NL-89-517, Monthly Operating Rept for May 1989 for Rancho Seco Nuclear Generating Station

Monthly Operating Rept for May 1989 for Rancho Seco Nuclear Generating Station

ML20244E607
Person / Time
Site:	Rancho Seco
Issue date:	05/31/1989
From:	Crunk S, Mueller M SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NL-89-517, NUDOCS 8906210013
Download: ML20244E607 (9)

Text

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SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S $treet, P.o. Box 15830, Sacramento CA 95852 1830,(916) 452-3211 '

AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA l

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l NL 89-517 June 13, 1989 1

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i U. S. Nuclear Reguiatory Commission Attn: Document Control Desk Hashington, DC 20555 Docket No. 50-312 1 Rancho Seco Nuclear Generating Station  !

License No. DPR-54 OPERATING PLANT STATUS REPORT l

Attention: George Knighton

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Enclosed is the May 1989 Monthly Operating Plant Status Report for the Rancho l Seco Nuclear Generating Station. The District submits this report pursuant to l Technical Specification 6.9.3. q Sincerely, f i

L. Mes Steve L. Crunk '

Manager Nuclear Licensing Enci (5) i I

cc: J. B. Martin, NRC, Halnut Creek A. D' Angelo, NRC, Rancho Seco i INP0  ;

R. Twilley, Jr.  !

.I hg N0033 090531 r

' 1 R ADOCK 0500032;;! ~)f }gy. .

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' o i RANCHO SECO NUCLEAR GENERATfNG STATION O '1444o Twin Cities Rote.'lierald. CA 95638-9799;(209) 333-2933' j

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SUMMARY

OF PLANT OPERATIONS Rancho Seco began the month at the 92% power level,~but power was reduced to 85% on May 1 (reached at 0540 hours0.00625 days <br />0.15 hours <br />8.928571e-4 weeks <br />2.0547e-4 months <br />) for predictive maintenance work on a heater drain pump. The plant operated between 85% and 92% power until reduced to 65% on May 13 (reached at 0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br />) to perform work on the "A" main feedwater pump controller. The plant remained at approximately 601 power the rest of the month.

SUMKARY OF CHANGES IN ACCORDANCE WITH 10 CFR 50.59 The plant staff accepted documentation packages in May 1989 for the facility changes, procedure changes and tests 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).

1. DCP R88-0085, Revision 0, installed a manual tilsabling (shut-off) valve with position switch in the selector valve pilot line for twelve C0 systems located in the Auxiliary Building. SurveillarceProcedure$P.8,

" Monthly' Surveillance of Fire Protecti_on System Valve Position,"

< Revision 3. Temporary Change TC-01, and Operations Administrative Procedure OAP-0099, " Operations Locked Valve List," Revision 2. Temporary Change TC-01, were implemented to reflect this modification.

This modification does not change the function of the CO2 system. It provides an alternate method of isolating the CO2 system but does not increase the probability of inadvertent actuation. The isolation of the CO, system is administrative 1y controlled. This modification did'not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in i the USul, nor was the possibility for nn accident or malfunction of a different type than any evaluated previously in the USAR created. This modification did not involve an Unreviewed Safety Question.

2. DCP 88-0088, Revision 0, installed a 900 MHz radio communication system.

Radio frequency interference potential from this change has been i determined to be no greater than that of the replaci ' equipment. No f

change in sequence of Operator actions or events re - !s from this change. Only type and frequency of radio equipment has changed, not its i

method of use. No assumptions made in the USAR are affected:by this .

( change. Thus, this modification did not increase the probability of i occurrence or the consequentos of an accident or malfunction of equipment q important to safety as previously evaluated in the U$AR, nor was the j possibility for an accident or malfunction of a different type than any evaluated previously in the USAR created. This modification did not involve an Unreviewed Safety Question.

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-  ;$UM4ARY OF CHANGE 3 IN ACCORDANCE HITH 10 CFR 50.59 (Continued)

3. Special Test Procedure STP.1199, Revision 1 "UHF Radio' Communications-Test," tested setup and verified operation and coverage for the 900 MHz radio communications system installed under DCP R88-0088. Three test deviations were noted and resolved. All acceptance criteria were met.

All testing was found to be satisfactory and 100% site coverage per 66

. selected locations in the repeater mode was achieved.

i Only the type and frequency of the-radio equipment changed. The field strength was no greater than the equipment previously in use. The only in+eraction with. existing equipment would have been through radio fiequency interference. The energy developed by the new system at 1 meter was.less than 7.7 volts per meter..which is 20% of the energy created within 1 foot of the 450 MHz radios previously used. This test did not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the USAR, nor was the possibility for an. accident or malfunction of a different type than any evaluated previously in the USAR created. This test did not involve an Unreviewed Safety Question.

4. Special Test Procedure STP.1201, "PV-36014A' Performance and Data Acquisition," was performed to monitor performance of PV-36014A during steam demand changes to power ascension and to acquire data to evaluate  ;

system response during power ascension. All acceptance criteria were met within steady-state conditions.

