, Revision 2 did not involve an Unreviewed Safety Question. (Log No. 997, Revision 1)

2. Under ECN R-0955, the power source of vital 120V ac buses S1 A, S1B, SIC and SID were changed from Auxiliary Building inverters lA, 18, 1C and 10, respectively, to Nuclear Service Electrical Building (NSEB) inverters S1A2, $182, SIC 2 and SID2, respectively. Regulating transformers, static transfer switches, independent vital 480V ac backup sources and manual bypass switches were added to each NSEB inverter to further increase the availability and reliability of these vital 120V as buses.

ECN R-0955 did not change the assignments of individual loads on their i respective inverter buses. It changed only the power sources to the  !

Auxiliary Building inverter buses. The design change and the associated l calculations verify that adequate power is available at the assigned sources and acceptable voltage levels are maintained, feeding either from the inverters or the backup sources. This design change improved the availability and reliability of the power supplied to the Auxiliary Building inverter buses without any detrimental effects on the power supplied to the NSEB inverter buses. This design provided two paths of standby ac power to each inverter bus backed with the second diesel generator of the same train. One path is through the standby battery charger, vital 125V dc bus and inverter; the second path is through the NJ voltage regulating transforcer to the vital 120V ac bus. Additionally, a systems interaction study of the "Two Diesel Generator Electrical Train Design (ERPT-E0179)" was performed and results confirm that there is very good "intratrain load separation" of electrical loads. No unacceptable system interactions were noted in cases where plant systems are supplied by both diesel generators in an electrical train.

A failure modes analysis is presented in the DBR for ECH R-0955. Two additional failure modes were introduced associated with the failures of the static switch or manual bypass switch. These failures are mitigated by the transferability of the inverter bus to the standby power source.

The implementation of ECN R-0955 does not create an Unreviewed Safety Question. (Log No. 851)

3. The refurbishment of 30 HOVs was completed as part of the MOV program performed in response to IE Bulletin 85-03 under Major ECN R-0914, Revision 3. Nuclear safety was improved as a result of the changes.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report was not increased because the H0Vs were modified or repaired to ensure that the valves operate within their design basis.

The possibility for an accident or malfunction of a different type than any evaluated previously in the saf ety analysis report was not created because the design basis of the valves was not changed.

The margin of safety as defined in the basis for any Technical Specification was not reduced because the program ensures that the valves will operate within their design bases; therefore, this is not an Unreviewed Safety Question. (Log No. 831, Revision 1)

4. Hajor ECN R-0968, Revision 4, was initiated to include all safety-related valves not included in the scope of IE Bulletin 85-03 and ECN R-0914.

Seventy-six safety-related HOVs were modified / refurbished. Nuclear safety was improved as a result of the changes.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report was not increased because the HOVs were modified or replaced to ensure that the valves operate within their design bases.

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

This is not an Unreviewed Safety Question. (Log No. 849, Revision 2) l

5. ECN R-0861 Revision 6, provided manual control for the Turbine By-Pass

• Valves (TBVs) and is an enhancement which improves plant stability during a loss of powe- 30 the Integrated Control System (ICS). It provides for operator action to control the position of TBVs following a loss of power to the ICS.

No new system failure modes are identified as a result of this modification. A new type of failure, which is fail-safe, was introduced and was a part of the modification design. The change did not increase the probability of occurrence of a failure, or reduce the margin of safety of any system defined in Technical Specifications, An Unreviewed Safety Question is not involved. (Log No. 833, Revision 4)

6. ECN R-1554, Revision 1 relocated the site meteorological tower 10 meter temperature sensors into a single aspirated ;hield away from road surfaces, installed an aspirator motor failure indication and disconnected both '

dewpoint sensors. This change provides the 'A' and 'B' temperature and delta-T channels with a more representative environment to "sample", and assures a closer matching of the temperature inputs to IDADS. The aspirator motor failure indication assures that repiir is initiated promptly if air flow is lost, to avoid capturing extended periods of unreliable c'ata.

j The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety was not escalated because the '

Meteorological System data retrieval is not linked to plant safety systems, and does not impact re.tual plant operations.

