ML20127G855

Monthly Operating Rept for May 1985

ML20127G855
Person / Time
Site:	Rancho Seco
Issue date:	05/31/1985
From:	Colombo R, Reinaldo Rodriguez SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
References
RJR-85-288, NUDOCS 8506260111
Download: ML20127G855 (12)

Text

.

MAY 1985

SUMMARY

OF PLANT OPERATIONS The plant has been in cold shutdown for refueling and plant modification for the entire month of May.

PERSONNEL CHANGES REQUIRING REPORT The Senior Nuclear Engineer position in Engineering and Quality Control (E&QC) has been vacated. Efforts are underway to fill this vacancy.

SUMMARY

OF CHANGES IN ACCORDANCE WITH 10 CFR 50.59(b)

The documentation for the following facility changes was completed in May:

1) The low differential pressure alarm (annunciator H2X window 30) from the spent fuel pool coolant filter was removed because, when a new filter is in-stalled, the differential pressure is so low that the low differential pressure alarm actuated and remained actuated for an extended period. The low differen-tial pressure alarm was already set so low that setting it lower was not a solution. The time required for the filter to becon.e backed up enough for the alarm to reset was significant. As a result, the operators had no n:ethod of determining whether the alarm was for a low or high differential pressure con-dition. Health Physics personnel make periodic checks of the filter and the fi.lter is generally removed due to a buildup of radioactive contamination.

Should the filter fail, it could go undetected for a considerable period of time; however, failure of the filter would not affect the pool cooling system.

The benefits of a valid differential pressure alarm are significantly greater than the risk of an undetected filter failure. Removal of the low alarm will not affect system safety.

2) Two independent, uninterruptable 120VAC power supplies (diesel and battery backed) were provided in the Nuclear Service Electrical Building (NSEB) to supply power fo. new computer systems added in response to NUREG 0737 and NSEB fire protection system requirements. Each power supply consists of a 600A, Class lE battery charger connected to a Class lE, diesel-backed load center; a 50 KVA, non-1E inverter; a non-lE battery sized for 30 minutes of operation for the computer system and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for t.he fire protection system; a 128 VAC, non-lE panel board; and a 50 KVA, non-lE, 480V-120V regulating, dry type transformer. No failure modes have been added by this change. Each system is separate and independent. Any failure in one battery system will not affect the operation of the other system. N

3) A safety grade reactor coolant system saturation temperature (T-SAT meter) indication was provided to the Control Room. The indication meets the require-ments of NUREG 0737 (Item II.F.2), Regulatory Guide 1.97 (Table 2, Type B variables), and NUREG 0578. The wide range temperature instrumentation utilized by the T-SAT meter are redundant and independent. Therefore, no single failure of the instrumentation added to the RPS cabinets will impair the capability to provide an indication to the T-SAT meter.

[A l l/l

4) A Safety Parameter Display System (SPDS) cabinet and interconnecting

' equipment were installed to meet the requirements of NUREG-0696. The SPDS was installed as an aid in the assessment of plant safety status by providing a concentrated set of selected plant parameters for use in the analysis of

. plant conditions. The SPDS information is obtained from the IDADS and

' existing-indications in the Control Room. Failure of the SPDS will only eliminate its own display system.

5) A change was made to allow the use of damping ratios of up to 20% in the seismic design of cable tray supports. Appendix 5B to the USAR lists damping values of 10 to 15% for the seismic analysis of bolted structures. The 20%

value has been in use since 1980 and is supported by Bechtel letter (No. FSL-3537) and Safety Evaluation Report (NUREG 0857), Supplement No. 4. This change clears NCR S-3579.

6) 'The design pressure of the reactor coolant system drain tank was lowered to 15 psig to eliminate the requirement for the tank to be code stamped. The tank possesses the same structural strength as the original tank; however, out-of-roundness conditions made it desirable to eliminate the code stamp re-quirement. The lower pressure rating does not alter the FSAR Analyses for Radioactive Releases.

7) An automatic notification system was installed to signal the State's Office of Emergency Services of emergency core cooling initiation and high radiation of the radioactive gas effluent stack. The system will multiplex the signals via a dedicated telephone line from the communications room at Rancho Seco to the State's Office of Emergency Services on Meadowview Road in Sacramento.

