NL-87-1322, Monthly Operating Rept for Sept 1987

Monthly Operating Rept for Sept 1987

ML20236B614
Person / Time
Site:	Rancho Seco
Issue date:	09/30/1987
From:	Little R, Karen Meyer SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
NL-87-1322, NUDOCS 8710260263
Download: ML20236B614 (17)

Text

{{#Wiki_filter:.y - E, : m ' OPERATING-DATA REPORT ' DOCKET NO. 50-312 DATE-9/30/87 COMPLETED BY R. Little TELEPHONE (916) 452-3211 ' OPERATING STATUS-i l '. Unit Name:' Rancho Seco Unit 1 Notes L2. . Reporting Period:. Seotember 1987 ~3.'

Licensed' Thermal. Power (MHt):=

2.772 Nameplate" Rating-(Gross MHe):. 963 " 4.. LDesignLElectrical Rating (Net MHe): 918

5.

6. Maximum Dependable Capacity'(Gross HHe): 917

7..

Maximum-Dependable Capacity (Net MHe): 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 MHe): 0 110. Reasons'For. Restrictions, If Any: NRC letter dated 12/26/85 j This Month Yr-to-Date Cumulative all. Hours 1in Reporting. Period 720 6.551 109.176' 12. Number of Hours: Reactor Has Critical 0 0 54.322 i

13. -Reactor. Reserve' Shutdown Hours 0

0 10.300.2 114. 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.- LGross Electrical Energy Generated (MHH) 0 0 41.523.197 ] 18.- Net Electrical Energy Generated (MHH) -3.338 -38.325 38.250.460-

19. Unit. Service' Factor.

0.0% 0.0% 46.1% 20.: 1 Unit Availability Factor 0.0% 0.0% 47.2% j

21. : Unit Capacity Factor (Using MDC Net) 0.0%

0.0% 40.1% l

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

0.0% 38.1%

23.. Unit Forced Outage Rate 100.0%

100.0% 42.1% 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: Indefinite 6

26. Units In Test Status (Prior to Commercial Operation):

Forecast Achieved j INITIAL CRITICALITY N/A N/A ) INITIAL ELECTRICITY N/A N/A [- COMMERCIAL OPERATION N/A N/A i fh ' D 0 O a m

'n 2.-

AVERAGE DAILY UNIT. POWER LEVEL DOCKET N0.

50-312 UNIT Rancho Seco Unit'l DATE 9/30/87 COMPLETED BY R. Little TELEPHONE (916) 452-3211 MONTH Seotember 1987 DAY. AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL-(MWe-Net) (MWe-Net) l 1-0 17 0 2 0 18 0 -3 0 19 0 I 4 0 20 0 l 5 0 21 0 l ] 6 0 22 0 7 0 23 0 l I 8 0 24 0 i 9. 0 25 0 j i 10 0' 26 0 11 0 27 0 12 0 28 0 13 0 29 0 14' O 30 0 15 0 31 I '16 0 INSTRUCTIONS j 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. 4

e f Se1 5 bt4 r o N ( 2 omt p e e 1 hei) s e l 3ctL6 'S s nt nF s si a - np 1 C o oa e) 0ae 9 R l iDiR c tcfL 5RSR( u E e e h l orL r v c g a ts o nf( i n O EEYE tc e i t no t e or h o I is r r t o MTBN t u t - tae r N AA D O on c o Gah r r e pe T N D &t t pS R E H Co e t a K T T P iR i e E 1 ct E e y. ib r yt E Pr n1 C N L L eAn u b h tnv1 I e6 s e r x O U P E u v d o EfEE0 D M T a e d r C P p e O i d C r e 4 t c r e e r r w o p o t p c p a i S e r C R T I S N m 7 O 8 u . a )n r I e,' O 8 r mci T 9 a t aSl C 1 U Ejgg s r p c cix r n St D aE e I E

laal m (

b R m duu oe onnt h R e haa ut E,po ~ B eMMAO v, E t t C W o O - m M1 234 e S P D H 3 N T A NO e a SN M e tnr 5 t s W T n e o 2 D O L R e v p O R c iE e 5 T P 8 U ER H S T .EI5 j3 i I N Iig 3 )n U ia j;. j lp x n E o tic (t D s 355 uT t e e r ig r lfiao e F R ec y ggE 0 tnn r g 2 ea o jEy 7 mnnt i el at pt el ai a

ninu u o uif ge s qa e e p a E. M R R e

,gH F RABCDEF 2 6 2 te 2 a 1 D d 5 8 le d u ed rc e h o c F S w. 1 FS 1 l l

SEPTEMBER 1987 i

SUMMARY

OF PLANT OPERATIONS l The plant was in cold shutdown for the entire month of September. The initial shutdown was due to the December 26, 1985, loss of Integrated Control System power event. PERSONNEL CHANGES REQUIRING REPORT 1 There were no changes in personnel which require reporting pursuant to Technical Specification Figure 6.2-2.

