ML20101F143

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Annual Rept for Braidwood Nuclear Power Station for June - Dec 1991
ML20101F143
Person / Time
Site: Braidwood  Constellation icon.png
Issue date: 12/31/1991
From: Simpkin T
COMMONWEALTH EDISON CO.
To: Murley T
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9206240297
Download: ML20101F143 (25)
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v Csmm:nwealth Edison -

1400 Opus Place

- Downers Grove, Illinois 60515

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l June 17,1992 Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Document Control Desk

Subject:

Braidwood Station, Units 1 & 2 10 CFR 50.59 Annual Report NRC DochtNoL50-A53_acG0-451

Reference:

A.R. Checca letter to Dr. T.L. Murley dated October 17,1991.

Dear Dr. Murley:

Pursuant to 10 CFR 50.59(b)(2), Commonwealth Edison is providing the

- required annual report for Braidwood Station (Facility Operating License Nos. NPF-72 and NPF-77). The annual requirement is based on the Unit 1 fuel load license (NPF-59) issuance date of October 17,1986.

This report covers the period from 6/19/91 to 12/18/91 and consists of the descr:ptions and the safety evaluation summaries for the following: changes to the facility described in the Safety Analysis Report, and changes to procedures as described in the safety analysis report. No tests or experiments were conducted which were not previously described in the Safety Analysis Report. Included also as part of this report are changes made to features of the Fire Protection Program not previously

- approved by the commission. No changes to procedures governed by 10CFR50.59 (a) were performed.

Currently, the UFSAR and Fire Protection Report revisions are submitted by December 18 of each year. This date is based on the anniversary of the Braidwood .

Unit 2 Operating License. Thece reports cover the period from June 19 of the previous year to June 18 of the year the' reports are due. The referenced letter informed the NRC of Commonwealth Edison's intent to shift Braidwood's reporting date for annual 50.59 reports to match the UFSAR and Fire Protection Report Revisions. To accomplish this we will submit the next 50.59 report by December 18,1992. That re aort w;il ec,ver the period from December 19,1991 through June 18,1992.!n ac dition, subsequent 50.59 reports shall be submitted annually on the same schedule as the UFSAR and Fire Protection Report revisions.

9206240297 DR 920617 g ADOCK 0500o456 PDR f I

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e Dr. Thomas E. Murley June 17,1991 Please direct any questions regarding this matter to this office.

Very truly yours, Q j v w ~k). ) ~ f T.W. Simpkin Nuclear Licensing Administrator ,

cc: R. Pulsifer Braidwood Pro.ect Manager, NRR.

B. Clayton-Chief, Branch ' -Rill S.G. Dupont Senior Braidwood Resident inspector

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Braidwood Nuclear Power Station 10CFR50.59 Annual Report June - December,1991 NRC Docket Nos. 50-456 and 50-457 License Nos. NPF-72 and NPF-77 i

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. I. EACJLLT_Y_CliANGES II. ERQQEQUf1E_ CHANGES

~A. - COREEEQESJGNS A. PROCEDURALUESAR CllANGES

1. Unit #2 Cycle 3 Reload 1. UFSAR DRP 3 005
2. UFSAR DRP 3 011 B. MlNORELANT_ CHANGES 3. UFSAR DRP 3-022 4, UFSAR DRP 3-023
1. MCR 20-0 91-660 5. UFSAR DRP 3 037
2. MCR 20 2-90-008 6. UFSAR DRP 3-042 C. MODJELCAIJONS Ill. IESISlEXEEBIMENTS
1. M20-0 89 011 A . None
2. M20 2 88-046
3. M20-2-88-056
4. M20-2-89-013
5. M20-2-89-023
6. M20-2-89 030

. M20-2-90-014 D. N1U.CLEAB_ WORK.BEQUESIS

1. NWR A48630 E. SCALINedSETEOINI_ CHANGES
1. SSCR 90-041 F. IEMP_QBABXALIEBAILONS
j. 1 TA 91-2-032 L

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COBEEEDESIGN Unit 2 Cycle 3 Reload DESCBlPllON Braidwood 2 Cycle 3 core redesign due to normal reload fuel requirements.

