ML19347D929
| ML19347D929 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 04/01/1981 |
| From: | Tubbs R COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML19347D928 | List: |
| References | |
| NUDOCS 8104140489 | |
| Download: ML19347D929 (23) | |
Text
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QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT MARCH 1981 COMMONWEALTH EDISON COMPANY AND IOWA-ILLINOIS GAS & ELECTRIC COMPANY NRC DOCKET NOS. 50-254 AND 50-265 LICENSE NOS. DPR-29 AND DPR-30
'810,414 0 ft$,
TABLE OF CONTENTS 1.
Int.roduction 11.
Summary of Operating Experience A.
Unit One B.
Unit Two Ill. Plant or Procedure Changes, Tests, Experiments, and Safety Related Haintenance i
A.
Amendments to Facility License or Technical Specifications B.
Facility or Procedure Changes Requiring NRC Approval C.
Tests and Experiments Requiring NRC Approval D.
Corrective Maintenance of Safety Related Equipment IV.
Licensee Event Reports V.
Data Tabulations A.
Operating Data Report B.
Average Daily Unit Power Level C.
Unit Shutdowns and Power Reductions VI.
Unique Reporting Requi rements i
A.
Main Steam Relief Valve Operations B.
Control Rod Drive Scram Timing Data
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Vll. Refueling Information Vill. Glossary i
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INTRODUCTION Quad-Cities Nuclear Power Station is composed of two Boiling Water Reactors, each with a Maximum Dependable Capacity of 769 MWe net, located in Cordova, Illinois. The Station is jointly owned by Commonwealth Edison Company and lowa-Illinois Gas & Electric Company. The Nuclear Steam Supply Systems are General Electric Company Boiling Water Reactors. The Architect / Engineer was Sargent & Lundy, incorporated and the primary construction contractor was United Engineers & Constructors. The con-denser cooling method is a closed-cycle spray canal, and the Mississippi River is the condenser cooling water source. The plant is subject to license numbers DPR-29 and DPR-30, issued October 1, 1971, and March 21, 1972, respectively, pursuant to Docket Numbers 50-254 and 50-265 The date of initial reactor criticalities for Units I and 2 respectively were October 18, 1971, and April 26, 1972.
Commercial generation' of power began on February 18, 1973 for Uni t 1 and March 10, 1973 for Unit 2.
This report was compiled by Becky Brown and Robert Tubbs, telephone number 309-654-2241, extensions 245 and 174.
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II.
SUMMARY
OF OPERATING EXPERIENCE A.
UNIT ONE March I-9: Unit One began the reporting period in a continuation of the Maintenance Outage that began on February 28, 1981. Major items performed
'during this outage were:
repaired leaking valves in the drywell ("B" feed-water check, "A" recirc pump suction, "A" recirc crosstle isolation, and head vent isolation valves), various SJAE valves, and miscellaneous valves in other systems. Due to difficulties in repairing the feedwater check valve, criticality was not achieved until 1610 on March 3 The generator was put on line at 0459 on March 4.
One hour later, controi rod 34-27 (J-7) overtravelled. Subsequently, the rod was inserted and electrically disarmed.
At 0625, a turbine trip was received due to a moisture separator high level.
Rods were manually inserted to shut the reactor down.
During this outage, control rod drive 34-27 was changed, repairs were effected on the moistura separator drain tank and two electromatic relief valves had now pilot valves installed.
The reactor was brought critical again on February 5, at 0950, and at 1912 the generator was put on line. Load was then increased at various rates, including a twelve hour xenon soak at 430 MWe, until load was held at 816 MWe on March 9 March 10-12: Over this three day period load was held at an average of 810 MWe.
March 13-16: On March 13 the circuit breaker for "A" recirc pump discharge valve was tripping due to an unknown cause. Therefore, at 1500 load was reduced to 200 MWe, and a dryuell entry was made at 2130. The problem was isolated to the cable inside the drywell. The cable was replaced by a temporary cable until a permanent repai r can be ef fected during the next outage. At 0445, on March 14, power was increased at various rates until a load of 817 MWe was held on March 16.
