ST-HL-AE-4268, Annual 10CFR50.59 Summary Rept for South Texas Project, Units 1 & 2, Describing Changes,Tests & Experiments Associated W/Required Safety Evaluations

Annual 10CFR50.59 Summary Rept for South Texas Project, Units 1 & 2, Describing Changes,Tests & Experiments Associated W/Required Safety Evaluations

ML20126A983
Person / Time
Site:	South Texas
Issue date:	12/16/1992
From:	Rosen S HOUSTON LIGHTING & POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ST-HL-AE-4268, NUDOCS 9212210326
Download: ML20126A983 (45)

Text

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The Light company linuston Lighting & l'ower _S ""'h I***' I"I"' 'I*'iric cenerating station P.o.Ison289 wadsworth Texas 77483 December 16, 1992 ST-HL-AE-4268 File No.: G20.01 G21.01 10CFR50.59 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Annual 10CFR50.59 Summary Renort Pursuant to 10CFR50.59, Houston Lighting & Power company (HL&P) and submits experiments this annual report which describes changes, tests, required safety evaluations.associated with the South Texas Project and the Note that there are gaps in the numerical sequence of the attached summaries. These represent safety evaluations that have been cancelled, were incomplete when this report was prepared, were submitted with the previous annual report, or have yet to be implemented.

If you should have any Mr. P. questions, please contact L. Walker at (512) 972-8392 or myself at (512) 972-7138..

M a S. L. Rosen Vice President, Nuclear Engineering PLW/ag Attacraents : Summary of Unreviewed Safety Question Evaluations 1

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llouston Lightiag & Power Company South Texas Project Electric Generating Station ST-HL-AE-4268 File No. G20.01, G21.01 Page 2 j cci l

Regional Administrator, Region IV Rufus S. Scott .

Nuclear Regulatory Commission Associate General Counsel '

611 Ryan Plaza Drive, Suite 400 Houston Lighting & Power Company Arlington, TX 76011 P. O. Box 61067 Houston, TX 77208 George Dick, Project Manager U.S. Nuclear Regulatory Commission INPO Washington, DC 20555 Records Center 1100 Circle 75 Parkway J. I. Tapia Atlanta, GA 30339-3064 Senior Resident Inspector c/o U. S. Nuclear Regulatory Dr. Joseph M. Hendrio Commission 50 Bellport Lane P. O. Box 910 Bellport, NY 11713 Bay City, TX 77414 D. K. Lacker J. R. Newman, Esquire Bureau of Radiation Control Newman & Holtzinger, P.C. Texas Department of Health 1615 L Street, N.W. 1100 West 49th Street Washington, DC 20036 Austin, TX 78756-3189 D. E. Ward /T. M. Puckett Central Power and Light Company P. O. Box 2121 Corpus Christi, TX 78403 J. C. Lanier/M. B. Loc City of Austin Electric Utility Department P.O. Box 1088 Austin, TX 78767 K. J. Fiedler/M. T. Hardt City Public Service Board P. O. Box 1771 San Antonio, TX 78296 Revised 10/11/91 L4/NRC/

Attachment 1 ST-ilL-AE-4268 Unreviewed Safety Question Evaluation #90-179 Subj ec t: Boric Acid Tank (BAT) Diaphragms

Description:

Delete reference to the presence of diaphragms in the Boric Acid Tanks because they are not installed and are not considered to be necessary as evaluated per RFA 90-1458.

UFSAR Section 9.3.4.1.2.5 will be changed as indicated.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of _

occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. Operating without BAT diaphragms could not increase the amount of dissolved oxygen (D0 2 ) in the makeup boric acid solution and thus into the RCS. D02 level in the RCS is maintained within Technical Specification limits. D02 concentration greater than the steady state but less than the transient limit will have no significant effect. Increased D02 concentration is not an abnormal operational transient or design basis accident nor is D02 concentration an initial condition or pertinent parameter for any plant condition not evaluated in Section 15 of the UFSAR.

Increased corrosion would not result from deletion of the boric acid tank diaphragms because the oxygen hydrogen recombination reactions occurring during normal plant operation keep the residual D02 in the RCS essentially zero. An increase in RCS D0 2 level does not play a direct role in mitigating the radiological consequences of an accident nor would it result in an increase in _

an accident dose.

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

No. D02 level in the RCS is controlled and maintained within Technical Specification limits to minimizo corrosion and the potential for leakage and failure, and to ensure structural integrity over plant lifetime Maintaining D02 less than steady state limits provides adequate margin to ensure structural integrity over plant lifetime. No other adverse effects are postulated.

USQ\92fM.001

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Attachment 1 'l '

ST-HL AE-4268 I

1 Unreviewed Safety Question Evaluation #90 179-(Con't)- j

3) Does the subject'of this evaluation reduce the margin of safety as defined in the basis for any technical specification?

No. Removal of the BAT diaphragms does not change or'otherwise impact the limit for D02 in the RCS. -There is no margin of safety directly related to use of diaphragms in'the BAT.

Based upon the above, there is no unreviewed safety question.

