NL-88-331, Monthly Operating Rept for Feb 1988

Monthly Operating Rept for Feb 1988

ML20150C246
Person / Time
Site:	Rancho Seco
Issue date:	02/29/1988
From:	Bosakowski P, Crunk S SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
NL-88-331, NUDOCS 8803180099
Download: ML20150C246 (11)

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, 'EEBRUARY 19M SU MARY OF PLANT OPERATIONS The plant was in cold shutdown for the entire month of February. The initial shutdown was due to the Decembar 26, 1985, loss of Integrated Control System power event.

PERSONNEL CHANGES REQUIRING REPORT Four management position changesf'equire reporting pu'rsuant to Technical Specification Figure 6.2-2. Paul' J. Bender is acting Manager, Hodifications; Paul J. Lavely is acting Manager,Mnvironmental Protection; H. John Sefick, Jr. is acting Manager, Nuclear Security; and Gene E. Hinsley and James H. Reese are sharing Radiation Protection Manager responsibilities.

Hr. Bender, acting Manager, Hodifications, has over 18 years of nuclear experience with Bechtel Power Corporation and the Sacramento Municipal Utility \

District. His experience includes supervisor / lead engineering positions in quality assurance / control, electrical engineering, and construction. He also has engineering experience in test, reliability, systems and startup.

i Hr. Bender has a Bachelor of Science degree in Electrical Engineering from the l University of Dayton and is a registered Professional Engineer.

Hr. Lavely, acting Manager, Environmental Protection Department, has been involved with health physics / radiation protection for over 18 years, is experienced in emergency planning, and has held several supervisory positions.

Prior to coming to the Sacramento Municipal Utility District, he worked for Impe11 Corporation, Detroit Edison Company, Illinois Power Company, Ingalls Shipbuilding Company and Purdue University. Mr. Lavely has a Bachelor of Science degree in Radiological Health from Purdue University.

Hr. Sefick, acting Manager, Nuclear Secueity, has 12 years experience in security for nuclear power facilitW, J years in law enforcement, and is experienced in security management. He has worked for the Clemson, South

! Carolina Police Department, Duke Power Company, South Carolina Electric and l Gas Company, and Tennessee Valley Authority and currently is a Nuclear Security r Consultant. He has a Bachelor of Arts degree in Interdisciplinary Studies from l University of South Carolina. Mr. Sefick is a member of the Domestic Safeguards Subcommittee of the Atomic Industrial Forum and an Edison Electric Institute Security Committee member.

l Hr. Hinsley and Mr. Reese are sharing the responsibilities of the Radiation l Protection Hanager. Mr. Hinsley and Mr. Reese are providing management and technical direction of the department, respectively. Mr. Hinsley has 15 years of nuclear experience including management, nuclear plant chemistry, radiation safety, environmental protection, and training. Mr. Minsley has a Bachelor of Science degree from Utah State University and currently works for General Electric Company. Hr. Reese has nine years of reactor health physics experience and has worked for Virginia Pater, GPU and the S:,cramento Municipal Utility District where he is currently the Radiological Health Superintendent.

He has a Bache *or of Science degree in Biology with minors in Health Physics and Psychology.

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ff SLM4ARY OF CHANGES IN ACCORDANCE HITH 10 CFR 50.59 The plant staff accepted documentation packages in February 1988 for the facility changes described below which required detailed safety analyses.

These changes were reviewed in accordance with the Technical Specifications by Review Committee (PRC) and the Management Safety Review Committee the (MSRC).gPlant;There were no documentation packages completed for procedure changes, tests or experirr.ints during February 1988. ,

1. ECN A-4711, Rev. 3, provided changes to correct fabrication deficiencies .

and improve performance and ALARA considerations for the Post Accident Sampling System (PASS). Changes included relocating the demineralizer water / nitrogen dilution manifold, and installation of an air conditioning 1 system,to cool control panel H3ESA and a stainless steel tank equipoed iwith a chiller unit.

