ML20148G779

Monthly Operating Rept for Dec 1987

ML20148G779
Person / Time
Site:	Rancho Seco
Issue date:	12/31/1987
From:	Bosakowski P, Croley B SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
DTS-88-016, DTS-88-16, NUDOCS 8801270031
Download: ML20148G779 (20)

Text

I DECEMBER 1987 i

SIM4ARY OF PLANT OPERATIONS The plant was in cold shutdown for the entire month of December. The initial shutdown was due to the December 26, 1985, loss of Integrated Control System power event.

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

StM4ARY OF CHANGES IN ACCORDANCE HITH 10 CFR 50.59 The plant staff accepted documentation packages in December 1987 for the facility changes 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 procedure changes, tests or experiments during December 1987.

1. The modifications in ECN R-0732, Revision 2, implemented the changes committed to the NRC to ensure the District's compliance with the NUREG 0737, Item II.B.3 requirement. These modifications provided additional shielding, instrumentation, and air particulate sampling of the Environmental Lab fume hood exhaust. The District committed to perform on-site diluted Reactor Building (RB) atmosphere isotopics analy-is and Reactor Coolant (RC) boron analysis. The T&R Environmental Lab was designated as the location for the performance of PASS RC and/or RB sample analysis.

These facility changes represent an enhancement by using the Environmental Laboratory on the 4th floor of the T&R Building for on-site diluted PASS grab sample analysis. This will ensure analysis of RC boron and chloride levels in a timely manner. The modifications are bounded by the licensing design basis. Three potential failure modes were identified. It was determined that there is no adverse impact on nuclear safety because the sample dilution and procedural controls will limit radiological contamination and personnel exposure.

l The possibility for an accident or malfunction of a different type, namely i radiological exposure, was not created. The margin of safety was not reduced because PASS operability is maintained per the program plan required in Technical Specification 6.18. The changes made under this ECN did not involve an unreviewed safety question. (Log No. 793, Revision 1) l l 2. New analyses were performed to account for the redistribution of loads in the piping, inconsistencies in the original calculations, as-built i configuration differences, etc., that were found as a result of the IE l Bulletin 79-14 close-out effort. The analyses determined that modifications to some supports were necessary to restore the stress

( design limits of the piping systems.

l hha -I-p

ECN R-0780, Revision 2, covered Seismic Class 1 piping systems that were reviewed under the IEB 79-14 program and involved approximately 100. pipe supports. Some of the work involved modifications to existing pipe.

supports on safety-related systems.

No new failure modes were introduced and no margins of safety were reduced; therefore, this change did not involve an unreviewed safety question. Technical Specifications were not affected by this change.

-(Log No. 805)

3. ECN R-1551, Revision 0, installed circuitry in respective multiplexers to generate analog signals to IDADS which are proportional to'the entire range of 125V DC bus voltage. The analog signals are also used for bus voltage alarms on IDADS for NSEB 125V DC panels SOA2, SOB 2, SOC 2,.and S002.

In the previous configuration, 125V DC voltage on the affected buses could not be monitored on IDADS; only a low voltage alarm for each panel was provided on IDADS at approximately 55-65V DC. An alarm at that level was meaningless since the voltage level of the alarm is below useful bus voltage. The modifications performed as part of this ECN served as operational enhancements to the NSEB 125V DC System.

The possibility for an accident or malfunction of a different type than any evaluated previously in the USAR was not created. The same input from the 125V DC system to the IDADS multiplexers as was previously used ,

is used by this modification.

The margin of safety as defined in the basis for any Technical Specification was not reduced since the functionality of both systems remained unchanged. The implementation of ECN R-1551.did not involve an unreviewed safety question. (Log No. 1038, Revision 1)

4. Previously, the CR/TSC Essential HVAC atmospheric chlorine detector could be actuated by rainfall, causing Essential HVAC System actuation, including isolation of the normal HVAC. A permanent splash shield was installed on each CR/TSC Essential HVAC chlorine detection probe, AE-54701 and AE-54702. The splash shields were designed to prevent water from striking the chlorine probe. This change (ECN R-0341, Revision 0) is an operational enhancement which eliminates unnecessary actuations of a safety system.

This change did not involve an increase in the probability or consequences of an accident or malfunction of equipment previously evaluated because the ability of the gaseous chlorine sensor to detect chlorine was not diminished.

No possibility of a new or different kind Of accident or malfunction of equipment from those previously analyzed was created by this change.

