ML20084S321
| ML20084S321 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 05/30/1975 |
| From: | Kalivianakis N, Kulivianakis N COMMONWEALTH EDISON CO. |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20084S324 | List: |
| References | |
| AO-50-265-75-17, NJK-75-302, NUDOCS 8306160678 | |
| Download: ML20084S321 (5) | |
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Hay 30, 1975 Director of Office of Nuclear Reactor Regulation U.S. Nuc1 car Regulatory Commission Washington, D.C.
20555 Refe rence: Quad-Cities Nuclear Power Station Docket No. 50-265, DPR-30, Unit 2 Appendix A, Sections 1.0.A.3, 1.0.A.5, 6.6.B.I.a Enclosed please find Abnormal Occurrence Report No. A0 50-265/75-17 for Quad-Cities lluclear Power Station. This occurrence was previously reported to Region lil, Directorate of Regulatory Operations by telephone on May 22, 1975 and to you and Region ill, Directorate of Regulatory Operations by telecopy also on May 22, 1975 This report is submitted to you in accordance with the requirements of Technical Specification 6.6.B.I.a.
Very truly yours, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION
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_l N.J. Kal ivianaki s hM
_ Station Superintendent J,V NJVJBBP/vmb cc: Region lil, Directorate of Regulatory Operations J.S. Abel 6
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REPORT NUMBER: A0 50-265/75-17 REPORT DATE: June i, 1975 d
OCCURRENCE DATE: May 22,1975 FACILITY: Quad-Ci ties Nuclear Power Station, Uni t 2 IDENTI FICATION OF OCCURRENCE :
(1) Unplanned of fgas release, y
(2) Probable fuel failure due to pellet-clad interaction.
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. CONDITIONS PRIOR TO OCCURRENCE:
Unit 2 was operating at 2120 ff.It and 700 MWe during a normal startup folle <-
ing a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> forced outage.
DESCRIPTION OF OCCURRENCE:
The sequence of events leading to the occurrence is as follows:
tuy 21, 1975 13:29 The generator was brought on the system at 100 MWe and power was increased by rod withdrawal at minimum pump speed.
18:40 At a load of about 400 MWe power increase was begun by core flow at a rate of 50 MWe/hr.
21:10 At a load of 560 MWe rod group 105 was withdrawn in-sequence
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(group 105 is an eight rod group withdrawn from position 36 to
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44).
22:00 LPRM Hi alarms were noted and the Unit 2 nuclear engineer was phoned.
He recommended insertion of group 105 to control the power peaking.
22: 10 Of the eight rods in group 105,
!J rod E-3 was inserted 44 to 36 rod N-5 was inserted 44 to 38 rod L-13 was inserted 44 to 38 i
The LPRM Hi's cleared at this point so rod insertion was dis-continued and the power increase was continued with core flow.
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AD 50-265/75-17 U (v Juna I, 1975 22:30 The uni t reached 700 We.
The average ramp rate from 400.We was therefore 60 We/hr.
24:00 The unit nuclear engineer was again phondd that there were LPRM Hi's.
The nuclear enoineer asked if any rods in group 105 were still at position 44.
He was told there were.
The nuclear en-gineer instructed the operator to insert group 105 to position
'40 and lef t home for the plant.
Pay 22,1975 00:20 The unit operator banked rod group.105 at position 40. This resulted in the following rod movements:
rod E-3 was withdrawn 36 to 40 rod N-5 was withdrawn 38 to 40 rod L-13 was withdrawn 38 to 40 rod C-Il was inserted 44 to 40 rod L-3 was inserted 44 to 40 rod N-il was inserted 44 to 40 rod E-13 was inserted 44 to 40 rod C-5 was inserted 44 to 40 00:40 The offgas Hi Hi alarm was received; the chimney valve timer started, and subsequently isolated.
Load was decreased to 400 We with flow and the rods in group 105 were inserted to position 32.
01:00 The of fgas isolation was reset.
DESIGilATI0rl 0F APPAREllT CAUSE OF OCCURRENCE:
(1)
Equipment Failure (fuel damage)
(2)
Personnel Error The cause of this occurrence was a combination of in-sequence rod with-dra.vals that produced abnormally high peaking at the bottom of the core due to a low xenon condition, followed by a power increase on flow.
The net ef fect was a local poaer level increase at a rate greater than that which would allow the fuel pellet to cladding stresses to relax without cladding failure.
Since no fuel safety limits were exceeded the primary cause must be at-tributed to equipment failure.
High exposure fuel is evidently subject to failure f rom the pellet-cladding mechanical interaction if the local rate of power increase is excessive.
Personnel error, however, must be designated as a contributing cause.
Mod-ification of the control rod sequence by the nuclear engineer prior to this startup or at a lower power Icvel during the startup to avoid exceed-ing the maximum previously experienced power shape, or redection of the rate of power increase may have prevented the fuel damage and resultant of fgas release.