SAEET_Y_ EVAL.UAIlotLSUMMABY

1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previcus'y evaluated in the Final Safety Analysis Report is not increased because as evaluated in section 3.1 of the Braidwood 2 Cycle 3 reload safety evaluation, the reload core does not impact LOCA limits. The revised Braidwood 2 Cycle 3 reload design has been verified to satisfy LOCA accident analysis limits and ,

assumptionsc The Braidwood 2 Cycle 3 reload parameters have been verified to be less limiting than the bounding values assumed in the (.OCA analysis of record. and the reload core does not adversely impact the design '

or operation of any other plant equipment

2. The possibility for an accident or malfunction of a different type than any areviously evaluated in the Final Safety Analysis Report la not created aecause the method and manner of plant operation s unchanged, and the reload core's structural, thermal-hydraulic and nuclear characteristics are not significantly different from previously installed equipment.

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3. The margin of safety, as defined in the basis, for any Technical Specification, is not reduced because the Braidwood 2 Cycle 3 reload safety evaluation / safety parameter interaction list process as documented in the reload safety evaluation / safety parameter interaction list master checklists and minutes demonstrate the new key parametern of interest do not exceed their associated limits.

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MLNQR.El.ANLC11ANGE MCR 20-0-91660

DESRBJPllON

Replace Magnetic tape drives with solid state memory devices (RAMDECKS). Current tape drives are becoming unreliable and obsolete. The RAMDECKS will be state of the ed and more reliable.

SAEETLEVALVATJON.SUMMABY

1) The probability of an occurrence on the consequence of rn accident, or malfunction or equipment impoi1 ant to safety as previously evaluated in the L

Final Safety Analysis Report is not increased because this equipment is not safety related. The equipment is not used or referenced in the accident analysis nor does the equipment affect other plant equipment.

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 seismic monitor system remains functionalin according seismic events. The seismic system itself is independent .. all other plant equipment, thus it cannot impact other plant systems or functions.
3) The margin of safety, as defined in the basis, for any Technical Specification, is not reduced because all changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative direction. Therefore, the actual acceptance limit need not be identified to determine that no function in margin of safety exists.

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MINOf1ELANLCHANGE MCR 20-2 90-008 DESCRIEIlON:

-Replace the seal injection filter outlet isolation valve. The existina valve is a 2" diameter Kerotest valve.~ The new valve will be a 2" M.S.B. BeliUws Sealed globe valve. The existing valve is prone to through leakage and steam leakage.

SAEEIY EVAL.UATJON.SUMMABY

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 new valve should be more reliable than the existing valve. The probability or the consequence of an accident will be unchanged.
2) The possibility for an accident normal function of a different type than any 3reviously evaluated in the Final Safety Analysis Report is not created 3ecause the new valve serves an identical function to the old valve. New accidents or malfunctions area created by its installation.
3) The margin of safety, as defined in the basis, for any Technical Specification, is not reduced because the new valve will perform all the same functions as the old valve. A modification test will confirm that the valve has satisfactory characteristics to retain Tech Spec safety margin.

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1 MODIFICATIONS M20-0 89 011 A DESchlPTION:

Relocate the free field seismic monitor, to facilitate cone'"iction activities for the new onsite training f acility, and decrease instrument proximity to adjacent vibration sources to preclude resulting ialso Indications of seismic activity.

SAEETYEVALVAIlON

SUMMARY

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 reliability of the affected setsmic nionitor for operation, as described in UFSAR Section 3.7.4.2.2,is enhanced by the implementation of this design change. The affected incirumentation provides indication used in evaluating nelsmic events only, for prompt determination of the seismic response of plant

'entures, and is not specifically required for cafe shutdown of the plant.

2) The possibility f-or an accident or malfunction of a different type than any 3rev4o0 sly evaluated in the Final Safety Analysis Report is not created accause no new failure modes are introduced by the installation of this I

modification to impact the ability of the free field triaxial acceleration sensor to peHorm its monitoring function, as specified in UFSAR Section 3.7.4.1.

