ML20195C119

Monthly Operating Rept for Apr 1986

ML20195C119
Person / Time
Site:	Rancho Seco
Issue date:	04/30/1986
From:	Colombo R SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
RWC-86-228, NUDOCS 8605290759
Download: ML20195C119 (11)

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APRIL 1986

SUMMARY

OF PLANT OPERATIONS The plant has been in cold shutdown for the entire month of April. Initial shut-down was due to the December 26, 1985 loss of ICS event.

PERSONNEL CHANGES REQUIRING REPORT There was one change in personnel which requires reporting pursuant to Technical Specification Figure 6.2.2. Mr. Robert McAndrew has been added to the Technical Support staff as the Senior Nuclear Engineer. Mr. McAndrew comes to the Distrit:t with 15 years of experience from Babcock and Wilcox where he was involved with fuel mechanical analysis, reactor internals design and analysis, plant startup #

testing and reload startup physics testing. Mr. McAndrew holds a Master of Science in Nuclear Engineering from Texas A&M University and a Bachelor of Scie r in Mechanical Engineering from Worcester Polytechnic Institute and is a registered Professional Engineer in Virginia.

SUMMARY

OF CHANGES IN ACCORDANCE WITH 10 CFR 50.59 The final documentation packages for the following facility and procedures changes were completed in April 1986. All of the changes have been subjected to the review and approval of the Plant Review Committee (PRC) and the Management Safety Review Committee (MSRC). No final documentation packages for tests or experiments were completed during April 1986,

1. Process liquid radiation monitors R-150ll, R-15012, R-15013 and R-15014 were removed from the liquid radwaste treatment system. The monitors were ineffec-tive in determining the effectiveness of the various process ion exchangers due to the carry over of non-soluble isotopes which caused the monitors to be in a constant state of alarm. These monitors did not perform a personnel protection or accident mitigation function, so their removal has no effect on any of the accidents described in the FSAR.

The physical removal of these four liquid process radiation monitors has re-sulted in another method of assessing the effectiveness of the ion exchangers.

This method currently is to perform a decontamination factor calculation for the ion exchangers periodically, whenever they are used.

2. The 120 upper core barrel bolts have been replaced with an improved design which incorporates a locking device which is remotely installed. The reason for this change was that some upper core barrel bolts have shown cracking failures at the head to shank transition. These bolts perform a design func-tion of attaching the core support shield assembly to the core barrel assembly.

The new bolts are of the same material and will be installed with less stress on each bolt than the original design. In addition, the manufacturing process (machined vs hot forged heads) used for the new bolts reduce the susceptibility of the new bolt to head failures.

3. The four mechanical snubbers on the auxiliary feedwater lines have, as part of the auxiliary feedwater header modification effort, been replaced with ITT Grinnel Figure 201 hydraulic snubbers with a rated load of 12,500 pounds The reason for the change is that all other snubbers in the plant are hydraulic and there existed no equipment or procedures for the testing of the p 8605290759 060430 1 PDR ADOCK 05000312  :

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Monthly Report mechanical snubbers. The hydraulic snubbers are sized to perform the same function as the mechanical snubbers and will be surveilled in accordance with existing snubber surveillance procedures.

4. The existing pressurizer relief valve accoustic leak monitoring system has been upgraded to Class 1 from Class 2. These modifications were required to meet NUREG 0737, Item II.D.3 and Regulatory Guide 1.97 fer a Category 2, Type D variable. The instrumentation and wiring were upgraded to Class 1 to pro-vide reliable indication of pressurizer relief valve position. These instru-ments are not used in automatic control of any plant equipment. This instrumentation is only used for indication of equipment status. The new configuration meets the single failure criteria, therefore, a redundant channel should always be available in the event of an accident.

5. The torque switch settings on the "A" and "8' loop decay heat system valves HV-20005 and HV-20006 have been changed from 2.75 (158 ft-lbs) to 3.0 (175 ft-lbs).