This change was a requirement of State Senate Bill 1184. System interfaces with Class 1 systems made by this change will be isolated. Thus, installation and operation of the system will not impair the operation of Class 1 systems.

8) A change was made to allow reactor operation with a reduced Tave at the end of fuel cycle 5. This. change allowed operation at a higher power level during the end of fuel Cycle 5 coastdown. Reduction of the reactor coolant system nominal Tave from 582*F to 577*F does not significantly affect the  ;

results of any FSAR Chapter 14 accident analysis nor any other accident analysis submitted to the NRC since the issuance of the operating license.

This was assured by resetting the RPS Hi-Flux Trip so the maximum power will not exceed 2568 MWt.

9) A system to continuously indicate containment water level and alarm on high containment emergency sump level was installed to address the requirement of

.R.G. 1.97 and NUREG 0737, Item II.F.1. The containment water level transmitters and~ level switches are installed in different locations using separate penetra-tions and power supplies. Therefore, no single failure in the instrument loop will compromise the ability to provide an indication of containment water level.

'10) New Class lE pressure transmitters were installed in the reactor coolant system to provide wide range pressure to the SPDS and plant computer for in-dication in the Control Room. This change was made to meet the R.G. 1.97, Table 2 requirement to provide Category 1 redundant indication of wide range pressure. All equipment supplied for the modification were seismically and environmentally qualified. No single failure in the instrument loop will impair the ability to provide wide range pressure to the Control Room.

(

11) A change was made to install piping necessary to provide for increased sampling capability of the Reactor Building atmosphere under post-accident

.onditions. Additionally, this change provided for a tie-in to the Auxiliary Building Radwaste Ventilation System for removal of any gaseous leakage within the Sample Collection and Analysis Station. Upgraded sampling capability under post-accident conditions was required by Item II.B.3 of NUREG-0737.

12) A change was made to provide demineralized water for sample dilution and component cooling water for cooling of the Post Accident Sampling System (PASS).

The system was designed to comply with NUREG-0737 and provides for remote opera-tion to maintain personnel exposure ALARA. Demineralized water is supplied to the PASS from the Miscellaneous Water Holdup Tank pump discharge header supplying the New Resin Slurry Tank. Control Rod Drive (CRD) cooling water is

- supplied to the PASS from the discharge of the CRD cooling water heat exchanger and returned upstream of the heat exchanger.

13) Four (4) safety grade instrumentation strings were added to the Reactor Protection System for the purpose of providing buffered reactor coolant signals to the T-SAT meter calculation. The equipment is qualified as Class lE in accordance with IEEE-323. Instrumentation and loops are independent and re-dundant; thus no single failure in the instrumentation portion added to the RPS cabinets will impair the capability of providing indication to the T-SAT meter.

14) A 1.emporary tower crane was erected for the construction of the Nuclear Service Electrical Building (NSEB). The crane was a fixed base "Pecco 1400" and had a fixed elevation jib with 360 horizontal rotation and a power lifting trolley for in - out and vertical load movement. It was equipoed with limit switches to prevent overload and had a main power disconnect at the tower base. The crane was designed to withstand all credible accident conditions and not contribute to tornado generated missiles which might damage any Class 1 equipment or system. Following construction of the NSEB, the crane was disassem-bled.

15) Bypass lines were added to the Main Steam to Reheaters Isolation Valves and Main Steam to Pegging Steam Isolation Valve. These lines are for warm-up only and the valves are locked closed during normal plant operation providing isolation of these valves. The failure of one of these bypass lines would result in a very small main steam line break which is bounded by the analysis for a main steam line break.

16) Section 9.8.2.2 of the USAR, " Loading and Removing Fuel," was updated to match Procedure M.2, " Reactor Intervals Removal and Replacement," regarding the removal of the upper plenum assembly. USAR Section 9.8.2.2 required flooding of the fuel transfer canal prior to removing the upper plenum assembly.

Procedure M.2 does not require the canal to be flooded.

