NL-89-379, Monthly Operating Rept for Mar 1989 for Rancho Seco Nuclear Generating Station

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

ML20244A870
Person / Time
Site:	Rancho Seco
Issue date:	03/31/1989
From:	Crunk S, Mueller M SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NL-89-379, NUDOCS 8904180240
Download: ML20244A870 (9)

Text

. - _ _ - . _ _ _ _ _ _ - _ -

.(

MARCH 1989 1

SUMMARY

OF PLANT OPERATIONS Rancho Seco left cold shutdown at 0324 hours0.00375 days <br />0.09 hours <br />5.357143e-4 weeks <br />1.23282e-4 months <br /> on March 7. Hot shutdown was I reached at 1156 hours0.0134 days <br />0.321 hours <br />0.00191 weeks <br />4.39858e-4 months <br /> on March 9. The reactor went critical at 0939 hours0.0109 days <br />0.261 hours <br />0.00155 weeks <br />3.572895e-4 months <br /> on March 13. The reactor reached 10 percent power but was reduced to a level slightly above criticality in order to replace one of six fans that cools electrical components and cables inside the closure head service structure on top of the reactor head. Power was raised to 3 percent but remained there pending resolution of issues involving position indicators on three control rods. Rancho Seco went back on line at 1903 hours0.022 days <br />0.529 hours <br />0.00315 weeks <br />7.240915e-4 months <br /> on March 17 and reached 92 percent power at 1603 hours0.0186 days <br />0.445 hours <br />0.00265 weeks <br />6.099415e-4 months <br /> on March 19. The reactor operated between 88 and 94 percent power until it automatically tripped on high pressure due to feedwater pump control fluctuations on March 28 at 1515 hours0.0175 days <br />0.421 hours <br />0.0025 weeks <br />5.764575e-4 months <br />. The plant remained in hot shutdown through month's end.

SUMMARY

OF CHANGES IN ACCORDANCE HITH 10 CFR 50.59 The plant staff accepted documentation packages in March 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 (MSRC).

1. Special Test Procedure STP.1214, Generator G-100A Starting Air System Upgrade Test, was run to demonstrate TDI Diesel Generator Starting Air System operability after modifications were made to the system. STP.1214 met all acceptance criteria and all test deficiencies cleared.

Ttis STP tested the operability of upgrades to the TDI Diesel Generator Starting Air System and ensured the upgrades meet Updated Safety Analysis Ra ort (USAR) criterion. Since the test was bounded by tests described in tia USAR, it did not increase the probability of occurrence or the cc nsequences of an accident or malfunction of equipment important to si'ety 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.

2. Special Test Procedure STP.1216, Manual Operation of MSS-032, verified the throttling capabilities of MSS-032 while transferring steam load from the auxiliary boilers to the Main Steam System. This test successfully proved that MSS-32 could be manually thr'ttled, allowing main steam to charge the auxiliary steam header resulting in a decrease in auxiliary boiler output.

i This test does not change the function of the system and does not involve equipment important to safety. Only Class 2 equipment and components were included in the performance of this test. Any conceivable accident that could have been caused by performance of this STP is within the envelope of accidents evaluated in the USAR. 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 Safetr Question. .

8904180240 890331 /

PDR ADOCK 05000312 PDC R

I

SUMMARY

'0F CHANGES IN ACCORDANCE HITH 10 CFR 50.59 (Continued)

- 3. Special Test Procedure STP.1221, Radio Coverage Test In The Aux Feed Pump Missile Shield (P-318), was run to determine the effect of. electrical and acoustic interference on the operation of 900 MHz hand held portable radios at a number of points inside the AFH pump missile shield. The satisfactory results of this test prove that clear and unbroken radio communications are available from within the missile shield during all operational modes of AFH pump P-318.

The USAR does not address voice radio communication capabilities and this test does not affect any assumptions made in the USAR. The conditions of radio usage during the test were not changed from normal operating procedures. This test therefore 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.1222, Pressurizer Spray Back-Up Valve Differential Pressure Stroke Test, was performed to demonstrate operability of pressurizer spray back-up valve PV-21509. The valve was cycled five times, four times in manual control and one time in automatic control. The valve functioned normally in all cases.

