ML20136F087
| ML20136F087 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/31/1985 |
| From: | Colombo R, Reinaldo Rodriguez SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| RJR-85-547, NUDOCS 8511220030 | |
| Download: ML20136F087 (10) | |
Text
r OCTOBER 1985
SUMMARY
OF PLANT OPERATIONS On October 1 at 2303 hours0.0267 days <br />0.64 hours <br />0.00381 weeks <br />8.762915e-4 months <br />, a reactor power reduction from 40% to 15'.' was initiated in order to perform a turbine overspeed trip test. At 0125 hours0.00145 days <br />0.0347 hours <br />2.066799e-4 weeks <br />4.75625e-5 months <br /> on October 2, the 220 KV breakers were opened to separate the generator from the grid in preparation for the test. At 0132 hours0.00153 days <br />0.0367 hours <br />2.18254e-4 weeks <br />5.0226e-5 months <br /> the reactor tripped on high reactor coolant system pressure at approximately 14% power and the reactor was subsequently brought to a hot shutdown condition. There was no further power generation for the month of October.
PERSONNEL CHANGES REQUIRING REPORT None
SUMMARY
OF CHANGES IN ACCORDANCE WITH 10 CFR 50.59(b)
The documentation for the following facility changes was completed in October. All of the changes have been subjected to the review and approval of the Plant Review Committee (PRC) and the Management Safety Review Comittee (MSRC).
1.
Procedure C.13A, Hot Shutdown from Shutdown Panel with a Fire in Control Room, was approved.
The procedure addresses 10 CFR 50.48 and 10 CFR 50, Appendix R, Section III.G.3 which requires that an alternative shutdown method be provided for a fire that occurs in any fire area where re-dundant equipment may be damaged (i.e., Control Room).
2.
A design change was made to provide diesel generator-backed emergency power to vital components of the Post Accident Sampling System (PASS) in the event of loss of offsite power coincident with an accident.
The design criteria for this change was based on the recommendations contained in NUREG 0737, Item !!.B.3, NRC clarification letter to SMUD, dated July 12, 1982, and NRC evaluation letter to SMUD, dated February 15, 1983.
3.
Thedecayheatremoval(DHR)systemoverpressureprotectioncircuitry was modified to enable the operator to reopen the associated DHR drop line block valve (HV-20001 or HV 20002) using normal Control Room equipment following loss of power to a channel.
The associated DHR drop line valve closes on a SFAS signal following the loss of 120 VAC power.
The circuit change ensures the capability to rapidly restore DHR cooling following a loss of power.
4.
Emergency Plan Procedure 500, Section 8, was revised (Rev. 17) to add the NRC to the monthly communications drill list (8.2.2.a) and to state that biannual offsite participation is required in the annual major exercise (8.2.6 a).
These changes were made in response to NRC Information Notice 85-44 and 85-55.
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An electrical cable raceway system was installed to support the power, control and instrumentation cabling required for the 1985 refueling outage TMI modifications. The cable was installed on a " bulk" basis with no unique system identification. Cable was pulled and coiled i
only under this modification.
No electrical terminations were made under this Engineering Change Notice (ECN A-4985),
i 6.
A change was made to provide for auctioneering at the NNI cabinets of l
1 the reactor coolant flow signals from RPS Channels A and B to ensure that a loss of power to Channel A or B does not cause a loss of flow signal to the Integrated Control System (ICS).
A false zero flow sig-nal to the ICS would result in main feedwater flow decreasing to zero at the maximum rate and, generally, a reactor trip on high reactor cool-ant system pressure.
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7.
SP 205.07 A/B/C/D and SP 214.01 were revised to reflect containment I
isolation valve stroke time changes resulting from IE Bulletin 79-018 modifications.
These changes are described in Technical Specification Amendment No. 72.
In addition, the stroke time of SFV-70001 was changed from 14 to 17 seconds as a result of the motor operator being changed.
i 8.
Conax Conductor Seal Assembly was installed at the hubs of the hydrogen purge system isolation valves (SFV-53615 and SFV-53616) and qualified Raychem splices made in the junction box to meet the LOCA requirements of 10 CFR 50, Appendix A, Criterion 4, and NUREG 0588, Category I.
These changes ensure that the eculpment will withstand the defined LOCA conditions of 286'F, 52 psig,100% relative humidity, chemical spray consisting of boric acid buffered with sodium hydrochloride, and gamma radiation of 1X108 rad (TID).
t 9.