No equipment important to safety was. involved in the performance of STP.1201. Class 2 equipment was tested after making improvements to prevent malfunction. Any accident possibly created by performance of this test (i.e., steam line break) falls within the envelope of. postulated high energy pipe breaks evaluated in the Safety Analysis Report. This test did not increase the probability of occurrence or the consequences of an accident or malfunction of' equipment important to safety as previously evaluated in the USAR, nor was the possibility for an accident or malfunction of a different type than ar.y evaluated previously in the USAR created. This test did not involve an Unreviewed Safety Question.

5. Radwaste Control Manual RP.309.II.02, " Container Selection and Packaging Requirements for Radioactive Material," Revision 1, reflects the use of High Integrity Containers (HICs) in response to IE Notice 89-27 and modifications to the compactor.

Use of radwaste containers does not involve association with safety equipment. Container strength and design of the HICs are superior to previously used container types. Use of these containers will increase the margin of safety due to container specifications to which the.HICs'are designed. Use of the HICs_will not increase the probability of occurrence or the consequences of an accident or malfunction of/cquipment important to safety as previously evaluated in the USAR,'nor will the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR be created. Use of HICs does not involve an Unreviewed Safety Question.

EU N RY 0F CHANGES IN ACCORDANCE WITH 10 CFR 50.59 (Continued) 6'. Special Test Procedure STP.1212r " Auxiliary Feedwater P-318 Performance Test," was' performed to prove Auxiliary Feedwater (AFH). System operability after the overpressurization on January 31,1989. One test deficiency was' identified during the performance of STP.1212; the K-308 turbine failed to reach rated speed while driven by auxiliary steam. PDQ 89-0265 wa's written to document this deficiency and was dispositioned to accept as is. All acceptance criteria were met with the exception that one of the turbine bearing vibration readings measured displacement in excess of 1.5 mils. PDQ 89-0326 was written to document this deficiency and was dispositioned to be not a nonconformance.

The consequences of.an AFH pump failure are already evaluated in the USAR. Thus, this test did not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the USAR, nor was the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR created. This test did not involve an Unreviewed Safety Question.

7. Temporary Modification 89-13 removed PSV-20544 from the main steam header and installed a blind flange in its place for power operation. This temporary modification.is to be in effect until completion of offsite testing and resolution of the effect of location on the valve performance.

The effect of this temporary modification is-to have seventeen of the eighteen main steam safety valves operable. .This mode of operation meets Technical Specification requirements and the USAR addresses overpressure protection with seventeen valves. Capacity is still adequate for 1127.

power. Thus, this temporary modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the USAR, nor is the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR created. This temporary modification does not involve an Unreviewed Safety Question.

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

1. Maintenance, tests and modifications during May 1989 included routine maintenance on the "A" and "B" decay heat pumps, "A" and "B" TDI diesel generators, "B" Bruce-GM diesel generator, "A" Control Room / Technical Support Center HVAC System, Nuclear Services Electrical Building HVAC Systems, large auxiliary boiler, and motor driven AFH pump; maintenance on the "A" heater drain pump, check valve in the "B" high pressure injection system and on the spent fuel cooling system; replaced a module on the "A" main feedwater pump controller.

2. DCP R89-0002 relocated instrumentation, added stiffeners for vibration reduction and deleted HC-36014B which is not required. These changes were made to reduce the consequences of vibration at the auxiliary steam reducing station PV-36014A.

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' MAJOR SAFETY-RELATED MAINTENANCE, TESTS' AND' MODIFICATIONS NOT REQUIRING ~ -l DETAILED SAFETY ANALYSES '-(Contilmed) .!

3. DCP R89-0010 removed' valves MSS 524'and MSS 525'and capped off the pipe l nipples. These valves were bypasses from motor-operated atmospheric dump valve (ADV) isolation valves which were a source of leakage and were normally closed and not used.

4. DCP R88-0042 added'new emergency lights powered from a battery .

backed / diesel backed Class 2125V DC bus to ensure adequate light in the Technical Support Center (TSC) during emergency conditions. New lighting j ballasts were installed in the Auxiliary Building Diesel Generator Rooms  ;

and the "A" and "B" Diesel Rooms to increase the illumination. level.

5. ECN R-0293, Revision 0, made a change to the interlock low level alarm circuit with the fuel handling bridge de-energized signal such that low.

level condition is not alarmed when the fuel handling bridge is y de-energized or not in use. Level switch LSL-27208 will alarm in the Control Room only when the fuel handling bridge is used for. fuel transfer.

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6. ECN R-3187, Revision 0, tied-in backup air bottles for the remaining four ADVs without motor-operated valves to provide Class 1 instrument air supply to open and close the ADVs and provide capability of isolating an ADV for maintenance.  ;

7. Special Test Procedure STP.1171, " Control Room /TSC Essential HVAC Performance Under Failure Conditions," verified design performance of individual essential HVAC units with simulated loss of power failure of-the opposite train.