The revised declgo is functionally and operationally unchanged from the existing design, as only the location of the subject instrumentation is changed by this modification.

3) The margin of safety, as defined in the basis, for any Technical Specitienin 1, is not reduced because the existing bases for related Technical Specification 3/4.3.3 tegarding the operab!lity of non safety related seismic instrumentation, is to ensure that sufficient capability is available to promptly determine the magnitude of a seistnic event and evaluate the response of those features 'mportant to safety. This capability is not impaired by relocating the ? se field ensmic monitor, as the function and operation of the associate instrumentation is not impacted by the proposed design change.

However, tha installation of this modification required a change to Table 3.3 7 of the Technical Specifications, as well as the associated bases for

! Hi'.ing Condition for Operation and Surveillance Requirements 3/4.3.3.3, to I,.acate the revised focatien of th9 free field sensor. The NRC issued Litonse Amendment 28 (411-91) addressing the Technical Specification changes. Accordingly, the margin of safety is unchanged.

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MODIElCATION M20 2 88-046 DESCBIEIJON ,

Provide an audible type alarm for the station operator,in addition to each of the division low low love: rtatus lamps, for the level of NAOH colution in the spray additive tank for the CS System by connecting a common window alarm to both of the low low level switches. Move the window alarms for high low levels in the spray additive tank and high percentage of hydrogen in containment down one tile each to provide proper top to bottom sequence for the new tile.

SAFET1EW1UAllONEU.MM6BY

1. The probabili of an occurrence or the consequence of an accident, or malfunction o equipment important to safety as previously evaluated in the Final Safety Analysin Report is not incronsod because the now alnrm circuit is only a source of low onergy, annunciatoi voltage out of a normally open level switch contact, and does not have the potential for c:reati,1g or alleviating the mitigation of a LOCA. 1
2. -The possibility for an acc! dent or rnalfunction of a different type than any 3rev ously evaluated in the Final Safety Analysis Report is not created 3ecause the analysis of failure modes and effects proved that this modification was not sub,Wt t; nct create any single failure event that could disable the indication for loc / low level or operation of the CS system and that it was not possible to create a di*forent type of accident or malfunction of equipment. ,
3) The margin of safety, as defined in the bask for any Technical Specification, is not reduced because the margin of safety whi r9 main the same due to the original operrbr actions to stop educalot flows for cetrolling a containment spray avent would begin upon recept of a low low level nicmtor lamp indication and that same action would be icpeated after the modification and include the now low low level alarms.

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MODIFICATION M20 2 88 056 DESCBlP. TION 3 "Two pen" recorders on Panel 2PM05J for the hot and cold leg temperature recorders of the reactor coolant system (RCS) are required to allow for a direct comparicon per loop for more efficient operation.  :

SAFEIY_ EVALUATION

SUMMARY

1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety no pteviously evaluated in the  ;

Final Safety Analysis Report is not increased through a review of UFSAR section 7.2.1.1.4. Modification testing will verify proper installation and i operation. No present or new occurrence or consequence was determined.

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2. The possibility for an accident or malfunction of a different type than any 3tev Analysis Report is not created ously evaluated aecause the equipment has no publicin the Final Safet[malth or oafety implication and the '

seismic qualification of the instruments and panel have been unchanged.

3. The margin of safety. as defined in the basis, for any Technical Specification, is not reduced because these recorders are not involved in or affect the margin of safety for Tech Spec 3.4.9 which describes the excessive cooldown late of the RCS. Tavg is used in the basis of this Tech Spec not the hot and cold 100 temperatures.

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MODmlCATION M20 2 89 013 DESCRIETION The subject modification provides a redundant,indept ' dent means of verifying Reactor Vessel Level Indication during refueling or reduced inventory conditions as required by NRC Generic Letter 8817," Loss of Decay Heat Removal". Inadequate determination of RCS level has been the root cause of many potentially significant loss of decay heat removal events. The modification provides ind cation in the control room, annunciation on low level, and a computer point.