The disposition of NCR S-4688 was approved. NCR S-4688 describes the falling of a plastic light co'.cr into the reactor vessel, breaking into several pieces, and locating in an inaccessible area at the bottom of the vessel. Based'on autoclave tests, material identification and evaluation, and the small mass of material involved, it was concluded that there is no adverse effect on nuclear safety, nor is an unroviewed safety question involved. The material was expected to melt and dis-integrate during plant heatup with no effect on reactor coolant system components.
- 10. Main steam pressure switches PSL-20601 through PSL-20608 were relocated and replaced with environmentally and seismically qualified pressure switches. The change was made to meet the requirements of IE Bulletin 79-01B and 10 CFR, Appendix A Critorion 4.
The new switches will pro-vide a more reliable indication in the event of a main steam line break l
(MSLB).
- 11. A computer alarm was installed to identify when the reactor coolant system pressure gets below 450 psig and low temperature overpressure protection I
(LTOP)isnotselected.
This change addresses the potential problem of a single power source failure disabling the overpressure protection l
i system alarm.
The alarm was installed on the Interim Data Acquisition and DisplaySystem(IDADS).
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ECN A-1710 which provides the mechanical portion of the radiation monitor R-15005 replacement task was completed. The replacement monitor, when operable, will be capable of measuring 1000 pC1/cc of Xe133 and Kr85 and will have sufficient range to measure the activity of the discharge from the waste gas surge tank. The electrical and instrumentation portions of the replacement will be performed under separate authorizations.
- 13. Two(2)isolationcabinetswereinstalledtoprovideadditionalsignal isolation capability for inadequate core cooling (ICC), Appenoix R (Fire Protection) and future work. The cabinets have been designated H451A and H45!B and are installed in the west and east switchgear rooms, respectively, i
14.
ECN A-5580 was prepared to allow the use of a ribbed mandrel type plug in the once through steam generator (OTSG) tubes. This type of plug 15 removable and allows sleeving of the tubes after the plug is removed.
The plug type is a Babcock & Wilcox design which utilizes Inconel 600 1
in the plug body and Inconel X-750 for the mandrel.
15.
The manganese-bronze nuts used for the fuel transfer tube to valve connec-l tions at SFC-500 and SFC-501 were replaced with 304 stainless steel nuts.
l The replacement was required because of cracking of the manganese-bronze nuts in service. The replacement was performed in accordance with ASME Section !!!, Subsection NC for Class 2 components, 1977 Edition with Addenda through summer 1978, 16.
New fire breaks were installed on cable trays inside the containment building and in the tank farm area. Material used in the fire breaks were Dow Corning silicone 3-6548 RTV and Marinite XL board which have i
been tested and qualified for fire resistance in accordance with ASTM E-119.
The breaks are designed to stop propagation of fire through the cable trays.
17.
SP 203.06 C&D were revised to include stroking closed decay heat suction
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isolation valves liv-2001 and HV-2002 and the recording of closing time.
The change was made in accordance with SMUD Ictter to the NRC dated July 18, 1979, inservice Inspection and Testing Program.
Stroking HV-2001 and HV-2002 closed will conform the surveillance procedure to the relief requested in the referenced letter.
18.
SP 211.10 ASB were revised to incorporate nuclear service eIcetrical i
building (NSED)essentialairhandlerfancurves.
These curves are used to verify proper NSEB essential HVAC systems A&B airflows.
The curves were plotted from data obtained from performing STP 176, NSED Emergency Air Handling Unit Flow Curve.
- 19. AnUnderwriter'sLaboratory(UL)approvedthree(3)hourratedbarrier was installed between auxiliary feodwater pumps P 318 and P-319 inside the misslie shield in the tank farm.
The barrier consists af a rolling counter fire door made of stainless steel with a removabic structural I
support.
The barrier provides separation between the pumps as required by 10 CFR 50, Appendix R.
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- 20. An electrical cable raceway system was installed to support plant modi-l l
fication to be completed during the 1985 plant outage. ii.e raceway was installed on a " bulk" basis with no unique system identification.
No I
electrical terminations were made under this Engineering Change Notice j
(ECNA-4768),
- 21. An air conditioning system was designed and installed to cool the corridor containing control panel H4ESA.
H4ESA contains vital components of the Post Accident Sampling System (PASS) which may not function properly if the ambient temperature exceeds 80*F.
In addition, a cabinet fan to circulate air was installed in H4ESA.
22.