Since the splash shields do not impair the CR/TSC Essential HVAC from performing its safety function, no unreviewed safety question was involved. (Log No. 729, Revision 0)

5. The normal and standby power supplies of the ICS cabinets were changed from the 120V AC SIC (vital) and S1J (non-vital) inverter buses, respectively, to 120V AC SIGB-1 non-vital inverter bus and SIGA non-vital inverter (through SlJ panel), respectively.

The normal and standby power supplies of the NNI cabinets were changed from the 120V AC SID (vital) and S1J (non-vital) inverter buses, respectively, to 120V AC SIGB-1 non-vital inverter bus and SIGA non-vital inverter (through SlJ panel), respectively.

The power source of bus S1J was changed from inverter IJ to inverter SIGA.

Inverters S1GA and S1GB provide normal power to buses SlJ SIGA-1, and

$1GB-1. These buses are supplied backup power by diesel generator backed 480V non-vital AC HCC buses through their respective regulating transformers and static switches.

The changes covered by ECN R-0927, Revision 0, represent improvement over the previous arrangement regarding the reliability and availability of power supplies to ICS and NNI. This facility change had no impact on Technical Specifications. Its implementation is within the plant's licensing design basis and did not create an unreviewed safety question.

(Log No. 836, Revision 1)

6. A now diesel driven compressor system was added to the plant air system.

The system consists of the compressor and its diesel engine, an aftercooler. air receiver, and a filter / dryer package. The engine /

compressor unit and aftercooler were skid mounted in the turbine laydown area at the NE corner of the Turbine Building. The air receiver and filter / dryer were mounted on their own skid just inside the building wall from the compressor unit.

The system is tied into both the IAS and SAS. The IAS header was altered to provide a tie-in point for the new system at line 91520-2"-HE2. The SAS connection was made at existing valve SAS-010 in line 90520-3"-HE.

The IAS isolation valve (IAS-971) is normally open while SAS 010 is locked closed under normal conditions to prevent inadvertent supply of non-breathing-quality air to the SAS header.

The diesel driven air compressor automatically starts on drop of IAS pressure to 85 psig and/or loss of 120V AC to compressor skid. Engine starting power is provided by a 12V battery. 120V AC power from panel SIC-3 is provided to a battery charger and engine block heaters.

Alarms on IDADS indicate compressor running and system malfunction. The scope of this ECN included interconnecting piping between the compressor, aftercooler, dryer and tie-in, interconnecting wiring and upgrade of some compressor components.

Nuclear safety is enhanced by ECN A-5233, Revision P.. No change to the Technical Specifications was needed, no new failure modes were introduced, and no unreviewed safety question was involved. (Log No. 527)

7. ECN R-1036, Revision 1, addressed a noted fire hazard by replacing existing pressure relief valves in the lube oil systems for HPI Pumps P-238A and P-238B, and for Hakeup Pump P-236. The relief valves installed have a larger flow capacity (22 gpm) than the previous valves.

This change was also an operational enhancement that revised the control logic for the HPI pumps and makeup pump controls'to allow each pump's auxiliary lube oil pump to automatically stop two minutes after the main shaft-driven lube oil pump has started and built up pressure. Revising the control logic involved rewiring and adding or replacing relays in MCC units S1Al-101, S2Al-102, S2B1-101, and S2B1-102. A cable was run between each lube oil pump's HCC and the switchgear for the respective main pump.

Additional alarm contacts were provided from each pump's lube oil pump control circuit to indicate that the auxiliary lube oil pump has automatically started on a low pressure signal or has not stopped after the two-minute time period. Installation of the alarm contacts involved running new Class I signals from auxiliary lube oil pump starters S2Al-102, S2B1-101, and S2B1-102 to Anatec multiplexers for annunciation on the IDADS computer.

This modification is an operational enhancement to the HPI and makeup pumps, and rectifies a known fire hazard. This modification does not change the Fire Hazards Analysis Report. There is no effect on or changes to the USAR or Technical Specifications. The probability of occurrence of an accident previously evaluated will not be increased. An unreviewed safety question is not involved. (Log No. 876)

8. ECN R-1993, Revision 0, and sub-ECN R-1993B, Revision 0, provided common alarm circuit annunciation for local fire alarm panels in the power block structures (panels H4FCP5, H4FCP6, H4FCP8, and H4FCP10). Jumpers or new circuitry, as required, were added to the listed local fire alarm panels.

These alarms annunciate on control room panel H3FPB.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR was not increased because the modifications improve the reliability of the equipment involved.

The possibility for an accident or malfunction of a different type than any evaluated previously in the USAR was not created since the functionality of the system and components does not change as a result of this modification.

