ML20207D274
| ML20207D274 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 11/30/1986 |
| From: | Colombo R SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| RWC-86-749, NUDOCS 8612300364 | |
| Download: ML20207D274 (10) | |
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NOVEMBER 1986
SUMMARY
OF PLANT OPERATIONS The plant has been in cold shutdown for the entire month of November 1986.
The initial shutdown was due to the December 26, 1985 loss of ICS 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 final documentation package for the following facility change was completed in November 1986.
This change was subjected to the review and approval of the Plant Review Committee (PRC) and the Management Safety Review Committee (MSRC).
There were no documentation packages completed for procedure changes, or tests, or experiments during November 1986.
1.
All Brown-Boveri, Inc. over-voltage relays on all Class lE busses were modified with vendor supplied modification kits.
Relays 45902A, 450-FA and other spare relays were found failed.
Tests indicate relay chattering and incorrect operating targets.
Spare relays showed the same characteristics.
Spara 459-3A worked properly and upon examination, its wiring was different from the defective ones.
Properly designed replacements that do not have the inherent problems associated with the original relay have been installed using the vendor supplied kits.
Referring to USAR Section 8.2.3.1.1 "The adequacy of the on-site distribution of power" was analyzed and designed with proper over/under voltage protection.
The modification of the defective relays with vendor supplied modification kits will achieve this same goal.
It should be noted that the relay chattering and relay flag indication concerns did not affect the ability of the relay to trip on an under-voltage or over-voltage condition as specified in the USAR.
The vendor representative approved the modification kits for the subject relays.
Furthermore, initial tests of relays which were repaired at Rancho Seco indicate that chattering and relay flag problems did not recur.
(Log Number 748, Revision 0).
8612300364 B61130 PDR ADOCK 05000312 R
PDR ZEh 44 i
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2.-
The_ Technical Specification additions mandated by the NRC Safety Evaluation of June 29, 1984 were included into Proposed Amendment 115 and approved as Amendment 82.
These are administrative controls which will prevent or l
reduce the likelihood of a Low Temperature Over-Pressure (LTOP) event.
1 1
A dual setpoint was installed on the EMOV (RCS i
Pressurizer power operated relief valve).
The circuit breakers for the four High Pressure 4
Injection (HPI) motor operated valves shall be opened and tagged.
The Make-up Tank water level will be maintained less i
than the high level alarm at 86 inches.
The pressurizer water level will be maintained at or below the high level alarm of 220 inches at system pressures above 100 psig and less than the high-high level alarm of 275 inches for pressures less than or equal to 100 psig.
These restrictions will not be imposed during RCS draining and filling.
1 The core flood tank discharge valves will be closed and the circuit breaker for the motor operators are racked out before the RCS pressure is decreased to 600 psig.
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When LTOP is required, only one HPI pump will be i
operated except during the transition 'of pumps that j
will supply Reactor Coolant Pump seals and make-up flow for the RCS.
i With the short overlap of the dedicated make-up pump and the HPI pump, no significant increase in pressurizer l
level will occur.
Therefore, a basis for resolving this issue as not being an unreviewed safety l
question can be justified, in that the margin of safety as defined in the basis for the Technical Specifications is somewhat reduced by the results are clearly within all acceptable criteria specified i
in the Standard Review Plan. (Log Number 615, l
Revision 0).
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,w 3.
ECN R-0832 installs approximately 250 Inconel-600 tube sleeves (80 inches long) per steam generator (OTSG) in tubes which are not considered defective.
Defective tubes are defined as tubes which have indications of >40%
through wall degradation.
Defective tubes will be plugged, not sleeved.
The tube sleeves were installed in accordance with B&W report #BAW-1832P, Revision 1, "OTSG Mechanical Sleeve Qualification" and B&W supplemental letter 51-1164120-00, "OD Tolerance Justification of OTSG Mechanical Sleeves."
This ECN was done as a preventive measure against OTSG tube failure to minimize steam generator tube failure outages and minimize the contamination of the secondary plant.
The effects of sleeving are minimal and do not alter the accident analysis.
The accident analysis for Rancho Seco remains bounding. (Log Number 821, Revision 1).
4.