The manufacturer recommends a minimum 173 ft-lbs to open these valves and the valve operators have a maximum setting of 3.25 (188 ft-lbs). The new setpoint valve is within the maximum stem torque values for proper valve operation without possible damage. ,

6. Surveillance Procedure SP 201.12, Revision 3, was changed to conform to License Amendment No. 57, dated October 30, 1984. The surveillance procedure now specifies an acceptance criteria for leakage of a sum total of 6.0 gallons per hour for both the decay heat and building spray systems. The NRC concluded that although the changes to the Technical Specifications increase potential LOCA doses, the total conservatively calculated doses are still within 10 CFR 100.11 dose guidelines.

7. The Rancho Seco Emergency Plan (AP.500), Section 8.0 (Rev. 18), has been revised to show a more flexible training program for the Emergency Response Organization groups. This change does not decrease the effectiveness of the Emergency Plan.

This change is in accordance with NRC commitments and is consistent with other Rancho Seco procedures.

8. Nuclear Engineering Procedure (NEP 5107.2) was developed to provide an adminis-trative control to determine " operability" of piping systems and supports. This procedure provides " operability" criteria where piping systems or supports are discovered through analysis or observation to be outside the design basis but can be determined to be operable in accordance with Technical Specification 1.3 until modifications restoring the system to code allowables are completed. The code allowable limits have a built-in margin to ensure that they are conserva-tive. By using this margin and establishing an " operability" limit less than ultimate, the piping system and supports are capable of performing their intended safety function until the next outage of sufficient duration at which time the system will be restored to within code allowable limits. If a system has been determined to be outside code allowables but is " operable" pursuant to this procedure, and that system experiences a dynamic event, then that system should be immediately reanalyzed prior to further operation to determine if it is still

" operable."

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Monthly Report 9. The Update Safety Analysis Report (USAR), Table 5.2-1, will be revised to re-flect that HV-32243 is a valve which serves to isolate containment penetration 41 (the "A" main steam line). HV-32243 was not included in USAR Table 5.2-2 because the pegging steam valves downstream of HV-32243 are closed during normal operation and only open when a turbine trip signal is received. Because of the piping class change at HV-32243 from Class 1, seismic category 1, ASME Nuclear Code 2 to Class 2, seismic category 2, this valve must be included in the USAR Table 5.2-2.

MAJOR ITEMS OF SAFETY RELATED MAINTENACE

1. The Steam Generatoi Inspection Program was completed in April 1986. The details of the tubes plugged and inspected are contained in enclosure 1.

2. Repair efforts on makeup pump P-236 continued through the month of April 1986.

Currently plans include replacing the gear box and rotating element and repair-ing the pump casing.

3. The process of performing preventative maintenance on various plant valves which are of particular interest to the Operations staff, continued through the month of April 1986.

4. The work to enhance the reliability and accuracy of the reactor building post-accident sampling system and reactor coolant post-accident sampling system has continued through the month of April 1986.

. 5. The "B" diesel generator's air start compressor (AC) was maintained by the i

adjustment and alignment of the motor to the compressor and the replacement of a pulley bushing.

REFUELING INFORMATION REQUEST i

1. Name of Facility Rancho Seco Unit 1 I

2. Scheduled date for next refueling shutdown: Sept. 13. 1987

3. Scheduled date for restart following refueling: .. Janua ry 13. 1988

4. Technical Specification change or other license amendsent required:

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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 13.1987

6. Important licensing considerations associated with refueling: N/A r

7. Number of fuel assemblies:

a) In the core: 177 b) In the Spent Fuel Pool: 316 >

8. Present licensed spent fuel capacity: 1080 1

9. Projected date of the last refueling that can be discharged to I 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 04-30-86 COMPLETED BY R. Colombo TELEPHONE (916) 452-3211 MONTH April 1986 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 0 17 0 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 16 0 INSTRUCTIONS 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.