Approved: 5/1/92 ,

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USQ\ 921H.001

-Attachment 1 ST-HL AE-4268 9

.Unreviewed Safety Question Evaluation #91-011

Subject:

Reload Safety Evaluation (RSE) for Unit 1-Cycle 4

Description:

This change incorporates licensing amendments for the Technical Specification MTC limit and source terms into the Unit 1 Cycle 4 RSE. This will allow continued operation of-Unit 1 Cycle 4 from 425 Effective Fuel Power Days of Operation (EFPD) to the total. cycle burnup of 498 EFPD.

Cycle 4 is designed to have an approximate full power cycle length of 456 EFPD plus a power coastdown of 42 EFPD, Seventy-six burned assemblies will be discharged from Cycle 3. These will be replaced with 76 fresh assemblies

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with 28 having a 3.60 w/o U235 enrichment, and 48 having a 3.80 w/o U235 enrichment. The Cycle 4 core has 1376 burnable poison rods (BPRs).

Safety Evaluation:

1) Does the subject of this evaluation #,acrease the probability of ~

occurrence or the-consequences of aa accident or malfunction of equipment important to safety previously evaluated in the safety analysis. report?

No. The proposed change supports the. fuel design _for Unit 1-Cycle 4 to the -total burnup (including coastdown) of 498 EFPD.

The UFSAR_ Chapter 15 analyses remains bounding for the fuel.

design, and there is no change in the radiological _ dose. ' Based upon the previous discussions, changes to the safety analysis are addressed by other USQEs, bounded by the existing safetyf analysis, or the license wa's amended and accepted,by'the NRC.

2) Does the _ subject of this evaluation createithe. possibility.'for an-accident or malfunction of a different-type than any_ evaluated previously in:the safety analysis report?.-

-No. The proposed change is-bounded by-existing analyses in the safety analysis report to the_ total _ cycle burnup _(including coastdown) of 498 EFPD.

3 ) -- Does the subject of this evaluation reduce' the ' margin of safety'as

defined _in the basis for any technical specification? : .

No. Since the UFSAR Chapter 15 analysis 1rerains bounding.to the total cycle _burnup (including coastdown) of f 498: EFPD, _ there f is , no

reduction in-the margin of safety as defined in the Bases'for any; Technical Specification.

Based upon the above, there is no unreviewed safety question._

Approved:-7/3/92 USQ\92!W 001-

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' Attachment 1--

ST HL AEa4268 Unreviewed Safety Question Evaluation #91014

Subject:

Motor-operated Valve (MOV) Breaker Ratings q 1

Description:

The' size of circuit breakers'in Motor Control Center MCC-E1A1, ElB1, and E1Cl-are to be changed from 10 amps to amps,- with the proposed setting increase from 84 amps: to 105 amps for the subject Motor-Operated Valves. This increase is to' avoid the nuisance trip during MOV diagnostic test' (MOVATS). i i

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of ,

occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. This change enhances the surveillance and maintenance. process by providing a reasonable setting to avoid the nuisance trip, while preserving the protection function of the circuit breaker to the whole circuit as well as the penetration conductor.- _j

2) ' Does the subject of this evaluation create the possibility for an -

accident or malfunction of a~different type than any evaluated previously in the safety analysis report?

No. The backup circuit breaker rated at 70 amps is: designed.to sustain the short circuit and Locked Rotor Current and to-protect-the penetration conductor. The new setting will not contradict with the-design intent and it will comply with1RGJ1.63.

3)- Does the subject of this evaluation: reduce- the: margin of safety as defined in the basis for any technical specification?

No. - The backup circuit breaker-rated at 70 amps is designed to sustain the short circuit and Locked Rotor Current and to protect the penetration conductor. The new setting will not contradict.

with the design intent and it will comply with RG 1.63.

Based upon the above, there 'is no unreviewed safety question.

Approved: 9/27/91 USQ\92FW,001 m

Attachment 1

! ST-HL AE 4268 Unreviewed Safety Question Evaluation u91-030

Subject:

Hydrogen Analyzer

Description:

Correct the demarcation for the hydrogen analyzer mode select switch (i.e. standby / operate). This also implements the necessary changes to provido physical hydrogen sample point location indication on Main Control Board control MCB ZCP002.

Safety Evaluation:

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1) Does the subject of this evaluation increase the probsbility of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. Revising the labeling indication does not alter the function or operability of the H2 Monitoring System. The change provides control room operators with better clarity of H2 monitoring controls and the as-built condition of the H2 sample points.

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

No. Operability and function of the existing systems are not being changed. Monitoring of H2 via the as-built sample points is adequate. Other enhancements are to aid the operators in determining the correct control functions for MCB instrumentation.

The present Hz sample points provide sufficient monitoring of H2-

3) Does the subject of this evaluation reduce the margin of safety as defined in the basis for any technical specification?

No. The change does not modify the actual function or operability of the system.

Based upon the above, there is no unreviewed safety question.

Approved: 9/27/91 USQ\92W.001

Attachment 1 ST.HL AE 4268

.Unreviewed Safety Question Evaluation #91 031

Subject:

Fuel Oil Supply

Description:

This change to the UFSAR clarifiesLthe method by which 1the seven day supply'of fuel oil is calculated; Each fuel oil-storage tank (POST) has sufficient capacity to operate the diesel generator at engineered safety features load requirements for at least seven days. Previously, the load ~

was described as being the " maximum connected load" Land "at-continuous rating."