The 80-inch sleeve installed in OTSG - A tube 77-17 does not meet the acceptance criteria for sleeves and will be plugged.
This tube, A77-17, is in an area that requires stabilization of plugged tubes per B&W Document number 51-1138302-02.
The_80-inch sleeve with an acceptable upper roll expansion is considered by B&W to be a fully qualified stabilizer for any tube location per B&W Document number 51-1156266-00, " Flow Induced Vibration Qualification of 177 OTSG Remote (Sleeve) Stabilizer."
The particular case of tube A77-17 is addressed in B&W Document number 51-1164747 which recommends that tube A77-17 be plugged and the sleeve be left "as-is" since the sleeve is qualified as a stabilizer with the existing tube sheet expansion. (Log Number 841, Revision 0).
5.
During modification and upgrade of the Diesel Generator GEA panels, portions of the GEB control wiring were identified as inadequately protected from a short circuit perspective.
This modification improves the ability of the Diesel Generator GEB to be restarted after an inadvertent control cabinet short circuit.
The Diesel has always been capable of reliable operation absent a chort circuit in the control cabinet.
Now fuses and circuit monitoring relays are installed, and undersized wires are replaced in portions of Diesel Generator GEB control wiring.
The reliability of Emergency Diesel Generator GEB is improved by this modification.
The new cables provide adequate ampacity.
The new fuses protect control circuits from short circuits during normal and transient operation.
The new monitoring relays permit improved identification of faults.
No new safety issues or failure modes were introduced.
No adverse safety impact was identified.
(Log Number 848, Revision 1).
6.
Penetrations (#60 and #67) for the Hydrogen Recombiner were installed in accordance with ASME B&PV Code, Section 1
XI, 1974 addenda through Summer of 1975.
The District is
)
- not required to, nor-authorized to apply the ASME Code
)
symbol to completed piping.
The flued heads for the two penetrations were machined from 7-inch thick A-516 plate material.
The ASME Code does not restrict the use of SA-516 to plate.
Reference to plate by the Code is only to designate the original product form and does not restrict the final use of the material to plate.
SA-516 material is superior in some respects to SA-105, but not all, since the through-thickness properties of the plate material are unknown and probably inferior to SA-105 for butt welding in the orientation used for these heads.
The configuration of the new penetrations is identical to the original penetration assemblies described in the USAR.
Current ASME Section III allows an exemption from impact testing for material less than 5/8" thick.
A drawing was available for dimensional checks.
Certified Materials Properties Test Report traceable to the part is also available. The raw product form was given single pass UT examination.
The machining process, by cutting across the cross section of the material along with visual examination have revealed no delaminations or other flaws.
Section 5.6 of the USAR is a general description of the a
original construction of the plant..It is not meant as a standard for material procurement, but as a historical guide to be used with present standards and practices for material procurement and constriction.
Section 5.2.3.1.2 1
does not describe all flued head penetrations as forged but states, "In the majority of cases..."
The new penetrations were found to maintain the margin of i
safety with respect to the USAR and did not involve a change to the Technical Specifications. (Log Number 421, Revision 0).
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MAJOR ITEMS OF SAFETY REIATED MAINTENANCE 1.
The Restart program schedule is being changed to reflect the corrective actions required as a result of pressurizer heater bundles damaged during the filling and venting of the RCS in November. The heater bundles damaged in November will be replaced prior to the start of the Decay Heat Outage.
Venting and pump sweeps as well as proceeding with the Decay Heat System temperature rise test (STP-1011) are pending reviews by the NRC.
2.
Vertical tubing installation in the Reactor Building for EFIC continues for the "A" train.
3.
The refurbishment of 163 motor operated valves is in
.p progress.
4.
The "100-hour" run of the Trans America Delaval (TDI)
Diesel Generator "A" was completed sucessfully.
The mini-teardown inspection is now in progress.
5.
Installed an new expansion joint on the Nuclear Service Raw Water Circulating Pump "B"
(P-476B) discharge line at the east spray pond.
In addition, a support bracket was added for pipe alignment (Work Request #123189).
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REFUELING INFORMATION REQUEST 1.
Name of facility Rancho Seco Unit 1 2.
Scheduled date for next refueling shutdown:
09/15/88 3.