March 17-20: An average load of 819 MWe was held during this four day perloo until 2030 on March 20.
At that time load was dropped at 200 MWa/
hour in preparation to trip the turbine off line.
March 21-23: On March 21, at 0040, the turbine was tripped off line.
However, the reactor remained in Hot Standby.
Repairs were then rade to a leak in an EHC oil supply line. Also worked, at this time was the packing
[
on the turbine stop valves. At 0610 the reactor was placed in RUN, and tha generator was placed on line at 0723. Load was increased at various rates until 815 MWe was held at 1000 on_harch 23
-March 24-28: During this five day period an average load of 810 MWe was held, until 2310 on March 28. At that time load was dropped at 100 MW / hour to 700 MWe.
e March 29-31: At 0010 load was held at 700 MWe to perform weekly turbJne tests and bi-weekly M5iV tests. During MSIV testing, 2A MSlV drifted past its test limit, initiating a Group I isolation and resultant reactor scram.
The problem was investigated and a faulty limit switch on the MSIV was 1
replaced. Also replaced a thermocouple on an electromatic relief valve and an accumulator on CRD 26-35 (G-10).
The reactor was brought critical at 1731, and on March 28, at 0233, the generator was synchronized.
Load then was increased at various rates, ending the reporting period at 660 MWe, increasing at 8 MWe/ hour.
B.
UNIT TWO
-March I-3: Unit Two began the reporting period dropping load at 100 MWe/
hour. At 0230 load was held at 400 MWe and the control rod pattern was changed. At 0430 load was increased using various rates until 817 MWe was held at 1800 on March 3 March 4-14: Load was held at an average of 794 MWe over this 11 day period, with the exception of March 8.
On that day the weekly turbine tests were performed resul ting in an average of 780' MWe.
March 15-18 On March 15, at 0015, load was dropped to 160 MWe-to perform scram timing on 89' control rods. The rod sequence was also changed at that time. At 1010 load uas increased untti March 18 at 0500 when a load of 800 MWe was held.
March 19-21: An average load of 800 MWe was held over this three day period.
i March 22-23: On March 22 load was dropped to 600 MWe to perform special rod moves for the Nuclear Engineer. Load was then increased at 5 MWe/ hour until 790 MWe was achieved and held at 0100 on March 24.
March 24-31: During this eight day period, load was held at an average of 788 MWe, except on March 28. On that day, there was an average load of 773 MWe, due to the weekly turbine tests being performed. The unit ended the reporting period holding a load of 796 MWe.
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PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED Ill.
MAINTENANCE Amendments to Facility License or Technical Specifications A.
Amendments 63 and 57 to Licenses DPR-29 and DPR-30, respectively On March 16, 1981, the NRC issued Amendment 63 to License DPR-29 and Amendment 57 to License DPR-30. These changes; (1) correct the minimum amount of diesel fire pump fuel oil in the day tank to be 150 gallons; (2) add the cable spreading room smoke detectors t.o Table 3 12-1; and (3) add the cable spreading room sprinkler system to Table 3 12-2.
Facility or Procedure Changes Requiring NRC Approval B.
There were no Facility or Procedure Changes requiring NRC approval for the reporting period.
C.
Tests and Experiments Requiring NRC Approval There were no Tests and Experiments requiring NRC approval for the reporting period.
D.
Corrective Maintenance of Safety Related Equipment The following represents a tabular summary of the safety related I
maintenance performed on Unit One and Unit Two during the reporting The headings indicated in this summary include: Work 4
period.
Request Numbers, LER Numbers, Components, Cause of Mal functions,
Results and Effects on Safe Operation, and Action Taken to Prevent Repetition.
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e-UNIT ONE MAIM ENANCE
SUMMARY
CAUSE RESULTS & ErfiCTS W.R.
LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVEl!T REPETIT107 Q11102 LPRM 16-49A 100 VDC Power LPRM went downscale; Tightened connection.
connector loose.
not affected.