>- These changes do not impact plant safety or involve a change in the Technical Specifications or plant operating procedures. An unreviewed Safety Question is not involved. (Log No. 462)

2. ECN A-5743, Revision 7, provides safety improvements to the instrument air system by providing' backup air in the case of the loss of rarmal and diesel driven compressor air. The Safety Category Instrument Air System (SClAS) compressed air bottle system automatically provides the air to the g( Atmospheric Dump Valves (ADVs). The SCIt.S has a provision for attaching replacement . bottles which could be used in support of 10 CFR 50, Appendix R requirementsi These modifications are a safety improvement to the Instrument Air System and are bounded by the Licensing Design Basis.

T})cprobabilityofoccurrenceoftheconsequencesofanaccidentor malfunction'ef equipment important to safety is not increased because the modifications are a safety improvement ti.at provides backup instrument air in the case of the loss of normal and diesel driven compressed air. The possibility of an accident or malfunction of a different type, namely f failure of the backup system, was not created due to provisions in backup air system design that mitigates the failure of the pressure reducing

' station by built-in pressure relief valves.

Technical Specification additions detailing operability and surveillance requirements for the installed backup air syste.t hr.ve been approved 5 s (Amendment No. 93). The margin of safety as defined in the basis for any Technical Specification is not reduced because the modifications are a safety improvement that provides backup instrument air; therefore, this modification does not involve an Unreviewed Safety Question. (Log No.

662, Rev. 5)

3. A backup instrument air system was installed under ECN R-0859, Revision 4, i as a safety improvement to overcome loss of normal ti strument air supply i S to assure operation anJ control of the Main, Startup and Auxiliary I

Feedwater (MFH, SFH and AFH) Valves, Turbine Bypass Valves (TBVs) and Compc,nent Cooling Hater (CCH) Safety Features Valves. These modifications are a safety improvement to the Instrument Air System and are bounded by the U censing Design Basis.

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/ The probability of occurrence of the consequences of-an accident or ' [

ma?fugction of equip;pnt important to safety is not-increased because the mdfications are a safety improvement that provides backup instrument air in the case of the loss of normal and diesel driven compressed air. The passibility of an accident or malfunction of a different type, namely ,

failure of the backu

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air system design at %p system,the mitigates is not created failur6'9f the due to provisions pressure reductionin backup i) station ty built-iipressure relief valveig I 7

.l The margin of safety in the basis for-any Technical Specification is not i reduced because the nodifications are a yfety improvement-that proddes

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backup instrument air, Thb' modifications covered by ECN R-0859, ReNih /s {'; ' '

do not involve an Un hviewed Safety Question. (LogNo.824,.Pav,.f4)

4. ECN R-1921, Revision 2, provided an alternate drain (West Decay bpat Pump ' ,

Room Sump) for the Spent Fuel Pool (SFP) liner leakage collectka / system. /

A rectangular opening with a hinged dever plate in tlte,SFP Li",er Cavity Cain cover plate was also provirjed to facilitate the monitgrirg' of SFP Liner leakage. These mod fications enhance the contrpl of dtaminated fluids and reduce. radiation exposure of plant personnel in the vicinity of the SFP.

The new configuration introduces the potential of overflowing the Hest Decay Heat Pump Room Sump. More fluid can be introduced into the sump than the installed pumps can remove. That accident is not included in the USAR, but is bounded by other accideAts in the USAR. The estimate of the whole body exposure to site boundarp personnel as a result of the change is bounded by calculations performed prior to the change. An Unreviewed Safety Question is'not involved. (Log No. 1028)

5. The RHUT 'A' and 'B' transfer pump impeller was upgradea'under sub ECN R-0775E to allow faster dump of the RHUTs to the b:. sins. A parallel pump was provided to ensure pumped discharge (vs. gravity drain) if P-683 is out of service. Also, all interconnecting piping required to z accommodate the changes specified in ECN R-0775, Revision 3 was installeo under this sub ECN. All wiring needed to connect the new pumps, RHUT equipment, sluiceable Demineralizer equipment, and effluent strainer to i plant power sources ar.d ical' control panel H4HR was installed under sub ECN R-0775I.