SUMMARY

OF CHANGES IN ACCORDANCE HITH 10 CFR 50.59 The plant staff accepted documentation packages in September 1987 for the facility changes described below. In addition, procedure changes from the last month are included. These changes were reviewed and approved by the Plant Review Committee (PRC) and the Management Safety Review Committee (MSRC). There were no documentation packages completed for tests or experiments during September 1987, 1. The Technical Specifications limit the activity in one waste gas storage tank to a maximum of 135,000 curies (Xe-133). The existing monitor did not provide the required measurement capability. Radiation monitor R-1500g was removed and replaced with a new monitor capable of measuring 2 x 10 uCi/cc of Xe-133 and 2 x 103 uCi/cc of Kr-85. The new monitor does not change the system configuration and meets the measurement requirements of the Technical Specifications. It will not create any failure mode other than those inherent in the monitoring system while providing the increased monitoring capability. The change (ECN A-4260, Revision 0) will 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 accident or malfunction of a different type than any evaluated previously in the FSAR, or reduce the margin of safety as defined in the basis for any Technical Specification. A change in the Technical Specifications or an unreviewed safety question is not involved. (Log No. 394). 2. ECN R-1025, Revision 0, installed two ISCO proportional flow composite water samplers in the Clay Creek effluent stream outfall. Additionally, this ECN directed the construction of a wooden housing for the sampler and the installation of electrical conduit to supply 120V ac power to the samplers. One of the samplers is to be placed by the State of California. The samplers (one belonging to SMUD and one to the State of California) were placed in accordance with DGM, Nuclear Directive, ND-86-15-A, " Commitments to Regulatory Agencies", dated August 29, 1986. The sampler is part of Rancho Seco's environmental surveillance program to ensure compliance with 10 CFR 50, Appendix I..

l The two portable cater samplers in the Clay Creek effluent stream provide no monitoring or alarm capabilities and the samplers are not capable of initiating any automatic actions in response to analytical results of the samples; therefore, the installation of the samplers did not impact plant Technical Specifications, nor did it represent a change to the facility as described in the USAR. There is no effect on any plant systems or their safety functions. Implementation of ECN R-1025 did not create an unreviewed safety question. (Log No. 893). 3. A time delay was installed in the control circuitry of each of the four emergency diesel generator breakers (4A202, 4B202, 4A08, and 4811) to delay reclosure following a manual trip. The purpose of this facility change (ECN R-0415A, Revision 0) was to introduce a time delay to prevent immediate reclosure of each diesel generator breaker following a manual trip. The time delay is necessary for safe breaker reclosure; it ensures sufficient bus auto-load shedding and decay of motor voltages. This change provides an improved method for performing periodic surveillance testing of the Emergency Generator System (EGS) pursuant to Technical Specification 4.6.2.B (4.6.3.3 after approval of Proposed Technical Specification Amendment No.147), and is an operational enhancement to the EGS. One new failure mode is introduced by this facility change. A hot short of the control circuit caused by the added relay could result in a failure to close a diesel generator breaker when called upon. The consequences of this failure are not more severe than the consequences of other previously analyzed single failures within the EGS. Each EGS train alone is sufficient to accomplish the onsite emergency power supply safety function. An unreviewed safety question is not involved. (Log No. 743A). 4. The radiation monitors installed under ECN A-4714 Revision 3, are additions to the Digital Radiation Monitoring System. Radiation monitor R-15017 (used for the Haste Water Disposal System) was replaced by retention basin liquid radiation monitors R-15017A and B. Liquid effluent radiation monitor R-15017 had a history of operational and maintenance problems. It acted as a crud trap which resulted in increased background radioactivity and increased minimum detectable levels. This modification is a radiological enhancement which increases the accuracy and reliability of the monitoring of plant liquid effluents. Control of plant effluents was improved by installation of an additional monitor upstream of the retention basin diversion valves. All components affected by this ECN are non-safety related, and the replacement detectors and associated equipment are QA Class II. The new monitors use non-Class 1,120V ac electrical power. The modifications performed under this ECN improve the functioning of the Haste Water System radiation monitoring system. -

