ML20215H844
| ML20215H844 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 05/31/1987 |
| From: | Andognini G, Colombo R SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| GCA-87-206, NUDOCS 8706240234 | |
| Download: ML20215H844 (17) | |
Text
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MAY 1987 jj,.
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SUMMARY
OF PLANT OPERATIONS The plant has been in cold shutdown for the entire month of May 1987.
The initial shutdown was due to the December 26, 1985, loss of Integrated Control System power event.
PERSONNEL CHANGES REQUIRING REPORT There were several changes in personnel which require reporting pursuant to Technical Specification Figure 6.2-2.
Rancho Seco has a new Manager, Nuclear Operations, Joesph Firlit; Plant
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Superintendent, Danny Keuter; Superintendent, Nuclear Operations, Dan Comstock; and Nuclear Electrical /I&C Superintendent, Robert Wichert.
Mr. John McColligan has taken on the responsibilities of the Engineering & Quality Control Supervisor until the position is filled with a permanent employee.
Manager, Nuclear Operations (AGM, Nuclear Power Production)
Joesph Firlit is a long-time nuclear veteran of the Consumers Power Company, most recently General Manger at the Palisades Nuclear Plant.
He has a B.S.
in Electrical Engineering from University of Michigan -- Ann Arbor.
He held titles of General Manager at the Midland Plant (three years) ; Director of Quality Assurance for all Consumers Power nuclear plants (one year);
Director of Management & Budget, Energy Supply for all Consumers Power circtric production plants (four years); General Superviser, Research and Testing Laboratory of a large staff (seven years); General Engineer, Field Services involved in the startup of a fossil plant and a nondestructive testing department (four years).
Plant Superintendent (Director, Nuclear Operations and Maintenance Department)
Danny Keuter is a long-time veteran of the Portland General Electric company, most recently as the Trojan -- Manager, Technical Services (three years).
That department included engineering, radiation protection, chemistry, and training.
He hhs a B.
S.
in Nuclear Engineering from Oregon State University.
He previously held titles of Senior Operator Evaluator (one year) ; Branch Manager, Operations (over two years of day to day plant operations responsibility) ; Shift Supervisor with SRO certification, including preparation of LERs.
For over three years Mr. Keuter was a plant engineer through the Trojan plant's original startup.
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.s-Superintendent, Nuclear Operations (Unit operations Superintendent)
Mr. Comstock has a Senior Operating License for the Rancho Seco plant.
He was hired as a plant operator in 1972 and six months later was promoted to shift supervisor.
His earlier experiences include a tour of duty in the Nuclear Navy; being a control rooni operator at a fossil facility for over a year, and; before coming to the District, a control room operator / shift supervisor at the H.
B. Robinson Plant for over three years.
4 Nuclear Electrical /I&C Superintendent (Nuclear Electrical Maintenance Superintendent and Nuclear I&C Superintendent)
Mr. Wichert has ten years of commercial nuclear industry experience, all gained at Rancho Seco during increasing responsibilities as an engineer with the Maintenance and Technical support Groups.
He is a degreed, Registered Professional Engineer by examination in California.
His duties at Rancho Seco included acting as Shift Technical Advisor; preparing plant transient reports; preparing responses to NRC I
Notices and Circulars; supervision of a technical staff; work l
request review for engineering requirements; disposition of i
Nonconforming Roports; and responsibilities as specified by the Rancho Seco Emergency Plan as the Assembly Point Coordinator and j
as the Engineering and Quality Control Coordinator.
Most i
recently, Mr. Wichert was the acting Engineering and Quality Control Superintendent.
Acting Engineering and Quality Control Superintendent (Manager, Plant Performance Department)
John McColligan is the Director, Nuclear Plant Support.
In this capacity, he is responsible for plant cupport activities for radiation protection, chemistry, plant performance, and training departments.
Mr. McColligan has a E.S.
in Mechanical Engineering from the University of Pittsburgh and is a California registered professional engineer in both mechanical and nuclear disciplines.
Mr. McColligan has twenty-six years of experience in commercial nuclear power at both Rancho Seco and Shippingport.