Operating personnel may also have minimized the damage had they more thoroughly understood the reactor conditions at the time and Inserted enough rods to. completely clear the high peaking.
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June I, 1975 ANALYSIS OF OCCURRENCE:
The TIP (Traversing in-core Probe) system was utilized by the operator to obtain an axial flux profile in a high power density region of the core prior to the midnight rod moves.
The calculated peak LHGR was 14.65 KW/ft In a 7 X 7 assembly and 13.26 KW/ft in an 8 X 8 assembly.
These values are less than the design LHGRs of 17 5 and 13.4 KW/f t.
and also less than the densification spiking penalty limits of 17.42 and 13 34 KW/f t, respectively.
As indicated in the following table the calculated values for MCPR and
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liAPLHGR were also less than thei r respective limi ts for both fuel types.
The 8 X 8 total peaking factor was 3.55 which existed f rom 21:10 when group 105 was first withdrawn until 00:20 when group 105 was banked to position 40.
The present APRM scram and rod block settings are conservative for peaking f
factors less tbn 3.14 for 7 X 7s and 3.10 for 8 X 8s.
Coasequently the settings were non-cor.servative for the 8 X 8 during this time interval.
iUEL TYPE LOCATtON CALCULATED LIMIT 7X7 47-42 LHGR 14.65 KW/ft 17.42 KW/ft node 4 (L=2.5 ft)
- 12. 78 KW/ f t 14.4 KW/ft (9 2 GWD/T)
TPF 3.03 3 14 MCPR 1.41 1,29 8X8 49-42 LHGR 13.26 KW/ft 13 34 KW/ft node 4 (L=2.5 ft)
APLHGR 11.08 KW/ft 11.16 KV/ft (0.4 GUD/T)
TP F 3.55 3.10 MCPR 1.42 1.35 Although the rods that were wi thdrawn in this startup had been withdrawn in previous cycle 2 startups, the peak local powers previously experienced in these nodes were approxinately 10 to 11 KW/f t compared to the 13.5 to 14.5 KW/f t experienced during this occurrence.
To follow the General Elec-tric recommendation for preventing fuel failure, the control rod sequence should have been modified earlier in the startup to reduce the bottom core peaking and the power increasc should have been slo.-cd at about 600 MWe.
As a result of the fuel failure there was an unplanned release of offgas.
The maximum relcase rate at the SJAE was estimated to be 1.5 ci/sec. At equilibrium conditions the corresponding chirney release would be 0.038 ci/sec. The instantaneous peak release at the chimney may have been five times greater than that, or 0.19 ci/sec, which is below the Technic _al Specification instantaneous release rate of 0.506 ci/sec based on E data
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g from April, 1975 The steady state release rate at the SsAE one week be-fo re t
the failure was approxinately 0.058 ci/sec at 2200 int and was approxi-mately 0.23 ci/sec four days af ter the failure at the same power level.
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There were no ef fects on the health and safety of the pu51ic related to this 3
occurrence.
j CORRECTIVE ACTION-I
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I Due to the short period of operating experience with the new charcoal filters, the results shown on the chimney monitor during this event were inconciusive.
Investi-gations are being continued to establish calibration data for these monitors while
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operating with the recombiner/ charcoal 3ystem.
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h initial corrective action to the offgas increase was to decrease reactor power i
The with flow and to insert control rods to reduce the power peaking at the bottom of 4
the core. The final corrective action will take the form of more detailed and fornal instructions and information for use by the operating department in deter-l mining when a power increase should be slowed or halted due to the possibility of l
causing fuel failure.
This will be accomplished by the following administrative improvements:
l (1) Whenever possible LPRM readings which represent the most recent maximum previously achieved power distribution will be provided to the reactor operator for reference during startups to serve as a guide in determining if the power envelope is being approached too soon.
(2) Written instructions from the nuclear engineer will be approved by the operating engineer and included in the Daily Order Book for maneuvers which have the potential'of exceeding the previous power envelope.
(3) Increased efforts will be made by the nuclear engineers to more accurately I
determine when the control rod sequence will require modification to stay within the previous envelope, especially on xenon deficient startups, fq Confideration will also be given to the possibility of reduced ramp rates or power
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soaks following outages of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more in order to allow the buildup of a larger xenon inventory during non-cmergency load conditions.
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l FAILURE DATA:
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f from rapid power increase has been experienced previously at both I
f Fuel failure Qua-: Cities and Dresden stations, as reported in Dresden report A0 50-249/74-38
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and as a result of a rod withdrawal error at 0.uad Cities reported September 21, 1973 and u verified by sipping results. Although differences exist in the circumstances j
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'f of tNse incidents, the cumulative experience indicates that significant fuel failure may result if the local rate of power increase is excessive.
The correct s,ve actiw s stated above in attempt to control administratively these local power increases should help eliminate further occurrences of this type.
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