ML20065P677
| ML20065P677 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 12/03/1990 |
| From: | Robey R COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9012140006 | |
| Download: ML20065P677 (12) | |
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.c RAR 90 87 Decenber 3,1990 U. S. Nuclear Regulatory Connission ATIN: Document Control Desk Washington, D. C.
20555
$UBJECT: Quad Cities Nuclear Station Units 1 sind 2 Changes, Tests, and Experiments Conpleted bjUocket Nes, 50 254 end 50 265 Enclosed please find a listing of those changes, tests, and experiments completed during the month of Noventer,1990, for Quad Cities Station Units 1 and 2, DPR 29 and DPR 30. A surrmary of the saf ety evaluations are being reported in compliance with 10CFR50.59 and 10CFR50.71(e).
Respectfully, COMMONWEALTH EDISON COMPANY QUAD
- CITIES NUCLEAR POdR STATION
[
R. A. Robey l
Technical Superintendent i
RAR/LTD/ktm 1
l Erslosure cci A. B. Davis, Regional ArJninistrator T. Taylor, Senior Resident Inspector i
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ADOCK 05000254 1647 yo ty
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l Safety Evaluation #90 783 Holst Limits Associated With New Refuel Bridge Most Design tieserlot ton I
Change FSAR section 10.1.3 to reflect current refuel-bridge mest desten. The FSAR describes the hoist limits for raising fuel as reflected by the old refuel bridge mast. The new l
mest has different holst limits. The old mest limited the minima depth to 9 feet of water above l
l active fuel. The new mest limits the depth to 81/2 feet for the " normal # position and 6.75 f eet for the minlaus possible depth.
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tvaluation i
1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment leportant to safety as previously evaluated in the Final Safety Analysts l
Report is not increased because the funetton and normal operation of the refuel bridge la unchanged. The refueling interlocks operate as before. The bundle drop accident evaluation is unchanged. The total distance a bundle enn drop'Is less than that analyzed in the FSAR.
2.
The posalbility for en accident or malfunction of a different type than any previously.
evaluated in the Final Analysis Report is not created because the function and normat operatton of the refuel bridge is unchanged. No new accident or malfunction type has Introduced.
l 3.
The margin of safety, as defined in the besla for any Technical Specification is not reduced because the margin of safety Is unchanged. The refueling interlocks operate I
as before.
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t Safety Evaluation #90 854 Control Rod Movements and Control Rod Sequences Deserlotion Provides additional controls for moving controt rode while there is no fuel in the vesset.
Eveluation I
1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipnent leportant to safety as previously evetuated in the Final Safety Analysis Report is not increased because all fuel will be removed from the reactor and stored in the Nigh Density fuel Rocks (NOFR) or fuel handling equipment. While the fuel is, removed f rom the vesset, control rod movements may be performed without ef fecting criticality or fuel Integrity. Therefore the probability of an occurrence or consequence of an accident is not increased.
2.
The possibility for en accident or malfunction of a difference type than any-previously evaluated in the final safety Analysis Report is not created because all fuel wlll be removed f rom the reactor and stored in the NDFR. The fuel will never be stored in a manor that is not consistent with analysis in the FSAR. Therefore, the possibility for en accident or malfunction different than previously evaluated is not created.
3.
The margin of safety, as defined in the bests for any lochnical speelficationi la not red.ced because control roo withdrawat will not effect reactivity or fuel integrity while the core is unloaded.
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tafety Evaluation #90 848
.OCAP Personnel Montiroing 1
i-EttrlDtlon to use TL0s instead of film badges and the use of electronic dosimetry instead of or along with lonisation chnabers.
b tvaluation f
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1.
The probability of an occurrence or the consepe of an accident, or selfunction of.
equipment inportant to safety as previoustt evaluated in the Final lafety Analysis l
Report is not increased because it does not involve equipment used to mitigate the i
affects of an accident.
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2.
The possibilty for en accident or malfunction of a different type than any previously k
evaluated in the Final Safety Analysis Report is not created because upgrading to state of the art dosimetry does not affect any accident precursor.
l 3.
The margin of safety, as defined in the basis for any Technical Specification,'Is not reduced because it is not safety related nor is it mentioned in fechnicet specifications.