SAFEIY.EVALUAIlON_

SUMMARY

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 does not affect any of the Single Failure Events or Design Basis Accidents analyzed in the FSAR.1he levelindicating system is used only when the plant is shutdown. The safety related piping is isolated from the non safety related piping during normal plant operations using ASME Section til valves and the p' pin 0 and components are seismically supported. In the unlikely event that the safety related piping f ailed during power operation, the failure would be bounded by the small break LOCA analysis (less than one square foot).
2. The possibility for an accident or malfunction of a different type than any arev;ously evaluated in the Final Safety Analysis Report is not created aecause the non safety related level Indicating system is used during refueling or whenever th9 RC level is required to be lowered. Loss of the levelindicating system would not prevent the systems from periorming their intended function since other methods are available for verifying level and proper RHR pump operation.
3. The margin of safety, as defined in the basis, for any Technical S3ecification, is not reduced because Technical Specification 3/4.9.8 required t mt at least one RHR loop be ir operation to ensure that sufficient cooling capacity is available to remove decay heat and maintain the RCS below 140 degrees F and ensures that sufficient coolant circulation is provided to minimize the effect of boron dilution incident and boron stratification. In as much as this modification provides a means of verifying adequate RHR pump NPSH thereby, improving RHR pump reliability / availability, the margin of safety as defined in Techn cal Specification is increased.

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MODIE10ATJON M20 2 89-023 DESCBIET10N This modification upgrades the existing RHR heat exchanger outlet temperature instrument loop to safety related and adds indication in the MCD.

SAEEIY_EVALUAllON_

SUMMARY

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 appropriate isolators will be installed to segregate safety from non safety components and an analysis has been performed to insure the sensing integrity of the equipment. No new accident or failure modes have been Identified or existing ones altered.
2. The possibility for an accident or malfunction of a different type than any 3rev ously evaluated in the Final Safety Analysis Report is not created

>ecause the appropriate isolators will be installed and an analysis has been performed to insure system and component integrity. No new accident or failure modes have been identified that have nol been previously analyzed.

3. The margin of safety, as defined in the basis, for any Technical Specification, is not reduced because the new Indication added to the MCB willlet the operator monitor tiw AHR system temperature from the control room. This will allow the operator to take the appropriate action upon recognition of a temperature deviation, which in effect increases the margin of safety due to early detection of the deviation, t

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MORIEICAT10N M20 2 89 030 DESCRIPllON Deletion of the autoclosure interlock (ACl) function on the RHR suction isolation valves 2RH8701 A/B and 2RH8702A/B. In place of the ACI function, an alarm will be provided on the main control board. Inputs for the alarm will be valve not fully closed (spare contacts in the limitorque operator) and RCS wide range pressure increasing (PT403/PT405).

SAFETY EVALUATION.

SUMMARY

t. 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 RHR suction relief valves are used as a means of cold overpressure protect;on. The cold overpressure protection system is designed to ensure the limits of Appendix G to 10CFR part 50 are not exceeded when one or more of the RCS cold legs are Ims than or equal to 350 degrees F. Transient analysis were performea to determine the wnrst case mass input and heat input events (refer to UFSAR, Section 5.2.2.11.2). Removal of the ACI does not impact the transient analysis. However, removal of the ACl helps ensure that the RHR r. action relief valves are available to mitigato potential overpressure transients. Additionally, removing the ACI reduces the potential for inadvertent isolation of the RHR system which can cause a Low Temperature Overpressure (LTOP) transient (reduced letdown combined with a loss of decay heat removal) while also isolating an overpressure mitigation path.

Therefore, removal of the ACI does not involve an increase in the probability of an occurrence or the consequence of an accident previously evaluated in the FSAR. In fact, removal of the ACI has a positive impact on LTOP mitigation, thereby, reducing the probability of an occurrence of an accident.