The possibility for an accident or malfunction of a different type, namely the cessation of dewpoint measurements, was not created. Dewpoint l

measurements are not used for dose calculations and are not required to be reported.

The margin of safety was not reduced because the modifications do not impair the ability for making dose projections for routine releases or for emergency response.

The measurement of wind speed, wind direction and atmospheric stability (temperature gradient with height) are being recorded at the appropriate

levels to determine diffusion and transport conditions for gaseous )

! releases. These measurements are sufficient to meet the reporting l 4 requirements relative to the District's demonstration of compliance with 10 CFR 50 Appendix I and for emergency response purposes. (Log No. 905)

7. ECH R-0767, Revision 1, modified the Diesel Generator Building fire alarm system to bring it into compliance with the applicable National Fire

, Protection Association (NFPA) code. The changes to the fire alarm system ,

were limited to panel H4FCP6, its power supply, the local fire alarm  !

bells, and the fire detectors. The sprinkler system changes were confined j to sprinkler piping, heads, and bracing, j The modifications performed under this ECN introduced no new failure modes. I

- It did not affect the safety function of any safety system. The probability of occurrence or the consequence of an accident or malfunction of equipment importhnt to safety previously evaluated in the safety i analysis report was not increased because the functionality of the fire l protection system was not changed. These changes brought the system into l conformance vith applicable NFPA codes.  !

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l 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 changes performed under this ECN did not change the functionality of the fire protection system, but brought the system into <

conformance with applicable NFPA codes.

The margin of safety as defined in the basis for any Technical Specification was not reduced since the changes to the fire protection system in the TDI diesel generator building resulted in the system performing as designed and as required by NFPA code. The implementation of ECN R-0767 did not involve an Unreviewed Safety Question. (Log No.

1019) 1

8. The Interim Onsite Storage Building (IOSB) was constructed under ECN A-3646, Revision 1. This is a shielded storage facility capable of 4

handling two and a half years of waste (generated at maximum historical rates). Solidified liquids, resins and compactable trash will be stored.

The 1058 represents an operational improvement by providing additional ,

onsite storage of low level radioactive wastes generated onsite.

Operation and waste storage in the IOSB is bounded by the Licensing Design Basis because compliance has been demonstrated to the design oDJctives i and criteria of NUREG-0800, Standard Review Plan 11.4.

The use of the IOSB will not increase the probability of occurrence or the r.onsequence of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report because of its physical isolation from site systems and ALARA considerations.

The possibility for an accident or malfunction of a different type than any evaluated praviously in the safety analysis report, namely radiological, was not created by the 10SB. Procedures are in place to minimize radiological exposure. The 1058 facility systems are designed to  ;

minimize startup, pre-operational check-out, and shutdown times to permit quick entry and initiation of facility operation from the normal unmanned l mode. All limits set forth in 10 CFF 20 and 40 CFR 190 are part of the  !

i design basis of the shielding calculations of the IOSB. Also, radiation l monitoring and controls are located throt,ghout the IOSB to facilitate i ALARA design bases.

The margin of safety as defined in the basis for any Technical

l. Specification was not reduced because procedures are in place to safely operate the 1058 facility. Inventory control procedures guarantee cott.pliance with dose requirements objectives and criteria of NUREG-0800, Standard Review Plan 11.4.

Use of the 1058 for the storage of low level waste does not involve an Unreviewed Safety Question. (Log No. 236, Revision 3) i

9. Plant fire protection systems for the Nuclear Service Electrical Building I (NSEB), Diesel Generator Building and the Technical Support Center were l provided/ modified under ECN A-3725, Revision 1. Major. Sub ECN A-3725A extended the CO2 system as a QA Class 2 system and required the NSEB include three-hour rated fire barriers between redundant divisions of safety-related equipment. This assures the same fire protection credit as other areas of the plant containing safety-related equipment.