This test was performed in hot shutdown under normal operating conditions. Over or under pressure conditions resulting from a failed pressurizer spray valve is bounded by the USAR. There is a block valve to stop flow for a stuck open spray valve and a spray valve in parallel for a stuck closed spray valve. This test did not revise any pressure or i temperature setpoints nor did it call for conditions that could cause temperature or pressure limitations to be exceeded. 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.

5. Rancho Seco Administrative Procedure RSAP-0101, Revision 3, updated the succession of responsibility to incorporate Assistant Nuclear Plant Manager and added responsibilities for the Assistant, changed the '

Director, Nuclear Quality and Industrial Safety to the Assistant General Manager level, expanded responsibilities of the Nuclear Plant Performance Department, and changed the title of the AGM, Nuclear Power Production to AGM, Nuclear Plant Manager.

Changes to RSAP-0101 enhance individuals / groups responsibilities or transfer responsibilities. These administrative changes do 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. These ,

changes do not involve an Unreviewed Safety Question.

l l

I I l 1

• e

SUMMARY

OF CHANGES IN ACCORDANCE WITH 10 CFR.50.59 (Continued)

6. Temporary Change Number 1 (TCN-01) to Plant Operations Procedure B.3, Normal. Operations, changed limits and precautions to allow startup physics testing at reduced RCS Tave. The design basis report,. Cycle 7 Reload Report, allows and requires reduction of Tave by approximately 5 degrees F. No adverse impact by the operation is identified either in the Reload Report or the USAR. The operation at reduced Tave is for a short period. Thus, this temporary change to B.3'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.

7.. Temporary Modification 89-20 removed fuse F159 from CRD Logic Cabinet H-4RS8 thus disabling the In and Out Limit relays and their logic functions for Group 4 Rod 11. Also, fuse F259 was removed from cabinet H-4RS8 thus disabling 0%, 25%, 50%, 75% and 100% reference lights from the reference panel for Group 4 Rod 11.

Temporary Modification 89-21 removed fuse F103 from CRD Logic Cabinet H-4RS7, thus disabling the In and Out Limit relays and their logic functions for Group 4 Rod 4. Also, fuse F203 was removed from cabinet H-4RS7 thus disabling.0%, 25%, 50%, 75% and 100% reference lights from the reference panel for Group 4 Rod.

Temporary Modification 89-22 removed fuse F107 from CRD Logic Cabinet H-4RS7 thus disabling the In and Out Limit relays and their logic functions for Group 6 Rod 3. Also, fuse F207 was removed from cabinet H-4RS7 thus disabling 0%, 25%, 50%, 75% and 100% reference lights from the reference panel for Group 6 Rod 3.

None 'of the accidents evaluated in Chapter 14 of the USAR take credit for the operation of the In and Out Limit lights or their associated interlocks. Inoperability of the lights therefore does not increase the probability of an accident's occurrence, nor does it increase the consequences of an accident. The In and Out Limit switches provide indication, ICS and boron control inputs only. The switches physically do not interact with the control rod drives; therefore, this change does not affect the probability of malfunctions of the CRDs, nor does it affect the operation / malfunction of any other safety-related function.

L These temporary modifications affect control, indication and monitoring functions. The equipment affected is physically isolated from all systems other than the CRD indication; it interacts only with the indication and interlocks. Additionally, the indication and interlocks are not included l

in the accident assumptions or analysis. Also, the changes involve removal of fuses in the circuits. Isolation of the signal, by removal of l the fuses, reduces the interaction between the switches, indications, and l interlocks. These temporary changes do not increase the consequences of a malfunction of equipment important to safety, nor do they create the potential for different accidents. The possibility of a different type of i

malfunction is reduced. The margin of safety presented in USAR Chapter 14 is not reduced.