Q09039 CRD J-11 "0" Rings & seals CRD drifts past "00";
Replaced drive, were worn.
Scram function not affected.
Q11366 1-203-3C Electro-Worn pilot seat &
Valve temperature Repaired pilot valva.
matic Relief Valve disc.
indication was high.
Valve was operab1c.
Q10791 1-8325 182 24/48 Loose capacitor Charger voltage was Soldered lead.
VDC Battery Charger Icad on voltage very erratic.
Battery regulator card.
system was operable, as were loads.
Q11353 81-6/03L l-203-3A Target Broken air hose The relief valve function Replaced stainless Rock Valve to operator.
of the Target Rock Valve steel hose.
was inoperable. Other relief valves were operable.
Q11399 M0-1-2301-3 HPCI Worn valve Steam leak from valve; Repacked valve.
Steam inlet Valve packing.
HPCI operability not af fected.
Q10289 l-2301-3 HPCI Steam Leakage!past the Steam was leaking past Overhauled valve.
Supply Valve valve disc & seat.
the valve to the drains.
Qll323 1-220-58B Feed-Leak from pressure Minor steam leakcge in Welded & machinec <;.;
water Check seal ring.
MSIV Room.
area--replaced seai ring.
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UNIT ONE MAINTENANCE
SUMMARY
o CAUSE RESULTS & EFFECTS W.R.
LER OF ON ACTION TAKEN TO NUMBER NUMBER COMP 0t!ENT MALFUNCTI0ll SAFE OPERAT!Cil PREVENT P.EPETITION Q11582 81-7/03L l-202-5A Recirc Grounded cable from LPCI loop. select Temporary cable run.
Pump Discharge breaker to valve inoperable.
Core Spray, Will replace during l
Diesels & Containment refuel outage.
Valve motor.
Cooling were operable.
Q11507 M0-1-1001-16B RHR Auxiliary contact The valve would not close; Replaced aux 111ary Heat Exchanger By-would nop pick up.
contacts.
pass Valve Q07718 LPRM 16-41A Went upscale.
LPRM was reading up-Replaced card--then scale; APRM scram, replaced connector under vessel.
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UNIT. Two _ MAINTENANCE
SUMMARY
CAUSE RESULTS & EFFECTS W.R.
LER OF ON ACTION TAY.EN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPET!TICH Q11304 2-595-125RRelay for The relay became The reactor cleanup Replaced burned out Cleanup System overheated and system was auto-relay.
Isolation shorted'put, isolated.
Q09944 LPRM 24-09A Faulty connector LPRM reads downscale; Replaced connector in LPRM power APRM RPS function not to power supply.
supply.,
a f fected.
Ql1583 5742-2A & B Worn seals in the The damper closed very R3placed air cylinder Reactor Building air cylinder.
slowly.
Inlet dampers seals.
closed satisfactorily Vent Dampers and SBGT auto-started.
Q11592 Rod Block Monitor Does not get input The LPRM input to the Found loose connector Ch. 7 from LPRM 08-33A.
RBM had to be bypassed, on 08-33 relay and Limiting C.R. pattern corrected.
did not exist.
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IV.
LICENSEE EVENT REPORTS The following is a tabular summary of all license event reports for Quad-Cities Units One and Two occurring during the reporting period, pursuant to the reportable occurrence reporting requirements as set forth in sections 6.6.B.1. and 6.6.B.2. of the Technical Specifications.
UNIT ONE Licensee Event Report Number Date Title of Occurrence 81-6/03L 03-02-81 Broken air line to Target Rock valve 81-7/03L 03-13-81 1A Reactor Recirc Pump Discharge valve inoperable UNIT TWO 03-12-81 Loss of Drywell to 81-6/03L
~-
Torus D/P 81-7/03L 03-22-81 No 1/2 scram on 2A HSIV while doing QOS 250-1, step 8 81-8/03L 03-26-81 Open fire stop RK20ll i
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DATA TABULATIONS f
The following data tabulations are presented in this report:
A.
