ML20128E421
| ML20128E421 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 04/30/1985 |
| From: | Colombo R, Reinaldo Rodriguez SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| RJR-85-243, NUDOCS 8505290382 | |
| Download: ML20128E421 (9) | |
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APRIL 1985 SUlHARY OF PLANT OPERATIONS-
.The plant has_been.in cold shutdown for refueling and plant modifications
-for the entire month of April.
' PERSONNEL CHANGES REQUIRING REPORT None~.
SUMMARY
-0F CHANGES IN ACCORDANCE WITH 10 CFR 50.59(b)
The following. facility changes ~were completed in April:
)1)
The allowable leakage from the Decay Heat Removal and Reactor Building Spray Systems was increased from 0.63 gph to 6.0 gph. This change resulted from the 0.63 gph' limit not being attainable and is consistent with the Standard Technical Specifications. The dose resulting from a maximum hypo-thetical accident (MHA) occurring at the 6.0 gph limit was calculated to be 7.21 rem thyroid and <0.01 rem whole body, which is still orders of magnitude.
less than the-10 CFR 100.ll(a)1 doseilimits.
This change required revisions to Technical Specification 4.5.3, FSAR Section 14.3.9.3, and surveillance procedure SP 204.07,LSP 201.12, and SP 203.09.
2)
A change was made to allow the pulling of Class-1 cables without the cable number being painted on the' jackets in the-associated protection channel color. code, as' described by USAR paragraph'7.1.4. The change permits the Class I cable jackets to be striped in the appropriate color and cable markers provided at'the ends of the circuit, which is the general industrial practice.
No Technical-Specification revisions or unreviewed safety questions.resulted
- from this change.
- 3) 5 Four.(4) out of one hundred and eight (108) lower core barrel bolts
^were selectively removed for testing purposes. These bolts will not be re-
.placed because of the extreme difficulty and expense involved..An evaluation determined that operation with the remaining 104. lower core barrel bolts is an acceptable configuration in accordance with the ASME Boiler and Pressure Vessel. Code., Calculations demonstrated that there was no significant increase in bolt loading from the removal of the four bolts.~ No Technical Specifications revisio'.or unreviewed safety question resulted from this change.
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.4)
Remote operating capability from the control. room'was added to valves DHS.001.and DHS 002'and the. valves were re-designated HV-20005 and HV-20004,
- respectively. Operation of these valves is needed to achieve cold shutdown of the plant and, under some. accident conditions,: high radiation could make sthese valves inaccessible. The remote operating capability was accomplished by replacing the existing spur gear manual operators on the valves.with Class 1E motor operators actuated from the control room. The motor operators fail "as is" on. loss of power. ' Both motor operators will be supplied from separate
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- Class lE-power-sources and no single failure will cause. failure of both valves. This change-did not involve a change in the Technical Specifications
.or an unreviewed safety question.
5)-
Procedure SP 207.01 (Inservice Inspection) was revised to approve the use of two (2) vendor eddy current inspection procedures at Rancho Seco. The vendor procedures are ISI-410.(Eddy Current Examination of Condenser Tubing, Rev. 5) and ISI-460 (Technical-Procedure for the Evaluation of Eddy Current Data of Nuclear Grade Steam Generator Tubing, Rev. 10). These procedures allow the. user to take advantage of new techniques and equipment related to Eddy Current Testing.
The procedures were reviewed by a qualified Level III technician from the vendor'_s staff ~as well as a qualified Level III member of the District's staff to ensure that applicable codes and standards were met.
6)-
Radiation monitors'(R-15056, R-15058, R-15059, R-15060, R-15061, and R-15062) which were installed on an interim basis to meet the requirements of NUREG-0578 2.1.8.b, were removed from service.
The functions of-these in-1struments were to monitor radiation levels at the reactor building vent,
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auxiliary building vent, main steam line A, and main steam line B following
-an accident. A new, permanent system has been installed to perform the-
. functions previously performed by the interim monitors.
c7)
Class lE power cables were installed between the Class lE power panels and the multiplexers for the Interim Data Acquisition and Display System (IDADS). The IDADS is an interim system designed to implement the essential fe.atures'of NUREG 0696 until the Plant Integrated Computer' System (PICS) is operable. This change does not introduce any failure modes that have not been analyzed previously.