At Rancho Seco he has held the title of Senior Power Plant Engineer, Supervising Mechanical engineer, Principal Project Engineer, and Assistant Manager -- Nuclear Plant.
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SUMMARY
OF CHANGES IN ACCORDANCE WITH 10 CFR PART 50.59 The plant staff accepted documentation packages for the following facility changes in May 1987 that are described below.
In addition, procedure changes from the last several months are included.
These changes were approved and reviewed by the Plant Review Committee (PRC) and the Management Safety Review Committee (MSRC).
There were no documentation packages completed for tests or experiments during May 1987.
1.
The District installed Rochester Instrument System qualified power supplies, instrument and isolator modules to replace existing components for qualification purposes and to make H4SCA and H4SCB compatible with the new cabinets H4SIA and H4 SIB.
Specifically, the power supplies in H4SCA and H4SCB were replaced. (ECN A-5620 A & B)
This change is a component replacement of vendor supplied electrical equipment which restores the cabinets H4SCA and H4SCB to their original intended Class 1 qualification.
This change improves plant safety by installing the required qualified replacement equipment and improving instrument calibration.
The equipment is internal to the cabinet and does not change the equipment function, and, therefore, does not affect the USAR or technical specifications. (Log No. 598, Revision 0).
2.
The District replaced Limitorque internal wiring for two Valves to meet the requirements of 10 CFR Part 50.49 in response to IE Notice 86-03. (ECN R-0699E)
HV-53620 -- H Recombiner isolation valve MOV HV-53621 -- H Recombiner isolation valve MOV The limit switch compartment space heater and its assoc!.ated wiring in HV-53620 were removed.
The limit switch compartment terminal block was removed.
Field wiring was connected directly to limit switch terminals.
Limit switch and torque switch jumpers were replaced with qualified 14 AWG SIS nuclear grade wire.
The previously installed splice on motor leads were replaced with qualified splices.
The limit switch compartment low I
point drain plug was removed. (Log No. 784E, Revision 1).
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The District completed (ECN R-0914HP, HS, & HU) the refurbishment?of:
HV-26007 -- Low Pressure Injection - HPI Cross Connect HV-26105 -- Reactor Building Sump Isolation SFV-25003 -- Borated Water Storage Tank Isolation for interim use as part of the motor-operated valve program performed in response to IE Bulletin 85-03. (ECN-R-0914HS & HU) Motor-operator components and operators are being replaced.
Environmental qualification and seismic requirements were maintained.. Also,. torque and. limit switch settings are being modified under a program based on the original response to IE Bulletin 85-03.
Equipment. classification and power' supplies remain unchanged.
The potential for failures on any of the MOVs is reduced as a result of the changes.
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The probability of occurrence or the consequences of an
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accident or malfunction of equipment important to safety L
previously evaluated in the safety analysis report is not
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increased because the Movs are being modified or replaced j
to ensure that the valves will operate within their i
p design basis.
The possibility of.an accident or malfunction of a i
different type than any evaluated previously in the i
L safety analysis report are not created because design I
L basis of the valves is not being changed.
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The margin of safety as defined in the basis for any j
Technical Specification is not reduced because the program ensures that the valves will operate within their j
design basis. (Log No. 831, Revision 1).
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4.
The District completed the refurbishment of:
HV-20001 -- Reactor coolant dropline isolation valve HV-20002 -- Reactor coolant dropline isolation valve HV-20003 -- Long term emergency core cooling SFV-26005 -- Low pressure injection loop
'A' HV-26038 -- Decay heat cooler by-pass HV-26046 -- Decay heat cross-tie line isolation valve HV-20570 -- Main steam
'A'
'B' drain to condenser HV-20571 -- Main steam
'A'
'B' drain to condenser HV-35069 -- Steam trap isolation SFV-26017 -- Cooling water to the decay heat coolers SFV-50005 -- Cooling water to reactor building emergency cooling unit SFV-50007 -- Cooling water to reactor building emergency cooling unit SFV-50008 -- Cooling water to reactor building emergency cooling unit SFV-50009 -- Cooling water from reactor building emergency cooling unit SFV-50010 -- Cooling water from reactor building emergency cooling unit SFV-50011 -- Cooling water from reactor building emergency cooling unit SFV-50012 -- Cooling water from reactor building emergency cooling unit SFV-29015 -- Reactor building spray additive to P-291A as part of the motor-operated valve program performed in response to IE Bulletin 85-03.