Operating Data Report B.
Average Daily Unit Power Level C.
Unit Shutdowns and Power Reductions 1
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OPERATING DATA P.EPORT DOCKET NO, 50-25^
UNIT ONE DATE April I, 1981 COMPLETED BY R C Tubbs TELEPHONE 309-654-2241 ext. I'74 OPERATING STATUS 0000 030181
- 1. Reporting perlod:2400 03310i_Grons hours in reporting period:
744
- 2. Currently authorized power level (MWt): 2511 Mux. Depend capacity (MWe-Net): 769* Design electrical rating (MWe-Net): 789
- 3. Power level to which restricted (if any)(MWe-Net): NA 4.
Reasons for restriction (if any):
This Month Yr.to Date C';:, la t i v e S. Hunber of hours reactor was critical 638.4_
1991.0 6%697.3
- 6. Reactor reserve shutdown hours 0. 0_
00
____j 421,9 7.
Hours generator on line 597.4_
1937,r. _
320,8 D.
Unit reserve shutdown hours.
0. 0.
0.0
..,._ 909.2
_ 2';24004 9.
Gross thernal energy generated (MWH) 1297175 4381916
- 10. Gross electrical energy generated (MWH) 426202 144414?
3: 723049
- 11. Het electrical energy generated (MWH) 3R9384 134556'?
3/ '02848
- 12. Reactor service factor GS.8 92,?
_80.5
- 13. %cc*ar avoilobility factor 85,8 92.
84,0
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Unit.:.avice futtor 80,3 89.t.
__ _. 76.8
- 1. 5. l.;a i t
..r. u i l ob il i t y facter 80,3 09 '
77,9_
t '>. linit ccyyrity factor (lising MDC) 63.1
- 81. '-
60,4 7
Uni:
rt a r.c i t y inctor (Uuing Des.M Je)
_6 6. 3_
79. t(
50,9 Un - '< ced.; u t trie ret 2 9,5 3_. "
7.7 C:u
.in sch.vbled
..tr naxt 6 nnnths ( T y p e,, Du t e, an d Durati.m ch):
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.. :...A at and o'
.:.o r. r t p e r 19 d., e '., t in n t e tl date of stcrt"p
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OPERATING DATA REPORT DOCKET NO.
50-265 UNIT TUD DATE April I, 1981 R C Tubbs GOMPLETED BY TELEPHONE 309-654-2241 ext. 174 CPERATING STATUS 0000 030101
- 1. Reporting period:2400 033181 Gross hours in reporting period:
744
- 2. Currently authorized poue.c level (MWt): 2511 Max. Depend capacity tuc-Ne t ) : 769% Design electrical ruting (MWe-Net): 789
- 3. Power level to which restricted (if any)(MWe-Net): NA 4.
Reccons for ret,triction (if any):
This Month Yr.to Date Canulative Nutiber of hours reactor was'criticc1 744,0 2106.2 6)?39.0 Reactor reserve shutdoun hours 0.0 0.0
,,,,_ _ 2 9 8 5. 8 s.
!'ours generat or on'line 744.0 2090.i_
_ S~.371.3 Unit reserve chuTr ovn h oves.
0. 0_
0. It 702.9 Gross thernal energy generated (MWH) 1750994 4870044 12ni?0452 Gross electrical energy generated (MNH) 557615 1547898
_ ~jf2S9449
. i., Net electrical energy genernted(MWH) 532344 1463787
_ 7.~ 2 0 7 3 9_
Renctor service inctor 300.0_
97.5 79.1
.a.
F.euctor availubility fuctor 100.0 97.5 83.0 Unit service fuctor 100.0 96.3
,,,,, _ 75.8
.5.
Unit availabilitj f ac t..r 100.n_
96.3 76.7_
'. Unit copociey fuctor (Us:ng MDC) 93.U 88.i 60.5 Unit capacity fuctor (Uning Des.MWe) 90.7.
