ML20042H001
| ML20042H001 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 05/03/1990 |
| From: | Robey R COMMONWEALTH EDISON CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| RAR-90-45, NUDOCS 9005170031 | |
| Download: ML20042H001 (13) | |
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[7 Commonwealth Edison-s1J
) ouad Cites Nuclear Power St: ten
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Cordova, luino6s 61242 9740
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RAR-90-45 L
L May 3,'1990 l-Director of Nuclear Reactor Regulations
.U. S. Nuclear Regulatory Commission Mail Station.Pl-137
' Washington, D. C.
20555 Enclosed please find a listing of those changes,' tests, and experiments completed during the month of April, 1990, for Quad-Cities Station Units 1 and 2. DPR-29 and DPR-30. A summary of the safety evaluations are being reported in compliance with:10CFR50.59 and 10CFR50.71(e).
Thirty _nine copies are.provided for your use.
.' Respectfully, COMMONWEALTH EDISON COMPANY-QUAD-CITIES NUCLEAR POWER STATION R. h.
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Robey Technical Superintendent RAR/LFD/jmt Enclosure cc:
R. Stols T. Watts /J. Galligan
'0027H/00612.
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Procedure Change,QOS 300-6.-Revision 4 i
Performance of Coupling Checks With Rod Withdrawal r
o Description This revision change allows for performance of coupling checks with numerous.
l rods withdrawn.
Evaluation f
IL[Theprobability'ofanoccurrenceorftheconsequenceofanaccident.
or malfunction of. equipment'important to safety as previously evaluated' t
in the Final Safety Analysis Report is not increased because the. con-(-
sequences of.an accident'are not changed since the coupling. check' 3--
' sequence is not changed.-
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- 2..The possibility for r,n' accident or malfunction of a different-type-l than any previously evaluated in the Final Safety Analysis Report.
j is.not created because this change-allows for doing copuling checks
'I with multiple rods withdrawn with no fuel in the vessel.
t 3.
The margin of safety, as defined'in the basis'for.any Technical Speci-l fication, is not reduced because the coupling check is not altered.
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Procedure Change'QOS 1600-13 Revision'8 PCI Group 2 and 3 Description Test the automatic, initiation of equipment with a simulated Primary Con-tainment 1 solation Group 2 and 3 signal..
Evaluation
- 1. -The. probability of an occurrence or.the consequence of.an' accident.:
or malfunction of equipment important to safety as:previously evaluated in the Final Safety Attalysis Report is not increased because-the changes to this test verifies the logic functions properly.
-2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report
.is'not created because this test opens the valves.and. verifies they.
close on a PCI Group 2 and 3 signal.
- 3. -The margin of safety. as defined.in the basis for any Technical Speci-fication. is not reduced because the containment isolation functions are still operable.
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i QTS 1104-1,1 Revision 12 SDM Calculations j
Description
-This revision change' adds steps and checklists to. evaluate corrections made in SDM calculations with respect to initial in-sequence criticals.
Evaluations 1.
The probability of an occurrence or the consequence of an accident, or malfunction of. equipment important to safety as previously cvaluated
.in the Final Safety Analysis Report is not increased because-this-change formalizes corrections to SDM calculations based on in-sequence critical determinations.
2..
The possibility for an accident or malfunction of a different. type =
than any previously evaluated in the Final' Safety Analysis, Report is not created because SDM requirements are still being maintained.
- 3. :The margin of safety, as defined in the basis for any Technical Speci-
-fication, is not reduced because SDM requirements still must be-maintained.
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J Modification M-4-2-87-017A j
Standby Condensate Booster Pump Auto Start Logic' 1
Y Description This modification revised the Standby Condensate / Condensate Booster Pump' l
(CBP) Auto Start Logic., The previous logic caused the standby CBP to auto start on low suction pressure (160 PS10)lto the Reactor Feed Pumps (RTP).
(This.
logic'was retained as a backup.to the new auto start logic.) The standby CBPf I
now starts on the loss'oflany running CDP. The modification was initiated'
.to eliminate the delay in starting the standby.CBP while waiting for the RFP suction pressure to' drop below 160 PSI.
Evaluation 1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important.to safety as previously evaluated
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in the Final Safety Analysis Report is not increased because this'
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modification does not alter any. equipment or, systems important to-safety as previously evaluated'in the FSAR.
In fact.'the CBP system reliability will be enhanced. -However, this would have no bearing on the probability or consequence of an accident or malfunction since analyses take no credit for this syctem.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated.in the Final Safety Analysis Report is not created because the potential. failure modes of theLCBP's (such as failure to auto start when needed or inadvertent start when not needed) are not different than the ones that presently exist.
Failure effects are bounded by existing failure anlaysis of the feedwater' control system, as provided in Section 11.3.3 of FSAR/UFSAR.
