ML18152A529
| ML18152A529 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 12/31/1990 |
| From: | Stewart W, Warren L VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 91-009, 91-9, NUDOCS 9101230205 | |
| Download: ML18152A529 (20) | |
Text
l.j I
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VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 232 61 January 15, 1991 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 MONTHLY OPERATING REPORT Serial No.
NO/RPC:vlh Docket Nos.
License Nos.91-009 50-280 50-281 DPR-32 DPR-37 Enclosed is the Monthly Operating Report for Surry Power Station Units 1 and 2 for the month of December 1990.
Enclosure cc:
U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.
Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Ho Hand NRC Senior Resident Inspector Surry Power Station r-
. 9 :I. 012::::0205 9012:::: 1 PDR ADOCK 05000280 R
.e VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION MONTHLY OPERATING REPORT REPORT II 90-12 APPROVED:
POW 34-04
e TABLE OF CONTENTS SECTION Operating Data Report - Unit No. 1 Operating Data Report - Unit No. 2 Unit Shutdowns and Power Reductions - Unit No. 1 Unit Shutdowns and Power Reductions - Unit No. 2 Average Daily Unit Power Level
- Unit No. 1 Average Daily Unit Power Level
- Unit No. 2 Summary of Operating Experience - Unit No. 1 Summary of Operating Experience - Unit No. 2 Facility Changes That Did Not Require NRC Approval Procedure or Method of Operation Changes that Did Not Require NRC Approval Tests and Experiments That Did Not Require NRC Approval Chemistry Report Fuel Handling - Unit No. 1 Fuel Handling - Unit No. 2 Description of Periodic Test(s) Which Were Not Completed Within the Time Limits Specified in Technical Specifications PAGE 1
2 3
4 5
6 7
8 9
13 14 15 16 16 17
i I
I e
e OPERATING DATA REPORT DOCKET NO.:
50-280 DATE:
01/05/91 COMPLETED BY:
L.A. Warren TELEPHONE:
(804)357-3184 x355 OPERATING STATUS NOTES
- 1. Unit Name:
Surry Unit 1
- 2. Reporting Period: Dec. 01-31, 1990
- 3. Licensed Thermal Power (MWt):2441
- 4. Nameplate Rating (Gross MWe):847.5
- 5. Design Electrical Rating (Net MWe):
788
- 6. Maximum Dependable Capacity (Gross MWe):
820
- 7. Maximum Dependable Capacity (Net MWe):
781
- 8. If Changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report, Give Reasons: ----------------------------
- 9. Power Level To Which Restricted, If Any (Net MWe): --------------
- 10. Reason For Restrictions, If Any: ----------------------
THIS MONTH YTD CUMULATIVE
- 11. Hours In Reporting Period 744.0 8760.0 158016.0
- 12. Number of Hours Reactor Was Critical 336.2 6723. 4 99474.2
- 13. Reactor Reserve Shutdown Hours 0
0 3774.5
- 14. Hours Generator On-Line 283.5 6657.0 97480.2
- 15. Unit Reserve Shutdown Hours 0
0 3736.2
- 16. Gross Thermal Energy Generated (MWH) 614311. 8 15103107. 3 226219910.3
- 17. Gross Electrical Energy Generated (MWH) 205315.0 5031420.0 73576823.0
- 18. Net Electrical Energy Generated (MWH) 195048.0 4772199.0 69783129.0
- 19. Unit Service Factor 38.1%
76%
61.7%
- 20. Unit Availability Factor 38.1%
76%
64.1%
- 21. Unit Capacity Factor (Using MDC Net) 33.6%
69.8%