17) A change was made to verify RB Spray Nozzle Air Flow by the method of thermography (Infrared Imaging System) at the spray nozzles +150' elevation.. .

Previously verification was performed by visual observation of ribbons. Thermo-graphic verification has been successfully demonstrated at several PWRs.

a

18) Anticipatory reactor trip equipment was added to the RPS cabinets to initiate an RPS trip on the tripping of both main feedwater pumps or on a turbine trip. Four (4) redundant and separate sensors with isolated inputs to the RPS are provided to monitor turbine and main feedwater pump status.

This change was required by I&E Bulletin 79-05B, dated April 21, 1979 and NRC letter, dated September 7, 1979.

19) Changes.were made to upgrade safety related electrical equipment to meet post accident environmental conditions as required by I&E Bulletin 79-01B, 10 CFR 50, Appendix A, Criterion 56 and NUREG 0696. The changes were made to ensure that the equipment would perform as required during the postulated accident without affecting the original failure mode analysis for the equip-ment.

20) Additional non Class lE battery capacity was'added to provide adequate spare storage capacity for new loads. Previously non Class lE batteries E and F did not have adequate ampere hour storage capacity to support the required duty cycle with the addition of new loads. No new failure modes were intpo-duced by this change.

21) Two (2) new panels were installed to house the Hydrogen Monitoring System remote control panel post-accident sample system valve switches, and the Auxiliary Building steam isolation reset switches. This change provides Class 1E panels that are seismically and environmentally qualified. Class lE devices are separated from non-Class lE devices in the panels by a physical barrier. The change was made to meet NUREG 0737 and R.G. 1.97 requirements.

22) A panel was installed in the Technical Support Center to house duplicate indications to those which are presently indicated on the remote shutdown panel and to house recorders that provide a record.of computer point values. The panel and associated ~ equipment are Class II. Failure of the panel will not impair the operation of the plant. The change was made to meet a requirement of NUREG 0696.

23) A modification was made to isolate the diesel' generator local control panel from Control Panel H2ES. This modification also isolates the control pushbuttons in the Panel HISS for the auxiliary feedwater pumps from the 4160V switchgear breakers and the 125V DC panel for the motor operated valve from the steam i driven auxiliary feedwater pump. The modification prevents a fire in H2ES from making the diesel generators inoperative and prevents a fire in Panel HISS from making the auxiliary feedwater pumps inoperative.

24) Temperature sensing devices were installed in the Auxiliary Building to pro-l vide signals to the auxiliary steam line isolation valves. An I&E Bulletin 79-01B. review of an auxiliary steam line failure in the Auxiliary Building showed that a severe temperature increase might cause the safe shutdown equipment in the building to be inoperable. This change allows the -20ft level of .he Auxiliary Building to be isolated in the event of an auxiliary steam line break.

. 25) An isolation switch was installed to isolate control of the make-up and high l pressure injection pump lube oil pumps from the Control Room and allow control l of the lube oil pumps from the 480V motor control center. This change prevents

.a fire in Panel HlRC from making the lube oil pump for the make-up and high

. pressure injection pumps inoperative.

I a

O

~

26) Contacts from the pump monitor were installed in parallel with the existing liquid flow switches for radiation monitors R-15009 and R-15010. This change prevents a flow fault alarm condition generated by the Nuclear Service Cooling Water Pumps not running from masking the operation of the fault alarm circuitry in monitors R-15009 and R-15010.

27) The inverter output voltage and frequency meters were rewired to monitor the actual inverter output instead of the vital bus output. This change allows the monitoring of inverters when maintenance is being done and the vital bus is being fed from another source.

28) A change was made to provide a skid-mounted and shielded Post-Accident Sample System (PASS) and an electrical control panel for remote operation of the PASS. PASS has the capability to collect samples, provide on-line analysis and dilute the sampled fluid as necessary for grab sample collection. PASS is normally de-energized and isolated during normal plant operation with the excep-tion of the containment atmosphere sample line heat tracing, and will not affect normal plant operation.