Safety Evaluation: -

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety-analysis report?

No. The sizing of the FOST meets all licensing commitments and ,

regulatory requirements.

2) Does the subject of this evaluation create the possibility.for~an-accident or malfunction of a different type.than any evaluated 1 previously in the safety analysis. report?

No. The sizing of the POST meets alll licensing commitments and-I regulatory requirements.

3) Does the subject of this evaluation reduce the margin of safety as-defined in the basis for any' technical specification?.-

No. Technical Specification 3.8.1.1 requires;that 60,500 gallons ,

be maintained in the standby diesel generator fuel oil storage l tank.

Based upon the above, there is no unreviewed safety question.

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Approved: 9/17/91 g

'USQ\92PW,001

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Attachment l' ST-HL- AE 4268 - ,

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Unreviewed Safety Question Evaluation #91-034

Subject:

Spray. Additive Tank Deletion

Description:

The sodium hydroxide containment spray additive tanks and associated piping are to be_ abandoned in place and replaced with_a system of stainless steel baskets filled.with -i trisodium phosphate located on the floor of the containment.

Safety Evaluation: ,

1) Does the subject of this evaluation increase the probability of - '

occurrence or the consequences of an' accident or malfunction of a equipment important to safety previously evaluated in the safety-analysis report? ,

No. The proposed modification will have the same-function as the present spray additive tank system which is to mitigate-the.

effects of a loss of coolant accident. The integrity of the containment spray system is not compromised. This? change was.

previously addressed in letter dated October 30, 1990.

(ST-Hl.- AE-3378) . The~Technica1' Specification change-associated with this change has been approved by-the NRC. *

2) Does the subject of .this evaluation create the po'ssibility' for an-accident or malfunction of a different type than any evaluated-previously in the safety analysis report?

No. This change was previously addressed in.ST-HL AE-3378. -The- 4 Technical Specification change associated with this change has been approved by the NRC.

3) Does the subject of this evaluation. reduce the margin of safety =as

• defined'inlthe basis for'any technical specification?..

No. This change was previously addressed in ST-HL-AE-3378. LThe ,

Technical Specification change associated -with _ this_ change. has

. been approved by the NRC,

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- Based upon the above, there is no.unreviewed safety question.

Approvedi'8/30/91 ".

4 USQi92IV.001

Attachment'l ST ilL AE-4268 Unreviewed Safety Question _ Evaluation #91-042 Subj ect: Removal of Contact Relay

Description:

A Potter & Brumfield mercury wetted contact relay associated with the Proteus computer is-to be removed because of chattering, causir.g the Proteus CPU to load excessively, resulting in multiple system halts. Removal is a temporary _ .

condition. ., .

Safety Evaluation:

1) Does the subject _ of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety-analysis report?

No. Pulling the relay opens a circuit which provides one of the inputs going to the rod position deviation' monitor; The other input is still capable of alarming (indicating),_the rod deviation annunciator. As a second check, operations will monitor digital-rod position indication (DRPI) system and demand position-indication system for a rod deviation per Technical Specification. J 4,13.2.1 There is no feedback provided by the circuit, so pulling' the' relay will-have no effect on the rest of the.DRPI system,

-2) Does the _ subject of this evaluation create the _ possibility for an -

accident or' malfunction of a different type than any evaluated previously in the safety analysis report?

No. Pulling the relay disables _the: automatic-updating of the-affected rod position and rod deviation computer points and. alarm.

The capability remains to manually po11'the affected-computer points which will update the associated-information and alarms.

Technical Specification 4.1.3;2: addresses the case should the-computer point:for rod position deviation monitoring be j inoperative.

3) Does the subject of this evaluation reduce the margin of safety _ as-defined in the basis for any technical specification?

No. The Proteus computer is-not specifically addressed _in the Technical Specifications as a means of monitoring or alarming rod position or rod deviation.

Based upon the above,.there is no unreviewed safety question.

, Approved: 8/28/91 q IfsQ\921W.001

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Attachment 1 -j ST-HL-AE 4268 Unreviewed-Safety Question Evaluation #91-043

Subject:

Security Force Coordinator

Description:

The position of " Security Force Coordinator" is to be changed to " Shift Supervisor on Duty and/or Health Physics Supervision" for control of keys to barriers and doors to areas adjacent to the fuel transfer tube in the FHB.

Safety Evaluation:

1) Does the subject of this- evaluation increase the1probab111ty of occurrence or the consequences of an accident or malfunction' of equipment important to safety previously evaluated in the safety analysis report?

No. The specific individual (s) responsible for administrative control of locked high radiation area keys does not increase the challenges to safety systems assumed to function in the accident <

analysis such that uafety system performance is degraded-below the-design basis. There is no impact cui the public health and safety.

2) Does the subject of this evaluation cr' ente the possibility for 'an accident or malfunction of a different type than any. evaluated previously in the safety analysis report?

l No. The-individual (s) responsible for administrative control of .

locked high radiation area keys does not create-a challenge to the-safety system assumed to function in the; accident analysis. such that safety system performance-is degraded below the design basis.