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4 Safety Eveluation #90 832 Change to Sectke 6 of Tech Specs Description The Nuclear Quality Programs approvet authority changed, and the head of Quality Programs and Assessnents title is changed, and the procedure section is rcplaced with the Stardard Technica!
Specification.
EvcLustion 1.
The probabilty of an occurrence or the consequence of an accident or malfuention of equipment important to safety as previously evaluated in the final Safety Analysis Report is not increased because the approved authority and title changes are 3dninistrative and do not af fect the occurrence probability or consequences of an accident or malfunction. Replacing the procedure section with the standard Tech spec will not decrease the quality of procedures or reviews and thus also will not effect accidents or modifications.
2.
The possibility for an accident or malfuention of a different type than any previously evaluated in the Final Safety Analysis Report is not created because the changes are basicatty adninistrative in nature and do not involve any new modes or methods of operating the plant. Thus, no new possibilities for accidents or malfunctions are created.
3.
The margin of safety, as defined in the basis for any Technical Specification, is not red.ced because the standard Tech Specs provide for a more appropriate review of procedares than current Tech Specs. Thus, the quality of procedures will not be decreased and no me-91n to safety will be reduced by thes6 normally adninistrative changes.
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I Safety Evaluation WO 773 Rector Recirculation and Reactor Water Cleanup Systern Decontamination DescrIDtion During the Unit 1 Refuel Outage decontamination of piping associated with the Reactor vessel was performed, the Reactor Water Cleanup Pintaa was performed with fuel in the vessel and the vessel head removed. The decontamination chemicals o'd not enter the vessel during this proces#.
The Recirculation Pump Suction and Discharge Piping was also decontaminated. This was done with the fuel removed from the vessel. The vessel head was in place but not tensioned. Water level in the vessel was maintained below the core area of the vessel. The decontamination chemicals we.re flushed from the vessel prior to reloading fuel.
Evaluation 1.
The probability of an occurrence or the consequence of an accident or malfunction of equipnent inportant to saf ety as previously evaluated in the Updated Final Safety Analysis Report (UFSAR) is not increased as a result of this job. The original matelial specifications for the recirculation system, annulus, and RWCU system allowed for general corrosion. Results of corrosion testing and analysis by GE, and EPR' ^nd reviewd by System Materiets Analysis (SMAD) and Chemistry Services indicate ths-the solvent corrosion rates are less than the original allowances.
2.
The possibilty for an accident or malfunction of a different type than pre.' wsty evaluated in the Update Final Safety Analysis Report is not created. Potential accidents or malfunctions of equipment were reviewed and addressed as follows:
The effects of residual solvent in the system was determined to be negligible.
Reactor Coolant is cleaned and returned to a conductivity and a TOC Level which is acceptable to the chemistry staff.
Gaseous releases from leaks / spills will be monitored via the normal Reactor Building Vent Monitor or by the Stand By Gas System if the release level exceeds the specification for the Reactor Building Vent system. Liquid spills "Ill be processed through normal redweste lines. The diluted solvent is compatible with the sts.tlon rodweste system.
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The margin of' safety, as defined in the basis for any Technical Specification, is not
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reduced. The decontamination project wlLL be performed in accordance with the
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existing technical specifications. The reactor will be maintained in the shutdown or refuel mode with att interlocks in the shutdown position.
Coolant cheelstry will be monitored regularly throughout the project and upon
- i completion of the decontamination, the coolant will be returned to a conductivity and a TOC Levet that is acceptable to stetton chemistry and redweste.
Liquid and/or geselous releases will be monitored rs normal and wlLL ad ere to technical specification limitatione.
The decontamination will be performed at 90 +/ 5 degrees Celsius (185 200 ' degrees FahrenhetO and at approximately atmospheric' pressure, both felt within the technical' specification limits for maintaining primary system integrity.
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Procedure Change GAP 300 2, Revision 28
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Deserlotion Procedure change to require throttle valves are given a 25 second Clost signal, requirements in seergencies to have procedures "on hand", putting equipment in PTL, and to correct DVR reference on resetting thermal trips.
Evaluation 1.
The probabilf'y 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 these changes are conservative or clarify requirements to be consistent with other procedures.
2.