The program was outlined to the NRC in letters dated May 16, 1986 and November 5, 1986.
The overall engineering package was implemented to prevent these valves from failing on demand in a common mode due to improper switch settings as described in IE Bulletin 85-03.
(Log No. 849, Revision 2).
5.
The District modified the emergency battery lighting system.
The change was required by Section III-J of Appendix R to 10 CFR Part 50 for emergency battery lighting for safe shutdown capability of the plant and the disposition of NCR S-5488.
The changes made were:
o Removing the photocell from all the Dual-Lite brand outdoor emergency battery lighting units, including units kept as spares Installing a test switch and replacing the charger module of the above fixtures with a new module which is able to monitor a loss of AC power Replacing the inoperable photocell controlled Chloride outdoor lighting fixtures with the available spare Dual-Lite brand outdoor lighting units.
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- The lamp heads'and fixtures which were inaccessible to I
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electrical maintenance personnel were relocated.
,vi fr A A number of changes relating to the Fire Protection k-System Emergency Lighting are made, resulting in
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_ effectiveness.g This change does require a change in the jl 3-0 EUSAR and in th@'to technical specifications is not Fire Protection Program Manual.
A change V
or'an addision required.
Tnis change is bounded by the plant's design
. 'l basis. (Log No. 856, Revision 0).
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A partial turnover'of
'A' train pipe supports.for Nuclear-Scrvice Raw Water'and Decay Heat Removal System (DHRS) was made from ECN R-0780, Revision 1 to allow
'A' DERS to be-declared operable. (Log No. 805, Revision 0).
7.
The District issued:
AP.305-13,' Revision 22, " Environmental Release of.
Liquid Radioactivity"
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4 AP.310-IL, Revision 4,
" Liquid Radioactive Releases g,q'_,
Compliance with 10 CFR 20" A
y AP.31022L, Rdrision 2,
" Liquid Radioactive Releases l
Compliance with'10 CFR 50" AP.310-3L,c. Revision 7,
" Liquid Radioactive' Release j
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Monitor setpoint Calculation"
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AP.310-2LA, Revision 0, " Calculation of Cumulative Off-l
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Site Dose'from the Release of Liquid Effluents by J
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Radiation Protection Personnel"
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'AF.306 V-13, Revision 3,
" Lower Limit of Detection Count Time Determination"
.u AP.306 III-H,~ Revision 13, " Liquid Systems" s
s to further. improve the overall compliance with Appendix I to 10^CFR Part GO, Appendix B to 10 CFR Part 20, and Technical Specifications 3.15, 3,17, and 4.21.
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.'The rewrite of'AP.305-13' incorporates the administrative
' controls established to ensure the limits of 10 CFR Part p
150:and 10 CFR Part 20-are met. ;The changes to the other
,six procedures'are being'madeJto' assure that they will be
.compatibleLand. supportive ofLthe administrative and
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l calculation method changes being made in AP.305-13.- An
' assessment ~of these changes was performed.
TheLeonclusion wasithat the' changes do!not constitute a major change to the. plant's radioactive wasteLtreatment. system, because no physical plant modifications'are performed pursuant to
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these procedure changes.
In addition, the procedural changes:do not: increase the amount of radioactivity being released off-site.- The changes being implemented are strictly procedural in nature and provide the plant with limproved administrative controls WDose. purpose is to_give added assuranceithat the plant's radioactive liquid.
releasess are maintained below the vaarly ALARA limits of Appendix I to 10 CFR Part 50,.thee concentration limits of:
Appendix B tof10 CFR 20, and the requirements of.
Technical # Specifications 3.15, 3.17, and 4.21. (Log No.
844,-Revision 0).
8.
The Distriht approved STP.1011, Revision 2, to determine the time 11t would take to heat up the reactor coolant 0
system.(RCS) to 190 F with decay, heat as the only heat input.
