ML20236N017
| ML20236N017 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 06/30/1987 |
| From: | Andognini G, Colombo R SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| GCA-87-345, NUDOCS 8708110337 | |
| Download: ML20236N017 (17) | |
Text
_
l.,
JUNE 1987 l
SUMMARY
OF PLANT OPERATIONS i
The plant was in cold shutdown for the entire month of June 1987.
The initial shutdown was due to the December 26, 1985, loss of Integrated Control System power event.
PERSONNEL CHANGES REQUIRING REPORT There were no changes in personnel which require reporting pursuant to Technical Specification Figure 6.2-2.
SUMMARY
OF CHANGES IN ACCORDANCE WITH 10 CFR PART 50.59 The plant staff accepted documentation packages for the following facility changes in June 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 June 1987.
1.
The existing coating of the Reactor Building Liner Plate was repaired (RE: ANSI N101.2-1972 and N5.12-1974).
Some of the maintenance coating work was accomplished during the February 1981 plant outage.
The remainder of the work was completed during June 1987.
The new coating was applied according to a procedure qualified under design basis accident condition, using hand tool cleaning and brush / roller applied coating.
The coating procedure was qualified so that the coating would be effective under irradiation and design basis accident conditions.
The coating and the procedure for its application were qualified at Oak Ridge National Laboratory, as documented in ORNL Log Book NO. A7562; A7-31-80, on August 14, 1980.
The coating system does not pose any safety related problems for a safe shutdown of the plant or interfere with the safe operation of the plant.
Limited abrasive blasting was performed with vacuum blasting under enclosures with HEPA filtering.
No adverse effects were encountered by other equipment inside the Reactor Building.
Resultant vapor was tested for activation and released through the Reactor Building ventilation system.
The vapor did not degrade the ventilation system. (Log No. 204, revision 1).
8708110337 870630 PDR ADOCK 05000312 g
2.
Pursuant to ECN A-4515, revision 2, ECN A-4615A, revision 1, and ECN A-4615B, revision 2, the District upgraded pressurizer safety / relief (SRV/PORV) valve discharge piping supports.
Fifteen supports were added; ten supports were removed, and thirteen supports were modified.
In addition, the reactor coolant pump motor service platform (for personnel access) had new bracing members and stiffening plates added.
NUREG-0737, item II.D.1 required licensees to test and evaluate their pressurizer safety and relief valves and associated piping for plant tLansients.
The original analysis was corrected to the new requirements and the SRV and PORV discharge piping was redesigned and modified to meet code requirements to withstand
-pressurizer blowdown transients identified in NUREG-0737.
The coolant pump motor service platform required reinforcement to accommodate the modified pipe supports.
The modification improves plant reliability and ability to mitigate plant accidents.
A potential failure mode is eliminated by this modification, therefore, nuclear safety is improved. (RE: Technical Specification Proposed Amendment 106). (Log No. 806 A&B, revision 0) l l
_ ____ _ __ __ _ -__ 91
'c 3..
The District completed (ECN R-0914HT, & HV) the refurbishment of:
HV-26008 -- Low Pressure Injection - HPI Cross Connect HV-26106 -- Reactor Building Sump Isolation
.as'part of the motor-operated valve program performed in response to IE Bulletin 85-03. (ECN R-0914HT & HV).
Motor-operator components and operators were replaced.
Environmental qualification'and seismic requirements were maintained.
Also, torque'and limit switch settings were 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.
The probability of occurrence or the consequences.of an accident.or malfunction of' equipment important:to safety previously evaluated in the safety analysis report is not increased because the MOVs were modified or replaced to-ensureLthat the valves will operate within their design basis.
The possibility of an accident or malfunction of a different type than any evaluated previously in the-safety analysis report were not created because design j
basis of the valves was not changed.
The margin of safety as defined in the basis for any Technical Specification was not reduced because the program ensures that the' valves operate within their design basis. (Log No. 831, revision 1).
4.