g 1 o The probability'of' occurrence;or=the consequences of an accident or . malfunction' previously evaluated in the. safety analysis report will not be increased. The modifications provide redundancy of signals to prevent ' unplanned,offsite. releases _and do not alter the functionality of the, system. L _ The margin of: safety as defined in the basis for any Technical. Specification is not reduced since the. capability of the replacement: ? components are an improvement on the capability of the existing components. An unreviewed safety question is not-involved. (Log No. 449,' Revision 2).- I5. Both Control. Room / Technical Support. Center (LR/TSC) Essential condensing. 1 units U-545A-and U-545B, were modified to provide a means for isolation - and calibration.of three pressure switches associated with each unit's. compressor..' Manual. isolation valves and test connections were installed between the pressure switches and the pressure switch connections to the j refrigerant system.. In each' unit a manual isolation valve was installed in the oil return line between the oil separator and the compressor. This facility change (ECN R-0789, Revision 1) was an operational enhancement to the CR/TSC Essential.HVAC System. The pressure switches could not be calibrated in place without loss of refrigerant from the system. This modification, by providing manual isolation valves and test connections, permits in place-calibration of the pressure switches while. preserving refrigerant system integrity. One new failure mode, the possible inadvertent closure or. blockage of the ' installed manual isolation valves, either before or after System actuation, was introduced by this facility change. This postulated. I failure could. result in either prevention of one Essentia1' HVAC-System train from. operating without operator corrective action, or prevention of one of the compressor protection functions from de-energizing the compressor as required in response to an abnormal operating condition. 'In either case, only one CP/TSC Essential HVAC System train is affected. Each train alone.is sufficient to accomplish the required CR/TSC~ l habitability' safety function. This is not a change to the facility as described in the USAR. A change or addition toLTechnical Specifications is not required. This modification is bounded by the Licensing Design Basis. An unreviewed -safety question is not involved. (Log No. 813). 6. This modification (ECN A-5672, Revision 1) provided Class 1 qualified r Reactor Coolant System hot leg temperature indication in the SPDS. This j was done to conform to the requirements of Regulatory Guide 1.97, Revision 3. Qualified Reactor Coolant System hot leg temperature signals (TE-21030 and TE-21033) were extended for Reactor Protection System cabinets (H4PR2B and H4PR2C) located in the CR to the SPDS/IDADS multiplexer (H4CDAR5 and H4CDAR3, respectively) located in the east and west 4160V switchgear rooms. i

1 _The signal cable was routed through Class 1 racetays to ensure that the required physical separation is maintained. SPDS/IDADS software was modified to accommodate the two new tag numbers (computer points T9012 and T9013) and their multiplexer addresses. Technical Specifications were not impacted and no new failure modes to the Reactor Protection System were introduced. The modification does not affect the USAR, nor does it change System Design Bases; therefore, an unreviewed safety question is not involved. (Log No. 651). 7. ECN R-1259, Revision 0, modified a structural angle on pipe support tag No. 1G-21581-N1 which supports the pressurizer high point vent piping. The support was notched to clear the new motor operator for valve HV-21505 (PORV isolation valve). The notched area was reinforced to restore the structural support capacity. This modification resolved the interference between the new motor operator for valve HV-21505 and the pipe support on pressurizer high point vent line #21581-1" CA. This change results in only a slight change to the support and does not affect the ability of the support to perform its intended functions. This change does not involve an unreviewed safety question nor affect the USAR or Technical Specifications. (Log No. 881). 8. Both NSEB Essential HVAC System refrigeration condensing units were modified to provide a means for isolation and calibration of three pressure switches associated with each train's compressor. Manual isolation valves and test connections were installed between the pressure switches and the pressure switch connections to the refrigerant system. Also, in each train, a manual isolation valve was installed in the oil return line between the oil separator and the compressor. This facility change (ECN R-1022, Revision 0) is an operational enhancement to the NSEB Essential HVAC System (NEH). The pressure switches could not be calibrated in place without loss of refrigerant from the system. This modification, by providing manual isolation valves and test connections, permits in place calibration of the pressure switches, while preserving refrigerant system integrity. The manual isolation valves in the oil return lines allow for the removal or maintenance of the compressors, while preserving system integrity. One new failure mode, the possible inadvertent closure or blockage of the installed manual isolation valves either before or after system actuation, has been introduced by this facility change. The consequences of this failure are not more severe than the consequences of other previously analyzed single failures within the NEH. The consequences of this failure, the loss of one train of NEH, and subsequent loss of one train of NSEB electrical equipment are mitigated by the proper functioning of the other train of NEH which will maintain sufficient NSEB electrical equipment to accomplish the associated safety function. Each train of NSEB electrical equipment and associated NEH equipment is sufficient to accomplish the safety function. _ _ _