This test:is to determine the lengthiof time available for the RCS to heat:up'when.the decay' heat.
removal (DHR). system ~is taken.out-of-service for:
. maintenance and repairs without exceeding cold shutdown s
limits.
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.STP.10ll'did not require a change'to technical specifications.
The. basis of Technical Specificaticn 0
3.1.1 states "(w) hen: T is below'280 F, a single a
' reactor coolant loop or.bIR' loop provices sufficient heat removal capability for removing decay heat; but single failure considerations require at least two loops be.
OPERABLE. Thus, if-the reactor coolant loops are not OPERABLE, this specification requires two DHR loops to be OPERABLE."
The STP had all four coolant loops operable throughout the test by intent.
This test did not' involve 1
an unreviewed safety-question. (Log No. 882, Revision'2).
9.
The District issued operating procedure B.2A, Revision 2,
" Plant Operation in Cold shutdown with Either or Both i
DHR Trains Not Operable."' In conjunction, the District issued' casualty precedure C.12, Revision 5,
" Loss of i
Decay Heat Removal System" to provide additional operator guidance upon the loss of operability of both decay heat q
loops, with the RCS filled and vented.
The preparation of B.2A is necessary to provide a normal operating procedure for the planned evolution for removing both decay heat loops from service for maintenance and repairs.
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The revision to C.12 is an enhancement which provides more detailed guidance for maintaining decay heat removal using the steam generators in the event that both DHR loops are inoperable with the RCS filled and vented and l
RCS temperature exceeds cold shutdown.
These procedure actions require a change to the USAR to address the use of steam generators for decay heat removal below RCS temperatures of 280 F.
A change or addition to technical specification s was not required.
i The probability of occurrence or the consequences of an t
accident or malfunction of equipment important to safety previously evaluated is not increased because the plant will remain in cold shutdown with required redundant methods of decay heat removal operable per Technical Specification 3.1.1.5.
The possibility for an accident or malfunction of a different type is not created because there are no changes to plant configuration or modes of operation not previously analyzed.
The margin of safety as defined in the basis for any technical specification is not reduced because the requirements of applicable technical specifications is maintained throughout the DHR outage.
Therefore, the procedure changes do not involve an unreviewed safety question. (Log No. 889, Revision 2).
10.
The District issued casualty procedure C.94, Revision 0, to govern the actions to be performed by operators in the event of a loss of all AC power (station blackout).
The establishment of natural circulation in the procedure is based on ATOG, part 2, Volume I (74-112746900).
The analysis of the results of a complete loss of all unit AC power is given in Section 14.1.2.8.4 of the USAR.
The environmental effects of a loss of all AC power event are presented in Section 14.3.2 and Table 14.3-1 of the USAR.
The limiting conditions for operation of the auxiliary electrical systems are addressed in Technical Specification 3.7.
This procedure is an outgrowth of a draft regulatory guide entitled " Station Blackout" and proposed 10 CFR Part 50.63 dealing with a station's capability to cope with a loss of all AC power.
Procedure C.94 does not require a change to technical specifications.
The actions required by the procedure maintain the plant in conditions bounded by the analysis in USAR Section 14.1.2.8.4, "Results of Complete Loss of All Unit AC Power."
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This procedure does not-increase the probability or consequences of an accident because is maintains the plant in a condition bounded by.the parameters of the accident analyzed in Section 14.1.2.8.4.
The procedure does not create the possibility of an l
accident of a different type than already evaluated t
because the procedure maintains the plant in hot standby.
The margin of safety is not reduced as defined in technical specifications because this procedure formalizes actions to be taken to comply with Technical Specification 3.7.
Procedure C.94 was developed for Rancho Seco in its f
present configuration.
The procedure does'not address l
Emergency Feedwater Initiation and Control or the addition of the Trans-America Diesel generators.
C.94 will be revised to reflect final plant configuration prior to restart. (Log No. 923, Revision 0).
11.
The District changed from using Portland Cement to Gypsum Cement for the solidification of radioactive waste.
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change was incorporated as part of procedure'AP.309 II-3,
" Radioactive Solid Waste Encapsulation," and AP.309 II-4,
" Solidification of Resin or Water Using Cement with PNSI (Pacific Nuclear Systems Inc.) Radwaste Solidification System."