85.o 59.0 Vait forced o u t u.,p r a t :-
0_. 0 0.a 0.9 c
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Sh '! t d o un s r,ch.: d o l.' <i <. u. c next /,., o n t h, (T y p e, Du t e,, an d D o r o t.i..n o
. il shutdoun cT cnd of e e n.ir t permoc,< tinuted date of starte;,, q.
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APPENDIX B AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
50-254 UNIT ONE DATE April 1, 1981 LCM?LETED BY R C Tubbs 7:1:anegg 309-654-2241 ext. T74 510NT H.
Nw ch 190i DAY AVERAGE DAILY POWER LEVEL DM AVERAGE DAILY ?CLTt. /EL (i!We-Ne t )
(MWc-W t) i.
-33._2 17.
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-30.3 19, 7 6 6,. ',_
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-29.9 20.
710. ti_
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6.0 21.
279.7 6.
427.7 22.
SyhL 7.
569.2 23.
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6.
680.0 24, 729.3 9.
75?.4 25.
760.*>. _ _ _.,
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763.1 26.
759.._n,_,
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742.8 27.
750 12.
754.3 28, 7v'.
.; 3.
6C.7 29.
11.
4h.. 6 30.
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640.7 31.
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APPENDIX D j
AVERAGE DAILY UdIT POWER LEVEL DOCKET NO.
BI)-P 63 UNIT TWO DATE April 1, 1981 COMI'LE TED BY R C Tubbs TELEP HONE 309-654-2241 ext. 174 111c 5 1901
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MY AVERAGE DAILY POWER LEVEL DAY AVER AGE DAI: Y P0ttER LEVEL (hWe-i.'c t )
(Mue-Net) 1.
529,6 17.
607. ? _
2.
645,4 18, 7 4_4.J,
3.
734.i 17.
765.3,_
750.1 20, 765.4, g,
773.(1 21.
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766 t 22.
577.A_,
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74S 0 24.
755.4 752.9 25.
743.9_.
3 763 3 26.
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,.. ; -4 APPENDIX D QTP 300-S13 UNIT SHUTDO*4NS A!;D PGWER iiEDUCTIONS P.cvi s ; a s
DOCKET NO.
50-254 March i% 5 IT NN4E quad-Citics One COMPLETED BY
?. C d -
CATE Aprii 1, 1981 REPORT MONTH MARCH 1981 TELEPHONE 309-654-224I.
ext. 174 a
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CORRECTIVE ACTIONS / COMMENTS p 5 Ei REPORT NO.
(HOURS) u.
NO.
DATE R
81-3 810228 5
77 0 B
4 ZZ ZZZZZZ Continuation of Maintenance Outage 81-4 810304 F
36.8
'A 1
HB XXXXXX Turbine trip on high Moisture Separator Drci-tank level 81-5 810314 5
0.0 8
5 81-07 CB INSTRU Load reduction for drywell entry to iiwest:
repair. problem with circuit breaker on 1/s re..
discharge valve 81-6 810321 5
6.7 8
5 HA INSTRU Turbine tripped to repair leak in LitC
.y - t e:
81-7 810329 F
26.2 A
3 C6 INSTRU Reactor scram due to MSIV drifting 3.w; c e routine test II i
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i APPENDIX D QTP 30,0-r.13 UNIT SHUTDOW?;S AND POWER REDUCTIO!iS nr.vi::c~ 5
.. r ;--.a 50-765 harch iS..,to
- i :L*#,E Quad-Citles Two -
COMPLETED BY Ej ~' <
Apri1 1, 1981 REPORT MONTH MARCH 1981 TEl.EPHONE 309-654-2241, CATE e'x t. I Fe l
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CORRECTIVE ACTIONS / COMMENTS p5g REPORT NO.
(HOURS) e u.
NO.
DATE R
81-4 81030!