C

,f 4

This isJnot a~ change'to.the facility as; described in-the'USAR. A change 1 m or addition 1to Technical: Specifications is not required. This- -J modification-is _ bounded by the Licensing Design. Basis. ' An unreviewed. 1 ~ 1 safety question is not involved. (Log No. 877). { 9. ECN R-1402,- Revision-0, provided new V-belt drives for CR/TSC Essential e HVAC air handling units AH-A-545A and AH-A-5458..The replacement V-belt drives reduce fan speed-approximately 20%. Previously, operation of the CR/TSC Essential HVAC System resulted in i undesirable sound levels in the CR..- Sound levels in the CR were;high enough to. impair communications. Additionally, air ' distribution was uneven. An' analysis of air distribution through the Essential HVAC System was . performed for the CR, computer' room, and TSC. Using the results of recent tests, refrigeration = calculations, and the analysis, the air supply quantities and temperatures of the CR/TSC Essential HVAC air distribution system were rebalanced to optimize air handling unit . performance'(provide adequate air flow to all areas while minimizing sound levels in the CR).. - The modification performed is an operational enhancement which lowers sound levels throughout the CR when'the Essential HVAC is operating and

improves air distribution to provide adequate cooling throughout the CR,

-. computer room,~and TSC. .This-change does not increase the probability of occurrence or consequences of an accident already evaluated because the replacement i equipment installed is similar in function.to the existing components. An accident or malfunction of'a different type than previously evaluated-is not created..This change does not reduce the margin of safety as defined in Technical Specifications, nor does it involve an unreviewed i ' safety; question'. (Log-No. 915, Revision 1). l 10. The standby battery chargers H4BAC and H4BBD were energized from MCC breakers S2B132 and S2A128, respectively. This power supply configuration represents a cross connection between Train A and Train B. i i -The cross connection of trains was eliminated by providing a power supply ~from the same train as the' power supplies energizing battery chargers H4BA i and H4BC and battery chargers H4BB and H4BD, respectively. The power i supply for the standby charger is from an MCC fed from the same train as the normal chargers to make the A and B trains of safe shutdown equipment completely separate and redundant, j Implementation of ECN R-1127, Revision 0, did not involve an unreviewed safety question. Equipment classification remains unchanged. Plant Technical Specifications were not impacted by the change. The i modification represents a change to the facility within the plant's j design basis. Plant USAR Section 8.2 will be modified to show this l change. (Log No. 898). j a h l

11. Both trains of CR/TSC Essential HVAC System ductwork eere modified (ECN R-0764, Revision.0) to provide a boundary which will prevent fire and smoke originating in the TSC from spreading to the TSC roof or the CR. Fusible link actuated 1-1/2 hour rated fire dampers were installed into l both the supply and return ductwork located in the TSC (Auxiliary Building) roof slab. Access doors were provided in adjacent duct sections to permit manual resetting of the thermal links following damper actuation. Separate smoke dampers were installed into both the supply and return ductwork below the fire dampers in the TSC. The smoke dampers are maintained open electrically and will be spring closed upon smoke detection in the TSC or upon loss of control power. Remote damper operation control is provided in the TSC. Power for damper operation and control is supplied from Class 2 battery-backed distribution panel SIN 1-1. Damper position indication is provided at the remote control switch and at the computer on IDADS. Fire dampers and smoke dampers were installed in the Essential HVAC ductwork in the TSC to protect CR habitability and to satisfy regulatory fire protection requirements. These dampers should prevent the loss of CR habitability in the event of a fire in the TSC. This modification is a change to the facility as described in the USAR. The USAR, the UFHAR, and the CR/TSC Habitability study require updating. A change or addition to Technical Specifications is not required. No adverse impact on nuclear safety has been identified. No unreviewed safety question is involved. (Log No. 803). 12. During modification and upgrade of the Diesel Generator GEA panels, portions of the GEA control wiring were identified as inadequately protected to ensure reliable operation. This modi fication (ECN R-1150, Revision 0) improves the reliability of Diesel Generator GEA. The 50A circuit breaker which was removed had been connected in series with another 50A breaker and was functionally utilized as a local disconnect at diesel control power panel H7J272. This functional application is no longer necessary. New fuses and circuit monitoring relays were installed, and undersized wires were replaced in portions of Diesel Generator GEA control wiring. One existing 50A circuit breaker between diesel engine control power panel H7J272 and the 125V de vital supply power distribution panel SOA was removed. The reliability of Emergency Diesel Generator GEA is improved by this modi fication. The new cables provide adequate ampacity. The new fuses protect control circuits during normal operation. The new monitoring relays permit improved identification of faults. No new safety issues or failure modes have been introduced. No adverse safety impact has been identified. This modification is bounded by the Licensing Design Basis. A change or addition to Technical Specifications is not required. No unreviewed safety question is involved. (Log No. 859). --