Analyses was also performed to confirm that one RHR relief valve has the capability of maintaining the RHR system maximum pressure within code limits (refer to UFSAR, Section 5.4.7.2.3). Removal of the ACI does not affect this analysis. Should a peak pressure occur while the RHR system suction isolation valves are open, the pressure effect on the low pressure RHR system would be mitigated by the RHR suction relief valves. The deletion of the ACI feature has no effect on the ability of the RHR system to survive pressure transients when the RHR system is connected to the RCS, since the RHR suction isolation valves are slow acting and no credit is taken for their actuation. Therefore, removal of the ACI will not involve an increase in the probability of an occurrence or the consequence of an accident previously evaluated in the FSAR.

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MQQlDCAItON M20-2 89 030, (continued)

The impact of removal of the ACI to Event V, LOCA outsido containment, frequency was also considered. Analysis demonstrates that the probability of the occurrence or consequences of an accident ale not increased. The dominant failure mode is rupture of the valve disc in each of the two series motor-operated valves (MOVs) in the RHR suction line when closed during normal power o aeration. This failure modo lo independent of the ACI Another less inf uential contributor to Event V frequency was found to be rupture of one valve while the other valve has f ailed open. The results demonstrate that, in this case, removal of the ACl is beneficial when compared to retalning it.

Analyses was performed to determine the impact of removal of the ACI on RHR system unavailability. The analysis indicates that the reliability of the RHR system is unchanged during RHR initiation and that it is improved during short and long term cooling. The ACI becomes more of a detrimental f actor as the length of time in which RHR is required to operate increases.

Therefore, the probability of malfunction of equipment important to sofoty as previously evaluated in the FSAR is not increased,

2. The possibility for an accident or malfunction of a different type than any areviously evaluated in the Final Safety Analysis Report is not created accause the effect of an overpressure transient will not change due to the ,

removal of the ACI. The RHR suction relief valves were designed to maintain tha RHR system aressure within design limits. Although the ACIisolates the RCS from the RF R suction relief valves on high RCS pressure, overpressure protection of the RHR system is provided by the RHR suction relief valves not by the slow acting suction isolation valves. The purpose of the Interlocks is to assure double isolation between the RHR system and the RCS when the plant is at normal operating conditions. The interlock prevents the possibility of an Event V due to operator error.

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  • 6 MODIElCATtON M20-2-d9 030 (continued)

Analyses were performod to demonstrate the impact of removal of the ACI on <

Event V frequency. RHR syntem reliability and ovoror9ssure transients. The analysis aerformed compared the results with and mthout the ACl. How9ver, i the resulls were contingent upon providing an alarm to aiort the operator that ,

a RCS RHR series suction isolation valve (s) is not fully closed and that  ;

double Isolation is not being maintained. The modificatlon will not impact the i opening circuitry, nor will it effect the MOV position Indication in the control room. The setpoint fr.r the alarm will be within the range of the open permisolve seipoint ressute and the RHR system design pressure minus the RHR aump setpolni oressure a the R -iR pump head pressure. nd the RHR system design pressure minusO the operator to take the necessary actions to close the open valve (if it is not closed), or if this is not possible, to return to the safe shutdown mode of operation. The analysin performed indicates an overallincrease in safety due to the removal of the ACl, implementation of the modification, and procedural changes. Therefore. the possibility of a new or different kind of accident from any previously evaluated is not created.

3. The margin of safety as defined in the basis for any Technical Specification is not reduced because deletion of the ACI has no effect on the ability of the RHR system to survive presoure transients when the RHR system is ,

connected to the RCS, since the RHR suction isolation valves are slow acting and no credit is taken for their actuation. However, removal of the ACl helas ensure that the RHR suction relief valves are available to mitigate potentia' overpressure transients. Additionally, removal of the AClimproves RHR system reliability. Therefore, the margin of safety is not reduced, in f act, the margin of safety is increased.

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i MODJEICATION M20-2 00 014 DESCRIPTION This modification replaces the existing opposite division power as D.C. operated " fall as left" solenoid operated valve in each train of the hydrogen monitoring system with a D.C. operated fall open" solenoid operated valvo. With this new configuration in place a loss of power in one ESF division will not leave a failed close valve in the opposite division.