These modifications improve the plant's ability to safely process and dispose of waste water following an OTSG tube leak. These changes result in better control of the radioactive waste water systems. Rancho Seco's design basis is not altered by these changes. The modifications do not affect previously analyzed plant failure modes and no new failure modes are introduced by the changes; therefore, the implementation of sub ECNs R-0775E and I does not involve an Unreviewed Safety Question.

6. The Emergency Feedwater Initiation and Control System (EFIC) was installed i under subs to ECN A-5415 Major, Revision 4, per NUREG-0737, Item II.E.1.2 requirements. EFIC and EFIC related AFH, MFH, and Main Steam System changes, including all process inputs to EFIC, were implemented under ECN A-5415.

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i EFIC is bounded by the design basis and the safety analysis as described in the USAR. EFIC does not increase the probability of occurrence or the-consequences of an accident or malfunction, nor does it create the possibility for an accident or malfunction of a different type because the September 26, 1983 NRC SER documents the acceptability of the design basis. It is also an upgrade of existing plant systems and enhances SAR accident analyses.

EFIC increases the margin of safety by upgrading the reliability of the existing plant systems. Amendment No. 93 to the Technical Specifications incorporates changes as a result of ECN A-5415, Revision 3; therefore,

.this modification is not an Unreviewed Safety Question. (Log No. 563)

7. Under ECN R-0415 Major, Revision 1, time-delay relays were installed on all four diesel generator breakers to delay the auto-closing of the breaker in the event of a manual trip during surveillance and periodic tests (sub ECN R-0415A). Also, synchronizing check relays were added to the control circuitry of both the emergency diesel generator breakers and the normal supply breakers (sub ECN R-0415B).

These modifications are operational enhancements to the 4160V Auxiliary System and the Emergency Generator System. The modifications provide:

1)-additional assurance that manual paralleling of non-synchronized 4160V bus sources does not occur, 2) a time delay in the control circuitry of all four diesel generator breakers to prevent immediate breaker reclosure following a manual trip, and 3) an improved method for performing required surveillance. The three new failure modes introduced by ECN R-0415 modifications are bounded by the Licensing Design Basis. No Unreviewed Safety Question is involved. (Log No. 743A and 8)

8. ECN R-1045, Revision 1, provided 4160V Class 1 bus overvoltage/

undervoltage protection modifications and 480V Class 1 bus overvoltage alarm removal. These modifications are operational enhancements to the Electrical Distribution System and are bounded by the License Design Basis. Failure modes were analyzed and it has been dett.rmined that there are no significant failures to compromise plant safety.

The probability of occurrence or the consequences of an accident or malfunction of Equipment important to safety previously evaluated is not increased because the elimination of the overvoltage trip reduces the

> probability of standby diesel generator actuation due to trant'ent

, overvoltage conditions while providing enhanced control (alarm instead of

, trip). The possibility of an accident or malfunction of a different type than any evaluated previously is not created because this modification is an operational enhancement and is within the USAR design basis.

4 The margin of safety as defined in the basis for any Technical Specification is not reduced because these modifications do not change the switchyard voltage allowable operating range and they reduce the probability of standby diesel generator actuation. These modifications do not involve an Unreviewed Safety Question. (Log No. 858, Rev. 4) i 4

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.' 9. ECN A-3660Z, Revision 2, modifications expanded the electrical distribution system to include tie-in of the new diesel generators to provide emergency power for Nuclear Services buses S4A2 and S482. These modifications to the emergency power system provide a positive affect on nuclear safety and do not introduce new failure modes.

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 because the addition of the TDI diesels is bounded by the analyses in the USAR. The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created because the installation of the TDI diesels increases the capacity of the emergency diesel generators while maintaining redundancy and independence of trains.