8)
Cycle 6 changes defined by the B&W Cycle 6 Fuel Cycle Design Report and Cycle ~6 Reload Report are:
a) Forty of the assemblies will incorporate axial blankets (Mark BAB) in.
the upper and lower 6" of the assemblies. This is an extension of the-
. Lead _ Test Assembly (LTA) irradiations initiated in Cycle 5.
b). The BPRA's installed in 40 Batch 8B assemblies and 8 of the Batch 8A assemblies will incorporate asymmetric poison stacks (117").
Sixteen of the BPRA's will contain BPRA's of the original design (126").
c)- The Batch 9B assemblies will incorporate the Mark B5 upper end fittings.
d) The Fuel Cycle will be designed with " gray" APSR's, i.e., the poison section will be replaced with Inconel to reduce the reactivity-worth of the APSR's.
-e) The design' length of Cycle 6 will be 345110 EFPD's.
The overall design and safety analyses are provided in the Cycle 6 reload report.
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3 9)
- Two out of three logic circuits were installed to initiate the
-nuclear.' service bus unloading and loading schemes on bus 4A and 4B on overvoltage/undervoltage conditions. This change was made to meet the NRC staff " Position 1: Second Level of Under or Overprotection With a Time Delay" from Enclosure 1 to the NRC's June 3, 1977. letter to the District (Reid to Mattimoe).
-10) -The auxiliary feed pump turbine casing drain configuration was modified,.
per vendor recommendation, to ensure proper, efficient operation. The high and low pressure stage drains were headered together to a common steam trap.
The low pressure drain should be open to atmosphere or exhaust with no devices creating backpressure located downstream. Proper low pressure casing drainage will' prevent excessive condensate buildup and possible turbine
. wheel tip corrosion. The modification, in effect, ensured that the low pres-sure drain will be open to atmosphere.
- 11) The Axial Power Shaping Rods (APSR) were replaced with newly designed APSR's.for' Fuel Cycle 6.
The-redesigned APSR's improved. resistance to creep and~ core power distribution characteristics. The previous design was found to be subject to irradiation induced creep mechanisms which could lead to failure or collapse of-the APSR cladding. The improved power distribu-tion characteristics were accomplished by replacing the former Ag-Cd-In poison section with Inconel. The change in operating characteristics was incorporated into operator training prior to and during Fuel Cycle 6.
-12) An overflow line from the Miscellaneous Water Hold Up Tank connecting to
,the overflow line from the Borated Water Storage Tank to the Reactor Coolant System Drain Tank was added to prevent unintentional releases of contaminated
' water to offsite. The overflow line piping is in accordance with piping specification HD and ANSI B31.1 Power Piping.
,13) Channelized power was provided to the Non-Nuclear Instrumentation (NNI) signal conversion cabinets so that loss of a single _ power supply will not cause loss of critical signals to NNI X and Y channels. This change was accom-plished by providing AC power for the non-lE portion of the H4SCA Signal Conversion cabinet from a source separate from the source supplying AC power to the non-lE portion of H4SCB.
Both AC sources are battery backed,-but are not QA Class 1.
-.14 ) Plant egress was improved by modifications to 'he' Personnel Access t
Portal.(PAP) Building Area. These changes included the relocation of the existing double door exit and modification of the badge-return window cnute.
-Also',-the personnel orientation room was enlarged to allow larger groups to be trained and badged during plant outage preparation. The Security Plan was.
L revised to reflect the changes. The changes will not reduce the effectiveness of the Security Plan.
MAJOR ITEMS OF SAFETY RELATED MAINTENANCE
- 1) The Diesel Generator Air Start Compressor Discharge Line check valve
.(EGS-021) was found to be leaking and thereby keeping the compressor loaded.
The valve was disassembled and the disc was damaged beyond repair. The disc, nut _pi.n, and gasket were replaced and the valve put back into service.