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Although the CO2 systen tas dorngraded to a QA Class 2, all existing QA records on the system were retained. CO2 equipment located above QA Class 1 equipment is supported to Seismic Category 1 standard and all ,

equipmt.it associated with the bypassing of CO2 in the Diesel Generator i rooms remained QA Class 1; thus, an inadvertent CO2 actuation will not prevent an engine from starting.

The District fire hazards analysis and the subsequent Safety Evaluation by NRC associated with License Amendmont No.19 gave no credit for the QA Class I design and this appears to constitute NRC review of the situation; thus downgrading the QA Class should not be considered an Unreviewed Safety Question.

The changes did not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR or create the possibility for an 1 accident or malfunction of a different type than any evaluated previously '

in the USAR or reduce the margin of safety as defined in the basis for any ;

Technical Specification. This change did not involve an Unreviewed Safety Question. (Log Nos. 279, 279-A Revision 1 and 279-C, Revision 1) f

10. The following sub ECNs related to the implementation of the Emergency I Feedwater Initiation and Control (EFIC) System:

ECN A-5415A installed level taps for the steam generator to meet accuracy requirements of EFIC.  !

ECN A-5415B. Revision 2, installed steam generator level instrumentation i tc provide input to EFIC, ECN A-5415T, Revision 2, installed remote analog control, indication, and EFIC channel A and B closing control logic for the auxiliary feedwater test valve FV-31855.

ECN A-5415X installed analog control for new auxiliary feedwater valves '

FV-20531 and FV-20532 and provided position indication in the Control Room for the valves.

ECN A-5415Y replaced the main steem failure logic vith Class I close signals from EFIC TIE cabinets to main feedwater valycs FV-20525, FV-20526, FV-20575 and FV-20576.

ECN A-5415Z modified the existing electrical and pneumatic control of the atmospheric dump valves to provide analog control only from the EFIC >

cabinets in the NSEB.

ECN A-5415AC installed Class 1 motor operators on two of the six manual i isolation valves for the Atmospheric Dump Valves (one on each steam generator loop).

ECN A-5415AD installed two Class I motor operated isolation valves for the

Turbine Bypass Valves (one for each steam generator loop) and Class 1 [

power to the new motor operators and their switchgear. l r

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EFIC is bounded by the d aign basis and the safety analysis as described

! in the USAR. I': enhances the USAR analysis; therefore, this modification l is not an Unreviewed Safety Question and increased the margin of safety.

l (Log No. 563)

11. Sub ECN A-4711I provided aa alternate source of diesel backed cooling and dilution / flush water for the Post Accident Sampling System (PASS).

This change did not impact plant safety or involve a change in the Technical Specifications. There are no unresolved safety questions. (Leg No. 462I)

12. Gub ECN R-0775A, Revision 1, isolated the make-up demineralizer vaste sumps from the polisher area sump, provided a new sump to collvet wastes and installed curbing necessary to prevent floor drainage entry into the new sumps.

This modificattun improves the plant's ability to safelv process and dispose of waste water following an OTSG tube leak. : an operational enhancement and an inprovement in radiological safety a m atrol.

Previously analyzed plant failure nodes were not affected a..d no new failures were introduced by the chang . The implementation of the sub ECN did not involve an Unrevie d Safety Question. (Log No. 807 Revision 3)

13. Sub ECN R-1672A, Revisica 0, installed two venturis in the Auxiliary Feedwater (AFH) System piping to the 01SGs. Existing flow elements were resized to match transmitter upper range Itmits of 1300 g;,m. The l flow-limiting venturis precluce the possibility of exceeding the minimum )

allowable flow to the OTSCs and reduce the potential for Reactor Coolant 1 System overcooling.