SAEEI1EVALUATIOttSitMhMBY

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 equipment has no function in an accident other than contalnment isolation. This function has not changed. The containment isolation function will still occur in the event of a safety injection actuation. The consequences of the accident will not be increased because containment integrity is maintained by the other isolation -

valves in series. This equipment performs a containment isolation function and does not affect other plant systems. The 3robability of a malfunction of equipment important to safety is not increasec because all other equipment is unchanged and the valve still performs its originalisolation function. The consequences of a malfunction of equipment important to safety does not increase because the equipment functions and systems remain the same except for the valves being modified which still perform containment isolation function.

2. The possibility for an accident or malfunction of a different type than any arev ously evaluated in the Final Safety Analysis Report is not created 3ecause the UFSAR assumptions for accident assumes that all containment isolation valves close and remain closed throughout and after the accident valves manually reopened. The modified valves will " fall open" u aon a loss of a D.C. ESF bus which is different than previously assumed. T1e-containment isolation will still be maintain!ng through the other valves in series. This configuration hat, been submitted to the NRC, evaluated by NRC and found acceptable. On April 19,1991 the NRC issued the supplemental Safety Evaluation which accepts this modification.
3. The margin of safety, as defined in the basis, for any Technical S3ecification, is not reduced because valves still provide containment isolation Lunction. In the event of a loss of one division of DC ESF power containment isolation is maintained and hydrogen monitoring is still achievable.

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NUCLEAR WORK REQUESI NWR A48630 l

DESCBIPTJON This Nuclear Work Request Inhibits the close intercept valve (CIV) function by adding a

,umper in the DEH control panel. This prevents the inadvertent closing of the turbine niercept valves which results in generator output transients.

SAEET_Y_ EVALUATION _

SUMMARY

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 close interceptor valve is not referenced in the accident section of the UFSAR and therefore will not have an affect.
2. The possibility for an accident or malfunction of a different type than any areviously evaluated in the Final Safety Analysis Report is not created '

]ecause the rel! ability of systems required for safe shutdowns described in the FSAR are not affected as no functional changes are required to these systems.

3. The margin of safety, as defined in the basis, for any Technical Specification, is not reduced because there are no Tech Specs affected by this Nuclear Work Request.

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SGAUNG!.SEIERINLCHANGE ,

SSCR 90-041 l DESCBIEllON Change the temperature switch 2(1)TS-VE003 for the miscellaneous electrical equipment room (MEER) exhaust fan 2(1)VE0SC so that SW-1 will close on the increase at 80 degrees F and so that SW-2 will open on the decrease at 70 degrees F.

The old setpoints for SW 1 and SW-2 were 90 degrees F and 60 degrees F, tespectively.

SAEEIX EVALVAIlotLSUMMABY l

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 new high temperature setpoint is a more conservative valve for the system to switch into equiouts'de airMEER.

pment in the mode, and the chan0e will provide greater protection for t

2. The possibility for an accident or malfunction of a different type than any areviously evaluated in the Final Safety Analysis Report is not created accause the basic operation of the system is unchanged, therefore, no different types of malfunctions are introduced.
3. The margin of safety, as defined in the basis, for any Technical Specification, is not eoiend because the ability of the system to maintain the temperature limits of Tech Spec 3/4.7.12 is enhanced by this change.

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4 4 TEMPORARY ALTERATION TA 912-032 DESCRIPTION Temporary Alteration to provide station fire arotection water to the sprinkler systems of temporary outage related structures to be p aced on the Turbine Deck during A2R02.

SAFETX EVALUAT!ON .

SUMMARY

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 no accident analysis is affected.
2. The possibility for an accident or malfunction of a different type than any neviously evaluated in the Final Safety Analysis Report is not created aecause fire protection has no safe shutdown tunctions as set forth in the Fire Protection Report, Section 2.4, as referenced by section 9.5.1 of the UFSAR. Fire Protection has no affect on safe shutdown, but serves to limit the consequences of a fire. The level of fire protection is not reduced. Fire Protection is not governed by the Technical Specifications.
3. The margin of safety, as defined in the basis, for any Technical Specification, is not reduced because Fire Protection is not applicable to the Technical Specifications.

