ML20238B368

Monthly Operating Rept for Jul 1987

ML20238B368
Person / Time
Site:	Rancho Seco
Issue date:	07/31/1987
From:	Andognini G, Little R SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
GCA-87-472, NUDOCS 8709010189
Download: ML20238B368 (12)

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.. . JULY 1987 o

SUMMARY

OF PLANT OPERATIONS

'The plant was in cold shutdown for the entire month of July 1987. The initial shutdown was due to the December 26, 1985, loss of Integrated Control System power event.

PERSONNEL CHANGES REQUIRING REPORT 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 PART 50.59 l

The plant staff accepted documentation packages in July 1987 for the facility changes described below. In addition, procedure changes from the last month are included. These changes were approved and reviewed by the Plant Review Committee (PRC) and the Management Safety Review Committee (MSRC). There were no documentation packages completed for tests or experiments during July 1987.

1. The battery room supply exhaust fan supports were upgraded to Seismic .

Class 1. In addition, Seismic Class 1 supports were added to the ductwork for the exhaust fans (ECN R-1002). These modifications were made to improve the reliability and availability of the exhaust fans following a l

seismic event.

The implementation of these enhancements does not involve any new failure modes. The probability of occurrence or severity of consequence of an l accident or malfunction is not increased as a result of this modification l since this change upgrades the fan and duct supports from Class 2 to Class 1. (Log No. 869, Revision 1).

2. The design pressure was increased for certain portions of the Decay Heat System (DHS). Affected piping and components were decertified as required. The design pressure of all DHS piping between the DH pumps and injection isolation valves SFV-26005 and SFV-26006, as well as the connecting piping up to the first normally closed isolation valves, was i

upgraded to 495 psig. The design pressure of the DH pumps (P-261A & B)

I and motor operated valves HV-26037, HV-26038, SFV-26039 and SFV-26040 was raised from 450 psig to 495 psig minimum and the nameplates for each were revised accordingly. Revised test calculations were provided for the DH pumps P-261A & B.

The decertification of the above equipment and piping was made in accordance with the rules of ASME Code,Section XI, 1974 edition with l addenda through Summer 1975, and in accordance with the rules of their l original Codes of construction.

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.. These changes have no effect on the normal er emergency function of the 4 DHS but will allow the piping and components to operate at higher local j pressures caused by elevation differences without being overpressurized. 1

- The probability of occurrence or severity of consequences of an accident or malfunction is not increased, because the functionality of the DHS is not affected. (Log No. 895, Revision 2).

3. Pipe support 1U-26528-14 on the common 14" pipe from the Core Flood and Decay Heat Removal systems to the reactor vessel nozzle was found with a wedge in one gap, larger clearances than specified, weld discrepancies, damage to the base plate and surrounding concrete, and different materials of construction than called for by the design drawing (ECN R-1137). In addition, the pipe was found to be cold sprung.

The pipe support is necessary for proper functioning of the Core Flood /

Decay Heat Removal systems. A new analysis was performed to account for the cold spring in the piping and to determine what modifications to the supports were necessary to restore system operability. The reanalysis for the DHS piping was performed in accordance with the rules of ASME Section III, 1980 Edition and the supports were evaluated in accordance l with ASME III,1977 Edition including the Summer 1979 Addenda, j This support is a seismic restraint which has no function during normal ,

operation. The support is necessary to maintain the system operability only during a seismic event. In the event of a seismic disturbance, the l piping support is necessary to maintain reactor coolant pressure boundary j integrity,- DHS function, low pressure safety injection function, core flood function, and long term post-accident core cooling.

The support was modified so as to restore the Core Flood / Decay Heat "

Removal system to its original design basis normal and accident functional requirements. The changes in the nozzle loads to the reactor vessel nozzle and the core flood tank nozzle were evaluated and found to be l acceptable. No new failure modes were introduced and no margins of safety ,

were reduced. (Log No. 839, Revision 1).