l j,' _ ECN R-1272, Revis' ion 0, deleted the NNI generated HPI= flow indication'on 13. m the Safety Features Panel (H2SFA and H2SFB) and replaced it with HPI flow indication -independent of NNI.s This' change'wasi recommended as a result'of the Deterministic Failure j Consequences' Analysis'(DFCA)..It provides Class 2 HPI flow indication i independent of'NNI and NNI-power supplies. This provides a more reliable i . power supply by eliminating the reliance on the NNI power supply (which 1 is in addition to.that for the instrument loop)..This also improves the District's. Regulatory Guide 1.97 position for. HPI flow indication. f This change reconfigure the HPI' flow instrument loops. It does not introduce any new or additional failure modes and eliminates a failure i mode-identified by the DFCA. It-does not require a change to the USAR and is within the Licensing Design Basis.' Rancho Seco Technical Specifications are not affected, an unreviewed safety question.is not i involved, and nuclear safety is not compromised. (Log No. 897). -14.- ECN A-2332 (mostly complete, but not.yet closed out) installed air handling unit A-546 and Nuclear Research Corporation norma 1' effluent monitor R-15546. Revision 1 of ECN A-2332 abandoned R-15546 in place. Normal effluent monitor R-15546 was~ abandoned prior to being placed in service due to concerns about its operational reliability and maintainability. I ECN R-1227, Revision 1, commissions. existing GA wide range gas monitor R-15546A as both the normal effluent monitor and post accident effluent monitor.for the Auxiliary Building Grade Level. R-15546A was originally installed as the post accident effluent monitor'under:ECN A-3683 to. satisfy the requirements of NUR.EG-0737, Item II.F.1. ECN R-1227 also adds high radiation signal to the State Notification System (SNS) .l required by California Senate Bill 1184. This facility change is limited .to procedures and designations, and the SNS.. q These changes have no discernible effect on the probability or i . consequences of malfunctions of. equipment important to safety. An -accident of a'different type than previously analyzed in USAR Chapter i 14.3 is not created since no new radiological sources have been created or identified. The margin of safety as defined in the bases of Technical Specifications is maintained. (Log No. 930, Revision 1). I l 1 1 t

15. The District completed the refurbishment of the follocing HOVs as part of-the MOV program performed in response to IE Bulletin 85-03.

s ECN Number Valve Number /Acolicable System R-0968AM,'_ Revision 0 HV-23801/ Seal Injection and Hakeup R-0968AN, Revision 0 HV-23802/ Seal Injection and Makeup R-0968BR, Revision C HV-53623/HVAC R-09688T, Revision 0 SFV-66308/Radwaste (Log No. 849, Revision 2) R-0914AA, Revision 0 SFV-20577/ Auxiliary Feedwater R-0914HK, Revision 0 SFV-24004/High Pressure Injection R-0914HP, Revision 0 SFV-25003/High Pressure Injection (Log No. 831, Revision 1) The M0V program was outlined to the NRC in letters dated June 16, 1986 and November 5, 1986. The overall engineering package was implemented to prevent the valves from failing on demand in a common mode due to I improper switch settings as described in IE Bulletin 85-03. 16. ECN A-3920. Revision 2, modified the Essential HVAC for the CR as follows: installed two redundant trains of Essential HVAC; integrated the TSC HVAC with the essential CR HVAC; and provided actuation of the new systems upon LOCA, radiological, toxic or high temperature conditions. The HVAC System for the CR was modified to comply with the requirements of NUREG-0660 and satisfies NUREG-0737 Item III.D.3.4. It was procured and installed to appropriate codes and standards as detailed in the Design Basis Report. The USAR and Technical Specification 4.10 will be updated to reflect the modification to this system. (Log No. 338, Revision 1).