The District completed the refurbishment of:
SFV-26016 -- Cooling water to the decay heat coolers SFV-29107 -- Reactor building spray pump discharge isolation valve SFV-29108 -- Reactor building spray pump discharge isolation valve HV-20005 -- Reactor Coolant System to decay heat removal train "A" isolation valve HV-20006 -- Reactor Coolant System to decay heat removal train "A" isolation valve SFV-26040 -- Decay heat removal cooler outlet isolation valve HV-26047 -- Decay heat cross-tie line isolation valve as part of the motor-operated vclve program performed in response.to IE Bulletin 85-03.
The program was outlined to the NRC in letters dated June 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 relocated backup (liquid and containment atmosphere Post Accident Sampling System (PASS) grab sample connections out of Room 106.
This modification improves the reliability of PASS and facilitates collection of backup liquid and containment atmosphere grab samples.
This modification will reduce radiation exposure during sampling operations, and allow the use of shielded flasks for shipment of material off-site.
The modification facilitates meeting the NUREG-0737 II.B.3 requirement (capability to obtain backup grab samples for analysis).
Samples are now easier to obtain.
There is no equipment "important to safety" in the room where these PASS samples were relocated to (Room 134).
PASS is a class 2 system having no equipment directly important to the safe operation of the pla:t. (Log No.
722, revision 0).
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'6.
ECN R-0380, revision 2, installs two-inch thick insulation over instrument lines for hot leg level i
transmitters and steam generator level transmitters.
The
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excess flow check valves on the instrument lines were s
also insulated.
.This change complies with the requirements of NURIG-0737, item II.F.2, " Instrumentation for Detection of Inadequate c
Core Cooling," and an NRC order' dated December 10, 1982.
This change is T. art.of a hot leg level measuring systen to-j 1
provide, reliable control Room indication of reactor j
coolant inventory trending with the reactor coolant pumps i
tripped.
j This change also complies with.the requirements of NUREG-l0737, item II.E.1.2, " Auxiliary Feedwater System Automatic Initiations and Flow Indication."
This ECN
.l installed insulation over portions of the instrument.
lines installed for EFIC steam generator level transmitters which will provide safety grade AFW 4 automatic initiation on low steam generator feeductor level.
q The instrument lines were insulated to protect them against ambient temperature transients during and after design basis events.
]
The additional weight of the insulation and stainless steel jacketing was included in the analysis to ensure.
that the existing seismic class 1 tray and tube supports maintain their seismic support capability.
The insulation placed on the QA class 1 sensing lines is commercial grade, with a certificate of compliance.
1 There are no new failure modes introduced by implementation of ECN R-0380.
The insulation has a-
)
stainless steel jackat which will prevent blocking of screens over the emergency sump (RE: NRC Generic Letter
.85-22).
USAR section 7.3.2.2.2.G was modified as part of this ECN to describe'the corresponding change.
The r.argin of safety as defined in the basis for any technical specification is not reduced because the changes implemented by ECN R-0380, revision 2, increase the accuracy of the existing hot leg and steam genere. tor level indication. (Log No. 741, revision 1) 7.
ECN R-0459 spared and abandoned in place cable 1RlSO4B6A.
That cable was found routed with a 480 VAC power feeder.
cable for motor-operated valve HV-20002. A new twisted shielded instrumentation cable was pulled through raceway system "B" to connect SFAS cabinet H4SAB2 and electrical penetration H7RP18.
The cable is used in the instrument loop for PT-21099 (RE: LER 85-16, and NCR S-5263).
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'The function'of the cable to provide an interlock permissive signal for HV-20002 based on RCS pressure.
That function is unchanged by this modification.
There was no effect on plant safety or plant operation due to this change.-
The new instrumentation cable eliminates spurious operation of.HV-20002, thus improving
.)
nuclear safety ~. (Log No. 742, revision 1) 8.
ECN R-0598, revision ~0, inserted pulsation dampeners (filters) into sensing lines for FT-26048 and FT-26049.
- These instruments measure flow in the decay heat removal system' train cross-tie pipe line.