57%
- 22. Unit Capacity Factor (Using DER Net) 33.3 69.1%
56.0%
- 23. Unit Forced Outage Rate 0%
4.4%
20.6%
- 24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):
- 25. If Shut Down at End of Report Period Estimated Date of Startup: -------
- 26. Unit In Test Status (Prior to Commercial Operation):
FORECAST INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 1
ACHIEVED
OPERATING DATA REPORT DOCKET NO.:
50-281 DATE:
01/05/91 COMPLETED BY:
L.A. Warren TELEPHONE:
(804)357-3184 x355 OPERATING STATUS NOTES
- 1. Unit Name:
Surry Unit 2
- 2. Reporting Period: Dec. 01-31, 1990
- 3. Licensed Thermal Power (MWt):2441
- 4. Nameplate Rating (Gross MWe):847.5
- 5. Design Electrical Rating (Net MWe):
788
- 6. Maximum Dependable Capacity (Gross MWe):
820
- 7. Maximum Dependable Capacity (Net MWe):
781
- 8. If Changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report, Give Reasons: ---------------------------
- 9. Power Level To Which Restricted, If Any (Net MWe): --------------
- 10. Reason For Restrictions, If Any: ----------------------
THIS MONTH YTD CUMULATIVE
- 11. Hours In Reporting Period 744.0 8760.0 154896. 0
- 12. Number of Hours Reactor Was Critical 744.0 7973.7 99172.3
- 13. Reactor Reserve Shutdown Hours 0
0 328.1
- 14. Hours Generator Ori-Line 716. 9 7921. 7 97570.6
- 15. Unit Reserve Shutdown Hours 0
0 0
- 16. Gross Thermal Energy Generated (MWH) 1452121.4 18525133. 5 228135468.3
- 17. Gross Electrical Energy Generated (MWH) 486700.0 6149515.0 74230114.0
- 18. Net Electrical Energy Generated (MWH) 459307.0 5837766.0 70378725.0
- 19. Unit Service Factor 96.4%
90.4%
63%
- 20. Unit Availability Factor 96.4%
90.4%
63%
- 21. Unit Capacity Factor (Using MDC Net) 79%
85.3%
58.3%
- 22. Unit Capacity Factor (Using DER Net) 78.3%
84.6%
57.7%
- 23. Unit Forced Outage Rate 3.6%
9.6%
15.2%
- 24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):
Refueling Shutdown, 04/05/91, 67 days
- 25. If Shut Down at End of Report Period Estimated Date of Startup: -------
- 26. Unit In Test Status (Prior to Commercial Operation):
FORECAST INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 2
ACHIEVED
DATE TYPE(l) 901201 s
F:
S:
(1)
Forced Scheduled UNIT SHUTDOWN AND POWER REDUCTION (Equal To or Greater Than 20%)
REPORT MONTH:
DECEMBER 1990 DOCKET NO.:
50-280 UNIT NAME: _S_u_r_ry_U_n_1_* t_O_n_e ____ -
DATE:
01/05/91 CO MP LE TED BY:
L.A. Warren TELEPHONE: --=-ao_4,__-3--5=-7_-3--1-=-8~4-x-3--5 __ 5_
DURATION (HOURS)
METHOD OF SHUTTING REASON(2) DOWN REACTOR(3)
LICENSEE EVENT REPORTI SYSTEM CODE(4)
COMPONENT CAUSE & CORRECTIVE ACTION TO CODE(5)
PREVENT RECURRENCE 460.5 C
4 N/A N/A N/A (2)
REASON:
A - Equipment Failure (Explain)
B - Maintenance or Test C - Refueling D - Regulatory Restriction E - Operator Training & License Examination F - Administrative G - Operational Error (Explain)
H - Other (Explain)
(3)
METHOD:
1 - Manual 2 - Manual Scram..
3 - Automatic Scram.
4 - Other (Explain) 3 Unit shutdown for refueling continued.
(Refueling outage began on 10-06-90).
Returned to online status 12-20-90 at 0431 hours0.00499 days <br />0.12 hours <br />7.126323e-4 weeks <br />1.639955e-4 months <br />.