3) Does the subject of this evaluation reduce the . margin of safety as defined in the basis for any technical specification?-

l No. The change in responsibility for administrative control 'of locked high radiation area keys meets the requirements of-Technical Specification 6.12.

i Based upon the above, there is no unreviewed safety _ question.

l, Approved: 7/12/92 USQ\92IV.001 -

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Attachment-1 y ST.HL AE-4268 1 Unreviewed-Safety Question Evaluation #91-044 '

Subject:

Reactor Pressure Vessel (RPV) Venting and Reactor Coolant System

-(RCS) Degassing i

Description:

A new vent line f om the RPV to the pressurizer will' allow better control during RCS filling and draining operations, and reduce the time and effort required for this operation. l The new degassing path from the Pressurizer Relief Tank (PRT) and the Reactor Coolant Drain Tank (RCDT) to the RCSVDS/CWPS will allow continuous purging and degassing of the RCS, and reduce the time and effort required for this operation.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of- ,

occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. These lines are used only during filling-and. draining operations in modes 5 and 6 only and not during normal. operation as for safe shutdown of the plant. These modifications do not affect the safety functions of the associated systems.

2) Does the subject of 'this evaluation create the' possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

No. The RPV venting and RCS degassing lines are not required.

during normal operation and are to be used during-draining and' filling operations.in modes 5 ar.d 6 only, Fuel' handling accidents

-are not affected by RPV venting or RCS degassing' operations.

3)- Does the subject of this evalu'ation reduce the margin of safety'as defined in the basis for any technical specification?

No. Technical Specifications do not. govern venting of the=RPV, or-degassing of the RCDT or PRT. Connecting.the vapor spaces of the l,  ;RPV and the pressurizer, or degassing of the- RCDT/PRT to the-RCVDS/GWPS.will not affect any of the Technical-Specifications.

Based upon the above, there is no unreviewed safety question.

Approved: 9/17/91

- USQ\ 92Pd. 001

Attachment if ST-HL AE 4268 Unreviewed Safety Question Evaluation #91-045 Subj ect: Reactor Coolant System (RCS) Vent Path

Description:

This modification provides for an alternative means of venting the RCS through a vent path. The new RCS vent' path provides a simpler alternative for satisfying specifications and precludes the need for removing the bonnet of the pressurizer PORV block valve.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of.

occurrence or the consequences of an accident or malfunction of-equipment important to safety previously evaluated in the safety analysis report?

No. The accidents evaluated in the UFSAR are not affected by the method used for satisfying the requirements of Technical Specifications 3/4.4.9.3, 3/4.8.1.2,-3/4 8.2.2 and 3/4.8.3.2 .

venting the RCS through an area greater than or equal to 2 square-inches. Addition of the 2-1/2" RCS vent path to the prnssurizer -'

spray piping meets the requirements of ASME Section III Subsection NB and NC and the South Texas Project Design Criteria. The vent piping modification does not invalidate the existing analysis of.

record for the pressurizer spray piping.

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than'any evaluated previously in the safety analysis report?

No. The new RCS vent path and the required supports are designed and installed to the same design conditions, Ccdes and-Standards, and Specifications ,as the-present-path and the RCS. Manual valves

! are not included in the Failure Modes and_ Effects Analysis _(FMEA) presented in the UFSAR. Therefore, FMEA of the RCS, as presented in the UFSAR,- is not affected.

i 3). Does the subject of this evaluation reduce the margin _ of safety as defined in the_ basis for any-technical specification?

No. Technical Specifications require that an.RCS-vent path-of greater than or= equal to a 2 square-inch area be available. The' new RCS vent path proposed =in this modification provides an-

. alternative.1CS vent path of greater than 2 square inches.

Based upon the above, there is no unrevievid! safety question.

Approved: 9/27/91 USQ)021W,001

l Attachment 1 ST-ilL AE-4268 Unreviewed Safety Question Evaluation u91-046

Subject:

Essential Cooling Water System

Description:

Six air inlet check valvi assemblies are to be added to the Essential Cooling Water (ECW) system piping to relieve water hammer pressures that occur in the ECW pump stoppage. Also, the standby diesel generator intercooler inlet and outlet expansion joints are being deleted, and replaced with a spool piece.

Safety Evaluation: _

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunctio of equipment important to safety previously evaluated in the sa sty analysis report?

No. The changes are designed to meet the original requirements for the ECW system as required by the ASME Code, and applicable standards and specifications. The stress analysis considered all the required design loading combinations including water hammer.

Addition of the air inlet check assemblies and d4 ?.etion of the expansion joints will not affect the function or operability of the ECW system. The design basis for the ECW system is not changed by this modification.

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

No. This change does not establish any system configuration which reduces or bypasses any of the protective barriers of the plant.

The change does not reduce the reliability of the ECW system or the standby diesel generator and its components. This change does not result in any system or component operating modes that are beyond tha capacity and qualification of the component.

3) Does the subject of this evaluation reduce the margin of safety as defined in the basis for any technical specification?

No. Technical Specifications address operability of the ECW system; however, this change does not affect the operability of the ECW system.

Based upon the above, there is no unreviewed safety question.