The District noted.
that the cross-tie flow indication would oscillate to mid-scale from its normal zero reading while decay heat removal. pump."A" was in operation.
When the two i
Instruments _were isolated from the system, the-oscillations ceased.
Pulsation dampeners work to eliminate oscillations in flow transmitters Cue to pump pulsation or flow noise.
The dampeners qualify as class 1 because they.are commercial grade and have a certificate of conformance.
The addition of the dampeners do not affect any' safety analysis for the plant, therefore,. nuclear safety is not.
Laffected by this change. (Log No. 769, revision 0) 9.
ECN R-0911, revision 0, moved the control power supply connection for Essential Filtration Units SF-A-7A and SF-A-7B from. electrical distribution panels S1A3/SlB3 to
. panels SlA2-1/S1B2-1.
This modification improves the reliability of the control power supply to the Essential' Filtration-Units.
The S1A2-1/S1B2-1 120 VAC vital power distribution panels.should maintain the vendor required voltage at the filtration unit terminals under all 1
operating conditions.
I Each of the two essential filtration units consists of a moisture eliminator, electric duct heater coil, two HEPA i
filter banks, two carbon filter be.nks,and a. booster fan.
The modification was only to the control power supply connection and corresponding cable routing for these units.
This modification does not impact the seismic or QA class 1 qualification of the filtration units.
The fire protection program was unaffected by this modification.
l This modification does not change any description providad in the USAP.
It is bounded by the Licensing j
L Design Basis.
A change or addition to technical specifications was not required.
(Log No. 843, revision 0)
)
1 10.
ECN R-1041 provides a buffered, isolated data link between five existing radiation monitors and the Safety Parameter Display system (SPDS).
The data link permits monitoring at the SPDS of radiation level inputs from the following:
1 1.
Radwaste Area Vent R-15546A 1
2.
Auxiliary Building Stack R-15045 I
3.
Containment Atmosphere / Reactor Building Purge l
Stacks R-15044 4.
Main Steam Line "A" Radiction Monitor RE-15047 5.
Main Steam Line "B" Radiation Monitor RE-15048 SPDS will process the logarithmic analog signal and provide displays in engineering units, i.e., uci/sec or mR/hr as appropriate.
The displays are part of the SPDS displays in the Control Room.
They are grouped under the existing alphanumeric
" Radiation Monitors" group in the form of a status line.
Monitors in alert status will appear in the " Alert Display."
This change is in accordance with NUREG-0737, item I.D.2 and NUREG-0696, item 5.5 to monitor plant gaseous effluent and main steam line vent radiation.
This information is now easily accessible and convenient from a human factors standpoint when displayed on SPDS.
All signals from the radiation monitors are supplied to the SPDS via the non-class 1 Anatec multiplexer --
H4CDAR1.
The multiplexer provides isolation between the SPDS and the radiation monitors.
In the event that either the SPDS or the multiplexer are lost due to failure, radiation levels are still available from the DRMC.
Implementation of ECN R-1041 does not involve an unreviewed safety question because the addition of a data link between existing non-class 1 radiation monitors and the SPDS is a human factors consideration only.
Plant technical specifications tre not impacted by the changes.
The modification will be reflected in a revision to the USAR.
ECN R-1041 is within the plant's licensing design basis.
(Log No. 892, revision 0) 11.
ECN R-1142, revision 1, disconnected the fire alarm trip circuit for the fans A-536, " Reactor Building Exhaust;"
A-542A, "Radwaste Exhaust," and; A-542B, "Radwaste Exhaust" from a smoke detector in plant Zone 20.
This modification does not affect the existing fan start sequence.
L This modification allows the uninterrupted monitoring of j
radiation potentially being released, even during a i
fire.
It will also permit airborne gaseous and particulate fission products to be filtered out because the fans remain in operation.
This modification will also help keep the Reactor Building at a relative negative pressure to ambient, if A-536 is in use during the postulated fire.
These features allow plant operators positive control of ventilation designed to mitigate radioactive releares.