i All mechanical components associated with the sub ICN are Quality Class I l and seismic class 1 and meet ASE code requirements. No additional l failure modes were introduced by this change. This change affected the I range of AFH flow indication cn SPDS and 1080S. l This change gid not increase the probability of occurrence or consequences of an accident already evsluated because the modification did not alter the func'ionality of the AFH System, nor dip ' irecte an accident or '

malfunction of a different type than previour) evaluated since the new

, and replacement components were Quality Class 1. '

Sub ECN R-1672A is covered by Ttchnical Specification Amendment 93, therefore, the: impicmentation of this sub ECN is not an Unreviewed Safety Question. (Log Nu. 982, Revision 1)

14. Special Teit Procedure (STP) 961, Revision 1 Loss of Offsite Power Test, conducts a test of the emergency power systerc. The procedure uses guidance provided in various Regulatory Guides dealing W1th emergency pow r requirements. These regulations and Technical Specifications require tnat tha diesel generators start, load in sequence, and add and/or reject the largest single load for a loss cf offsite power and SFAS signal. The tests will also demonstrate tW.' independencs of the diesel generator trains.

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F-The probability of occurrence or the consequences of an accident previously evaluated in the safety analysis report will not be increased because under existing plant conditions, and with the compensatory measures taken, there will be an equivalent degree of protection in the maintenance of an adequate heat removal capability for the reactor core. The consequences of the loss of forced circulation are bounded by the USAR Chapter 14, "Loss of Electric Power" accident.

The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created because both the loss of normal power supply (for test purposes) and the loss of decay heat removal are bounded by the "Loss of Electric Power" accident analyzed in USAR Chapter 14.

The margin of safety as defined in the basis for any Technical Specification is not reduced because the capability for the reactor core will be maintained, based on compensatory actions and plant conditions.

The perf^rmance of Special Test Procedure 961 does not involve an Unrevieweo Safety Question. (Log No. 1032, Revision 5)

MAJOR SAFETY-RELATED MAINTENANCE, TESTS AND MODIFICATIONS NOT REQUIRING DETAILED SAFETY ANALYSES

1. Filling and venting of the Reactor Coolant System were completed except for the pump sweeps.

2. Cable-pulling for the Auxillary Feedwater system flow indication modifications has been completed.

3. All cold shutdown testing of the Post Accident Sampling System has been completed.

4. Purification and Letdown System logic and main feed pump testing have been completed.

5. ECN R-0914PX, Revision 0, set motor operated valve (MOV) SFV-23604 limit l and torque switches in accordance with engineering documents and drawings.

6. ECN R-1672, Revision 0, installed two flow limiting venturi (FX-20501 and FX-20502) in the Auxiliary Feedwater (AFH) System supply piping to the OTSGs (1800 gpm) to improve AFH flow indication and to limit flow.

7. ECN R-1079, Revision 2, modified fire alarm power supplies for panels H4FCP1 and H4FCP2 in order to eliminate false signals.

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8. ECN A-5619, Revision 2, replaced LSL-47001 and LSL-47002 with appropriate range switches.

9. E94 R-2120, Revision 0, provided power and co'nmunication cables to the Auxiliary Building Hot Laboratory in order to make radioactive effluent data accessible from different locations.

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10. ECN R-1822, Revision 0, replaced the Borated Hater Storage Tank Low-Low level switch LSLL-25008B with a new pressure switch with sufficient accuracy.

11. ECN R-2233, Revision 2, resulted in catA' being rerouted, wrapped with I hour fire wrap and pulled in accordance (ith 10 CFR 50 Appendix R, Section III G in fire areas 17, 19, 20 and closed NCR S-7028.

12. ECN R-0914HY, Revision 0, reset the torque switch for HV-20569 and HV-20596 in accordance with information provided in Engineering drawings.

13. ECN R-2267, Revision 0, moved the dewpoint probe HE 8410A on the hydrogen dryer inlet to the vertical portion of the piping to prevent moisture buildup on the dewpoint probe.

14. ECN R-2220, Revision 0, repairet the damaged heat trace heater on pipe 48225 and replaced the insulation over the exposed heat trace heater for the entire pipe.