Approved' -/6/92 USQ\921V.001

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Attachment 1 ST-HL-AE 4268 Unreviewed Safety Question Evaluation #91-049 Subj ect: Spent-Fuel Pool Boiling Dose Analysis

Description:

This updates the UFSAR spent fuel pool boiling analysis to incorporate the effects of full core offloads rather than one third core offloads.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of ,

occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated- in the safety analysis report?

No. The possibility of a spent fuel boiling accident has already' been addressed. Changes in assumptions made in evaiusting the radiological consequences of the accident are bounded by those given in the SER. Since this is a limiting analysis for offsite doses only, there is no impact on equipment qualification, Assumptions in malfunction of equipment important-to-safety have not changed.

2) Does the subject of this evaluation create the possibility for an -

accident or malfunction of a different type than any-evaluated previously in the safety analysis report?

No. The change proposes using a different assumption than previously used which analyzes the_ effects of a full core offload

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in the spent fuel: boiling offsite dose calculation. .The1 change in -

assumption does not create the possibility of- a;different accident, or failure mode.

3) Does the subject of this- evaluation reduce the margin of safety as defined in the basis for any technical specification?

No. The acceptance limit is-30 rem-thyroid as defined in the SER.

However, the exclusion and low population zone boundary doses are below the acceptance limit of 30 rem-thyroid.

p Based upon the above, there is no unreviewed safety question.

L -Approved: 9/17/91 i.

i USQ\92IW.001

Attachment 1

> ST HL AE-4268 l

Unreviewed Safety Question Evaluation #91 050

Subject:

Unit 1 Cycle 4 Core Operating Limits Report (COLR)

Description:

The COLR has been expanded to satisfy Amendment 27 to the Unit 1 Technical Specifications in the areas of limits for HTC, peaking factor limits, rod insertion limits, etc. This COLR revision implements Amendment 27 to the Unit 1 Technical Specifications which relocates several LCO limits to the COLR.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. Since none of those limits are changed, there is no real change to the facility. All of the COLR limits are generated using an NRC-approved methodology. Because the action statements and surveillance requirements remain unchanged, there is no change in operation of the plant. Since the LCO limit values are not changed, safety analyses remain bounding.

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

No. See response to (1).

3) Does the subject of this evaluation reduce the margin of safety as defined in the basis for any technical specification?

No. See response to (1).

Based upon the above, there is no unreviewed safety question.

Approved: 9/27/91 UsQis2Ew.001

Attachment li ST HL AEa4268 l l

_Unreviewed Safety Question Evaluation #91-053

Subject:

DC Power Supply Dropping Resistor

Description:

STP experienced a failure of the DC Power Supply Droppingf -

Resistor in the Woodward Governor on inril 6, 1989 at which time the voltage dropping assembly was-replaced._ The examination concluded that the failure of the resistor was a-random occurrence attributable to either improper handling ~ ~!

or stresses induced during the manufacturing _ process.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment impor..".nt to safety previously evaluated in'the safety analysis report No. If the Woodward Governor de power supp'y l dropping resistors are defective, there is no effect on the probabilityLof occurrence. The resistors have no role-in accident initiation.

The resistors are components of equipment used for accident . .

mitigation. However, the resistors, governor, and standby diesel generators (SDC) are expected to operate as required for accident mitigation. The apparent primary contributors to-resistor failure-elsewhere'(heat and voltage) are not_significant'at STPEGS. In addition, if the resistors were to cause SDG failure, such failure has been considered in accident analysis.

2) Does the subject of this evaluation create the: possibility for an accident or malfunction of a different type than-any evaluated previously in the safety analysis. report?

No. The- resistors are not a component with potential for: accident initiation. While-the resistors would affect functionality-of-the' SDGs, their failure has already been-_ considered.-

3) Does 'the subject of this evaluation reduce the margin of safety'as

- - - defined in the basis Ear any_ technical _ specification?

No, These resistors'are not governed by. Technical Specifications.

While the SDGs are governed,.the_ assumptions for them have not changed.

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! Based upon the above, there is no unreviewed safety question.

Approved: 9/27/91 L

USQ\92 W .001 t

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ST-HL. AE 4268:

Unreviewed Safety Question: Evaluation **91-054

Subject:

Reload Safety Evaluation (Unit 2, Cycle 3)

Description:

The fuel loading pattern for Unit 2 Cycle 3 differs from that described in the UFSAR. The results of this evaluation-and the COLR evaluation show that the Unit 2 Cycle 3 fuel-design will not constitute a USQ for all modes of operation up to 354 Effective Full Power Days (EFPD). +

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report? ,

No, The proposed change supports the fuel design for Unit 2-Cycle 3 up to 354 EFPD of operation. The subject of this USQE does not change any plant equipment (other than-the fuel) or-procedures. Changes to the safety analysis are addressed by other USQEs, bounded by the existing safety analysis, or the license was amended and accepted by the NRC. The Chapter-15 analyses remain bounding for Unit 2 Cycle 3 up to 354 EFPD of operation,1so there is no-change in the radiological dose due to accidents. The burnup assumed for the source terms of safety-related systems _and components is bounded for operations up to 388 EFPD.

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2) Does the subject of this evaluation create the possibility for an-accident or malfunction of a different type than any. evaluated..

previously in the safety analysis report?