(Log No. 894, revision 0) 12.
The District issued AP.305-22, revision 0,
" Radiation Protection on-the-job Training Guidelines."
The procedure furnishes instruction and guidance for the implementation of the oJT segment of the Radiation Protection Technician Training program.
The procedure is an operational enhancement which clearly documents the District's compliance to training program commitments in USAR Section 12.3.6.
The issuance of this procedure will be followed by a revision to USAR Section 12.3.6.
These USAR changes are administrative only and not substantive, reflecting organization restructuring.
The-USAR changes do not represent a change in the basic operating philosopny of the Radiation Protection department.
The technical specifications are not impacted. (Log No. 988, revision 0) 13.
Procedure AP.305-30, revision 4,
" Temporary Lead Shielding," was issued by the District.
The procedure is applicable only to temporary shielding and not permanent shielding or shielding that is to be installed via design modifications.
The revision encompasses the INPO guidelines set forth in INPO Report 86-006 (Good Practice TS-411) February 1986, " Temporary Lead Shielding."
The procedure provides for an engi. leering evaluation of the temporary shielding stress loading requirements and l
installation guidelines that ensures the shielding I
configuration design is in accordance with seismic classification and design criteria.
Missile protection review is also required in the procedure.
Guidelines are provided in AP 305-30 that reduce radiation l
exposure in a reasonable and effective manner in l
accordance with ALARA objectives.
As a minimum, documentation will be made in Part IV of the Temporary Shielding Req 1est that will identify all
)
components bearing load, all locations of nearest j
supports, all equipment impacted by a dropped load, and j
all possible missile effects due to the shielding installation.
l
i Because of the discussion above, the implementation of the revision to Rancho Seco Procedure AP 305-10 does not j
involve an unreviewed safety question. (Log NO. 986, l
revision 0)
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l 14.
The District issued new surveillance procedure SP.321A,
" Monthly Test of Nuclear Services Bus 4A2 Voltage l
Protection," a;,d SP.321B, " Monthly Test of Nuclear i
Services Bus 4E2 Voltage Protection."
)
i This SP supports the new electrical configuration which is accompanying the installation of the two TDI diesel generators.
The technical specification for the new electrical system add requirements for new systems, improve requirements for existing systems, and incorporate NRC recommendations.
The changes demonstrate the operability of required systems to ensure safe operation of the plant.
l The changes improve the determination of operability and, therefore, preserve the margin of safety. (Log No. 760, revision 3) l
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MAJOR ITEMS OF SAFETY-RELATED MAINTENANCE 1.
The cable raceway tracking system verification program sample size was increased to approximately 600 cable inspections due to the discovery of additional misrouted cables.
Currently 242 cables have been inspected.
2.
The installation of the new batteries for the "B" nuclear service bus was completed in June 1987.
3.
ECN A-3881, revision 3, added an annunciation of a diesel generator breaker lock-out (486 relay) on the "Not Ready for Auto start Alarm."
This made the diesel alarm scheme consistent and logical, and assured all required alarrs are on the "Not Ready for Auto start Alarm."
4.
ECN R-0163, revision 1, re;.nstalled heat tracing and insulation on pipe 31822-2y" for pump P-318 (Auxiliary Feedwater) test line freeze protection.
Heat tracing for this line is in a mild environment, and not required for accident mitigation.
5.
ECN R-0314, revision 0, removed FSL-54512 from the cold duct of Control Room HVAC Air Handler AH-A-1 and installed it at the return air duct of AH-A-1 so total flow into or from the Control Room can be monitored correctly and AH-A-1 is not spuriously shutdown.
6.
ECN R-0588, revision 1, reduces wire size.in SFAS panel H4SDB1 so that lugs will fit on terminal blocks, and identified an unlabeled power terminal block.
ECN R-0605, revision 1, added identification of an unlabeled power terminal block in SFAS digital cabinet #3, H4SDA3.