The margin of safety as defined in the basis for any Technical Specification is not reduced. The implementation of the modifications did not involve an unreviewed safety question. (Log No. 1034)

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

_ ECN Number Valve Number /Acolicable System R-1511F, Revision 0 SFV-22005/ Purification & Letdown R-1511I, Revision 0 HV-22008/ Purification & Letdown (Log No. 1015)

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

ECN Number Valve Number /Acolicable System R-0914AG, Revision 0 IN-20596/ Auxiliary Feedwater R-0914HE, Revision 0 SFV-22005/High Pressure Injection R-0914HF, Revision 0 SFV-22006/High Pressure Injection R-0914HG, Revision 1 SFV-22023/High Pressure Injection R-0914HI, Revision 0 HV-22007/High Pressure Injection R-0914HJ, Revision 0 HV-22008/High Pressure Injection R-0914HQ, Revision 0 SFV-25004/High Pressure Injection (Log No. 831, Revision 1)

R-0968H Revision 0 HV-26011/ Decay Heat R-0968AH, Revision 0 HV-21515/ Reactor Coolant R-09688K, Revision 0 SFV-53504/HVAC R-0968BP, Revision 0 HV-53621/HVAC R-09688Y, Revision 0 HV-26514/ Core Flood R-0968CG, Revision 1 SFV-26016/ Nuclear Service Water SFV-26017/ Nuclear Service Hater SFV-26018/ Nuclear Service Hater SFV-26019/ Nuclear Service Water (Log No. 849, Revision 2)

11. The Main Distribution Frame (HDF) located in the Communications Room was expanded in order to provide service to the T&R Building. Two 100 pair cables were installed from the new section of the MDF to a Branch Distribution Frame located on the third floor of the T&R Building.

ECN A-4882, Revision 0, did not increase the probability of occurrence or

! the consequences of an accident or malfunction of equipment important to l safety previously evaluated in the USAR or create the possibility for an accident or malfunction of a different type than any evaluated previously.

The margin of safety as defined in the basis for any Technical Specifications was not reduced.

! The modification did not involve a change in the Technical Specifications or an unreviewed safety question. (Log No. 473)

12. The upper and middle pressurizer heater bundles were replaced with identical replacement bundles. (ECN R-1372, Revision 1) l 1

During a recent plant evaluation, the pressurizer heaters were operated with the pressurizer liquid level below the top of the immersion heater elements. As a result, many of the heater elements failed. Subsequent visual inspection revealed that elements in the upper bundle had experienced severe bowing, which could possibly lead to a breach of the primary system boundary (NCR S-6097). One element in the middle bundle was bowed, with other elements discolored though not distorted (NCR S-6164). The upper and middle bundles were replaced to restore the plant to its design condition.

This change restored the pressurizer to its intended design condition following recent heater damage. This change did not impact the USAR or Technical Specifications. No unaddressed safety issues or new failure modes have been identified. This change is bounded by the licensing design basis. An unreviewed safety question is not involved.

(Log No. 900)

13. The internal wires of Limitorque motor valve operators were replaced with qualified SIS nuclear grade wire. Space heaters and low point drain plugs were removed, and motor splices outside the limit switch compartments were replaced with qualified splices. Existing terminal blocks were removed an.d control wires laid down directly onto the limit switches.

The design changes specified in ECNs R-0699A, B, C, D, and E were made in response to IE Notice 86-03. The changes upgraded the internal wiring and established the qualification per 10 CFR 50.49 of Limitorque motor operator wiring and components. (Log No. 784)

14. ECN R-0953, Revision 0, extended various measured parameters to multiplexers to be used as SPDS inputs. Six 253.2 ohm resistors were added to cabinet H4SIA and multiplexer H4CDAR1 (3 each) to convert the existing 4-20 ma DC signals into 1-5V DC signals for use in the multiplexers' 0-5V DC cards. Additionally, an SPDS software change was made to enable the SPDS to recognize the new multiplexer points.

The purpose of this modification was to make variables required for hot shutdown available to the SPDS in the event of loss of power to NNI.

This modification enhances plant operation because it makes hot shutdown variables, which are available to SPDS, independent of NNI power.

This modification is an operational enhancement that increases the dependability of displayed variables on the SPDS. The six parameters addressed by ECN R-0953 are required for hot shutdown, and are made available to SPDS even in the event of loss of power (AC or DC) to NNI.