(4)
Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report (LER)
File (NUREG 0161)
(5)
Exhibit 1 - Same Source e
e
DATE TYPE(l) 901221 s
901223 F
901228 F
(1)
F:
S:
Forced Scheduled UNIT SHUTDOWN AND POWER REDUCTION (Equal To or Greater Than 20%)
REPORT MONTH:
DECEMBER 1990 DOCKET NO.:
50-281 UNIT NAME: --=s-u_r_ry_lJ=n-1-=-* t_Tw_o __
DATE:
01/05/91 COMPLETED BY: ---L-.-A-. _W_a_r_r_e_n ___ _
TELEPHONE:
804-357-3184 x355 DURATION (HOURS)
METHOD OF SHUTTING REA.SO:N(2) DOWN REACTOR(3)
LICENSEE EVENT REPORT#
SYSTEM CODE(4)
COMPONENT CAUSE & CORRECTIVE ACTION TO CODE(5)
PREVENT RECURRENCE 0
0 27.1 (2)
REASON:
B 4
A 4
A 4
A - Equipment Failure (Explain)
B - Maintenance or Test C - Refueling D - Regulatory Restriction N/A N/A N/A E - Operator Training & License Examination F - Administrative G - Operational Error (Explain)
H - Other (Explain)
TA EL EL (3)
METHOD:
1 - Manual V
BDUC BDUC 2 - Manual Scram.
3 - Automatic Scram.
4 - Other (Explain) 4 Ramped Unit to 79.5% power, 670.e in order to perform 2-PT-1 (Turbine Governor Valve Freedom Test).
Ramped Unit from 90% to 54% to reduce heat load on isolated phase bus duct.
Failure of ground straps on bus duct created elevated temperatures on the bus duct.
Ramped Unit down and opened the generator output breakers in order to replace failed ground straps.
e (4)
Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report (I.ER)
File (NUREG 0161)
(5)
Exhibit 1 - Same Source
e AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.:
50-280 UNIT NAME:
Surry Unit 1 DATE:
01/05/91 COMPLETED BY:
L.A. Warren TELEPHONE:(804)357-3184 x355 MONTH:
DECEMBER 1990 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
20 199 5
0 21 428 6
0 22 522 7
0 23 736 8
0 24 778 9
0 25 786 10 0
26 788 11 0
27 788 12 0
28 783 13 0
29 788 14 0
30 784 15 0
31 783 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.
5 J
AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.:
50-281 UNIT NAME:
Surry Unit 2 DATE:
01/05/91 COMPLETED BY:
L.A. Warren TELEPHONE:(804)357-3184 x355 MONTH:
DECEMBER 1990 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MWe-Net) 1 700 17 697 2
684 18 698 3
695 19 696 4
694 20 698 5
696 21 675 6
699 22 694 7
699 23 584 8
699 24 403 9
699 25 406 10 699 26 409 11 698 27 408 12 698 28 363 13 699 29 0
14 699 30 602 15 697 31 697 16 699 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.
6
SUMMARY
OF OPERATING EXPERIENCE MONTH/YEAR:
DECEMBER 1990 Listed below in chronological sequence by unit is experiences for this month which required load significant non-load related incidents.
UNIT ONE a
summary of operating reductions or resulted in 12/01/90 0000 This reporting period started with the Unit at CSD due to the ongoing Refueling Outage.
12/17/90 2213 Commenced Reactor start up.
2347 Reactor was critical.
12/18/90 0140 Commenced Low Power Physics Testing.
1837 Low Power Physics Testing completed.
12/19/90 0312 Reactor was at 2% power and holding.
12/20/90 0345 Started ramp up; 2% power.
0431 Unit on line.
12/23/90 1245 Unit at 100% power, 810 MWe.
12/31/90 2400 This reporting period ended with the Unit operating at 100%
power and 825 MWe.
7
SUMMARY
OF OPERATING EXPERIENCE MONTlll!/YEAR:
DECEMBER 1990 Listed below in chronological sequence by unit is experiences for this month which required load significant non-load related incidents.
a summary of operating reductions or resulted in UNIT TWO 12/01/90 0000 This reporting period started with the Unit operating at 90%
power and 740 MWe due to the inoperable M-12 Control Rod.
12/02/90 1027 The Unit experienced a Runback from 90% power and 740 MWe to 83% power and 700 MWe as a result of the loss of "F" Transfer Bus.
1710 Started ramp up; 84% power, 695 MWe.
1907 Stopped ramp; 90% power, 740 MWe.
12/21/90 0807 Started ramp down to perform 2-PT-29.1; 90% power, 740 MWe.
0918 Stopped ramp; 79.5% power, 670 MWe.
1400 Started ramp up; 79% power, 650 MWe.
1557 Stopped ramp; 90% power, 740 MWe.
12/23/90 1055 Started ramp down to reduce the circulating currents causing overheating in ground straps on the Isolated Phase Bus Duct; 90%
- power, 740 MWe.