(RE:
4 NCR S-5410 and NCR S-5393 respectively) 7.
The Training and Records Building computer rooms were isolated from the building's HVAC system, and provided with a separate system (ECN R-0675, revision 0).
8.
ECN R-0804, revision 1, removed the cross-tie between the high pressure N2 system and the Instrument Air System on che west side of the "A" main circulation pump (P-402A) on personnel safety grounds to eliminate the potential for inhaling N '
2 9.
ECN R-0904A, revision 0, removed a hot gas bypass line between an inactive coil section and the compressor suction for the CR/TSC Essential HVAC refrigeration system.
This modification eliminates compressor overheating.
10.
ECN R-0938, revision 2, added CR/TSC HVAC pressure taps to facilitate system testing and verify the satisfaction of positive air pressure requirements.
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ECN R-ll60, revision 0, added new receiver level gauges and removed insulation to allow indication of refrigerant
.in-the receiver for Essential HVAC Condensing Units U-545 n
A&B (Control Room / Technical Support Center (CR/TSC)) and
'U-503 A&B (Nuclear Services Electrical Building [NSEB]).
1 12..
ECN R-ll59, revision 0, added two new refrigerant sight glasses in the refrigerant liquid lines to indicate the condition of liquid' refrigerant in each HVAC unit described above.
13.
ECN R-1260, revision 0, provided a deeper liquid trap seal for the CR/TSC HVAC condensate drain line to provide a means to prime traps and measure level of the water seal.
14.
ECN R-1261, revision 0,-installed security barriers in openings into vital areas to conform with the Security Plan.
The change added barriers to a fire hose cabinet
' penetration.
i 115.
ECN R-1343, revision 0, adds piping drops to three pendant sprinkler heads in corridor 103 (grade level Auxiliary Building, Zone 43) wet pipe sprinkler system.
.This modification eliminates interference with conduits and radiation sampling equipment in the hallway.
The sprinklers.can now provide a water discharge pattern with a water density sufficient to adequately control'a fire.
16.
ECN L-1361, revision-0, installed a new valve to fulfill the function of PSV-30125 (Gland Steam Spillover Relief Valve) which has the 40 psig set pressure specified by Westinghouse.
ECN R-0976, revision 0, altered and refurbished HV-30124 (Gland Steam Spill Bypass Valve) to ensure availability and development of baseline data.
17.
ECN R-1401, revision 0, increased the length of piping
. drops to four pendant sprinkler heads in corridor 105 (grade' level Auxiliary Building, Zone 42) wet pipe sprinkler system and also depositions one of them.
This modification eliminates interference with cable trays and new cable tray supports conduits in the hallway.
The sprinklers can now. provide a water discharge pattern with a water density sufficient to adequately control a fire.
18.
ECF R-1486, revision 0, removed twenty-nine protective cable tray covers in NSEB Cable Shaft Zones 83 and 82.
The fire deluge spray nozzles can now reach cables as intended.
19.
.ECN R-1493, revision 0, (RE:NCF S-6053) removed motor-operator space heater from valves HV-23004, FV-24019, FV24020, FV-24021, and FV-24022.
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ECN.R-1508, revision 0, reversed the wiring on the front face' incandescent. lamp sockets on eight 3ailey 885 system modulosLin the RPS cabinets to assure the minus 15 VDC supply.to the socket is on the center pin of the socket and.not on the socket sleeve.
(RE: 'LER 87-07) 21.
ECN R-1751, revision.0, provided channel isolation as required by IEEE-279-1971, per the design criteria of the reactor protection system (RPS).
This modification resolves one commitment.made in LER 87-23.
22.
On an interim basis, the Dietrict placed radiation monitor R-15049 in-service prior to'taking R-15050 out-of-service during June 1987. (RE: ECN R-0295, revision 1).
23.
On an interim basis,.the District connected new 480 VAC
. power feeds for standby battery' charger H4BBD using MCC breaker S2B313; feeder cable 121B313A/B, and; MCC brea.;'.er' S 2 A12 8.