l 1

l

g

• e

- hAJORSAFETY-RELATEDMAINTENANCE,TESTSANDMODIFICATIONSNOTREQUIRING DETAILED SAFETY ANALYSES

1. Maintenance, tests and modifications during March 1989 included l' surveillance runs on the "A2" TDI and "A" Bruce-GM diesel generators, AFH System testing, local leak rate test on Reactor Cuilding penetrations, repair of pressure indicator on the core flood tank, post maintenance test on steam driven AFH pump P-318, repairs on motor driven AFH pump, block valve work for one of the turbine bypass valves, MOVATS testing of four valves, Furmanite repair of three reactor coolant pumps, repair of faulty fan motor which cools electronic equipment mounted in the support structure over the reactor head, changeout of position indicator tubes and, replacement and testing of a feedwater heater valve.

2. ECN A-4764 Major, Revision 2, changed Auxiliary Building and NSEB doors including vital doors, emergency exits, and fire door instrumentation. A card reader was added to the Training and Records Building door TR604 and to the Technical Support Center door. Also, three CCTV cameras were relocated in order to view areas that had been blocked by new construction.

3. ECN A-5266 major, Revision 1, added a temporary resin disposal facility for once-through resin and added a-resin addition system to the' existing resin separation tank.

4. ECN A-5682, Revision 1, replaced pressure gauges PI-43003A, B and C on canal station pumps with Duplex' pressure gauges because the old gauges were ranged for positive pressure and were being destroyed by occasional vacuum created by pumps.

I

{

l

. REFUELIi4G INFORMATION REQUEST

1. Name of Facility Rancho Seco

2. Scheduled date for next refueling shutdown: September 14. 1989

3. Scheduled date for restart following refueling: Decernber 13. 1989

4. Technical Specification change or other license amendment required:

a) Removal of Cycle Specific Core Limits and Approval of Core Operating Limits Report - submitted as Proposed Amendment 179 on March 31, 1989.

b) Reactor Vessel Level - To be submitted as Proposed Amendment 178 on July 1, 1989.

' 5. Scheduled date(s) for submitting proposed licensing action: Item 4.b - 7/1/89

6. Important licensing considerations associated with refueling: Technical Specification chanae reauired for reactor vessel level

7. Number of fue'l 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

l

)

_ -___m-m.

2 :e AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 312 UNIT Rancho Seco DATE 3/31/89 COMPLETED'BY, Maria Mueller TELEPHONE (916) 452-3211-MONTH March 1989 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MHe-Net) (MWe-Net) 1 0 17 0 2 0 -18 463 3 0 19 741 4' 0 20 780 5 0 21 763 6 0 22 764 7 0 23 820 8 0 24 821 9 0 25 831 10 0 _

26 840 11 0 27 839 12 0 28 517 13 0 29 0 14 0 30 0 15 0 31 0 16- 0 l -- INSTRUCTIONS On this format, list the average daily unit power level in MHe-Net for each day in the reporting month. Compute to the nearest whole megawatt.  ;

OPERATING DATA REPORT DOCKET NO. 50-312 )