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  • l UFSARRROCEDURAl. CHANGE UFSAR DRP 3-005 RESOBIPllON This UFSAR change to section 15.0 reflects revision 1 A of the Westinghouse Owners Groua emergency procedures. The operator actions were summarized for certain even! s.

SAFETY EVALUATION

SUMMARY

1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment impor1 ant to safety as previously evaluated in the Final Safety Analysis Reporils not increased because the changes do not affect the accident analyses in cha ater 15. There is no change in plant operation. The revision provides e artfication of Westinghouse Owners Group procedure guidance.
2. The possibility for an accident or malfunction of a different type , a any 3rev ously evaluated in the Final Safety Analysis Report is not created -

accause the change provides a description of station procedures used as additional requirements for operator action. Operator action is still from approved station procedures, which undergo a 10CFR50.59 safety evaluation prior to use. ,

3. The margin of safety, as defined in the basis of any Technical Specification, is not reduced because the change does not affect the basis for any Technical Specification.

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UFSAR PROCEDURAL CHANGE UFSAM DRP 3 011 DESCRIPTION This UFSAR change to section 6.2 involves changing the isolation time from 10 minutes to 30 minutes for the Auxiliary Feedwater (AF) flow from a depressurized steam generator. An alternative method was developed to isolate Auxilimy Feedwater flow in the event of a main streamline break. This was necessary because the AF013 valves in their present condition would not develop enough thrust to ensure full closure under worst caso dP conditions.

SAFETY EVALUATION

SUMMARY

1. The probability of an occurrence or the consequence of an accident, or mnl function of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not ine: eased because the valves in question perform a mitigative function. Under normal conditions the valves are open with flow secured in the line. The AF system is in a standby condition. In such a condition, the inability of the valves to close under full pump dP is not a f actor. The condition of the valves is unrelated to the probability that a chapter 15 event will occur. If a streamline break insido containment occurs.

AF would be secured within 10 minutes of the accident initiation. If Instrument Air is available the AF005 valve can be used to isolate the f aulted steam generator if I A is unavailable the pump in the affected train can be isolated until the AF013 valve can be closed, and then restarted. Alternative steps can be taken to isolate AF

2. The possibility for an accident or malfunction of a different type than any

>rev ously evaluated in the Final Safety Analysis Report is not created aecause no new equipment is being introduced and installed equipment is not being operated in an abnormal manner. Procedures outline methods to ensure that the AF013 valves are subjected to a dP large enough to adversely affect their ability to isolate.

3. The margin of safety, as defined in the basis of any Technical Specification, is not reduced because several options are available to ensure that the assumptions of the limiting analysis are met, in the case where a delay in AF isolation to a f aulted generator is postulated, the peak pressure is well below the design pressure, and redundant means of pressure control are available.

This will ensure that the margin of safety is maintained.

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UFSARfROCEDURALCHANGE UFSAR DRP 3 022 4

'9SCBlP_TMN i nis UFSAR change to section 9.1 allows the new fuel elevator to be used for moving objects other than new fuel assemblies, including irradiated fuel assemblies.

SAEEIY. EVALVAllGN.S_UMMARY

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 minimum shielding distance specified remains the same. Irradiated assemblies are inherently less reactive than new ones for which the elevator is already designed. A i failure involving a dropped fuel assembly is bounded by existing analyses.
2. The possibility for an accident or innifunction of a different type than any 3reviously evaluated in the Final Safety Analysis Report is not created aecause the change addresses protec'ive t measures to handle uncontrolled upward movement of an irradiated assembly. Other postulated failures are bounded by the analyses of a fuel handling accident n Section 15.7.
3. The margin of safety, as defined in the basis, for any Technical Specification, is not reduced because the new fuel elevator is not addressed in the basis for any Technical Specification.