S 0.0 H
5 RB CONROD Load reduction to change control rod pattern RB CONROD' Load reduction to perform scram ti.nin0 ar.d ch...ce 8I-5 810315 S
o.0 8
5 control rod sequence 81-6 810322 S
0.0 H/B 5
RB CONROD Load reduction to perform special rod r. oves, turbine weekly, and reverse condenscr flow e
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UNIQUE REPORTING REQUIREMENTS The following items are included in this report based on prior commitments to the commission:
A.
MAIN STEAM RELIEF VALVE OPERATIONS Relief valve operations during the reporting period are summarized in the following table. The table includes information as to which relief valve was actuated, how it was actuated, and the circumstances resulting in its actuation.
VALVES NO. & (YPE PLANT UNIT DATE ACTUATED ACTUATIONS CONDITIONS DESCRIPTION OF EVENTS 1 03-03-81 1-203-3A 1 Manual RX PRESS Surveillance T.S.
1-203-38 i Manual 400 4.5 0.1.b.
1-203-3C 1 Manual 1-203-3D 1 Manual 1-203-3E 1 Manual 1 03-05-81 1-203-3D 1 Manual RX PRESS Post Maintenance 1-203-3E 1 Manual 400 (Replaced pilot solenoid valve)
B.
CONTROL ROD DRIVE SCRAM TIMING DATA FOR UNITS ONE AND TWO The basis for reporting this data to the Nuclear Regulatory Commission are specified in the surveillance requirements of Technical Specifications 4.3.C.1. and 4.3.C.2.
The following table is a complete summary of Units One and Two Control Rod Drive Scram Timing for the reporting period. All scram timing was per-formed with reactor pressure greater than 800 PSIG.
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RESULTS OF SCRAM TIMING MEASUREMENTS PERFORMED ON UNIT I c _2_ CONTROL ROD DRIVES, FROM l-1-81 TO 12-31-81 AVERAGE TIME IN SECONDS AT %
Max. Time INSERTED FROM FULLY ULTHDRAWN For 90%
Inr.artion DESCP.lPTION Technical Specification 3 3.C. l &
NUMBER 5
20 50 90 0E OF RODS 0.375 0.900 2.00 3.5 7 sec.
3.3.C.2 (Average Scram insertion Time' 3-3-il 1
0.26 0.49 1.0-1.76 Unit 1 Cold Scram Time J-7 (34-27)
P.od replaced--coupling probicm 3-6-!!
1 0.31 0.69 1.51 2.62 J-7 Hot 3 '.5-i.
89 0 32 0.69 1.46 2.57 2.91(F-12)
Unit 2 "B" Sequence Hot l
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Vll. REFUELING INFORMATION
-The following information about future reloads at Quad-Cities Station was requested in a January 26, 1978, licensing memorandum (78-24) from D. E.
O'Brien to C. Reed, et. al., titled "Dresden, Quad-Cities, and Zion Station--NRC Request for Refueling Information" dated January 18, 1978.
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QTP 300-s32 Revision 1 QUAD-CITIES REFUEllHG March 1978
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INFORMATION REQUEST
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Unit:
1 Reload:
6 Cycle:
7 9-12-82 (Shutdown EOC6) 2.
Scheduled date for next refueling shutdown:
12-5-82 (Startup BOC7) 3 Scheduled date for rest'rt following refueling:
4.
Will refueling or resumption of operation thereafter require a technical specification change or other license amendment:
No, Plan 10CFR50.59 reloads for future cycles of quad Cities Unit 1.
The review will be conducted in August, 1982.
Scheduled date(s) for submitting proposed itcensing action and supporting 5
informa tion: August, 1982 for 10CFR50.59 related changes rv 90 days prior to shutdown.
Important licensing considerations associated with refueling, e.g., ne': or
. 6. ' different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuci design, new operating procedures:
New fuel designs:
7 The number of fuel assemblies.
a.
Number of assemblics in core:
724 b.
Number of assemblics in spent fuel pool:
820 8.
The present licensed spent fuel pool storage capacity and the size of c:rf increase in licensed storage capacity that has been requested or is p!nn ud in number of fuel assemblies:
i Licensed storage capacity for spent fuel:
1460 i
a.