The overheating was a result of ground strap failures.
1925 Stopped ramp; 54% power, 440 Mwe.
12/28/90 2007 Started ramp down to replace ground straps on Isolated Phase Bus Duct; 54.5% power, 440 MWe.
2241 Unit off line, Reactor maintained critical.
12/29/90 2338 Started power increase; 7
-6 x 10 amps on Intermediate Range.
2355 Reactor was at 2% power.
12/30/90 0147 Unit on line and ramp up continued.
1004 Stopped ramp; 90% power, 740 MWe.
12/31/90 2400 This reporting period ended with the Unit operating at 90%
power and 740 MWe.
8
DCP 88-01 EWR-89-713 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTl!l!/YEAR.:
DECEMBER 1990 INSIDE RECIRCULATION SPRAY PUMP FULL FLOW TEST LINE UNIT 1 This Design Change provides recirculation test piping and a
temporary sump to allow full flow testing of the IRS pumps.
The piping was designed to ANSI B31.l code requirements and includes flow and pressure instrumentation to meet ASME XI requirements for pump testing.
The recirculation test piping and temporary sump will be connected/installed during Unit shutdown conditions only.
This modification will not affect the operation of the safety-related IRS pumps.
It will not affect the design basis of the IRS system to mitigate LOCA or SLB accidents.
The modified IRS pump discharge piping will meet the original station design criteria.
This modification increases the reliability of the IRS pumps since the periodic testing will be used to assess the flow capacity and operational readiness of the pumps.
The only permanent change to the IRS system piping installed under this design change is the addition of a flanged spool piece in the pump discharge line.
This piping section has been designed to meet safety related and seismic requirements.
There are no valves installed in the pump discharge piping which could be misaligned (closed) and prevent initiation of spray flow.
Therefore, an unreviewed safety question is not created.
ENGINEERING WORK REQUEST UNITS 1&2 (Safety Evaluation #N89-0036) 12/01/90 The valve operator spring precompression for valve l/2-RC-PCV-1455/2455 A&B will be increased such that the valve will begin to operate at 40 PSIG versus the current setting of 38 PSIG.
The increase of valve operator spring precompression is required to prevent the valve from being forced opened due to Reactor Coolant Spray (RCS) pressure during PT-11 testing and to provide additional seating force margin during normal operations.
An unreviewed safety question does not exist because the design basis operation, function and level of integrity of the valves remains unchanged.
9
EWR-89-042 EWR-89-458 EWR-89-529 TM-Sl-90-052 e
e FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTI:ll/YEAR:
DECEMBER 1990 ENGINEERING WORK REQUEST UNIT 1 12/02/90 An electrical cable, l-CC-TC-114A, in the Component Cooling system was damaged and required replacement with a
different qualified cable type.
An existing cable, mark number NVB, is to be replaced with a new
- cable, mark number NGB-35.
Both cables are one pair, number 16 AWG, shielded.
Only the jacket material is different.
An unreviewed safety question is not created.
ENGINEERING WORK REQUEST UNIT 1&2 12/02/90 During recent testing of the Bottled Air
- System, the existing pressure control
- valves, l-VS-PCV-531/532, functioned erratically and did not properly provide the design flow rate of 300 cfm.
To assure that the design flow rate is achieved, new Fisher pressure control valves were installed as replacements.
The consequences or probability of an accident were not created due to the component replacement with safety related equipment which performs in accordance with design basis requirements.
ENGINEERING WORK REQUEST UNITS 1&2 12/03/90 This request was performed to reduce spurious alarms being received on the Chemical and Volume Control System (CVCS) heat trace annunciator in the control room.
CVCS heat trace annunciator under temperature backup control setpoints are being reviewed eliminate spurious alarms.
No changes are being existing equipment and the alarm will continue problems associated with the eves heat tracing.
safety question is not created.
TEMPORARY MODIFICATION (SE 90-288) setpoint and and reset to made to the to communicate An unreviewed 12-04-90 This temporary modification installed a test plug.