8. Special Test Procedure STP.1195, "LSL-27208 Alarm Verification," verified proper function of LSL-27208, spent fuel pool level Hi-Lo alarm which annunciated when the spent fuel pool level is below 38'-6" +-3" during l fuel movement in the spent fuel pool.

9. Special Test Procedure STP.1218, " Diesel Generator G-886A 24 Hour and Load Rejection Test," performs the function of a surveillance procedure.

Testing proceeded normally as directed by the procedure with ten exceptions. All acceptance criteria were met and there are no outstanding test deficiencies. j

10. Special Test Procedure STP.1240, " Auxiliary Boiler Chemical Addition j System Functional Test," was conducted to perform initial startup and operational verification of the Auxiliary Boiler Chemical Addition System. Two test deficiencies were declared following the failure of two flow indicators. The flow indicators were deemed unnecessary and were removed. Acceptance criteria were satisfied. All injection pumps worked per design. The system was free of leaks and the indicator lights on the  ;

local control panel functioned properly. -

11. Special Test Procedure STP.1228, " Post Modification Test of Air Dryer Y-910 A/B," verified the correct operation of the Instrument Air Dryer Y-910 A/B following the modification to its control air.

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• RFUILINGINFORMATIONREQUEST 1

.1.. .Name of Facility Rancho Seco '!

2. Scheduled date for next refueling shutdown:

• Scheduled date for restart following refueling:

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4. Technical Specification change or other license amendment required:

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5. Scheduled date(s) for. submitting proposed licensing actioni *

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6. Important licensing considerations associated with refueling: * -

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

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• Plant shut down June 7 following negative outcome of public vote regarding continued operation of Rancho Seco by SMUD.

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OPERATING DATA REPORT DOCKET NO. 50-312' DATE 5/31/89 COMPLETED BY Marla Mueller TELEPHONE (916) 452-3211 OPERATING STATUS

1. Unit Name: Rancho Seco Notes:

2. Reporting Period: May 1989

3. Licensed Thermal Power (MWt): 2.772

4. Nameplate Rating (Gross MWe): 963

5. Design Electrical Rating (Net MWe): 918

6. Maximum Dependable Capacity (Gross MWe): 917

7. Maximum Deptndable 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 MHe): N/A

10. Reasons for Restrictions, If Any: N/A _,

This Month Yr-to-Date Cumulative

11. Hours in Reporting Period 744 3.623 123.Z.91

12. Number of Hours Reactor Has Critical 744 2.200.Q__ 62.061 8

13. Reactor Reserve Shutdown Hours 0 36.0 10.336.2

14. Hours Generator On-Line 744 2.062.9 57.656.5

15. Unit Reserve Shutdown Hours 0.0 36.0 1.246.2

16. Gross Thermal Energy Generated (MWH) 1.496.259 4.324.904 14jji1JJZ___

17. Gross Electrical Energy Generated (MWH) 519.139 1.485.026 46.133,251

18. Net Electrical Energy Generated (MHH) 487.218 1.362.622 42.406.505

19. Unit Service Factor 100.0% 56.9% 46 th%

20. Unit Availability Factor 100.0% 57.9% 47.f%

21. Unit Capacity Factor (Using MDC Net) 75.0% 43.1% 39.2%

22. Unit Capacity Factor (Using DER Net) 71_23% 41.0% _,

37.3%

23. Unit Forced Outage Rate 0,01_ 42.51_ 42.H%___

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

25. If Shut Down At End Of Report Period, Estimated Date of Startup: *

26. Units In Test Status (Prior to Commercial Operation): Forecast Achieved INITIAL CRITICALITY __hl6 N/A INITIAL ELECTRICITY N/A __b'/A COMMERCIAL OPERATION N/A N L&___

• Plant shut down June 7 following negative outcome of public vote regarding continued operation of Rancho Seco by SMUD.

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AVERAGE DAILY UNIT POWER LEVEL j DOCKET NO. 50-312 1 UNIT Rancho Seco 1

DATE 5/31/89 COMPLETED BY- Marla Mueller ,

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TELEPHONE (916) 452-3211 MONTH May 1989 DAY AVERAGE DAILY POWER LEVEL- DAY' AVERAGE DAILY POWER LEVEL (MHe-Net) (MWe-Net) ]'

1 819 17 557 2 778 18 578 3 766 19 580  ;

4 760 20 584-5 757 21 583 6 765 22 583 7 811 23 __

578 8 814 24 577 9 821 25 '530 10 825 26 512 11 825 27 552 12 812 28 580 13 632 29 580 14 553 30 556 15 579 31 546 16 .530 l INSTRUCTIONS On this format, list the average daily unit power level in MWe-Net 'or each day in the reporting month. Compute to-the nearest whole megawatt.

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