None b.
Planned increase in licensed storage:
9 The projected date of the last refueling that can be discharged to th.:
spent fuct paal assuaiing the prcr.cnt licensed capacity: S e p t em're r. 1 :' N (end of batch dischart;u capabiiit/)
W pi> n 0 V E D
~
(
i
QTP 300-532
)
Revision 1 March 1978 QUAD-C! TIES REFUELING f-INFOR!iAT!3N REQUEST
,\\
1.
Unit:
2 Reload:
5 Cycle:
6 l
8-30-81 (Shutdown EOCS) 2.
Scheduled date for next refueling shutdown:
t 12-20-C1 (Startup 80C6) 3 Scheduled date for restart following refueling:
L 4.
Will refueling or resumption of operation thereafter require a technical specification change or other license amendeent: No, Plan 10CFR50.59 Reloads for future cycles of Quad Cities Unit 2.
The review will be conducted by early August, 1981.
Scheduled date(s) for submitting pro,sosed licensing action and supporting
~
5
[
information: Early August, 1981 for 10CFR50.59 related changes es90 days g
prior to shutdown.
O f
6.
Important licensing considerations associated wi'th refueling, e.g., new or fuel design or supplier, unreviewed design or performance anal'ysis
'different methods, significant changes in fuel design, new operating procedures:
New Fuel Design:
1.
Barrier Fuel 2.
Control Cell Core u
.a i
'l 7
The number of f uel a.ssemblies.
a.
Number of assemblies in core:
724 f
s b.
Number of assemblies in spent fuct pool:
_672 1
licensed spent fuel pool storage capacity and the slac of,:w,*
8.
The present Increase in licensed storage capacity that has been requested or is pl...::d in number of fuel assemblies:
Licensed storage capacity for spent fuel:
1461 a.
b.
Pinnned increase in licensed storage:
tione 9
The projected date of the last refueling that can be discharged to tb l
spent fual poal assuming the present licensed capacity:
sof ter.-htr.,1; ;\\
(Fnd of botch discharga capability) 7
,,. e i:
\\/ EC [>
L (tNi y o. 378.-e
- s. Ft.
Vill. GLOSSARY The following abbreviations which may have been used in the Monthly Report, are defined below:
Atmospheric Containment Atmospheric Dilution / Containment ACAD/ CAM Atmospheric Monitoring American National Standards Institute ANSI ATWS
- Anticipated Transient Without Scram Boiling Water Reactor BWR Control Rod Drive CRD Electro-Hydraulic Control System EHC E0F
- Emergency Operations Facility Generating Stations Emergtacy Plan GSEP HEPA
- High-Ef ficiency Particulate Fil ter High Pressure Coolant injection System HPCI HRSS
- High Radiation Sampling System IPCLRT Integrated Primary Containment Leak Rate Test Intermediate Range Monitor
. l RM In-Service inspection 151-Licensee Event Report LER LLRT
- Local Leak Rate Test LPCI
- Low Pressure Coolant injection Mode of RHRS Local Power Range Monitor LPRM MAPLHGR
- Maximum Average Planar Linear Heat Generation Rate MCPR.__ - Minimum Critical Power Ratio Maximum Permissible Concentration MPC MSIV
- Main Steam Isolation Valve NIOSH
- National Institute for Occupational Safety and Health PCI
- Primary Containment isolation PCIONR
- Preconditioning Interim Operating Management Recommendations' RBCCW
- Reactor Building Closed Cooling Water System RBM
-' Rod Block Monitor RCIC
- Reactor Core isolation Cooling System RHRS
- Residual Heat Removal System RPS
- Reactor Protection System RWM
- Rod Wbrth Minimizer SBGTS
- Standby Gas Treatment System SBLC
- Standby Liquid Controi SDC
- Shutdown Cooling Mode of RHRS SDV
- Source Range Monitor TBCCW
- Turbine Building Closed Cooling Water System TIP
- Traveling incore Probe TSS