4. The District replaced the existing Class 1 power cable for nuclear service raw water (NSRW) pump P-472B with a new Class 1. cable (ECN R-0328). The associated raceway system was modified and properly sized junction boxes, pull box and conduit were installed. This modification was made because the sidewall pressure and bending radius of the previous power cable exceeded the manufacturer's recommended values. The cable was pulled to replace a previously failed cable.

The modification to the existing raceway system (approximately 245 feet) fot the NSRH pump cable allowed the new cable to be routed without being damaged. The new junction boxes were sized to ensure the new cable would ,

not be subjected to less than the minimum allowed bend radius of 12.8 i inches. A cable tray, section H35BA1, was added to maintain expected )

pulling tension below the 2000 lb. limit for the new cable.

<. The change assures the power cable supplying the 'B' NSRH pump will be able to perform its function during shutdown and accident conditions beyond the licensed operating life of the. plant. . Nuclear safety is improved by the change, and no new failure modes are introduced.

(Log No. 771).

5. The District completed the refurbishment of:

SFV-23812 - High Pressure Injection (HPI) Isolation Motor Operated Valve (MOV)

SFV-23810 -

HPI Isolation MOV SFV-30801 - Auxiliary Feedwater Pump Turbine Isolation MOV as part of the MOV program performed in response to IE Bulletin 85-03 (ECN R-0914 HD, HB & AE). Motor operator components and operators were replaced. Environmental qualifications and seismic requirements were maintained. Torque and limit switch settings also were modified under a program based on the original response to IE Bulletin 85-03. Equipment classification and power supplies remain unchanged. The potential for failures on any of the MOVs is reduced as a result of the changes.

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

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

The margin of safety as defined in the basis for any Technical Specification was not reduced because the program ensures that the valves operate within their design basis. (Log No. 831, Revision 1).

6. During MOVATS testing, operators for MOVs SFV-25003 and SFV-25004 were replaced with SB units. A reanalysis for seismic impact indicated that the yoke would need to be stiffened to meet generic 'g' levels and resonant frequency requirements. Stiffener plates were added to the yokes of the valves to distribute the added two hundred pounds due to the new SB type operators.

In addition, the District completed the refurbishment of the following MOVs as part of the M0V program performed in response to IE Bulletin 85-03.

Valve Number Description SFV-26006 Low Pressure Injection Loop "B"  !

HV-53622 Hydrogen Recombiner Outside the Reactor. Building Isolation HV-53620 Hydrogen Recombiner Inside the Reactor Building Isolation SFV-26019 Cooling Hater from the Decay Heat Coolers HV-21514 Pressurizer Cross-Tie Sample l t

. 7 The program was outlined to the NRC in letters dated June 16 01986 and November 5, 1986. The overall engineering package was implemented to prevent the valves from failing on demand in a common mode due to n improper switch settings as described in IE Bulletin 85-03. (Log No.

849, Revision 2).

7. ECN A-5034 Revision 0, permitted the rework of the Air Handling Unit (AHU) A-546 controls. The existing controls were not adequate to provide steady state air flow. A pneumatic transmitter and reverse acting controller were installed in place of the original controls.

These new fan controls do not pose any safety-related problems. Failure of the controls will result in shutdown of the fan. This modification does not involve an unreviewed safety question. (Log No. 488).

8. Isolation valves NSH-015 and NSH-017 were deteriorated and no longer provided proper isolation for maintenance purposes. The pipe was cut to remove the valves from the existing pipe configuration. To facilitate future maintenance, flanges were added in the piping adjacent to the valves. The flanged joints conform to Pipe Spec HE2 requirements for 150 psig, 200'F service.

The change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, nor does it introduce any new failure mode. (Log No. 602).