Additionally, four of the six signals provide redundancy of information due to the duplication of existing parameters. (Log No. 867)

15. ECN R-0824, Revision 0, reduced the following ICS transient response runback rates for:

a. an MFH pump trip; from 50%/ min. to 25%/ min.

b. an RC pump trip; from 50%/ min, to 25%/ min,

c. an asymmetric rod condition; from 301/ min, to 31/ min.

To accomplish this change, the existing contact logic was rewired and three circuits on signal generator module 3-10-14 were adjusted. No modules were added or removed. Blocking contacts were added to allow the existing tracking runback (20%/ min.) to bypass the asymmetric rod runback when both runbacks are simultaneously required.

By decreasing the ICS transient response runback rates for an MFH pump trip, an RC pump trip, and an asymetric rod condition, a smoother plant transient will result if one of the three conditions described snould occur. This change is an operational enhancement and will improve plant safety following a transient.

No new failure modes are introduced by the change. Reducing ICS runback rates allows operators time to react to various transient situations and gain better control of the plant. Since the Rancho Seco accident analyses do not take credit for ICS plant runback following a trip of HFH pumps, a trip of one or more RC pumps, or an asymmetric rod condition, there is no significant adverse impact on plant safety by reducing the runback rates.

The ICS runback rate reductions implemented in ECN R-0824 are recommended by the B&H Owners Group Transient Assessment Committee. Toledo Edison (Davis-Besse) and FPC (Crystal River) have modified the ICS HFH pump trip runback rate, are in the process of modifying the ICS runback rate for an asymmetric rod condition, and are evaluating the RC pump trip runback rate recommended by B&H. The other B&H plants have not modified their runback rates, but are evaluating these B&H recommendations. (Log No.

817)

16. The existing Class 2 Pressurizer Hater Temperature signals were upgraded to Class 1 to be used in the SPDS to provide Class 1 Temperature Compensated Pressurizer Level indications per Regulatory Guide 1.97.

(ECN A-5429, Revision 1)

Failure of one isolation cabinet (i.e., loss of power, fire, etc.) would cause the loss of one of the Pressurizer Hater Temperature signals; however, the redundant cabinet will provide the proper signal to IDADS and NNI. Operator action is required to manually select the remaining operating signal for use in the NNI.

In addition to the above, the failure of the new Class 1 dual element would cause the loss of the Class 1 Temperature Compensated Pressurizer Level indication from the SPDS. This Class I signal provides indication only. (Log No. 525)

17. ECN R-0825, Revision 0, altered feedwater flow inputs to the ICS. One signal generator module in conjunction with the existing RC pump status contacts in the ICS was added to develop equivalent flow signals corresponding to the number of pumps running.

The present total RC flow signal which limits the Unit Load Demand was deleted. This allows the removal of the runback on loss of RC flow.

The startup feedwater flow and temperature signal corrections to the main feedwater flow signal were deleted from the ICS.

l l

The changes improve ICS reliability and reduce or elicinate plant transients and reactor trips due to RC flow signal failures, startup feedwater flow signal failures, and feedwater temperature signal failure.

Since the changes reduce transient and trip frequency, plant safety will be improved.

Nuclear safety is not adversely affected by the change. The change acts to increase the reliability of the ICS and decrease the probability of a failed input signal to the ICS resulting in a plant transient and reactor trip. (Log No. 816)

18. NCR S-5577 identified an inconsistency between plant Operating Procedure A.6 and USAR commitments regarding the positioning of steam supply valves to the main feedwater pump turbines. While the USAR states that one of the motor operated valves on the supply side of the main feedwater pump turbines "is maintained in a normally closed position by administrative procedures" (USAR 10.2.1), in fact, plant operating procedure A.6 calls for both valves to be open. P&ID H-530 also reflects this open valve positioning.

USAR Section 10.2.1, Page 10.2.3, Paragraph 4, 3rd line will be revised to read: "One of these valves is closed after 15 percent reactor power is attained. Thus, after 15 percent reactor power has been attained, the failure of a single valve in this system will not cause a blowdown of both steam generators."

P&ID H-530, Sheet 2 was revised to show valve HV-20565 as normally closed. Operating Procedure B.2 was revised to state that HV-20565 be closed at 157. power following heatup and startup. Emergency Operating Procedures E.05 and E.06 were revised to provide instructions to the operators that both isolation valves (HV-20560 and HV-20565) should not be closed at the same time. Procedure A.6 was not revised.

The change did not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR.

The margin of safety as defined in the bases for the Technical Specifications was not reduced; precautions to prevent or mitigate a double steam generator blowdown are not addressed in Technical Specification bases.