15. ECH R-1018, Revision 1, drilled and tapped holes to accommodate temporary installation of vibration probes on make-up and HPI pumps.

16. ECN R-0968CH, Revision 0, reset the torque switch for HV-20560, HV-20565, HV-20598 and HV-20597 in accordance with the information provided in the Engineering drawings.

17. ECN R-0285, Revision 0, replaced leaking valves and added new isolation valves for the Post Accident Sampling System Decay Heat Removal System to improve maintainability.

18. ECN R-2015, Revision 1, retagg- 'ay sections and revised the vias to alleviate tray overweight probleo .

19. ECN R-2491, Revision 0, replaced pressurizer heater control boards with upgraded boards with gain and bias controls to allow for full range control of heaters. 1

20. ECN R-1991B, Revision 0, installed new supports as required on the TOI Diesel Generator Train "A" and Train "B" to meet seismic design requirements.

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21. ECN R-0032, Revision 0, changed the power source of existing emergency lighting units and relocated head lights inorder to have all emergency lights available in the event of loss of normal area lighting.

22. ECN R-1907 Hajor, provided heat trace installation supplying freeze protection for EFIC Main Steam Line instrumentation and instrument tubing.

23. ECN R-2473, provided new cables for main FH turbire speed indication and recorder to eliminate false readings.

24. ECN R-0559, Revision 1, consisted of changes in hardware and flow paths to improve reliability and enhance the maintainability of the Post Accident Sampling System.

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25. ECN R-0277, Revision 1, provided interface and first out trip annunciation to K-307A and K-307B Hain Feed Pump Turbines to facilitate analysis of feedwater problems.

26. ECN R-0775H, Revision 1, rerouted the secondary sample waste discharge to the polisher sump and secondary sample cooler service water drains to the oil / water separator, and provided a sampling basin at the polisher sump to improve the ability to process and dispose of waste water following an OTSG tube leak.

27. ECN R-1987, Revision 0, replaced the reactor coolant drainline in order to eliminate radiation in the drainline, a high radiation area, and the need for temporary shielding.

28. ECN A-4968, Revision 1, eliminated the level drainlines and associated Motor Operated Valves (H0Vs) and added four manual isolation valves to the demineralizer drain path to allow for enhanced isolation capabilities and to eliminate system pressure loss and resin loss through leaking valves.

29. ECN R-0976, Revision 3, extended the program of HOVs refurbishment i required by IE Bulletin 85-03 to include Class 2 and 3 H0Vs. Fifty-seven HOVs have been refurbished under this ECN.

30. ECN R-2604, Revision 0, replaced pressure relief valve PSV-26003 to reset the relief setpoint to the Decay Heat System design pressure to which the sample line is exposed.

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s REFUELING INFORMATION REQUEST

1. Name of Facility Rancho Seco Urit 1

2. Scheduled date for next refueling shutdown: _ March.l. 1989 _

3. Scheduled date for restart following refueling: June ?J_,_JS89

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: January 15. 1989

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: 1080

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

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AVERAGE DAILY UNIT P0HER LEVEL DOCKET MO. 50-312 UNIT ... Rancho Se.co Unit 1 DATE 1/31/88 COMPLETED BY P. Bosakowski _

TELEPHONE (916) 452-3211 MONTH January 1988 DAY AVERAGE DAILY P0HER LEVEL DAY AVERAGE DAILY P0HER LEVEL (HHe-Net) (MHe-Net) 1 -Q_ 17 0 2 0 18 0 3 0 19 0 4 0 20 0 5 0 _

21 0 1

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 INSTRUCTIONS i

On this format, list the average daily unit power level in HWe-Net for each day in the reporting month. Compute to the nearest whole megawatt.