No. -The proposed change is bounded by thelexisting analyses in-

the safety analysis report up to 354 EFPD of operation.

3) Does the subject of this evaluation- reduce the margin of safety as defined in the basis for any technical specification?

No. Since the safety analysis remains bounding for.up to 354 EFPD of operation, there is no reduction in the margin ~of safety as defined in the bases for any-Technical Specification.-

Based upon the above', there is no unreviewed safety question.

Approved: 10/07/91-USQi92N.001

Attachment i i

ST HL AE 4268 Unreviewed Safety Question Evaluation st91-055

Subject:

Valve Replacement

Description:

A packless metal diaphragm Y-type globe valve is to be replaced with a standard Y-type globe valve.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. Calculations / analysis associated with this valve have been reviewed to verify that the change of valve type will not af fect associated design documents. The change does not affect the function or operability of the Reactor Coolant / Reactor Vessel Head Vent System. Assumptions or conclusions previously made in evaluating the consequences of an accident are not altered by this change. Replacement will meat the design specification for materials and construction.

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

_ No. This change does not establish any system configuration which reduces or bypasses any of the protective barriers of the plant.

The change does_not reduce the reliability of the RCS or Reactor Vessel Head Vent System. This change does not result in any _

system or component operating modes that are beyond the capacity and qualification of the equipment.

3) Does the subject of this evaluation reduce the margin of safety as defined in the basis for any technical specification?

No. This change does not affect the operability of the Reactor Coolant System.

Based upon the above, there is no unreviewed safety question.

Approved: 11/20/91 USQi92 W.001

Attachment 1 ST-HL-AE-4268 Unreviewed Safety Question EvrJ- ition #91-056

Subject:

Control of Heavy Loads

Description:

A new safe load path is being added to the procedure for control of heavy loads to move the Reactor Head Missile Shield over Residual Heat Removal Train C.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety T analysis report?

No. Procedures administrative 1y prohibit movement of the Reactor Head Missile Shield over new or spent fuel stored in the In-Containment Storage Area. Sufficient safety factor (10/1) is used at the interfacing lift points to ensure a load will not affect operation of the Residual Heat Removal system. Addition of this load path does not affect the design of any safety-related equipment.

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety anclysis report?

No. The only type of accident affected by moving the Reactor Head Missile Shield is fuel assembly damage. This type of accident is prevented by administratively prohibiting movement of the Reactor Head Missile Shield over fuel stored in the In Containment Storage _

Area. A different type of malfunction is not created because of the 10/1 safety factor at interfacing lift points.

3) Does the subject of this evaluation reduce the margin of safety as defined in the basis for any technical specification?

No. Fuel handling accidents are not affected because the Reactor Head Missile Shield is administrative 1y prohibited from moving

. over the In-Containment Storage Area. There is no impact on RHR System operability because of the 10/1 safety factor at the interfacing lift points. Shutdown margin is ensured because the procedure contains requirements to either have control rods inserted when moving the Reactor Head Missile Shield, or declare control rods not inserted to be inoperable and perform applicable Technical Specification surveillance to ensure adequate shutdown margin.

Based upon the above, there is no unreviewed safety question.

Approved: 10/24/91 USQ\92FW.001

a I

I Attachment 1 ST-HL AE-4268 Unreviewed Safety Question Evaluation #91-058-Subj ec t: Actual Locd on the Standby Diasel Generators

Description:

This change will update motor power factors, efficiencies and kilowatt loads to reflect actual data, as well as' revise-electrical load totals and clarify load descriptions. These changes will make UFSAR Table 8.3-3 consistent with other i plant design documentation.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of-

-equipment important to saiety previously evaluated in the safety analysis report?

No. It does not require a physical change.to any plant equipment or alteration of any equipment operating procedures,:and only.

editorial alterations to Plant Emergency Operating Procedures; therefore, the plant design basis is not affected. ._ The proposed -

change results in a net reduction in documented auxiliary bus and SBDG steady-state loadings and a. negligible impact on SDG. -

transient. loading criteria, The subject of this review does not-change the ability of the SDG to start in the event of a LOOP or SI-signal and does not affect the ability of the diesel generator to automatically start and accept ESF loads within the specified time limits. .The accident analyses presented in Chapter 15.of the UFSAR remain bounding,

2) Does the subject ~of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

No. Tne changes resulting-from the subject-represent, conditions which were analyzed in the original plant design. 1The subject of' ,

this review results.in a negligible impact to existing plant:

safety design margins.

3) Does the subject of' this evaluation reduce the margin of safety as .

defined in the basis for any technical specification?

No. !The proposed change is a general-revision to UFSAR Table'8.3-3 which does not impact;the diesel generator-availability requirements-or periodic-testing requirements as described:in Technical Specification ?/4.8.1. Thus, there.is no numerical or intent change-to the Technical Specifications.

Based upon the above, there-is-no unreviewed safety question; Approved: -11/7/91 USQ\921W.001--

Attachment 1

'ST-llL AE-4268 Unreviewed Safety. Question Evaluation #91 062

Subject:

Elimination of Essential Cooling Water!(ECW) System Expansion Joint

Description:

An expansion joint at the 6" inlet nozzle to _ the Standby Diesel Generator (SBDG) #11 intercooler is to be eliminated. ,

A vacuum breaker is to be installed'to relieve water hammer.