! 9. The District issued Revision 16 to procedure AP.305.14, which allows for I

the venting of pressurized atmosphere through an unmonitored release point during the containment integrated leak rate test (ILRT). The release will be preceded by the independent analysis of two samples and will be followed by the analysis of a post-release sample. Periodic sampling will take place during the venting of the containment atmosphere, i The sampling and analysis that take place before, during and after the release ensure compliance with General Design Criteria, Technical Specification limits and regulatory requirements; therefore, the possibility of an accident or malfunction is not increased and the margin '

of safety as defined in the basis for any Technical Specification is not reduced. (Log No. 1026).

MAJ6.? ITEMS OF SAFETY-RELATED MAINTENANCE

1. "B" Decay Heat Train was tested and declared operable during the month of July 1987.

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2. Hydrogen recombiners have been installed.

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3. Vendor Acceptance Tests of the S1A2-SID2 inverters on the 120 volt AC system were begun as were Transformer and Channel Functional Tests.

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. 4. The '35 Start' Test of the 'A' and 'B' Emergency Diesel Generators is on hold due to problems with lube oil piping and a . relay in field flash circuits. Corrective maintenance of the lube oil system has been completed. Recabling work to resolve separation problems is in progress.

5. ECN R-1887, Revision 0, provided high point vents for the supply and 4 return lines in the Nuclear Servica liter (NSW) System, thus eliminating I the air trapped in the high pehm of the NSW System.

6. ECN R-1759, Revision 0, incorporated recommended changes to a new thrust i bearing nut tab design which will preclude a new self locking nut from 1 preventing the set screw design from loosening up the thrust bearing nut i of main feedwater pumps 'A' and 'B'.

7. ECN R-1751, Revision 0, provided channel isolation for Reactor Protection System Total Reactor Power signals for channels 'A' and 'B' and channels

'C' and 'D' as required by IEEE 279-1971.

8. ECN R-1607, Revision 0, changed annunciator windows H2YSB-51 and 62 from '

N.0. (normally open) to N.C. (normally closed). This modification will reduce the potential effect of induced voltages in the interconnecting cables to annunciator.

9. ECN R-1508, Revision 0, reversed the wiring on the front face incandescent-lamp sockets on certain Bailey Auxiliary Relay 885 Modules in the Reactor Protection System. The possibility of shorting out the ~15V DC power supply when the lamps are changed out as directe, in ODR 87-05 is thus eliminated.

10. ECN R-1326, Revision 0, added a .05 uf filter capacitor across Hi-Hi alarm )

output to filter noise, thereby preventing noise spikes, which caused alarms in the High alarm circuit for R15020 and diverted liquids from the Retention Basin.

11. ECN R-1300C, Revision 1, upgraded the Plant Communication System by installing sound powered telephones in the new Diesel Generator building.

12. ECN R-1262, Revision 1, modified the existing platform grating by adding a hinged grating opening. This modification provides access to the Gantry Crane hold-down hooks.

13. ECN R-1230, Revision 0, added pipe unions to emergency pump room air cooler A-529E, thus facilitating accessibility for inspection and future 4 servicing. A similar change was performed on pump room air cooler A-529A l by ECN A-3466.

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14. ECN R-1113, Revision 0, allowed all December 26, 1985, transient related outage electrical cable to be pulled prior to the release of the individual system ECNs so as to expedite plant maintenance.

15. ECN R-0891, Revision 0, relocated regenerate hold-up tanks discharga monitor's (R15020) sample point. The relocation to a point downstream of the newly installed effluent filtering system (ECN A-4954) will provide l R15020 with a resin free representative sample, j l

.. 16.' ECN R-0741, Revision 0, provided for wiring changes in the limit switch compartment of valve SFV-46907.

17. ECN R-0614., Revision 1, eliminated excessive terminations, clarified device identification, and provided identification for unidentified resistors on panel H2SFB.

18. ECN R-0675, Revision 0, modified the Heating, Ventilation and Air Conditioning (HVAC) system in the third and fourth floor computer rooms of the Training and Records (T&R) Building. The previous HVAC system was inadequate for removing the high heat loads generated in these rooms.