The change does not involve an unreviewed safety question.

l l

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

1. The District completed functional testing of the pressurizer heaters.

2. The District modified the TDI diesel generator fuel oil, lube oil and air systems to correct vibration problems.

3. The District completed functional testing of the CR/TSC HVAC System.

4. ECN A-5541, Revision 1, pulled a new cable from MCC S2C6 to the MET tower to replace one which failed to ground. The new cable runs entirely in existing conduit. The old cable was abandoned in place.

5. ECN A-5735, Revision 1, replaced the existing Dionex Program Controller located in the PASS control panel H4ESA with a new manually operated control panel. The new controller, designed specifically to meet the needs of PASS, facilitates the operation, reduces the analysis time, and simplifies the operating procedure.

6. ECN R-2185, Revision 0, rewired switch 131 (Core Flood Tank A to reactor valve HV-26513) to existing terminal block. This change results from the disposition of NCR S-5430.

7. ECN R-0828, Revision 1, added valve position indication lights on the H1RI console for ADVs, TBVs, MFH and Startup Feedwater valves. The installation required new cables, conduits, and other associated work and materials to make the system functional.

8. ECN R-1300, Revision 0, improved and expanded the existing Communication System under the scope of the following sub-ECNs:

l e R-1300A - Plant Telephone System l

• R-13008 - Plant UHF Radio System l

• R-1300C - Miscellaneous Communication Modifications

9. ECN A-5710, Revision 1, replaced the existing alarm relays being used for annunciation of under voltage on 120V AC vital buses SIA, SIB, SIC and S10. The existing relays were not functioning properly.

10. ECN R-2275, Revision 0, replaced existing battery ground fault detectors in panels SOA2, SOB 2, SOC 2, S002, and SONI. The existing Brown Boveri (ITE-278) relays had a generic design problem,

11. ECN R-0717 Revision 0, removed the Dead Band Monitor module for and associated power supply from MFH Pump Turbine Controls. The high rate of failure caused several reactor trips in the past, thus necessitating removal of the module.

12. ECN R-1322, Revision 0, eliminated leakage in unions and provided connections for leak testing of 0TSG. Threaded unions from both sides of 32 steam traps were replaced by welded connections. Connections consisting of a tee, valve, and flange for filling and leak testing were added to lines 35053-2"-EAl and 35055-2"-EA1.

_9-

13. ECN R-1119, Revision 1, removed the four sheet metal blank-offs in i essential air units AH-N-2A and AH-N-28. The blank-offs were installed per disposition of NCR S-4790 pending further evaluation. It has been  :

determined that air flow to meet the Technical Specification 4.31 limit  !

can be achieved by adjusting the existing volume dampers in the ductwork.

14. ECN R-2025, Revision 0, disconnected 120V AC power cables 1R2I150A and H2X, and H7RP41 and 1R2Il50AA at H7RP41 and FY15050, respectively. The cables were providing 120V AC power to FY-15050; however, associated flow transmitter FT-15050 has not been functional for at least three years.

FY-15050 was therefore disconnected without affecting plant operations.

15. ECN A-4932, Revision 1, removed existing loose parts monitor TEAC recorder. Mechanical problems due to age necessitated removal because of low input impedance attenuating the signal to an unreadable level.

16. ECN R-2137, Revision 0, reversed the configuration of AFH Pump P-318 motor so that it rotates in the opposite direction.

17. ECW R-0033, Revision 0, authorized removal of the reactor coolant pump (RCP) seal leakage prefill valve and line, thus allowing better access to the RCP seal cavity area. Integrity of the seal leakage chamber is maintained with a plug in the RCP leakoff cover.

18. ECN R-0878, Revision 1, provided for continuous modulation of auxiliary steam header pressure control valve PV-36014A independent of ICS power.

19. ECN R-1218, Revision 2, replaced the oil atomizer on Auxiliary Boiler E-360. To support the new oil atomizer installation, a force draft fan duct baffle and atomizing steam pressure regulating valve were also installed.

20. ECN R-1615, Revision 0, revised the control scheme of the steam generator and the high point pressurizer venting solenoid valves to prevent the valves being energized due to a fire-induced hot short.

21. ECN A-5667, Revision 0, installed a small junction box and pulled 8 feet of new cable to transmitter LT46501 per disposition of NCR S-4132, Revision 1.

22. ECN R-1496, Revision 0, modified the circuit to accept a new version of the H-12 module in the feedwater turbine signal processor.