i DATE 3/31/89 l l

COMPLETED BY Marla Mueller TELEPHONE (916) 452-3211 OPERATING STATUS

1. Unit Name: Rancho Seco Notes: l

2. Reporting Period: March 1989 l

3. Licensed Thermal Power (MWt): 2.772 4

4. Nameplate Rating (Gross MWe): 963

5. Design Electrical Rating (Net MWe): 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): N/A

10. Reasons For Restrictions, If Any: N/A i

This Month Yr-to-Date Cumulative

11. Hours in Reporting Period 744 2160 122.328

12. Number of Hours Reactor Was Critical 365 tE_ 902.6 60.768.5

13. Reactor Reserve Shutdown Hours 0 0.0 10.300.2

14. Hours Generator On-Line 260.2 772.2 56.365.7

15. Unit Reserve Shutdown Hours 0.0 0.0 1.210.2

16. Gross Thermal Energy Generated (MWH) 640.522 1.743.312 139.072.140

17. Gross Electrical Energy Generated (MWH) 208.554 584.069 45.232.294

18. Het Electrical Energy Generated (MWH) 183.648 523.168 41.567.051

19. Unit Service Factor 35.0% 35.81_ 46.1%-

20. Unit Availability Factor 35.0% 35.8% 47.1%

21. Unit Capacity Factor (Using MDC Net) 28.3% 27.7% 38.9%

22. Unit Capacity Factor (Using DER Net) 26.9% 26.4% 37.0%

23. Unit Forced Outage Rate 65.0% 64.2% 43.3% s

24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each): J Refuelina shutdown scheduled to begin Seotember 14. 1989.

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

Reactor startuo on Aoril 7. j

26. Units In Test Status (Prior to Commercial Operation): Forecast Achieved l INITIAL CRITICALITY N/A N/A INITIAL ELECTRICITY N/A N/A COMMERCIAL OPERATION N/A N/A 1

l l

l l

l L . .

n1 13 b M c 1 - a 2 U e 2 r f N S1 e o e (

3 w5 4 pr oe s .ht s

. ee l e

2 o nt anF si 1 hh e) np sm oa ce) 3 c cl 6 i oao

- nr e1 m l rt iDiR tc fL 0a oe . c u E 5R%W(9 -

e.

irc e

c n

tt s sn oss t e r

ts nf(

orL o e eyo ci Ino is r t

O.EEYE e "rr uSi eil t t o

N MTBNAADO ru'ce r u d ra t ut p d ap

- ar ee Gah p

e T

E ND E H T P C"R

&'t nec wt i eos mu t ie b r pSR yt )

KT E E e^ ne oaf rt ipr n1 h r t e6 CN I

L EL s d wi uur x o nv1 O u v td c sae EfEE0 DU P e MT a C r uee s w -

P h ep ed o O sFS rep 4 C ps .

x yl ud s Rra S ae p ac Cct m lii R a u sr p al n uih h pa nxc gmpW aue i ue F MAT H pS M S .

N m O .

a )n r

C mci

• . d I

T V C aSl a C c* gEg r cp U 9 . P T cix D 8 StaE l

E 9 1

duu oe la a m (r R onnt h haa ut h R c +, ?U BB J eMMAO t

E S - - - -

r E gm SA M1234 W a O M P 3

/

D H .

N T A N S O e # 1 4 M et N s n n t

r 0 0 W T e o - -

R e c v p 9 9 O iE e 8 8 D O L R T P U E H R S

T r. = h5 gS -

I N " )

n Ygil j

~ 2 3 .i U  ;

a

.93j lp x n

• E o A A (t

e es ic t

3l5 i luTr r ig r

ts iao e

_ 8 F e R

7. ).0. )h t c nn yT r

- ggE 0r3 c 0 ea g o

- 7b0 r 8 mn int 53f 6 e4 a pt el el ai at a F M :ninu u o uif g

( (

s q a e e aEMR.R-e - -

,gH F F RABCDEF 2

- 9 9 8 8 te - -

a 1 8 '

D - 2 d 2 - le 3 d u ed rce h

- o c FS

- o 6 7 FS N

1 1

_ I

. ' 1 I l:!

I

>~,sMun j l

t '"J SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S Street, P.O. Box 15830, Sacramento CA 95852-1830,(916) 452-3211 i AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA j i

NL 89-379 April 14, 1989 I

U.'S. Nuclear Regulatory Commission Attn: Document Control Desk Hashington, DC 20555 Docket No. 50-312 Rancho Seco Nuclear Generating Station License No. DPR-54 OPERATING PLANT STATUS REPORT

' Attention: George Knighton l

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

Sincerely, v

Steve L. Crunk Manager Nuclear Licensing Enci (5) cc: J. B. Martin, NRC, Halnut Creek A. D'Angelo, NRC, Rancho Seco INP0 R. Twilley, Jr.

i b

'\

RANCHO SECO NUCLEAR GENERATING STATION O 1444o Twin Cities Road, Herald, CA 95638-9799;(209) 333-2935

_-_