The contract for solidification services was awarded to Pacific Nuclear Systems which uses Envirostone (Gypsum Cement) as stabilization media for water, resin, and filter elements rather than Portland Cement, which was previously used as a solidification media, j
The use of PNSI does not represent a major change in the District's radwaste handling as defined in Technical Specification 6.17.
There are no changes to be made in the plant USAR, or technical specifications as a result of the vendor change to PNSI.
Evaluation the change against Technical Specification 6.17, the use of the PNSI solidification system does not utilize any significantly different process equipment or effluent monitoring instrumentation from that use by the previous vendor Chem-Nuclear systems, Inc., or from that described in the USAR.
The process for solidification by PNSI uses cement as the solidification media, which is consistent with the previous vendor and what has been documented in the USAR.
The PNSI solidification system for radwaste handling poses no safety concern.
Extensive testing of the solidification media Envirostone, as represented in topical reports provided to the NRC, show that the cement meets all the requirements of 10 CFR Part 61 guidelines.
The probability of occurrence or the consequences of an I
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accident or malfunction of. equipment important to safety E
will not be. increased.because the solidification J
equipment is not connected or in-line with any safety piping' systems.
There is no possibility of an accident or malfunction of a different type-than any evaluated.previously in the safety analysis.
Radiological exposure is the most serious accident in the operation of the solidification system, however precautions are cited in AP.309 II-4, i
AP.309 II-3, and AP.305 II to minimize exposure.
i There is no reduction in the margin of safety as defined in the technical specifications from the use of the PNSI system or Envirostone cement.
All testing assures the PNSI system and the solidification media are within the required guidelines of 10 CFR Part 61.
The process of solidification media have been submitted and reviewed by the NRC. (Log No. 929, Revision 1).
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MAJOR ITEMS OF SAFETY-RELATED MAINTENANCE 1.
The dual train decay heat system outage ran from May 2, until May 6, 1987.
Eight decay heat system related i
snubbers were rebuilt during this time, l
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The regular semi-annual Health Physics Drill was held on i
May 7, 1987.
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The '24-hour' loaded run of the
'A' TDI diesel generator I
was completed on May 14, 1987.
4.
The
'A' decay heat removal system was returned to service on May 27, 1987.
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5.
ECN R-0066, Revision 0, modified the plant security electrical distribution system to eliminate a problem with a power supply for communication equipment.
6.
ECN R-0465, Revision 0, installed a dissolved gas stripping pump for the Post Accident Sampling System
(" ASS).
Previously, the accuracy of dissolved gas analysis was not very good because there was no agitiation to strip gas out of solution.
7.
ECN R-0976D, Revision 0:
Revised the existing drawing to eliminate the data for MOV HV-30124 (Gland Steam Spill Bypass Valve) and added a sheet using the new format to provide the data for the valve.
The valve's electrical elementary diagram and wiring diagram was revised to implement the changes for l
resolving IE Notice 86-29 concerns.
In addition, the limit and torque switch settings were adjusted according to the information provided in i
engineering calculations and drawings.
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ECN R-0871, Revision 0, established annunciator window H2WC-16 as "IDADS TROUBLE" and deleted the previous
" ELEVATOR TROUBLE" annunciation.
Part of the ECN was to ensure that there is sufficient information in the elevators for calling for assistance in case of emergency.
9.
ECN R-1323, Revision 0, installed a new electrical power distribution system for the new load at the west wing of the second floor of the Training & Records Building.
A new computer system for plant cabling and raceway tracking is being installed at the ground level office of the Administration Building.
The personnel and machinery previously at the ground floor of the Administration building were relocated to the west wing of the T&R i
Building, second floor.
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ECN.R-0956, Revision 0,. installed one-hour fire rated-wrap materials, qualified by test to the. requirements of
. ASTM E814 around portions of the. existing: raceway system' which'were not'in compliance with Appendix R wrap criteria.
The addition of the fire protection material-did'not change the function of the various systems, to Lwhich the eight subject raceways belong,:nor did itialter the interface withLthe existing' raceway system.