The possibility for en accident or malfunction of a different type than any previously evaluated in the Final Analysis Report is not created because giving guldance on when equipment can be put in PTL will not cause an accident different from the FSAR types.
3.
The margin of safety, as defined in the basis for any Technical'$pecification, is not reduced because theses changes are consistent with Tech Specs.
Procedure Change QCS 1300 3, Revision 7 Description 00$ 1300 3 is now incorporated into QCOS 1300 3.
The major change is per the Writers Guide (format). Also steps were made more information and functional for operator use.
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 RCIC system will still function as stated in the FSAR. The new Q00$ verifles monthy valve operability and does not change the reliability or the function of any RCic motor operated valve which would increase the probability or consequence of an accident previously evaluated in the FSAR.
2.
The possibility for an accident or malfunction of a different type than any previous,,.
evaluated in the Flant Safety Analysis Report is not created because QCOS 1300-3 now Incorporates 00$ 1300 3 and did not change the configuration of any valves, instrtrnents, or controls that could put the RCIC system in an unanalized conditions not previously addressed in the FSAR.
3.
The margin of safety, as defined in the basis for any Technical Specification is not reduced because the Tech Spec requirements are still being satisfied by the new QCOS 1300 1 and RCIC motor operated valve operability is maintained to verify the margin of safety is not reduced.
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1 Procedure Change oos 13001, Revision 14 Monthly RCIC P g Operability Test Deserlotim 00S 1300 1 is now incorporated into oCOS 1300 1.- The major change la per the Writers Guide (format). - Also steps were made more information and functf unal for operator use.
Evaluation 1.
The probability of an occurrence or the consequence of an accident, or malfunction of
- equipment inportant to safety as previously evaluated in the Final Safety Analysis Report is not increased because the RCIC system will stlti function as stated in the FSAR. The new QCOS verif fes monthly puy operability and does not change the reliability or function of any RCIC conponent which would increase the probability or consequence of an accident previously evaluated in the FSAR.
2.'
The possiblLity for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because QCOS 1300 1 now incorporates Cos 1300 1 and did not change the configuration of any valves, instruments, or controts that could put the RCIC system in an unanalized condition not previously addressed in the FSAR.
3.
. The margin of safety, as defined in the basis for any Technical Specification, is not re<beed because the Tech Spec requirements are still being satistfled by the new QCOS 13001 and RCIC pump operability is maintained to verify the margin of safety is not reduced.
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Procedure Change QFP 100 1, Revision 24 Master Refueling Procedure Description This revision adds additional guidance and controls to ensure safe refueling operation.
Requirements are added regarding the use of the SRM shorting links, the blockage of control. rod motion, and the use of the Fuel Handling Verifier for second verification for all fuel moves.
Gulden 6e and control on raising the main hoist above the "normat up" position is added. A prerequisite is'added to verify proper Indication of the grapple selsyns.
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 Incrossed because the function and normat operation of:the refuel bridge is unchanged. Additionet' procedural controls are added to help ensure safe operation, proper consu11 cation, and proper transferring of fuel. The refueling interlocks operate as before and the bundle drop accident evaluation is unchanged. The total distance that a bmd!e can drop is less than that analyzed in the FSAR 2.
The possibility for en accident or malfunction of a different type than any previously evaluated in the Final safety Analysis Report is not created because the function and normat operation of the refuel bridge is unchanged. No'new accident or malfunction type is introduced.
3.
The margin of safety, as defined in the basis for any Technical specification, is not reduced because the margin of safety is unchanged. The refueling interlocks oeprate as before.
I Safety Evaluation #90 819 18 Core Spray Motor inspection Description Installation of a teaporary steet plate on a Secondary containment hatch.
Evaluation 1.
The probability of an occurrence or the consequence of an accident, or malfunction of j
equipment leportant to safety as previously evaluated in the finat safety Analysis Report is not increased because the teaporary plate has been analyzed by engineering to ensure that it meets secondary containment requirements. Reference SESR #4 0345.
2.
The possibility of an accident or malfunction of a different type than any previously evaluated in the Final-Satety Analysis Report is not created because the plate.
1 provides for secondary containment. No other system is af f acted.
3.
The margin of safety, as defined in the besla for any Technical Specifications, is not-reduced because secondary containment wiLL be maintained. There will be no increased--
risk of a radioactive release to the environment.
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