23. ECN R-2212, Revision 0, replaced deteriorated taped terminations at the Reactor Building electrical penetration for cables 162A03C and D i (RCP P-210A), 162803C and D (RCP P-2108), 162A02C and D (RCP P-210C) and 162802C and D (RCP P-2100) using Okonite tape.

24. ECN R-1107, Revision 1, installed thermostats in electrical equipment rooms 232 and 233 to prevent exceeding the equipment qualification temperature. The existing room thermostats TSHH-50135 and TSHH-50136 in switchgear rooms were rewired to provide fail safe operation.

l l I

1

25. ECN R-0361. Revisien 1, repaired and relocated valve RSS-706 in a horizontal portion of line 70057-1"-CA and added another check valve to prevent back leakage in Decay Heat sample line 26001-1"-GD from RCS sample line 70059-1"-CA.

26. ECN A-3062, Revision 1, reworked the existing steam trap configuration which had proved inadequate to handle condensate load on steam supply to AFH pump turbine K-308.

27. ECN R-0042, Revision 1, added a drain line and additional globe valve in each drain line from drain valves for 87420-4"-HC and 87421-4"-HC. This change allows condensate to be easily drained from the vapor extraction lines due to improved access.

28. ECN R-1003, Revision 0, revised the piping configuration for the moisture separator reheater relief valves sealing steam by abandoning the existing steam supply (gland steam header 9 4 psig). The steam is obtained from the auxiliary steam header via the service steam layup connections no longer in service.

29. ECN R-2350, Revision 0, rectified the problern of gland connection in panel H2YS. All drain wires of the field side were directly connected to vertical ground bus. A jumper wire was connected on the field side of TB11 and TB12 and also connected to the same ground bus so that panel sides of the shielding are grounded.

30. ECN R-1778, Revision 0, replaced various relief valves such that the seating material (originally brass to brass) is of a softer material (viton), thus relieving seat failure for EDG Start Air Relief Valves.

31. ECN R-2157, Revision 0, deleted the CBS high flow alarms on H2SFA and H2SFB. The condition which the alarms were intended to protect against (CBS Pumps P-291A and P-291B approaching pump runout) was not proven possible by calculation.

32. ECN A-4954, Revision 1, added a self contained, skid-mounted system to separate whole resins and resin particles from the water downstream of the RHUTs (T-950A and B). This change prevents contamination from being carried offsite.

33. ECN R-2240, Revision 2, installed safety drop lugs on the trolleys of the A-1 Monorail to prevent the hoist mechanism from falling off the monorail if the trolley wheels or axles were to fail.

34. ECN R-0838, Revision 1, provided for additional penetration seals and fire rated enclosures to be reworked to meet the fire endurance ratings required by 10 CFR 50 Appendix R.

35. ECN R-0488, Revision 2, installed additional insulation as required to l

minimize heat loss at all pipe supports that act as heat sinks.

l Fiberglass insulation installed was at least 1" thick.

36. ECN R-1925, Revision 0, modified the T&R Building QA Vault Halon System to comply with recommendations made by ANI. Notice of Violation 86-22 documented the deficiency with regard to storage and retention of QA '

records.

37. ECN R-1034, Revision 0, restored nuzerous breached fire barriers identified by surveillance activities.

38. ECN R-0669, Revision 0, properly spared unused cables per disposition of NCR S-5426.

39. ECN A-3795, Revision 4, replaced the existing HPI and MU pump lube oil coolers with stainless steel coolers and installed provisions for connections of differential pressure instrumentation in the coolant inlet / outlet piping.

40. ECN R-1540, Revision 1, modified thirteen pipe supports to: a) permit normal valve maintenance activities, b) account for heavier valve / operator replacements as part of the HOV program, and c) repair a discrepant support per disposition of NCR S-5936 and EAR PL87-068.

41. ECN A-4822C, Revision 0, installed wiring (Class 2) from the exceu flow check valve position switches through a junction box inside containment to Anatec tr.ultiplexer. This change complies with the NUREG 0737 Item II.F.2 requirement to provide indication of Inadequate Core Cooling in the Main Control Room.

42. ECN R-0968CI, Revision 0, changed the spring pack and reset the torque switch on H0V FV-95401.

43. ECH R-1300A, Revision 0, provided an alternate system to the existing telephone system for emergency communications in case of a single fire in the Communications Room. The existing system was modified for expansion to meet additional demand.

44. ECN R-19078, Revision 0, exchanged freeze protection heat tracing power supplies and alarms on AFH suction and discharge pressure transmitters and tubing due to different channel classifications between the power source and respective AFH pumps.