The margin of safety as defined in the basis for any Technical Specification is not reduced by the deletion of the SFAS signal to start the CR/TSC Essential HVAC System because the signal was an electrical necessity only which is no longer needed, and the modifications to the electrical systems maintain the single failure criteria for auxiliary electrical systems. Implementation of the modifications under this ECN does not involve an Unreviewed Safety Question. (Log No. 307, Rev. 10)

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

1. HVAC air balance testing for the Control Room and Technical Support Center has been completed.

2. TDI diesel modifications have been completed.

3. The Loss of Offsite Power Test has been completed.

4. ECN R-2079, Revision 1, provided for TDI Diesel Generator Trouble Indication in the Control Room. Eight Diesel Generator trouble alarms in the local annunciator of the Diesel Engine Panel were brought up to Panel H2EH in the Main Control Room.

5. ECN R-2733, Revision 0, provided Pitot traverse ports in purge and vent stacks in order to get the accurate airflow data required to perform stack flow profile for sampling and to demonstrate compliance with 10 CFR 20 and 10 CFR 50, Appendix I.

6. ECN R-2715. Revision 0, replaced the reactor head structure ventilating fan motor.

7. ECN R-2697, Revision 0, added local position indicators en three motor operated valves in order to provide accurate unambiguous local valve position indication.

8. ECN R-2663, Revision C. replaced NNI and ICS fuses with longer time delay fuses because the existing fuses were marginal to cover the transient.

9. ECN R-2348, Revision 0, removed the Class 2 Lo-Lo switches from the Class 1 control circuits of NSH pumps P-482A and P-482B becaes9 those switches degraded the reliability and overrode the Safety Feature Actuation of the pumps.

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/ 10. ECN R-2629, Revision 0, installed a jumper to bypass 42/b contact on MCC breaker S2C314 used for indication.

11. ECN R-2612, Revision 1, changed the demineralized water (DHH) system valve line-up to ensure that the miscellaneous water outlets inside of containment are available for use by the fire brigade during a fire inside of containment.

12. ECN R-2527, Revision 0, installed a higher pressure range level transmitter for LT-21503D to meet cystem design requirements.

13. ECN R-2509, Revision 1, fabricated and installed instrument thermal enclosures for RCS pressure transmitters PT-21050 and PT-21051 to reduce the peak ambient temperature seen by the transmitters during accident conditions, thereby maintaining instrument error to an acceptable level.

14. ECN R-2492, Revision 0, added a manual valve in line 95704-10"-LE downstream of FV-95401 to prevent potentially contaminated water from entering Hadselville Creek.

15. ECN R-2255, Revision 0, replaced the portions of the AFH dischargo piping that were carbon steel with stainless steel and provided Class 2 freeze protection replacement. (NCRs S-6081 and S-7176)

16. ECN R-2090, Revision 5, provided refrigeration controls for operation at low load and to minimize wind effect to stabilize operation of the essential refrigeration system for the HVS-NSEB and CR/TSC Essential HVAC.

17. ECN R-1950, Revision 2, added a boric acid flow indicator for ASHE Section XI testing.

18. ECN R-1590, Revision 1, installed bulk raceway to support the modification work to be accomplished during the present outage.

19. ECN R-0546, Revision 0, installed bulk raceway for plant modifications related to the present outage.

20. ECN R-1991, Revision 0, installed and tested new supports to the TDI Diesel Generator Train 'A' and 'B' turbochargers to bring measured vibrations within acceptable limits.

21. ECN R-2713, Revision 0, replaced the carbon steel reducer between valves FV-35109 and HCH-081 with a low alloy steel reducer to provide increased erosion resistance and to improve HCH system reliability.

22. ECN R-2781, Revision 0, replaced the water column drain valve (ASC 746) located at the auxiliary steam boiler.

23. ECN R-1078, Revision 1, resulted in the carbon dioxide fire suppression systems in the NSEB being disabled and compensatory measures initiated in accordance with Technical Specifications. Some water spray nozzles in cable rooms 148 and 149 were plugged to prevent inadvertent water spray into switchgear rooms 146 and 147 through fire damper openings. All NSEB fire detection systems are operable.

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.' 24. ECN R-2070, Revision 0, upgraded the electrical installation in the NSEB and TDI Building for improved Regulatory Guide 1.75 compliance by maintataing minimum separations between redundant channel raceways or ,

Class 1 and Class 2 raceways. l

25. ECN R-2205, Revisica 1, installed four flow transmitters and thermal enclosures. The differential pressure range of the new transmitters provides better indication accuracy at low Auxiliary Feedwater flow rates.