OPERATING DATA REPORT DOCKET NO. 50-312 DATE 1/31/88 COMPLETED BY P. 80sakowski TELEPHONE (916) 452-3211 OPERATING STATUS

1. Unit Name: Rancho Seco Unit 1 Notes:

2. Reporting Period: January 1988

3. Licensed Thermal Power (MHt): 2.772

4. Nameplate Rating (Gross HHe): 963

5. Design Electrical Rating (Net HHe): 918

6. Maximum Dependable Capacity (Gross HHe): 917

7. Maximum Dependable Capacity (Net HHe): 873

8. If changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report, Give Reasons: N/A

9. Power Level to Hhich Restricted, If Any (Net HHe): 0

10. Reasons For Restrictions, If Any: NRC letter dated 12/26/85 This Month Yr-to-Date Cumulative

11. Hours in Reporting Period 744 744 112.128

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 (HHH) 0 0 127.861.688

17. Gross Electrical Energy Generated (HHH) 0 0 41.523.197

18. Het Electrical Energy Generated (HHH) -6.668 -6.668 38.225.358

19. Unit Service Factor 0.0% 0.0% 44.9%

20. Unit Availability Factor 0.0% 0.0% 46.0%

21. Unit Capacity Factor (Using HDC Net) 0.0% 0.0% 39.0%

22. Unit Capacity Factor (Using DER Net) 0.0% 0.0% 37.1%

23. Unit Forced Outage Rate 100.0% 100.0% 44.0%

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

N/A

25. If Shut Down At End Of Report Period Estimated Date of Startup: 3/20/88

26. Units In Test Status (Prior to Commercial Operation): Forecast Achieved l INITIAL CRITICALITY N/A N/A INITIAL ELECTRICITY N/A N/A COMMERCIAL OPERATION N/A N/A l

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F DOCKETNO. 50-112 UNIT SHUTDOWNS AND POWER REDUCTIONS UNITNAME Rancho Seco DATE January 31. 1988 C9MPLETED BY P. Bosakowski REPORT MONTH January 1988 TELEPHONE (916) 452-32 Q c

Er en .3 en :

5X Licensee c 't.

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g. Cause & Corrective

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\" DJ'c i 33 g 255 Event  ::: y o.3 Action to

$5 5 j di y Report a <N 0 Fievent Recurrence d

1 85-12-26 F 744 D 3 85-25 CB Instru Reactor trip due to high RCS pressure Trip preceded by a total loss of ICS power.

I 2 3 4 i I~ Fos ced Reason: Method: Exhibit G - Instructions S S(heduled A Equipment Failure (Explain) I Manual for Preparation of Data B-Maintenance of Test 2-Manual Scram. Entry Sheets for Licensee C-Refueling 3-Automatic Scram. Event Repor (LER) File (NUREG-D-Regulatory Restriction 40ther (Explain) 01611 E-Operator Training & Ucense Examination F-Administrative 5 G Operational Error (Explain) Exhibit I - Same Source s'8/77) Ilother (Explain) 1

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(JSMU-SACRAMENTO MUNIC1FAL UTIUTY DISTRICT C 6201 S Street, Po. Box 15830. Sacramento CA 95852 1830 (916) 452 3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA i

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NL 88-223 February 15, 1988 U. S. Nuclear Regulatory Commission Attn: J. B. Martin, Regional Administrator Region V Office of Inspection and Enforcement 1450 Maria Lane, Suite 210  ;

Halnut Creek, CA 94596 Docket No. 50-312 Rancho Seco Nuclear Generating Station License No. OPR-54 i OPERATING PLANT STATUS REPORT I

Dear Hr. Hartin:

i Enclosed is the January 1988 Honthly Operating Plant Status Report for the Rancho Seco Nuclear Generating Station. The District submits this report pursuant to Technical Specification 6.9.3.

Sincerely, Steve L. Crunk Manager, Nuclear Licensing Enci (5) cc: I&E Hash (12)

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

INP0 G. Kalman R. Twilley, Jr.

4 RANCHO SEOO NUCLEAR GENERATING STATION O 1444o Twin Cities Road, Herald, CA 95638 9799;(209) 333 2935