The stresses in the pipe have been evaluated using:the original design basis and are found acceptable per ASME Code requirements.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated'in_the safety analysis report?

No. -The ECW system for the-SBDC was evaluated on the basis of a moderate energy line break. The pipe / flange meets the system-piping specification for weight and' stress. Dae possibility _of a

-break or flooding will decrease with installation ofipiping due-to; [

elimination of a set of flanges and the expansion bellows. ; System integrity will be assured via hydrostatic testing and installation

. procedures.

2) Does- the subject of this evaluation create the possibility for an -:

accident or malfunction of a different' type than any. evaluated previously in the safety analysis report?-'

No. This repair changes _the connection of the SBDG intercooler -

piping and the-ECW-piping. Calculations determined the stresses are acceptable. A critical crack ini the moderate energy expansion.

=

joint _ is the only credible failure mechanism;f therefore,-

elimination of the expansion-joint and installation of system.

-specified pipe equal to lor better :than- thel expansion joint .will not create any different type of accident or malfunction from that previously evaluated in the UFSAR.

~

"3) Does the subject of this-evaluation reduce the margin of safety as defined in the basis for any technical specification? >

No. The Technical Specifications address operability of the SBDG's and-the ECW system. The piping material and the expansion I joints' are not addressed.

Based upon the above, there is no unreviewed safety question.

Approved: 12/30/91 USQ)D2FW.001

Attachment l' ST-HL-AE 4268

.c

--Unreviewed Safety Question Evaluation #91-063 Subj ect: Change in Nuclear Engineering Organization

Description:

The Nuclear Engineering organization has been changed. The UFSAR is to be revised to reflect new management

-responsibilities.

Safety Evaluation:

1) Does the subject of this evaluation increase _ the probability of  !

occurrence or the consequences of an accident or malfunction of-equipment important to safety previously evaluated in the safety analysis report?

No. The structure of the engineering organization has no impact on such probability or consequences. All cngineering functions will continue to be performed under the new structure.

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated-previously in the safety analysis report?

No. The structure of the engineering organization has no bearing on such a possibility. Engineering functions will still be performed in the new organization.

3) Does the subject of this evaluation reduce the margin of safety as defined in the basis for any technical specification?

No. The structure of the engineering organization is not a factor in the margin of safety in the basis of any Technical Specification.

Based upon the above, there is no unreviewed safety question.

Approved: 1/9/92

- USQ\92 W.001

. Attachment 1 ST HL AE-4268 Unreviewed Safety Question Evaluation #91 064

Subject:

Operations Quality Assurance Plan

Description:

Chapter 1.0 is to be revised to-reflect the organization of the Nuclear Engineering Department and define'the Manager, Nuclear Engineering responsibilities.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident 1or malfunction of equipment important to' safety previously evaluated in the safety analysis report?

No. This is an organizational change only, and the structure of the engineering organization has no impact on such probability or consequences. Engineering functions will continue to'be performed under the new structure.

2) Does the subject of this evaluation create the possibility for an.

accident or malfunction of a different type than any evaluated previously-in the safety analysis report?

. No. This is an organizational change only .and the structure of the engineering organization has no bearing on such a possibility, j Engineering functions will still be performed in the new organization.

3) Does the subject of this evaluation reduce the margin of safety as defined in the basis for any technical specification?

No. The structure of the engineering organization is not a factor

, in the margin of safety in the basis of any Technical E

Specification.

Based upon the above, there is no unreviewed safety question.

I l

~

Approved: 1/9/92 i

USQW2fW. 001 l -

. , _ =_-____ 3

Attachment 1 ST-HL-AE-4268 Unreviewed Safety Question Evaluation #92-001

Subject:

Health Physics Organization

Description:

The Health Physics organization as described in the UFSAR is being revised to reflect the reorganization concerning assignment of the Health Physics Operations General Supervisor and the Radiation Protection Supervisors, Safety Evaluation:

1) Does the subj e>:t of this evaluation increase the probability of __

occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. The change is administrative in nature and does not increase the challenges to safety systems assumed to function in the accident analysis such that safety system perform:nce is degraded below the design basis. The change does not decrease the ability to mitigate the radiological consequences of accidents.

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

No. The change is administrative and does not create the potential for creating a challenge to the safety systems assumed to function in the accident analysis such that safety system performance is degraded below the design basis.

3) Does the subject of this evaluation reduce the margin of safety as defined in the basis for any technical specification?

No. The change is administrative and does not reduce the number of Health Physics personnel on site to below that required by Technical Specification 6.2.2. This change does not adversely affect the ability to control access to High Radiation Areas as required by Technial Specification 6.12.

Based upon the above, there is no unreviewed safety question.

Approved: 1/24/92 USQ\92iv.001

2 Attachment 1--

ST-HL AE 4268- __ t 1

Unreviewed Safety Question Evaluation #92-004

Subject:

Residual 10 it Removal (RHR) Pump Differential Pressure -Gauges -

Description:

A local differential pressure gauge is'to be installed-across each of the three RHR pumps in each unit. The gauges-are to provide a more reliable and accurate means of ,

measuring RHR pump differential pressure. The existing RHR' pump suction pressure gauges will be relocated outside the RHR pump cubicles.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of' occurrence or the consequences of an accident or malfunction of equipment important to safety-previously evaluated in the safety analysis report?.