45. ECN R-1673C, Revision 0, replaced the initiating spring of breaker handle mechanisms in Hestinghouse MCC cubicles S2A317, S28317, and S2A206. In their previous condition, the breakers did not give positive trip indication.

46. ECN A-3795G, Revision 0, changed the heat exchanger for E-2398 and added connections for differential pressure instrumentation in the inlet / outlet piping of E-237A and E-239A and B.

47. ECN A-4697B and C, Revision 0, replaced transformers X43C2 and X43DI, respectively, with new dry-type transformers. This change was made in response to tap changer failures, possible PCB hazards, and indications of other internal problems in the original station service transformers.

48. ECN A-5415H, Revision 1, provided digital and analog inputs to IDADS and SPDS computers for monitoring EFIC parameters. This ECN also added an audible alarm to IDADS so it will perform similarly to the Main Control Room annunciator.

49. ECN A-5415Q, Revision 1, aligned the power s8urces for the AFH pump pressure indication with that powering the respective pumps. Class 1 AFH pump discharge pressure and AFH flow indication was added in the Main Control Room on HISS.

50. The District completed the refurbishment of the following H0Vs as part of the MOV program performed in response to IE Bulletin 85-03.

ECN Number Valve Number /Acolicable System R-0976AJ, Revision 0 PV-21520/ Reactor Coolant R-0976AK, Revision 0 HV-22005/ Component Cooling Water R-0976AL, Revision 0 HV-22006/ Component Cooling Hater R-0976AN, Revision 0 HV-21907/ Pressurizer Relief Tank R-0976AR, Revision 0 HV-21922/ Pressurizer Relief Tank R-0976AS, Revision 0 FV-36060/ Auxiliary Steam R-0976BA, Revision 0 HV-43011/ Site Reservoir

51. ECN R-0775A, Revision 0, installed a new sump; ECN R-0775D, Revision 0, installed the new RHUT; ECN R-0775E, Revision 2, buried piping, and ECN R-0775I, Revision 2, buried conduit for secondary waste water segregation.

52. ECN R-0472, Revision 1, relabeled breakers S1 and S2 to enhance readability and clarity.

EDDY CURRENT EXAMINATION RESULTS An eddy current inspection of the Rancho Seco Steam Generators was initiated on September 9, 1987 and completed on September 25, 1987. Due to the identification of tube defects in the initial sample of both the A & B-0TSG, a second random sample was run in each steam generator. A total of 16 tubes were plugged based on the inspection results (9 tubes plugged due to defects greater than or equal to 401. TH and 7 tubes as licensee option).

TUBES INSPECTED A-0TSG B-0TSG l Sleeves - 253 (All) Sleeves - 254 (All)

Full Length - 1414 Full Length - 1434 Partial Length - 1646 Partial Length - 1766 1

TUBES PLUGGED A-0TSG Row / Tube -Elevation 1I8 4-29 7th TSP 65%

8-57 UTS+6.16" 38%

10-1 10th TSP 49%

14-72 14th TSP 74%

72-15 UTS+8.14" 62%

75-96 13th TSP +10.08" 66%

82-13 13th TSP + 6.21" 54%

97-2 15th TSP + 7.34" 38%

97-75 3rd TSP 38%

150-23 10th TSP 38%

In addition, tube 2-3 was plugged at the upper tubesheet. This tube was plugged at the lower tubesheet in a previous outage. Tube 2-3 did not contain any eddy current indications.

B-0TSG Row / Tube Elevation 118 48-9 8th TSP 56%

56-8 8th TSP 58%-

76-3 UTS+ .53" 75%  !

18-2 UTS+19.14" 38%

48-25 10th TSP +30.5" 31% -

84-65 3rd TSP +32.37" 38%

REPORTABLE INDICATIONS A-0TSG Row / Tube Elevation 118 2-18 UTS+9.17" 33%

4-6 UTS+10.41" 22%

11-67 15th TSP 24%

24-95 14th TSP 24%

25-96 14th TSP 32%

30-2 UTS+4.28" 36%

30-3 UTS+5.25" 231 50-123 10th TSP 23% ,

52-3 8th TSP 22%

69-64 3rd TSP +34.35" 35%

121-84 4th TSP 28%

I t

B-0TSG Row / Tube Elevation 118 l 4-7 9th TSP +29.47" 22% i 26-9 14th TSP 261 28-7 14th TSP 351 39-116 9th TSP 30%

44-4 UTS+7.54" 29%

52-2 UTS+14.14" 27%

64-7 .UTS+7.21" 37%

70-10 UTS+7.52" 20%

75-68 3rd TSP +35.93" 33%

82-13 UTS+2.55" 30%

83-64 UTS+0.50" 29%

98-44 6th TSP +16.41" 28%

130-69 15th TSP +11.13" 28%

145-53 8th TSP +21.69" 32%

AFH HEADER GAP MEASUREMENT An AFH header gap measurement was performed on the tubes identified in Rancho Seco Technical Specification Table 4.17-3. All tubes were determined to have at least 1/4" clearance from AFH header.