19. ECN R-0586, Revision 1, removed the lighting and switch, and convenience outlet from Safety Features Actuation System relay cabinet H4SDB-00. The lighting and switch, and convenience outlet were unused and non-operating.

20. ECN R-0295, Revision 1, provided resplicing of BCN connectors at penetrations H7RP21 and H74P63 using qualified nuclear grade splices.

This resplicing will prevent moisture-induced degradation and ensure the integrity of specified radiation monitor circuits and radiation detectors.

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, . . REFUELING INFORMATION REQUEST

l. Name of Facility Rancho Seco Unit 1

2. Scheduled date for next refueling shutdown: September 15. 1988 1

3. Scheduled date for restart following refueling: January 15. 1989

4. Technical Specification change or other license amendment required:

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

5. Scheduled date(s) for submitting proposed licensing action: March 15. 1988

6. Important licensirig 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 l

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-312

{ UNIT Rancho Seco Unit 1 DATE 7/31/87 COMPLETED BY R. Little

. TELEPHONE (916) 452-3211 MONTH July 1987 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL-(MWe-Net) (MWe-Net)

! 1 0 17 'O 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 0 15 0 31 0 16 0 INSTRUCTIONS i 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 l l

DOCKET NO. 50-312 DATE 7/31/87 COMPLETED BY R. Little TELEPHONE (916) 452-3211 QPERATING STATUS

1. Unit Name: Rancho Seco Unit 1 Notes

2. Reporting Period: July 1987

3. Licensed Thermal Power (MHt): 2.772

4. Nameplate Rating (Gross MHe): 963

5. Design Electrical Rating (Net MHe): 918

6. Maximum Dependable Capacity (Gross MWe): 917

7. Maximum Dependable Capacity (Net MWe): 873

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

9. Power Level to Which Restricted, If Any (Net MWe): 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 5.087 107.712

12. Number of Hours Reactor Was 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 Re:erve Shutdown Hours 0 0 1.210.2

16. Gross Thermal Energy Generated (MWH) 0 0 127.861.688

17. Gross Electrical Energy Generated (MHH) 0 0 41.523.187

18. Net Electrical Energy Generated (MWH) -5.313 -30.686 38.258.099

19. Unit Service Factor 0.0% 0.0% 46.8%

20. Unit Availability Factor 0.0% 0.0% 47.9%

21. Unit Capacity Factor (Using MDC Net) 0.0% 0.0% 40.7%

22. Unit Capacity Factor (Using DER Net) 0.0% 0.0% 38.7% .___

23. Unit Forced Outage Rate 100.0% 100.0% 41.1%

24. Shutdorns 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

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 i

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,, RECEIVED SACRAMENTO MUNICIPAL UTILITY DISTRICT O P. O. Box 15830, Sacramento CA 9585hRg30, (916) 452-3211 i l

AN ELECTRIC SYSTEM SERVING TfREggp OF CALIFORNIA AUG 181987 1931 AUG 20 P l2: 4l l GCA 87-472 l

U. S. Nuclear Regulatory Commission Attn: J. B. Hartin, Regional Administrator Region V ,

Office of Inspection and Enforcement '

l 1450 Haria Lane, Suite 210 I Halnut Creek, CA 94596 l DOCKET NO. 50-312 l RANCHO SECO NUCLEAR GENERATING STATION l LICENSE NO. DPR-54 l OPERATING PLANT STATUS REPORT

Dear Mr. Martin:

Enclosed is the July 1987 Monthly Operating Plant Status Report for the Rancho  :

Seco Nuclear Generating Station. The District submits this report pursuant to  !

Technical Specification 6.9.3.

Sincerely, Chief Executive Officer, i Nuclear Encl (5) cc: I&E Hash (12)

MIPC (2)

INP0 g g -; vf e 110

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_D~~k RANCHO SECO NUCLEAR GENERATING STATION O 1444o Twin Cities Road, Herald, CA 95638 9799;(2o9) 333 2935 r J