3.
The margin of safety, as defined in the basis for'any Technical Speci-fication, is not reduced because this modification does not alter or affect any equipment described in the Technical Specifications..
Therefore, the margin of safety will not be reduced.
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M-4-2-88-22B Reactor Drain Line is Description-Installed a tee with a blind flange in line 2-1265-2"-A to facilitate decontamination of RWCU line 2-1202-6"-A between valves MO 2-1201-2 and-MO 2-1201-5.
Evaluation 1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final' Safety Analysis Report is not increased because the modified piping has been analyzed, constructed and supported to meet the requirements of the FSAR.
Furthermore, no new systems are being added so the occurrence or the consequence of an accidet.t or malfunction F
of equipment important to safety as previously evaluated in the FSAR
-is not increased.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the Final ~ Safety Analysis Report-is not created because this modification does'not affect any of the bounding conditions in the FSAR accident analysis.. Because all bounding conditions remain the same, no new accidents are introduced by this modification.
3.
The margin of safety, as defined in the basis for any Technical Speci-fication, is not reduced because this modification does not reduce any of the existing margins of safety defined in the Technical Specifi-cations.
Since the intended function'of the modified system'1s not changing and the modified portion is designed to the'same standard of the existing system, the margin of 1fety'is not reduced.-
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' l' Modification M-4-2-88-101D Description The purpose of this modification was to comply with post accident monitoring requirements detailed in'the NRC Regulatory Guide-1.97. Partial D of this modifi-cation added seismic support to the existing 2-640-27 recorder on the 902-5 panel to comply with the Regulatory Guide 1.97 seismic requirements.
s Evaluation-
'1.
The probability of an occurrence.or the consequence of an accident, Jor malfunction of equipment important.to safety as previously evaluated-in the. Final Safety Analysis Report is.not increased because remounting-the existing recorder 2-640-27 on seismic mounts will retain.these devices-in place during seismic events.'therefore, the addition of the mounts has.not increased the probability of Design Basis Analysis or a Single Failure Event.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in'the Final Safety Analysis Report is not created because seismically mounting of recorder'2-640-27. mitigates component failures.
Failure of mount'is mitigated by the availability of the other division's redundant system.
Such failure does not create a new accident or malfunction not previously analyzed in the FSAR.
3.
Ttui margin of safety, as defined in.the basis for any Technical Speci-fication, is not reduced because the addition of seismic mounts to recorder 2-640-27 has no known impact;to the existing Technical Speci-fications. The-margin of safety remains unchanged.
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c t-Safety Evaluation #89-703 M-4-1(2)-89-071 Off Gas Cohdenser Level Controller Replacement Description Existing GE MAC offgas condenser level controller and indicator were replaced with Moore products digital controllers.
Evaluation
- 1..The probability of an occurrence or the consequence of-an accident,
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or malfunction of equipment important to, safety as previously evaluated in the Final Safety Analysis Report is not increased because this-modification replaces.the existing controllers with a digital controller.
The design function of'the equipment-is not altered.
2.
The; possibility for an accident or malfunction of a different type than any,previously evaluated =in the Final Safety. Analysis Report is not created because replacement of the controllers does not change the design function of the equipment or the eystem. so no new accidents or malfunctions are created.
3.
The margin of safety, as defined in the basis for any. Technical Speci-
'fication, is not reduced because replacement.of the level controllers will not reduce the margin of safety for any Technical Specification since the design function of the equipment and system is not changed.
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M-4-2-89-165 RHR Swing MCC Description This modification added protective relays to the LPC1/RHR Swing MCC 28/29-5.
The modification was implemented as a result of Licensee Event Report (LER)
$0-254/89-018. This LER identified a scenario in which a particular failure of the unit diesel generator voltage _ regulator during a loss-of-coolant. accident.
concurrent with a loss of offsite power event, could lead to a loss of all but one loop of the low pressure emergency. core cooling systems.
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This modification added. relays to monitor one phase of the incoming MCC power. The relays will actuate on under/overvoltage or under/overfrequency condition when the unit-diesel generator is supplying power to the MCC.
This will then cause the MCC to transfer its feed to the 1/2 Diesel Generator.
Evaluation 1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated-in the Final Safety Analysis Report is not increased because reliability of the power source of the injection valves Jof the LPC1_ mode of the RHR system will be enhanced.
This has 1) no bearing-on the probability of an accident, 2) the effect of reducing the probability of a mal-function of equipment important to' safety (specifically the Swing:
MCC), and 3) no effect on the radiological consequences of an accident.
or malfunction, since existing analyses which assume failure of the LPCI mode of the RHR system are not impacted.
2.