29) USAR, page 5.2-45, stated that all air operated reactor building isolation valves are held open by air pressure and closed by a compressed spring. The letdown RB-isolation valve (SFV-22009) has no spring and is closed by air. This valve has a pneumatic relay which closes the valve by using a separate air supply (local accumulator) if the normal air supply fails. The accumulator has been shown by test to be sufficient to close the valve if the normal air supply fails. The installation was determined to be acceptable as is.

30) The RCP Seal Return RB Isolation Valve (SFV 24013) is listed in FSAR Table 5.2-2 as failing closed on loss of air. SFV-24013 was installed without a spring force to close option. Instead, it was installed with air accumulators to ensure operability following a loss of instrument air pressure. An evaluation showed that the installed accumulators will provide ten (10) valve strokes following a loss of instrument air pressure. This is sufficient to ensure that the valve will close upon an SFAS actuation. The installation was determined to be acceptable as is.

s

31) An analysis ~was performed to show that an 18 month calibration frequency for. area radiation monitors R-15025, R-15026 and R-15027 was adequate to maintain the reliability of the equipment.

Previously, a quarterly calibration of this instrumentation was required. The calibration frequency change was made because cf ALARA concerns resulting from the inability to purge the containment. A safety evaluation was performed by TERA Corporation to support the reduced calibration frequency.

32) A procedure for volumetrically examining the Reactor Coolant. Pump casing weld was added to the Inservice Inspection Program. This technique has been used successfully on two identical pumps at a similar plant, ar.d will be implemented by personnel experienced in the method. The resulting procedure has been re- ,

viewed by a qualified Level III technician from a vendor's staff as well as a qualified Level III member of the SMUD staff to ensure that the requirements

.of the applicable codes and standards are met.

MAJOR ITEMS OF SAFETY RELATED MAINTENANCE

1) RC pump (P-2108) stuffing box and motor stand were assembled. Work included installing shaft, bearings, seal and bolts.

2) The upper primary manway cover on the "B" 0TSG was replaced. Work consisted of cleaning stud holes-and-gasket surfaces, obtaining new gasket and torquing

'of manway cover nuts with hydraulic tensioning equipment.

'3) A damaged seat'for high pressure turbine valve TV-3 was replaced.

4) The following 0TSG tube plugging was performed during the 1985 refueling outage.~

A '0TSG '

Row - Tube Location  % Defect Type Plug 33-5 Bundle 58 MS 46-7 Bundle 56 MS 49-5 Bundle 60 MS 101-10: Bundle 36 M 146 Bundle 90 MS 65-2 Wedge / Lane 91 MS 75-12 Wedge / Lane 39 MS 75-18 Wedge / Lane 34 MS 2-23 Periphery 68 MS-3-22 Periphery 36 MS 12-65 Periphery 55 MS 28-2 Periphery 67 MS 29-3 Periphery 76 MS 32-4 Periphery 78 MS 45-1 Periphery 81 MS 45-2 Periphery 37 MS 46-1 Periphery 54 MS 50-121 Periphery 74 MS 50-122 Periphery 93 MS-59-3 . Periphery 46 MS 119-108 Periphery 69 MS 120-106 Periphery 78 MS 149-22 Periphery 40 MS

.83-130 Proximity to AIW HDR -

MS88-126 -

MS

'94-129 -

MS 105-1 -

MS 2-3 Misplaced Plug - Explosive 29-2 Replugged (Bottom Only) - "

29-3' -

72-4 -

74-27 -

29-1 -

, =

~l~

Row - Tube' Location- .% Defect ' Type Plug 77-4'- Lane- 99 MS

• 22-89 ~ Periphery 44 MS 115 ~ Periphery. 34 MS 122-1 -~ ' Periphery 38 MS

.124-1 Periphery 37 MS

133-82 Periphery- 63 MS 136-76 --Periphery 40 MS 6 Replugged (Bottom Only) - Explosive

"~

6-51. " -

47 "- 60-431 '" ~ "

73-5, -

<74-3 -

75-13 .

75 -

75-19 -

75-27 -

77-17 -

77-18 -

~82-12 -

97-1 -

104-1 -

105-1 -

146-1 -

MS - Mech'anical plug with stabilizer M - Mechanical plug only-

5) 'The tubes and outside casing of auxiliary boiler E-360 were replaced. Work-included replacement of 1400 2 X 21" tubes, replacement of mud drum heater bundles, replacement of all insulation and refractory material and hydro test-ing of the boiler.-

O

~g .