The Type 'C' test valve used for penetration leak testing of the 0-rings for an electrical penetration was broken off and temporarily replaced with a test plug until the valve was replaced.
Containment integrity continued to be met during this installation.
The test plug only replaces the valve which is used to perform the 0-ring leakage testing.
The plug does not affect the ability of the 0-ring/flanges to perform their pressure retaining functions.
An unreviewed safety question was not created.
10
EWR-88-587 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH:l!/YEAR:
DECEMBER 1990 ENGINEERING WORK REQUEST UNITS 1&2 12/06/90 This request provided generic guidance for Motor Operated Valve maintenance on spring packs.
The request also established a record of correct assembly of spring packs after disassembly and cleaning.
This request does not create an unreviewed safety question.
SAFETY EVALUATION (SE 90-289) 12-06-90 This safety evaluation analyzed the loading of known failed fuel assemblies in dry storage casks.
The evaluation examined the known effectiveness of vacuum drying to remove water from failed rods and the potential for release of radioactive isotopes from the fuel rod gap during the drying process.
As a related issue, the release of radioactive isotopes resulting from the failure of a
fuel rod during the vacuum drying of the cask cavity was evaluated.
An unreviewed safety question does not exist because the vacuum drying process for dry storage casks is effective in removing any water from the gap of the failed fuel rods.
In addition, gas isotopes will not be present in the gap of failed fuel rods at levels detectable during the vacuum drying process.
If a fuel rod should fail during vacuum drying, the release of Kr-85 would result in calculated doses of 0.36 rn.Rem (skin) and 0.004 mRem (whole body) at the Station Boundary.
AC-Sl-90-1207 ADMINISTRATIVE CONTROL (SE 90-290) 12-07-90 Administrative Control was established over certain normally closed and locked manual containment isolation valves while in an open position to allow supply of instrument air from the turbine building Instrument Air system.
Administrative control will be established to ensure that containment integrity can be maintained in accordance with the Technical Specification (TS) definition in TS 1.0.
Specifically, an operator will be stationed with instructions to close the valves immediately (within 30 seconds) upon notification from the control room.
In
- addition, the administrative control will ensure the ability to maintain the limits for containment partial pressure in accordance with the limits of TS 3.8.
Therefore, an unreviewed safety question is not created.
11
TM-Sl-90-59 TM-Sl-90-60 TM-Sl-90-61 e
FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONT.lffi/YEAR:
DECEMBER 1990 TEMPORARY MODIFICATION (SE 90-292A) 12-16-90 Electrical jumpers will be installed to provide continuity through the low flow reactor trip circuitry during replacement of relay FC-434XA.
The jumper is required to maintain train 'A' low flow circuitry in the energized condition while normal continuity is broken for relay replacement.
Replacement of relay FC-434XA using this jumper will affect only Train 'A' reactor trip circuitry.
- Further, the jumper will provide electrical continuity through the Train 'A' circuitry and failure of the jumper will result in a reactor trip.
Therefore, an unreviewed safety question is not created.
TEMPORARY MODIFICATION (SE 90-293A) 12-17-90 to provide an failed relay in the of the relay is reactor protection the jumper will be removed before This temporary modification is necessary electrical jumper to allow replacement of a reactor protection circuitry.
Replacement necessary to provide the third operable system circuit for RCS low flow.
Failure of result in a reactor trip.
The jumper will unit startup.
Therefore, an unreviewed safety created.
question is not TEMPORARY MODIFICATION (SE 90-293) 12-17-90 Electrical jumpers will be installed to provide continuity through the low flow reactor trip circuitry during replacement of relay FC-lXB.
The jumper is required to maintain train 'B' low flow circuitry in the energized condition while normal continuity is broken for relay replacement.
Replacement of relay FC-lXB is necessary to provide redundant relay failure protection for RCS Loop 'A', Train 'B' RPS low flow protection logic.
The jumper will maintain the RPS circuitry energized during relay replacement.
Failure of the jumper will result in a reactor trip.
Therefore, an unreviewed safety question is not created.
12
l-PT-18.3A PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:
DECEMBER 1990 SURVEILLANCE TEST PROCEDURE (SE 90-291) 12-09-90 This change involved the testing methodology employed for testing the Safety Injection (SI) system check valves to the RCS cold legs.