The work t ~-
was,done without de-energizing the:affected' circuits.
11.
.ECN R-1009,. Revision 0,' improved the' maintainability and operability of;the Control. Room / Technical Support. Center g
Air conditioning System by removing duct liner. material in the vicinity of eight isolation. dampers.
12.
ECN R-1555, Revision 0, provided exposed ground cable to panels H4CD-04-and H4 CAL-4 in the computer room to replace-embedded ground cable that was cut while core drilling work for conduit was being. performed.
j S-6237) '
'(RE:.NCR 13..
- A partial release of ECN R-1128E, Revision 1, was made j
for isolation switches for breakers 52-4A01 and 52-4A08 1
.and bus 4A load shedding (4ACC).
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REFUELING INFORMATION REQUEST e
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Name of Facility Rancho Seco Unit 1 2.
Scheduled date for next refueling shutdown:
Seotember 15. 1988 3.
Scheduled date for restart following refueling:
January 15. 1989 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:
March 15. 1988 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:
December 3. 2001 i
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-AVERAGE' DAILY UNIT POWER' LEVEL-DOCKET NO..
50-312
' UNIT' Rancho Seco Unit l' DATE 5/31/87 COMPLETED BY R.' Colombo l
TELEPHONE-(916) 452-3211 MONTH May 1987 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
l 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.
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a OPERATING DATA REPORT DOCKET NO.
50-312 DATE 5/31/87 COMPLETED BY R. Colombo TELEPHONE (916) 452-3211 OPERATING STATUS 1.
Unit Name:
Rancho Seco Unit 1 Notes 2.
Reporting Period:
May 1987 3.
Licensed Thermal Power (MWt):
2.772 4.
Nameplate Rating (Gross MWe):
963 5.
Design Electrical Rating (Net MWe):
918 i
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 letter dated 12/26/85 This Month Yr-to-Date Cumulative
- 11. Hours in Reporting Period 744 3.623 106.248
- 12. Number of Hours Reactor Was Critical 0
0 54.322 13.
Reactor Reserve Shutdown Hours 0
0 10.300 1,
- 14. Hours Generator On-Line 0
0 50.363.8
- 15. Unit Reserve Shutdown Hours 0
0 1.210.2
- 16. Gross Thermal 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)
-4528
-21227 38.267.558
- 19. Unit Service Factor 0.0%
0.0%
47.4%
- 20. Unit Availability Factor 0.0%
0.0%
48.5%
- 21. Unit Capacity Factor (Using MDC Net) 0.0%
0.0%
41.3%
- 22. Unit Capacity Factor (Using DER Net) 0.0%
0.0%
39.2%
- 23. Unit Forced Outage Rate 100.0%
100.0%
40.0%
- 24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):
N/A i
25.
If Shut Down At End Of Report Period, Estimated Date of Startup:
IIIIefinite
- 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 J
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($1SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT U 6201 S Street. P.O. Box 15830. Sacramento CA 958521830.(916) 452-3211 AN ELECTRIC SYSTEM SERV 4NG THE HEART OF CALIFORNIA GCA 87-206 i
June 15, 1987 I
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J B MARTIN REGIONAL ADMINISTRATOR REGION V OFFICE OF INSPECTION AND ENFORCEMENT U S NUCLEAR REGULATORY COMMISSION 1450 MARIA LANE, SUITE 210 WALNUT CREEK, CA 94596 OPERATING PLANT STATUS REPORT DOCKET NO. 50-312
Dear Mr. Martin:
Enclosed is the May 1987 Monthly Operating Plant Status Report for the Rancho Seco Nuclear Generating Station.
The District submits this report pursuant to Technical Specification 6.9.3.
Sincerely, Zt
..t' P]f/AL I
Car Andorfnini Chief Executive Officer, Nuclear Encl (5) cc:
I&E Wash (12)
MIPC (2) 7, INPO (3
R. Twilley Jr.
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G. Kalman
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UJo 3E4 RANCHO SECO NUCLEAR GENERATING STATION C 14440 Twin Cities Road, Herald, CA 95638-9799;(209) 333 2935 I L
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