OPERATING DATA REPORT DOCKET NO. 50-312 DATE 04/30/86 COMPLETED BY R. Colombo TELEPHONE (916) 452-3211 OPERATING STATUS

1. Unit Name: Rancho Seco Unit 1

2. Reporting Period: April 1986

3. Licensed Thermal Power (MWt): 2.772

4. Nameplate Rating (Gross NWe): 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): 0

10. Reasons for Restrictions, If Any: NRC limit pursuant to letter dated 12/26/852 _

This Month Yr-to-Date Cumulative

11. Hours in Reporting Period 719 2.879 96.744

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 Reserve Shutdown Hours 0 0 1.210.2

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

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

18. Net Electrical Energy Generated (MWH) -2.823 -15.010 38.315.092

19. Unit Service factor 0.0% 0.0% 52.1%

20. Unit Availability Factor 0.0% 0.0% 53.3%

21. Unit Capacity Factor (Using MDC Net) 0.0% 0.0% 45.4%

22. Unit Capacity Factor (Using DER Net) 0.0% 0.0% 43.1%

23. Unit Forced Outage Rate 100.0% 100.0% 32.7%

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

N/A

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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

T DOCKET NO. 50 317 .

UNIT SHUTDOWNS AND POWER REDUCTIONS

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UNIT NAME Rwhn 4cn " nit 1 -

DATE 4/30/R6 APRIL 1986 cOeetETED BY R_ Col mhn

• REPORT m TELEPHONE (916) 452-3211 ,

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!? 3 jIE Licensee ,E*, ja, Cause & Correctise

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? $ 2s& Event g? g3 Action to 35 E j;g g Report

• mO g' Prevena Recurrence 6

1 85-12-26 F 719 A 3 85-25 CB INSTRU Reactor trip en high pressure pre-ceded by a loc of ICS power.

Corrective a: cas being implemented.

I 2 3 4 F Ferad Reason: Method: Exhibit G-Instructions S ScheJuled A-Equipment Failure (Eaplain) 1-Manual for Preparation of Data B-Maintenance of Test 2-Manual Scram. Entry Sheets for Licensee C-Refueling 3-Automatic Scram. Event Report (LE R) File t NUREG-D-Regularory Restriction 4-Other (Explain) 01611 E4)perator Traming a lxense Exansnation F-Administrative 5 G4)perational Error (Emptam) Eshibit I - Same Source sW77) Ilf)ther (E uplam)

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ENCLOSURE 1 A OTSG - ALL INDICATIONS ARE 00 5 cm SMFLE f TU8ES EXTENT SCOPE R0W-TUBE  % TWO s d IST 927 F/L 6% (of 1 SG) Run in '85 - - - 0 2ND 2171 F/L Lane / Wedge & Drilled 15 TSP 5-3 51 14 X 33-1 42 UTS X 57-126 69 15 X 68-1 74 15 X 71-26 43 UTS X 73-8 67 UTS X 119-107 51 15 X 148-26 62 15 X 2-18 21 UTS 4-39 36 14 10-1 25 10 11-67 28 15 22-39 34 14 23-93 25 14 23-94 31 14 24-95 37 14 25-96 34 14 26-99 36 14 28-101 37 14 30-2 25 UTS38-111 37 14 49-3 30 8 52-3 22 8 72-2 27 UTS 75-22 25 15 80-25 24 9 81-1 25 UTS 82-13 33 13-14 83-10 31 UTS 97-2 29 15-U 108-1 34 13 112-1 31 13 148-19 36 4 151-2 21 14 3RD 1216 F/L Expanded Lane & Center 63-26 49 3 X 81-65 47 UTS X 4TH 803 F/L All Previous Indications 23-83 40 3 X 36-6 67 UTS X 121-84 30 4 TOTALS 5117 27 12 (720-<40)

• Enclosure 1-(continued)