The possibility for an accident _or' malfunction of a different_ type.
than any previously evaluated in the Final Safety Analysis Report is not created because individual component failures'within the swing MCCs control logic could either 1) prevent transfer to' alternate-upon loss of normal, or equivalently, disconnect both power feeds which would disable the injection valves of the LPCI mode of-the RHR system fed from this MCC, or 2) cause an, inappropriate. transfer to alternate, which upon future loss of this alternate, would also disable the same injection valves. These failure modes exist in both the present and the modified configuration. There is no new failure mechanism created.
3.
The margin of safety, as defined in the basis for any Technical Speci-fication, is not reduced because_the Tech Spec. bases for the LPCI mode of the.RHR system (3.5) and Aux. Power (3.9) systems are not changed by the design covered by this modification which enhances availability of the emergency power to the valves of the LPCI mode of the RHR system.
The margin of safety remains unchanged.
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M-4-2-89-166 Jet Pump Instrumentation (JPI)-Drain Lines-Description The 2-263-39A, B and 40A, B valves have been leaking through their packing
-increasing ~drywell leakage..The lines associated with the valves existed to
' drain the Jet Pump. Instrumentation Nozzle and were never.used. The lines were removed from the reactor and the drain header which goes to the sump tank.
The ends were capped on'both ends.
Supports to-the remaining pipe stubs were added as necessary.
Evaluation 1.
The' probability of an occurrence or the consequence of an accident or malfunction of equipment important to safety as previously evaluated.
in the Final Safety-Analysis Report-is not increased because'this modification consists of removing two unnecessary drainage lines within the drywell.
Therefore, there.is no impact on the probability of
- an occurrence or the consequences of an accident.
2.
The possibility for an~ accident or malfunction of'a different type than any previously evaluated in the Final Safety Analysis Report is not created because no new types of malfunctions are created by this modification, because the.JP1 drain lines meets its design basis which includes identified plant transients.
3.
The margin of. safety, as defined in the basis for any Technical Speci-fication, is not reduced because as described above, this modification has no impact on the probability of an occurrence of an accident since it involves removal of equipment, and no new types of malfunctions are created. Therefore, the margin of safety as defined in the basis of any Technical Specification is not reduced..
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t Special Test #1-129 c
Description Special Test No. 1-129 was completed on April 2, 1990.
This test-consisted of lowering a neutron source with integral thermal neutron detectors into the Unit One reload high density fuel rack cells using the refuel bridge grapple.
The purpose-is to determine the amount of boraflex loss in the fuel racks-and, hence, to assess the suberiticality of the fuel pools.
Evaluation 1.
The probability of an occurrence or the consequence'of an accident, or malfunction of-equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because no fuel:
will be moved as a part of this test.
In addition, all work will t
take place in the Unit One spent fuel pool in accordance with approved station procedures.' Therefore, the refueling accident, as described in FSAR Section 14.2.2, is not applicable.
- 2..The possibility for an accident or malfunction of a different type than any previously evaluated in the Final, Safety. Analysis Report is not created because no activities in conjunction with this test '
will create accident potentials greater than the dropped fuel bundle accident considered in the FSAR.
3.
The margin of safety, as defined in the basis for any Technical Speci-fication, is not' reduced.because none of the activities associated with thin test could potentially cause radioactivity releases in excess of those assumed for any Technical Specification.
The purpose of this test is to assure compliance with' Technical Specification 5.5.B,.which requires the spent fuel pool K gt to be 1 95.
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Special Test'$1-142-
'i Safety Evaluation #90-251' i
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CRD M-3 Special Maneuver j
s Description-j
.Special Test No. 1-142 was completed on March 29, 1990. The purpose'.
l of this.special test was to determine whether control rod drive M-3 would j
respond to control rod movement signals generated by the rod motion control system and. execute-proper inward and outward rod notch motion.'
This-special test attempted to re-create the conditions under which-control rod M-3_was reported to insert continually for several notches when given a single notch-in signal.
7 Successful completion of this special test may be used by the operating engineer to evaluate the,previously described performance of control rod M-3.
This test is diagnostic only and directs rod motion'in accordance-with QOP 280-1 Reactor Manual Control. This test does not require the-installation of any permanent or temporary test equipment.
Evaluation it was determined that no 10CFR50.59 Safety-Evaluation was required.
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3 Special Test'#2-96 I-Safety-Evaluation #90-310 m-Installation of Strain Gauges.on the Unit TVo o
i-Reactor Vessel' Head Description Special Test'No. 2-96 war ompleted'on' April 23, 1990. The-purpose-
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s of this test was to measure the stresses on the Unit Two reactor vessel
-head. Strain gauges were insta11es and monitored during' reactor vessel-hydrostatic ~ test.
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. Evaluation-i It_was determined.that'no'10CFR50.59 Safety. Evaluation was required-for this-special test.
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