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REFUELING INFORMATION REQUEST 1.
Name of Facility Rancho Seco Unit 1 2.
Scheduled date for next refueling shutdown:
Sectember 15.198L 3.
Scheduled date for restart following refueling:
Januarv 15. 1989 4.
Technical Specification change or other license amendment required:
a)
Change W Rod 11dex vs Power Level Curve (TS 3.5.2) b)
Change to Core Imbalance vs Power Level Curve (TS 3.5.2) c)
Tilt Limitr (TS 3.5.2) i 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:
1030 9.
Projected date of the last refueling that can be discharged to tiie Spent Fuel Pool:
December 3. 2001
- AVERAGE DAILY UNIT POWER' LEVEL
> ^
DOCKET NO.
50-312 l
UNIT.
Rancho Seco Unit l'-
DATE 6/30/87 l
COMPLETED BY R. Colombo 1
. TELEPHONE (916) 452-3211 I
. MONTH.
June 1987
)
'o DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MHe-Net) 1 0
17 0
2' O
18 0
1 3
0
'19 0
4 0
20 0
j 5:
0 21 0
1 6
0 22 0
7-
~0 23 0
i
'8-0 24 0
9-0 25 0
1 L10 0
26 0
11 0
27 0
I
'12 0
28 0
13 0
29 0
l 14
-0 30 0
i
?
15 0
31
)
i
.16 -
O i
-INSTRUCTIONS i
On-this format, list the average daily unit power level in MWe-Net for each day in the repcrting month.- Compute to the nearest whole megawatt.
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1
0PERATING DATA REPORT L
DOCKET NO.
50-312 DATE
.6/30/87 COMPLETED BY-R. Colombo i
TELEPHONE (916) 452-3211 1 PERATING STATUS 1
1.
Unit Name:
Rancho Seco Unit 1 Notes
'2.
Reporting Period:
Sune 1987 3L Licensed Thermal Power (MHt):
2.772 4.
Nameplate Rating (Gross HHe):
963 I
5.
Design Electrical Rating (Net MHe):.
918 6.
Maximum Dependable Capacity (Gross MWe):
917 7.
Maximum Dependable Capacity (Net MHe):
873._
G.-
.If changes Orcur 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.. seasons For Restrictions, If Any:
NRC letter dated 12/26/85 m,-
This Month Yr-to-Date Cumulative
- 11.. Hours'in Reporting Period 720 4.343 106.968
- 12. Number of Hours Reactor Has Critical 0
0 54.322-13.
Reactor Reserve Shutdown Hours 0
0 10.30012_
- 14. Hours Generator On-Line 0
0 50.363.8 L15. Unit Reserve. Shutdown Hours 0
0 1.-210.2
- 16. - Gross Thermal. Energy Generated (MHH) 0 0
127.861.688
- 17. Gross Electrical. Energy Generated (MHH) 0 0
41.523.187
- 18. Net Electrical Energy Generated (MHH)
-4146
-25373 38.263.412
- 19. Unit Service Factor.
0.0%
0.0%
47.1%
.20.
Unit Availability Factor
-0.0%
0.0%
48.2%
- 21. Unit Capacity Factor (Using MDC Net) 0.0%
0.0%
41.0%
- 22. Unit Capacity Factor (Using DER. Net) 0.0%
0.0%
39.0%
- 23. Unit Forced Outage Rate.
100.0%
100.0%
40.5%
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:.
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 l
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$)SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT C P. O. Box 15830, Sacramento CA 95852 1830,(916) 452-3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA GCA 87-345 gc July 15, 1987 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 DOCK 3T NO. 50-312
Dear Mr. Martin:
Enclosed is the June 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,
/n/16p2 G.
Ca Andtgnini Chief' Executive Officer, Nuclear Encl (5) cc:
I&E Wash (12)
MIPC (2)
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W N1 Ilh,7::: 8 RANCHO SECO NUCLEAR GENERATING STATION O 14440 Twin Cities Road, Herald, CA 95638 9799;(209) 333 2935