17. The following modifications were made under ECN A-5773, Revision 1:

the power supply, cryogenic flask, and intrinsic germanium element used to analyze radio isotopes was moved from its present location near the SCAS (Room 106) to the main entry (Room 102) near the PASS Control Panel; vibration dampening material was installed to insulate the Geli detector and cryogenic flask from building vibrations; the existing spectroscopy amplifier was replaced with a new Canberra amplifier; and a gross radiation monitor was added to the reactor coolant sample line entering the SCAS panel. These modifications improve the reliability and performance of Geli detector AE-7001, reduce operational exposure, provide determination of dilution ratios, and facilitate the reliability of sampling operations. These changes do not alter the facility as described in the USAR. An unreviewed safety question is not involved and the changes do not affect the Technical Specifications. (Log No. 664). - _ - _ _ _ _ -

1 s 18. ECN R-0955, Revision 0, changed the power source of vital 120V ac buses S1A, SIB, SIC and SlD from Auxiliary Building inverters 1A, 1B, 1C and 1D, respectively, to NSEB inverters S1A2, S182, SIC 2 and S1D2, respectively. A regulating transformer, static transfer switch and independent vital 480V ac backup source was added to each NSEB inverter to increase the availability of the vital 120V ac inverter bus. A manual bypass switch was added to each NSEB inverter to make the vital 120V ac q buses operable during inverter / static switch outage or maintenance. l ECN R-0955 does not alter the assignments of individual loads on their respective inverter buses; it alters only the power sources to the Auxiliary Building inverter buses. The change and the associated 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 change improves 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. Two paths of standby ac power are provided to each inverter bus backed with the second diesel generator of the same train. One path is through the standby battery charger, vital 125V de bus and inverter; the second path is through the voltage regulating transformer to the vital 120V ac bus. Additionally, a systems interaction study of "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 generators in an electrical train. The USAR and Technical Specification Section 3.7 will be updated to reflect this change (Log No. 851). 19. Sub-ECN R-1993A provided a common alarm circuit annunciation for each of the local fire alarm panels in the power block structures (panels H4FCP1, H4FCP2, and H4FDC02A). This involved the adding of jumpers or new circuitry, as equired, to the listed local fire alarm panels. These alarms annunciate on control room panel H3FPB. ECN R-1993 and its sub-ECNs simplified the monthly unsupervised circuit test to satisfy the requirements of Technical Specification 4.18.1.3, and removed the need for fire watches at the local fire alarm panels. The fire alarm system is classified as Project Class 22; therefore, the modifications implemented by this major ECN have no effect on any safety system. There is no change in power supply to the panels affected. There are no new failure modes associated with this major ECN. The existing failure mode (loss of CR annunciation) does not affect the activation of any automatic fire suppression system or auxiliary function. This change does not affect the Technical Specifications or the USAR for Rancho Seco. _

" ;,7 4 1.n ~. m. L20. Surveillance Procedure SP.26' " Refueling Interval'Hain Feedwater System Test for. EFIC Actuation,". wasLapproved by the PRC., SP.26 implements the ?-

requirements of Proposed Technical Specification' Amendment No;'152 Table-n L4.1-2, Item 16, Main'Feedwater Isolation Valves' Functional.