Detailed inspection results, data analysis, graphics, tubesheet maps, and equipment and personnel certifications are available in the Eddy Current Examination report, dated September 1987, by Conam Inspection. This report is located in the Plant Performance Department at Rancho Seco.

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: 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 i

1 I

I

AVERAGE DAILY UNIT P0HER LEVEL DOCKET NO. 50-312-UNIT Rancho Seco Unit 1 DATE- 12/31/87 COMPLETED BY P. Bosakowski TELEPHONE (916) 452-3211 MONTH December 1987 ,

DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY PONER LEVEL (MHe-Net) (We-Net)'

1 0 17 0 2 0 18 0 3 0 19 0  !

4 0 20 0 l I

5 0 21 0 6 0 22 0

7. 0 23 0 ,.

i. 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 l l

INSTRUCTIONS  ;

l On this format, list the average daily unit power level in MWe-Net for each day '

in the reporting month. Compute to the nearest whole megawatt. '

i

OPERATING DATA REPORT DOCKET NO. 50-312 DATE 12/31/87 COMPLETED BY P. Bosakowski TELEPHONE (916) 452-3211 OPERATING STATUS

1. Unit Name: Rancho Seco Unit 1 Notes:

2. Reporting Period: December 1987

3. Licensed Thermal Power (MHt): 2.772

4. Nameplate Rating (Gross HHe): 963 S. 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 8.760 111.384

12. Number of Hours Reactor Has Critical 0 0 54.322

13. Reactor Reserve Shutdown Hours 0 0 l'J . 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. Net Electrical Energy Generated (HHH) -10.337 -56.759 38.232.026

19. Unit Service Factor 0.0% 0.Q%__ 45.2%

20. Unit Availability Factor 0.0% 0.0% . 46.3%

21. Unit Capacity Factor (Using HDC Net) 0.0% 0.0% 39.3%

22. Unit Capacity Factor (Using DER Net) 0.0% 0.0% 37.4%

23. Unit Forced Outage Rate 100.0% 100.0% 43.5%

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: 2/28/88

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

UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 50-312 UNIT NAME Ranchn Seen

' DATE December 31: 1987 COMPLETED BY P. Bosakowski REPORTMONTH December 1987 TELEPHONE (916) 452-3211 "i -

E .-

O es 5

.3 se gsX Licensee g-to 5*

g. Cause & Correctise

%.. Date g 3g  ;; jg5 Event u7 o.3 Action to

$5 5 j5y Report # <N 0 0 Prevent Recurrence 3

1 85-12-26 F 744 0 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 F Forced Reason: Method: Exhibit G-Instructions 5 Scheduled A Equipment Failure (Explain) 1-Manual for Preparation of Data B Maintenance of Test 2-Manual Scram. Entry Sheets for Licensee C Refueling 3-Automatic Scram. Event Report (LER) File (NUREG-D-Regulatory Restriction 4-Other (Explain) 01611 E-Operator Training & License Exar,ination F-Administrative 5 Goperational Esror (Explain t Exhibit I-Same Source a8/77) llother (Explain)

gSMUDSACRAMENTO MUNICIPAL UTILITY DISTRICT C 6201 S Street, P.O. Box 15830. Sacramento CA 958521830 (916) 452 3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA DTS88-016 January 15, 1988 U. S. Nuclear Regulatory Commission Attn: J. B. Hartin, 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. DPR-54 OPERATING PLANT STATUS REPORT

Dear Mr. Martin:

Enclosed is the December 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. Also included in this report are the Eddy Current Examination results.

Sincerely, f' -

Bob G. Croley L;:

Director, Technica Services 9

. . T' ~QR

.. t Enc 1 (5) s

,Mr.:

cc: I&E Hash (12) .C.

F. J. Miraglia, NRR, Bethesda -

HIPC (2) E>

INP0 G. Kalman R. Twilley, Jr.

t\

S . .! y' RANCHO SECO NUCLEAR GENERATING STATloN O 1444o Twin Cities Road, Herald, CA 95638-9799;(2o9) 333 2935