Seven (7) self-contained breathing apparatus (SCBA) have been placed in racks within the Control Room boundary for use by Control Room personnel in the event of a hazardous chemical release.
Additional spare bottles located in racks near the Control Room will provide six (6) hours of breathing air to each of five personnel in the Control Room.
The addi-tion of the SCBA equipment was required by the Control Room Habitability Report in response to NUREG 0660.
- 23. A change was made to provide modulating control of the auxiliary feed-water cross-tie isolation valve (HV-31827) from shutdown panel H2SD in j
order to facilitate natural circulation cooldown of the plant in the event of a fire in the Control Room.
The change did not introduce any new fiilure modo into the circuitry.
- 24. Paragraph 3.2 of SP 205.02, Local Component Leak Rate Surveillance Test-ing, was changed to read:
" Piping in direct communication with the test volume boundary valves shall be depressurized during testing. When system pressure on the opposite side of the valve under test cannot be l
reduced to atmospheric preswre, then test pressure shall not be less than 52 psig plus system pressure." This change assures a minimum of 52 psi test pressure across penetration boundary valves and is in accordance with 10 CFR 50, Appendix J.
- 25. PASS strainer Y-209 was relocated from line 26001-1"-GD to 70057-3/8"-CA, which is common to both lines 26001-1"-GD and 70059-1"-CA.
The reloca-tion of the strainer provides maintenance capabilities for both linas for elimination of debris prior to entering the sample collection and analysis station.
26.
The existing Pressurizer Relief Valve Acoustic Leak Monitoring System was upgraded from Class 2 to Class 1.
The existing instruments were procured to meet Class 1 requirements.
However, some relocation and modifications were required to meet the requirements of NUREG 0737/II.D.3 and Regulatory Guide 1.97 for a Category 2, Type D variabic,
- 27. The tape wrap on the reactor building spray pump motor leads was replaced I
with Raychem tubing, and a spacer block for support and separation of the i
motor leads was installed.
Both the Raychem tubing and spacer block l
materials of construction will sustain LOCA conditions defined by IE Bulletin 79-010.
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l 28.
SP 201.13 was revised to allow moving the boraflex sample coupon spect-men holder (BSCSH) over irradiated fuel assemblies.
The total weight of the BSCSH is approximately 50 pounds. This change reduces the number of spent fuel movements in the spent fuel pool, thus lessening the potential for an irradiated fuel assembly accident.
29.
Environmental seals were installed on the Clns 1 Namco limit switches, Rosemount transmitters snd equipment important to safety located in the turbine building and tank farm area.
This change was made to ensure that equipment environmental qualification is maintained and the MSLB require-ments of IE Bulletin 79-01B are met, t
- 30. The vent lines on reactor coolant pump motor lube oil collection tanks T213 A&B were increased from 1" to 3" = ' rerouted down to floor eleva-tion - 25'-o".
This change was made to preclude a potential fire hazard which could have existed in the previous configuration.
- 31. The Control Room page was relocated and provided isolation near the safe shutdown panel.
The change will protect the operability of the l
shutdown panel page override in case of a fire in the Control Room.
32.
the operability of reactor New procedure SP 204.01E was written to verify (SFV-29015 and SFV-29016) b building Na OH chemical addition check valves use of a full stroke test under design flow conditions.
This test will be conducted during refueling outages.
MAJOR ITEMS OF SAFETY RELATED MAINTENANCE 1.
Auxiliary feedwater pump P-319 was inspected to determine why the outboard pump bearing temperature was high after 10-15 minutes of operation (SP 210.018).
The as-found oil level was 3" below the pump shaft cen-terline(acceptable)andthethrustandradialbearingsweredamaged.
The oli slinger ring was also found out of place.
The thrust and radlal bear-ings, bearing removal ring, thrust bearing ring and thrust bearing ring pin were replaced, i
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AVERAGE DAILY UNIT POWER LEVEL 00CKET NO.
50-312 UNIT Rancho Seco Unit I-I DATE 10-31-85 COMPLETED BY R. Colombo TELEPHONE (916) 452-3211 r
MONTH October 1985
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DAY TVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(NWe-Net) 1 307 17 0
2 0
18 0
3 0
19 0
i 4
0 20 0
5 0
21 0
l 6
0 22 0
7 0
23 0
l 8
0 24 0
t 9
0 25 0
10 0
26 0
11 0
27 0
12 0
28 0
13 0
29 0
i 14 0
30 0
15 0
31 0
16 0
INSTRUCil0NS
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l On this format, list the average daily unit power level in NWe-Net for each day in the reporting month.