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UESAFLPROCERUBAL CHANGE UFSAR DRP 3 023 DESCBIP_IlON  ;

This UFSAR change to Table 7.5 2 involves changing the full out axial park position of the rod cluster control assemblies (RCCA) from 228 steps to 231 steps. This is to mitigate possible fretting wear on the RCCA rodlet cladding surf ace caused by flow induced vibration between the rodlets and the gulde cards.

SAEEIY_EVALUEDORS_UMMARY ,

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 Recorils not increased because the rod drop time increase due to the addihonal 3 steps is negligible. Rop drop times will now be verified at 231 steps.
2. The possibility for an accident or malfunction of a different type than any arev ously evaluated in the Final Safety Analysis Report is not created accause the RCCAs are not being operated in a different manner. The change in the bank separation as well as bank overlap distancos resulting from repositioning does not violate the rod insealon limits.

3.

The is notmargin reduced of safety, because theas roddefined insertioninlimitsthe (RIL basis of any) given Technical Specificatio in Technical Specification 3.1-1 only limits the control bank's positions in relation to core 3ower and not with respect to other specific bank positions. Control banks at 3raidwood are typically positioned well above the allowed RILs. The change increases the margin to the RlLs for all control banks except for the lead bank, thus providing greater shutdewn margin. The impact of a one step deviation in the relative positions of two sequential control banks is negligible.

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  • l UESAR. PROCEDURAL CHANGE UFSAR DRP 3 037 DESCRIP_ TION Th!s UFSAR change to Ser,tions 9.1 and 15.7 allows use of a secondary restraint on the fuel building crana M meet the single f ailure proof guidelines in NUREG 0012 and 0554. This allows thu 125 Ton load block (hook) to remain on the crane during all phases of fuel handling activities.

SAEETX.EVALUATIOR

SUMMARY

1. The probability of an occurrence or the consequence of an accident, t  ;

malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because the restriction in section 15.7 to keep loads from travelling over the spent fuel storage area is maintained. This prevents objects from dropping on spent fuel.

2. The possibility for an accident or malfunction o! a different type than any 3rev ously evaluated in the Final Safety Analysis Report is not created aecause the secondary restraint is passive and there is no effect on plant operation, if a single failure were to occur, the secondary restraint would be taut to ensure no acceleration forces are possible. No new f ailures are 3 created because single failure proof guidelines in NUREG-0612 are used.
3. The margin of safety, as defined in the basis, for any Technical Specification, is not reduced because the limits in Technical Specification 3/4.9.7. are maintained. The secondary restraint meets the single failure proof criteria, the' 'Se load block is no longer classified as a load or heavy load and is al! owed over the spent fuel pool. The bases for fechnical Specification 3/4.9.7 are unchanged,

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UESAR PROCEDURAlmCHANGE UFSAR DRP 3-042 DESCHIPllON This UFSAR change involves the addition of clarifications to sections 7.1 and 7.3 of the UFSAR to more completely describe testing the reactor trip and engineered safety features actuation system.

SAEETJJVALUAIlON SUMMABY

1. The probability of an occurrence or the conseque' s of an accident, or malfunction of equipment impor1 ant to safety as previously evaluated in the Final Safety Analysis Report :s not increased because there are no changes to plant operation or testing requirements or methods. Any f ailure of the system to maintain valve position has been analyzed previously.
2. The possibility for an accident or malfunction of a different type than any oreviously evaluated in the Final Safety Analysis Report is not created because maintaining the valves in a closed position is assumed in the basis for the analysis and that the Technical Specification lequirmnents are initial satisfied. - No new f ailures are introduced.
3. The marain of safety, as defined in the basis, for any Technical Specification.

is not reduced because the valves remain in their safoguards actuated condition listed in LCO 3.6.1.7. The Technical Specification manualinitiation surveillance requirement for trip actuating device operational tests for phase A, phase B, and manual safety injection is consistent with LCO 3.6.1.7.

Because of this, there are no interactions or adverse effects.

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