The change allowed the testing of the valves at full flow conditions with the RX head bolted.
The integrity of the RCS was the major issue evaluated.
RCS temperature effects were evaluated by the NSSS vendor (Westinghouse).
During the test 45 degree Fahrenheit water from the RWST will be injected into the RCS at a rate of 600 gpm at CSD conditions.
The evaluation concluded that the Over Pressure Mitigation System (OPMS) and Technical Specification requirements for OPMS operation (TS 3.1.G) can be relied upon to relieve RCS pressure during the test in combination with establishing an atmospheric vent for the RCS prior to the test.
Plant operators have instructions to terminate the test if pressurizer level exceeds 80% to avoid solid plant conditions.
Therefore, an unreviewed safety question is not created.
13
e TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTill/YEAR:
DECEMBER 1990 NONE DURING THIS REPORTING PERIOD 14
VIRGINIA POWER SURRY POWER STATION CHEMISTRY REPORT MONTil!/YEAR:
DECEMBER 1990 PRIMARY COOLANT ANALYSIS UNIT NO. 1 MAX.
MIN.
AVG.
Gross Radioact., µCi/ml 7.0lE-1 3.23E-4 9.25E-2 Suspended Solids, ppm 0.0 0.0 0.0 Gross Tritium, µCi/ml 4.82E-2 4.82E-2 4.82E-2 Iodine-131, pCi/ml 3.96E-5 2.60E-3
- 1. 71 Boron - 10, ppm*
445 178 329 Oxygen, (DO), ppm 3.000
- s;0,005 0.202 Chloride, ppm 0.014
<0.001 0.003 pH@ 25 degree Celsius 6.55 4.66 5.74
UNIT NO. 2 MAX.
MIN.
- 2. OlE-1
- 1. 27E-2 0.0 0.0 l.35E-l l.06E-l 2.66E-3
- 1. 41E-4 0.16 0.06 33.2 25.2 2.34 1.47 90 32
- s;0,005
- s;0,005 0.004
- s;0,001 7.31 6.07 Unit One:
Lithium concentration was out-of-spec (high) from 12/27/90 at 2020 hours0.0234 days <br />0.561 hours <br />0.00334 weeks <br />7.6861e-4 months <br /> until 12/28/90 at 1705 hours0.0197 days <br />0.474 hours <br />0.00282 weeks <br />6.487525e-4 months <br />.
The Lithium was 2.38 ppm; the limit is 2.35 ppm.
The Lithium concentration was out of spec for 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> and 45 minutes.
15 AVG.
l.27E-l 0.0 l.19E-l 6.32E-4 0.10 27.6
- 1. 76 43
- s;0.005 0.002 7.17
UNIT 1&2 NEW OR SPENT FUEL SHIPMENT II N/A DATE STORED 12/19/90 e
FUEL HANDLiltiG JITTJMBER OF ASSEMBLIES ASSEMBLY ANSI PER SHIPMENT NUMBER NUMBER CASK CASTOR/
V/21 500.13 PIS E02 LM0074 E08 LM0078 E09 LM0076 ON3 LM06FO ON7 LM06FY ON9 LM06GD lNl LM06FZ 1N3 LM06GH INS LM06GO 1N6 LM06ES 1N8 LM06FX 1N9 LM06FK 2NO LM06G3 2Nl LM06FM 2N2 LM06EV 2N4 LM06F7 2N5 LM06G7 2N6 LM06GE 2N7 LM06FB 3N3 LM06FE 16 DATE: DECEMBER 1990 NEW OR SPENT INITIAL FUEL SHIPPING ENRICHMENT CASK ACTIVITY LEVEL 3.11 N/A 2.61 2.61 2.61 3.41 3.41 3.41 3.41 3.41 3.41 3.41 3.41 3.41 3.41 3.41 3.41 3.41 3.41 3.41 3.41 3.41
e DESCRIPTION OF PERIODIC TEST(S) WHICH WERE NOT COMPLETED WITHIN THE TIME LIMITS SPECIFIED IN TECHNICAL SPECIFICATIONS MONTH/YEAR:
DECEMBER 1990 NONE DURING THIS REPORTING PERIOD 17