Proposed' Amendment No. 152~ incorporates changes to the Rancho Seco ~ >Technica1 Specifications required because of modifications to the -Auxiliary Feedwater System and the addition of Emergency Feedwater-Initiation and Control:(EFIC). EFIC.is bounded by the design basis and the safety analysis as described ' 'in the Rancho Seco USAR. EFIC provides additional-margin to the existing, USAR Chapter :14 analysis. The September 26, 1983, NRC SER documents the acceptability of.the design basis. The system is bounded by the NRC-approved SER dated September 26,1983; however, the reduction in required AFH flow,.until approved by the NRC, involves an Unreviewed-Safety Question. EFIC does:not increase'the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR. It is also'an upgrade of existing plant systems and enhances SAR accident analyses. EFIC does not create.the possibility for an accident or malfunction of a different type than any evaluated previously in.the SAR. Erd does n'ot reduce'the margin of safety. as defined in the bases for' the Technical Specifications. It increases the margin of safety by upgrading .the reliability of the existing plant s9 stems. The reduction in the minimum AFH flow rate '(under worst case conditions)'below that required in-the present Section 4.8 of: Technical Specifications reduces "the margin of safety as defined in the basis for Technical Specifications.' The reduction-of the reqLired AFH flow rate, coupled with the' installation of flow-limiting venturis, is an increase in the margin of safety since it provides the;necessary balance between . protection against design basis undercooking events and protection -against non-design basis overcooling events. j 'Since Technical Specification Bases must be revised to reflect this change in AFH flow rates, Proposed Amendment No. 152, Revision 2, involves an Unreviewed Safety Question'as defined by 10 CFR'50.59(a)(2). q (Log'No. 842, Revision 2). 1 21. A.70, Revision 13, " Safety Features Actuation System," was approved by the'PRC. This revision incorporates changes in which equipment is EFIC actuated rather than SFAS. The SAR for SP.26 (Log No. 842, Revision 2) .is also applicable for this procedure. l 1 J __:_=__

i 22. A.69, Revision 9, " Reactor Protection System," was approved by the PRC. Step 3.7 was added to the procedure due to the installation of EFIC. "If a channel of the RPS is in maintenance bypass, only the corresponding channel oylf.FIC can be bypassed. No testing shall be performed on an RPS' instrument string which supplies an output to EFIC without placing that RPS channel in bypass." Six related changes were made and justified. This procedure revision was approved in conjunction with changes to the Technical Specifications. The RCS high pressure reactor trip setpoint and the anticipatory reactor trip on turbine trip arming threshold have be m increased to 2355 psig and 451. reactor power, respectively. Specification 2.2, " Reactor System Pressure Safety Limits," and Specification 2.3, " Protective Instrumentation Limiting Safety System Settings," have been updated appropriately. These changes meet the established regulatory requirements for the RPS and ARTS. Specifically, compliance has been demonstrated with NUREG-0737, Items II.k.3.2 and II.k.3.7. (Log No. 711). MAJOR ITEMS OF SAFETY-RELATED MAINTENANCE 1. The District completed Helium and Eddy Current Testing of the Once Through Steam Generators (OTSG's). Eleven tubes in the "A" OTSG iere plugged and five tubes in the "B" OTSG were plugged. 2. The Integrated Leak Rate Test (ILRT) was completed. 3. The Post Accident Sampling System (PASS) test failed. The District is developing a recovery plan and will re-test. 4. ECN R-0788, Revision 1, installed two test manifolds in each of the CR/TSC HVAC Essential filtration units (SF-A-7A and SF-A-78). These manifolds will provide uniform sampling during tests of the upstream HEPA and charcoal filters and uniform distribution of the tracers to test the downstream HEPA and chart.oal filters. 5. ECN R-1158, Revision 0, provided means to check for sufficient refrigerant charge under operating conditions. One new refrigerant sight glass was added to each liquid line between the double solenoid valve piping inside of the receiver compartment and the isolation globe valve outside of the unit housing. 6. ECN R-1898, Revision 0, resolved inconsistencies between electrical design documents and equipment configuration for Bruce GM Diesel Generator "B". r 1.f 7. ECN A-4687, Revision 2.-replaced existing circuit breakers and_' main breakers on dc panels 50A2,'S082, SOC 2 and S002 to provide breaker- ~ osition positive. status indication in the CR. Bypass switches were p ' installed to prevent anLalarm in the CR when the circuit breakers.for-the above' mentioned' panels are not in service or are placed under clearance. 8. ECN R-1465, Revision 0, replaced the unqualified tape on certain tags ~ ith qualified Okonite. tape. w

9. ' ECN.R-0022, Revision 2, made security modifications to certain doors and Ladded a telephone to the fourth floor of the'NSEB (between doors NS401-and NS402).

'10. ECN R-1616. Revision 1, installed racking gear guard on 480V loadEcenter . breakers 3A206 and 3B205 to prevent contact between the control wire harness and racking. gear. 11. ECN R-1762, Revision 1, provided ground fault protection for 150A breakers 20244~and 2C208 feeding the H8TJ and H8TK battery building transfer. switches. .12. ECN R-1655, Revision 0, reworked existing cable lY3PC0AC and installed two new cables to provide communications link for the new IBM PC in the TSC. IDADS data was then transferred to floppy disk so as to be available for off-line analysis.