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4 E2) Primary coolantLwa's detected seeping from temperature element'TE-21034.
- Verification.that:the. leakage wasinot coming from the welded area'was made.
1 Replacementof the flexitallic: gasket stopped the leakage.
3)i'TheRBSumpDIdolationValve(SFV-66309)wasfoundtobeleakingapproximately
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88.0.to'108.6.sccm past its seat. The valve was" disassembled, new Teflon
. seat:and gasket. installed, and reassembled and stroked to verify proper
. operation.
L4).The three1(3) Ton Bridge Crane motor froze in place when-the hoist was 7
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- overloaded during a surveillance procedure. The rotor ring was turned down
--o and the brushes were replaced to correct the problem. Following.this repair,.
ithe: crane ~was successfully functionally tested, s
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REFUELING INFORMATION 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 amendnent 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 i
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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
50-312 UNIT Rancho Seco Unit 1 DATE 04-30-85 COMPLETED BY R. Colombo TELEPHONE (916) 452-3211 MONTH April 1985~
DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(Mde-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.
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OPERATING DATA REPORT DOCKET NO.
50-312 DATE 04/30/85 COMPLETED BY R. Colombo TELEPHONE (916) 452-3211 OPERATING STATUS NOTE:
1.
Unit Name:
Rancho Seco Unit 1 2.
Reporting Period:
April 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-Date Cumulative
- 11. Hours in Reporting Period 719 2.879 87.984
- 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 11 7. Gross Electrical Energy Generated (MWH) 0 1.366.846 40.798.809 l
- 18. Net Electrical Energy Generated (MWH) 0 1.289.938 37.881.184
< - 19.. Unit Service factor 0.0%
56.21%
- 56.01%
- 20. Unit Availability Factor 0.0%
56.21%
57.39%
- 21. Unit Capacity factor (Using MDC Net) 0.0%
51.33%
49.32%
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- 22. Unit Capacity Factor (Using DER Net) 0.0%
48.81%
46.90%
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- 23. Unit forced Outage Rate 0.0%
8.8%
29.4%
- 24. Shutdowns Scheduled Over-Next 6 Months (Type, Date, and Duration of Each):
Refueling - 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 l
INITIAL ELECTRICITY N/A N/A COMMERCIAL OPERATION N/A N/A
- Corrected from March 1985 report.
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UNIT SIlUTDOWNS AND POWER REDUCTIONS DOCKET NO.
50-312 UNIT NAME RANCHO SECO UNIT l DATE.4/30/85 REPORT MONTil APRIL 1985 COMPLETEI)IlY R. Coinmhn TELErilONE (916)4591711 "b
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g.s -g Licensee
$*s, c.
Cause & Corrective
.E e
!E No.
Daic g
gg j!j
, g oc Event u7 83 Action in H
$E jd Report #
i# u ju Prevent Recurrence Q
5 4/1/85 S
719 C
1 N/A ZZ ZZZZZZ Shutdown for Refueling I
2 3
4 F: Forced Reason:
Method:
Exhibit G Instructions S: Scheduled A liquipment Failure (Explain) 1-Manual for Preparation of Data ll.Maintenante of Test 2-Manual Scram.
Entry Sheets for I.icensee C-Itcructing 3-Automatic Scram.
Event Report (LER) File (NUREG.
D. Regulatory itestriction 4-Oiher (Explain) 0161) li.Operstm Training A License Examination F Aelministrative 5
G Operational Error (Explain)
Exhibit I - Same Source (9/77)
Il-Other (Explain)
$ SMUD SACRAMENTO MUNICIPAL UTILITY 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 85-243 May 8, 1985 DIRECTOR OFFICE OF INSPECTION AND ENFORCEMENT U S NUCLEAR REGULATORY COMMISSION WASHINGTON DC 20555 OPERATING PLANT STATUS REPORT DOCKET N0. 50-312 Enclosed is the April 1985 Monthly Plant Status Report for Rancho Seco Unit One.
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R. J. Rodrig z Assistant Gen. al Manager, Nuclear cc:
I&E Washington (9)
Region V MIPC (2)
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