B OTSG - ALL INDICATIONS ARE 00 5 =

SAWLE 6 TUBES EXTENT SCOPE R0W-TUBE 5 TWO a d Special 642 Above 14 TSP Lane / Wedge 69-8 51 UTS X 69-17 47 UTS X 70-20 51 UTS X 71-17 78 UTS X 72-3 35 UTS X 72-5 70 UTS X 72-10 93 UTS X 75-11 36 15 X 77-20 42 15 X 77-23 34 15 X 77-25 49 15 X 77-30 35 15 X 80-12 57 UTS X 81-13 43 UTS X 81-18 43 UTS X 81-25 42 UTS X 81-26 54 UTS X 81-28 33 UTS X 82-21 48 UTS X 82-22 90 UTS X 83-2 74 15 X 83-3 94 15 X 70-10 31 UTS 70-19 36 UTS 75-14 20 15 75-16 24 15 7F-22 31 15 77-10 28 15 77-11 32 15 77-12 20 15 77-16 30 15 77-19 26 15 77-28 23 15 77-31 25 15 80-9 25 UTS 82-1 29 15 82-13 26 UTS 1ST 929 F/L 6% (For 1 SG) 0 Center 77-66 43 UTS X 76-4 90 3-4 X 79-64 41 UTS X 80-64 68 UTS X 82-58 95 UTS X 106-1 55 15-U X' 108-93 48 6 X 81-64 30 UTS 120-1 24 13 123-1 33 14 145-53 32 8-9 2ND 1855 F/L 12% (For 1 SG)54-127 57 15 X 70-25 88 UTS X 77-68 56 3-4 X 142-38 55 15 X 143-34 64 2 X

ENCLOSURE 1 (Continued)

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~8 OTSG - ALL INDICATIONS ARE 00 g SAMPLE # TUBES EXTENT SCOPE R0W-TUBE 1 TWO  !! Ef 2ND continued 144-57 61 14 X 150-25 80 15 X 39-116 33 9 - ~ .

108-119 26 10 122-45 31 2 125-86 25 8-9 143-29 35 14 143-48 29 9 3RD 4200 F/L 24% (For 1 SG) Expanded Lane / 13-73 43 15 X Wedge & Drilled 15 TSP 13-74 39 15 X 36-8 80 LTS X 36-18 40 LTS X 48-1 40 15 X 58-1 67 15 X 59-124 95 15 X 60-129 40 15 X 73-71 50 3-4 X 81-39 62 UTS X 82-25 75 UTS X 82-26 40 UTS X 82-55 38 UTS X 82-56 39 UTS X 82-57 70 UTS X 82-61 41 UTS X 84-66 55 3-4 X 90-17 40 15 X 91-17 42 15 X 91-124 39 9 X 95-2 35 UTS X 111-7 81 15-U X 2-12 24 10 27-99 25 8 28-7 31 14 28-101 32 LTS 29-6 22 14 32-106 24 15 38-1 29 13 44-4 27 UTS 48-9 35 8 48-25 37 10-11 53-1 32 12 56-8 33 8 62-129 23 15 64-7 29 UTS 68-69 37 3-4 68-72 24 LTS 71-54 27 8 75-68 35 3-4 76-5 27 3-4 82-53 24 9 96-2 28 4 130-89 26 14 141-2 27 13&14 TOTALS 6984 F/L 48 58 (pf0,<40)

esuun SACRAMENTO MUNICIPAL UTILITY DISTRICT 6201 S Street. P.O Box 15830 Sacramento CA 95852-1830.1916)452 3211 AN ELECTHIC SYSTEM SERVING THE HEAR T OF CALIFORNIA RWC 86-228 May 14, 1986 0

J B MARTIN REGIONAL ADMINISTRATOR REGION V 0FFICE OF INSPECTION & ENFORCEMENT ~

U S NUCLEAR REGULATORY COMMISSION 1450 MARIA LANE SUITE 210 -

WALNUT CREEK CA 94596 OPERATINC PLANT STATUS REPORT DOCKET NO. 50-312 Enclosed is the April 1986 Monthly Operating Plant Status Report for Rancho Seco Unit No. One, kOThubo R. W. COLOMB0 REGULATORY COMPLIANCE SUPERINTENDENT Enci 5 cc: I&E Washington (12)

MIPC (2)

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RANCHO SECO NUCLEAR GENERATING STATION : ,14440 Twin Cities Road, Herald, CA 95638 9799;(209) 333 2935