13..ECN R-0904C, Revision 0, installed three adjustable pressure switches to monitor. suction pressure at each refrigeration compressor.

Each of the three compressor unloader valve solenoids were reconnected to a separate . pressure switch to load the compressor in steps as suction pressure rises. L

14. The District completed the refurbishment of the following MOVs as part of the MOV program performed in response to IE Bulletin 85-03.

ECN Number Valve Number /Acolicable System 'l R-0976A. Revision 0 HV-20583/ Gland Steam & Condensate R-0976AD, Revision 0 FV-24019/ Purification & Letdown R-0976AE, Revision 0 FV-24020/ Purification & Letdown R-0976AF, Revision 0 FV-24021/ Purification & Letdown i R-0976AX, Revision 0 HV-98301/ Pressurizer Relief Tank R-0976BD, Revision 0 HV-43402/ Site Reservoir 15. ECN R-1785, Revision 0, rerouted instrumentation cables associated with Decay Heat System cross-tie flow transmitter FT-26049 (Train A). This configuration change corrected the intermixing of power / control and instrumentation cables which is contrary to USAR Section 8.2.2.11.H.5. l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _

.m\\q, . v), v :.; g& L .'nU N q.7 t, .q r.. o..

a.1 REFUELING, '

INFOR,MATION REQUEST, y

.,4 A i -{ N

I

.j ). .y, ,1 , j, i - 1.. Nametof Facility - D1ncho seco'U Q ) a .i Scheduleddatefor'next.refuej'ingshutdown:- Septe'aber 15. l.R8L

2.>

'd Kt 3. Scheduled'date for restart?foi-lowing refueling: _Ji,n6d.. <r ry 15. 1989 s, 4 j ty 0 1

4. -

Technical. Specification change or other license amendment required: i- { (6 + ik)._. Change to' Rod 4Index vs Power Level Curn f7S 3.5.2) .'v'*

j' 1

. b)~ Change.to Core Imbalance vs Power Level.C.i,rve (TSf3,.5.2) h[ A c) TiltLimiN.(TS3.5.2) i e A a. - 1 y Ei Scheduled date(s) for subm,1 ting proposed licenping act; ion: March 15. 1988 'lG.,.ffmportantlicensingcorsf6drationsfassociatedwithrefueling: N/A ,, a t s; 97. Number of. fuel asse,mbl'ies: ,d \\ a) In the core: 1T22 l <z. b) In the Spun $ Fu l Pool: i 316 1 \\. Present licensed's) pent fuel tapacity: 1080' 8. N0 .i.. s Projecteddateohthelastrefuelingthatcanbedischargedto 9. 3 the Spenh rfuel Pool: . December 3. 2001 i 9' / 4 .i ?

f.

~, ,' V i 4., t y; y' c' i 1 .}_y_ u 1 .e [5 '( !'t j 4 g. 5 1 3 4 l .a ,t e f/; 9 a y + I o 'f ^ J { r ) \\ l i /[Q,Iy. l, ]' c i a f 't g ,M[ \\ g !'{/ : 4 ,( ( ., ;j. e ' .>g. .s $l \\ l i. y p 3 y 1 t i 'c y, _6 - '(

>[.i %SMUD "Y SACRAMENTO MUNICIPAL UTILITY DISTRICT O P. O. Box 15830, Sacramento CA B - 830,(916) 452-3211 EgT FqALIFORNIA AN ELECTRIC SYSTEM SE C F NL 87-1322 October 15, 1987 g a R c-E N a U. S. Nuclear Regulatory Commission 7 Attn: J. B. Martin, Regional Administrator y en 'd l Region V Office of Inspection and Enforcement -1450 Maria Lane, Suite 210 Halnut Creek, CA 94596 j i l DOCKET NO. 50-312 ) RANCHO SECO NUCLEAR GENERATING STATION l LICENSE NO. DPR-54 OPERATING PLANT STATUS REPORT

Dear Mr. Martin:

i Enclosed is the September 1987 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, / /' / ,4_ 1 A. Meyer Manager, Nucley Licensing i ? Encl (5) cc: I&E Hash (12) F. J. Miraglia, NRR, Bethesda MIPC (2) INPO G. Kalman R. Twilley, Jr. l 7 jb"*'ol Y k ' RANCHO SECO NUCLEAR GENERATING STATION O 1444o Twin Cities Road, Herald, CA 95638 9799;(209) 333-2935 ,l

}}