Compute to the nearest whole neg4 Watt.
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OPERATING DATA REPORT l
DOCKET NO.
50-312 3
l DATE 10/31/85 i
COMPLETED BY R. Colombo TELEPHONE (916) 452-3211 l
l OPERATING STATUS NOTE:
1.
Unit Name:
Rancho Seco Unit 1 2.
Reporting Period:
October 1985 3.
Licensed Thermal Power (18dt):
2.772 4.
Nameplate Rating (Gross NWe):
963 5.
Design Electrical Rkting (Net NWe):
918 6.
Maximum Dependable Capacity (Gross NWe):
917 7.
Maximum Dependable Capacity (Not NWe):
873 8.
If changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report, Give Reasons:
N/A l
9.
Power Level to Which Restricted, If Any (Net NWe):
N/A 10.
Reasons For Restrictions, If Any N/A
+
This Month Yr-to-Date Cumulative 11.
Nours in Reporting Period 745 7.296 92.401
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- 12. Number of Hours Reactor Was Critical 25.5 1.793.4 53.240.8 13.
Reactor Reserve Shutdown Hours 0
110.5.
10.300.4 i
- 14. Nours Generator On-Line 25.5 1.672.8 49.336.3
- 15. Unit Reserve Shutdown Hours 0
0 1.210.2 i
16.
Gross Thermal Energy Generated (MWH) 32.243
_4.131.372*
125.737.774*
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17.
Gross Electrical Energy Generated (MWH) 8.318 1.383.700
_40.015.743 l
- 10. Net Electrical Energy Generated (MWH) 0 1.289.988 37.881.184
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- 19. Unit Service Factor 3.45 22.95 53.45
- 20. Unit Availability Factor 3.45 22.95 54.75
- 21. Unit Capacity Factor (Using M0C Net) 0.05 20.35 47.05 22.
Unit Capacity Factor (Using DER Net) 0.0s _
19.3s 44.7s 23.
Unit forced Outage Rate 96 di__,
34.45 30.05 24.
Shutdowns 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
_ November 2. 1985 l
26.
Units In Test Status (Prior to Conumrcial Operation):
Forecast Achieved I
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INITIAL CRITICALITY
_ N/A
__N/A INITIAL ELECTRICliY
_ N/A N/A COMMERCIAL OPERAT10N
_ N/A
__N/A i
- Figures corrected due to calculational error in Dec 84, Jan 85 and March 85.
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LNT MARIE Rancho Seco Unit 1 DATE 10/31/85 OCTOBER CoasPLETED BY R CoI*
steromTa00NTN TELErt00NE (916) 452-3211
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1 85-19 ZZ ZZZZZZ Reactor trip on high pressure. Trip l
preceded by "A" PEP trip and a low condenser vacuum. An " action items" l
list with a timetable for completion was subciitted to the NRC on 10/18/85.
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REFUELING INFORMATION REQUEST 1.
Name of Facility Rancho Seco Unit I 2.
Scheduled date for next refueling shutdown:
Jan 1. 1987 3.
Scheduled date for restart following refueling:
May 1.1987 4.
Technical Specification change or other license amendment required:
a)
Change to Rod Index vs Power Level Curve (15 3.5.2) b)
Change to Core Imbalance vs Power Level Curve (15 3.5.2) c)
Tilt Limits (TS 3.5.2) 5.
Scheduled date(s) for submitting proposed licensing action:
Aug 1. 1986 6.
Important Ilcensing considerations associated with refueling:
N/A 1
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 i
the Spent Fuel Pool:
. December 3. 2001 I
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esuun SACRAMENTO MUNICBML UTILITY DISTRICT U 6201 S Street, PO. Bom 15830, Sacramento CA 958521830,(916) 452 3211 AN ELECTRIC SYSTEM SEHVING THE HEART OF CALIFORNIA RJR 85-547 November 15, 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 October 1985 Monthly Plant Status Report for Rancho Seco fitiETpir O e ating Station Unit No.1.
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R. J. R00 IGUEZ ASSISTANT GENERAL F
- GER, NUCLEAR Enclosures (5) cc:
I&EWashington(9)
Region V MIPC(2)
INP0 4
MANCHO BECO NUCit AM GENtMATING STATION I i 14440 Twin Cities Romf, Herald, CA 95030 9199;(2001 333 2036