26. ECN R-2761, Revision 0, provided weepholes on EQ junction boxes located inside containment to equalize pressure in the event of a design basis pressure transient.

27. ECN R-2736, Revision 0, added local turbine tripped indicator to turbine triplever to allow for positive local indication of a turbine trip. .

28. ECN R-1054, Revision 0, reconstructed the northern floor area of the Administration Building to support a computer facility. Modi fications included providing fire protection and electrical power requirements for the computer equipment.

29. ECN A-2679, Revision 1, replaced the controller for the motor driven fire pump P-440 with one listed by UL and FM and erected a shelter for the new controller.

30. ECN R-2401, Revision 1, modified HV-50104 (Train 2) and HV-50105 (Train A) from "N0", "F0", to "N0", "FC" dampers so that dampers will close on loss of instrument air and essential HVAC will serve essential electrical equipment adequately.

31. ECN R-2561, Revision 0, replaced breaker DC control fuses in Class 1 4kV and 480V switchgear with proper fuses in order to provide adequate protection and coordination with the upstream devices.

32. ECN R-0952, Revision 2, upgraded SPDS to include Regulatory Guide 1.97 Category 1 variables.

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33. ECN R-2397, Revision 1, replaced the Sky stress cone kit for circulating witer pump P-402A cable 142E038 with a Raychem 15kv stress cone kit.

34. ECN R-2784, Revision 0, replaced time delay relay 62A6 with two qualified Class 1 relays. The replacement relays do not alter the circuit function but continue to provide and maintain a reliable undervoltage protection scheme.

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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: lantarv 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 1

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AVERAGE DAILY UNIT P0HER LEVEL DOCKET NO. 50-312 UNIT Rancho Seco Unit 1 DATE - 2/29/38 COMPLETED BY NW P. Bosakowski TELEPHONE (916) 452-3211 MONTH February 1988 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY P0HER LEVEL (MWe-Net) (MHe-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 l

15 0 31 16 0 l'

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

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e l OPERATING DATA REPORT DOCKET NO. 50-312 DATE .2/29/88.

Mboak wX-COMPLETED BY P. Bosakowski TELEPHONE (916) 452-3211 OPERATING STATUS

1. Unit Name: Rancho Seco Unit 1 Notes:

2. Reporting Period: February 1988

3. Licensed Thermal Power (MHt): 2.772

4. Nameplate Rating (Gross MHe): 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 Which Restricted, If Any (Net MHe): 0

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

11. Hours in Reporting Period 696 1.440 112.824

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 (MHH) 0 0 127.861.688

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

18. Net Electrical Energy Generated (MHH) -7.173 -13.841 38.218.185

19. Unit Service Factor 0.0% 0.0% 44.6%

20. Unit Availability Factor 0.0% 0.0% 45.7%

21. Unit Capacity Factor (Using HDC Net) 0.0% 0.0% 38.8%

22. Unit Capacity Factor (Using DER Net) 0.0% 0.0% 36.9%

23. Unit Forced Outage Rate 100.0% 100.0% 44.4%

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: ___3/20/88

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

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SACRAMENTO MUNICIPAL UTluTY DISTRICT C S201 S Street, P.o. Box 15830, Sacramento CA 95852 1830,1916) 452 3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CAllFORNIA NL 88-331 E R 1 r GSS U. S. Nuclear Regulatory Commission Attn: J. B. Hartin, Regional Administrator Region V 1450 Haria Lane, Suite 210 Halnut Creek, CA 94596 Docket No. 50-312 Rancho Seco Nuclear Generating Station License No. DPR-54 OPERATING PLANT STAWS REPORT Dear Mr. Hartin; Enclosed is the February 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,

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

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

INP0 G. Kalman R. Twilley, Jr.

RANCHO SECO NUCLEAR GENERATING STATloN D 14440 Twin Cities Road, Herald, CA 95638-9799;(209) 333 2935