Scheduled date for restart following refueling:
01/15/89 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:
03/15/88 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:
12/03/2001 l
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n-o AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
50-312 UNIT Rancho Seco Unit 1 DATE 11/30/86 COMPLETED BY R. Colombo TELEPHONE (916) 452-3211 MONTH November 1986 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MWe-Net) 1 0
17 0
2 0
18 0
3 0
19 0
4 0
20 0
5 0
21 0
6 0
22 0
7 0
23 0
8 0
24 0
9 0
25 0
10 0
26 0
11 0
27 0
12 0
28 0
13 0
29 0
14 0
30 0
15 0
31 16 0
INSTRUCTIONS On this format, list the average daily unit power level in MWe-Net for each day in the reporting month.
Compute to the nearest whole megawatt.
OPERATING DATA REPORT DOCKET NO.
50-312 l
DATE 11/30/86 i
)
COMPLETED BY R. Colombo r
TELEPHONE (916) 452-3211 OPERATING STATUS 1.
Unit Name:
Rancho Seco Unit 1 Notes 2.
Reporting Period:
November 1986 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):
0 10.
Reasons For Restrictions, If Any:
NRC letter dated 12/26/85 This Month Yr-to-Date Cumulative
- 11. Hours in Reporting Period 720 8.016 101.881
- 12. Number of Hours Reactor Was Critical 0
0 54.322
- 13. Reactor Reserve Shutdown Hours 0
0 10.300.2 14.
Hours Generator On-Line 0
0 50.363.8 15.
Unit Reserve Shutdown Hours 0
0 1.210.2
- 16. Gross 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)
-3.694
-37.766 38.292.336
- 19. Unit Service Factor 0.0%
0.0%
49.4%
- 20. Unit Availability Factor 0.0%
0.0%
50.6%
l 21.
Unit Capacity Factor (Using MDC Net) 0.0%
0.0%
43.1%
- 22. Unit Capacity Factor (Using DER Net) 0.0%
0.0%
40.9%
23.
Unit Forced Outage Rate 100.0%
100.0%
36.9%
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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'l 50-312 UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO.
UNIT NAME Rancho Seco DATE 11/30/86 November 1986 COMPl.ETED BY p rninmhn REPORT MONTH TELEPHONE (916) 452-3211 "k
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7.s W Licensee g't, c.
Cause & Corrective
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Date g
gg g
,g g 5 Event u?
E3 Action to
$3 5
jiji j Report #
N0 V
Prevent Recurrence 6
1 85/12/26 F
720 A
3 85-25 CB Instru Rx trip following loss of ICS Power I
2 3
4 F I'orced Reason:
Method:
Exhibit G Instructions S Scheduled A Equipment Failure (Explain) 1-Manual for Preparation of Data B-Maintenance of Test 2-Manual Scram.
Entry Sheets for Licensee C-Refueling 3-Automatic Scram.
Event Report (LER) File INUREG-D Regulatory Restriction 4-O:her (Explain) 0161)
E Operator Training & Ucense Examination F-Administrative 5
G-Operational Error (Explain)
Exhibit 1 Same Source l'#17 7 )
Il-Other (Explain)
s SMUD g/ SACRAMENTO MUNICIPAL UTILITY DISTRICT O P. O. Box 15830, Sacrament 95852-1830,(916) 452-3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA m8 RWC 86-749 g
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December 15, 1986 g
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J B MARTIN REGIONAL ADMINISTRATOR REGION V OFFICE OF INSPECTION AND ENFORCEMENT U S MUCLEliR REGULATORY COMMISSION 1450 MARIA LANE, SUITE 210 WALNUT CREEK, CA 94596 OPERATING PLANT STATUS REPORT DOCKET NO. 50-312 Enclosed is the November 1986 Monthly Operating Plant Status Report for the Rancho Seco Nuclear Generating Station pursuant to Technical Specification 6.9.3.
R. W.
COLOMBO REGULATORY COMPLIANCE SUPERINTENDENT Encl (5) cc:
I&E Wash (12)
MIPC (2)
INPO i
pm RANCHO SECO NUCLEAR GENERATING STATION L 14440 Twin Cities Road, Herald, CA 95638-9799;(209) 333 2935 l/
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