REFUELING INFORNATION REQUEST

1. Name of facility Rancho Seco Unit 1

2. Scheduled date for next refueling shutdown: Sept 15.1986

3. Scheduled date for restart following refueling: Jan 15.1987

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: April 9. 1986

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: Dec 3rd: 2001 l

i J

t

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

DATE 05-31-85 CONPLETED BY R. Colombo TELEPHONE (916) 452-3211 40 NTH Nay 1985 UA. 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 0 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 05/31/85 COMPLETED BY R. Colombo TELEPHONE (916) 452-3211 OPERATING STATUS NOTE:

1. Unit Name: Rancho Seco Unit 1

2. Reporting Period: May 1985

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 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 This Month Yr-to-Cate Cumulative

11. Hours in Reporting Period 744 3.623 88.728

12. Number of Hours Reactor Was Critical 0 1.624.5 53.071.9

13. Reactor Reserve Shutdown Hours 0 0 10.189.9

14. Hours Generator On-Line 0 1.618.2 49.281.7

15. Unit Reserve Shutdown Hours 0 0 1.210.2

16. Gross Thermal Energy Generated (MWH) 0 4.055.333 125.665.601

17. Gross Electrical Energy Generated (MWH) 0 1.366.846 80.798.809

18. Net Electrical Energy Generated ~(MWH) 0 1.289.988 37.881.184

19. Unit Service factor 0.0% 44.66% 55.54%

20. Unit Availability factor 0.0% 44.66% 56.90% _

21. Unit Capacity Factor (Using MDC Net) 0.0% 40.78% A8.90%

22. Unit Capacity Factor (Using DER Net) 0.0% 38.79% 46.51%

23. Unit forced Outage Rate ,

0.0% 8.8% 29.4%

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

Refuelina - March 15.1985 - June 15.1985 Three Months _

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

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 L

m- _ _

c . _ . _ . _ _ . - . . - _ _ _ _ _ . _ . _ . _. _ _ _ . , , _ _ ___.,e

. 7. ..

' UNIT SilUTDOWNS AND POWER REDUCTIONS DOCKET NO. sin 119 UNIT NAME RANPHO 9FCO IINTT I

, DATE' G /11/ AR COMPLETED ltY R~. Colombo REPORT MONTil May 1985 TELErilONE (916)452-3211 n.

- E e No. Date g.

.! E 3 ij "h

.y ) .E g& E Licensee Event Eg gg h

n.1 Cause & Coerective Action to u!O H

$E 5 3 di e Report # v> U Pievent Recusicuce n .

o q 5 5/1/85 S 744 C 1 N/A ZZ ZZZZZZ Shutdown for Refueling

! 2 3 -

4 F: Forced Reason: Method: Exhibit G Instructions S: Scheduled A-Etiuipment Failure (Explain) I-Manual for Pacparation of Data 11 Maintenante of Test 2 Manual Scram. Entry Shecis for I.icensee C-Refueling 3-Auf omatic Scram. Event Report (LliR) File (NUREG-D Regulatory Restriction 4-Other (Explain) 0161) li-Operato Training & License Examination F.A.iniinistrative 5 G-Operational Error (Explain) Exhibit I - Same Sotirce

,(9/77) II-Other (Explain)

,j '

'$SMUD SACRAMENTO MUNICIPAL UTlWTY DISTRICT O 6201 S Street, P.O. Box 15830. Sacramento CA 95852-1830,(916) 452-3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA RJR 85-288 June 13,1985 DIRECTOR OFFICE OF INSPECTION AND ENFORCEMENT U S NUCLEAR REGULATORY COMMISSION WASHINGTON DC 20555 OPERATING PLANT STATUS REPORT DOCKET NO. 50-312 Enclosed is the May 1985 Monthly Plant Status Report for Rancho Seco Unit

+

R. J. R0 RIGUE ASSISTANT GENERA MANAGER, NUCLEAR cc: I&E Washington (9)

Region V MIPC (2)

INP0 h

d g

'y t t

.