No. The gauges and their associated sensing lines will bo 7 installed seismically and in accordance with existin5 site procedures and practices. These instrument installations are designed to-maintain the'.r pressuro boundary during a seismic event.

2) Does the subject of. this evaluation' create the possibility for 'an-accident or malfunction of a different. type than any evaluated previously in.the safety analysis report?

No. Installation of local instrumentation is being performed.in accordance with existing site procedures and practices,-utilizing.

existing process connective points. These instruments are used

~

strictly for local indication and. provide no communication interface between any other component or piecq of equipment. fThey-have the same capability to maintain their pressure boundary and -

seismic integrity. ,

3) Does the subject of this evaluation reduce the margin of safety as #

defined in the basis for any technical specification?

No. The new' local-instrumentation.does not impact RHR systemL operability. (They will be used for.ASME Section XI testing.)

' Based upon the above,-there'is no unreviewed safety question.

Approved:'2/20/92 USq\921W,001

-. . -. .-.-..- --.- - - ._ - - . - . - - -- - - _ . . . -.~.- . - _ .

Attachment 1 ST llL.AE.4268 l I

r l

Unreviewed Safety Question EvaPsation n92 007 i

Subject:

Valves-Wide.Open Test (Unit 2)

Description:

This test is to dott raine the decrease in electrical power  ;

associated with reduction in hot leg temperature. The test  !

will determine the hot leg temperature correspanding to the governor valves in the open position and associated reduction in electrical power. The results, will be used in }

sn evaluation to determine the economic fe asibility of cperating with a reduced hot leg temperature. >

Safety Evaluation:

1) Does the subject of tnis evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety ,

analysis report?

No. The situations that may occur are adequately addressed in the test procedure. All chapter 6 and 15 accidents that consider dose are bounded. Evaluations determined that the original design  !

4 criterin: established in the FSAR and ASME Section III continue to be met. No new performance requirements are being imposed,

• Equipment qualification is not impacted. An increase in dose may occur should a steam generator tubo rupture occur; however, the=

increase is below the acceptance limit.

2) Does the subject of this evaluation create the possibility- for an accident or malfunction of a different type than any evaluated previously in the safuty analysis report?

No. The evaluation identified that_the margin to pressurizer becoming solid under accident conditions-may decrease. However, this decrease is acceptable since the acceptance limit is still~

satisfied. Performance _ of major components and equipment _ under- .

transient and steady. state conditions remains-within the design basis. Analysis of the equipment and components.is_either bounded by the current analysis, or remains within the design basis.

3) Does the subject of this evaluation reduce the margin of safety _ as .

defined in the basis for any technical specification?

No.- Impact _of the reducedLTavg conditions has been evaluated and.

the-reduced Tuo conditions were found to not reduce the. marsi n of safety, Based upon the-above, there is no unreviewed safety question.

Approvedi 3/25/92 l[

USQ\921M.00t

. . . _ . , . _ . . - - - , , . . - - , . - - - . . ,_ _ _ , .-,,.- 4 ,,_. . _

. - -- . - .-- ~ . - - . _ . _ . - .- . -.- . . - _ - _ - . . - _ . - ~ ..

Attachment 1 ST HL AE 4268 i i

Unreviewed Safety Question Evaluation #92-009 4

Subject:

Storage of Low Level Radioactive Waste ,

Description:

The UFSAR is to incorporate a revised description of the Onsite Staging Facility (OSF) for Low Level Radwaste. The i design of the OSF has been simplified from tne original  !

description.

Safety Evaluation:  ;

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. Storage of low Level Radioactive Vaste (LUW) for the revised .

Onsite Staging Facility (OSF) design is physically located outside -i the Protected Area. Storage of LLW does'not interact with any of i

the systems or equipment evaluated for accidents in the SAR or equipment important to safety. The calculated annual dose to.the ,

public duo'to LLW storage in the OSF is bounded by the consequences for accidents 6 1ven in UFSAR Section 15.7. Accidents =

involving handling of LLW will be evaluated later and are not part of the OSF design review. Fire is not considered to result in a

, possible radiological hazard. Seismic criteria are not a-design- ,

roquirement for storage of this LLW. There are no radiological hazards due to environmental events.

2) Does the subject of this evaluation' create-the possibility for an accident or malfunction of a different type than any evaluated previously-in the safety-analysis report?

No. Storage of LLW in the OSF does not interact with'any= safety -

related equipment or systems,

3) Does the subject of this evaluation reduce the margin of safety as defined in the basis for any technical specification?  ;

No. Technical-Specification ~3/4.11.3 (Solid Radwaste) has no applicable margin of safety in the requirements or in the' Basis, Based upon the above, there-is no unreviewed' safety question.-

Approved: 4/20/92- }

s s

y gr w e tvom vre.r+ 74- e,m-- e w - e ,r r e-e- ew'-- - e'**=ew =---"E 5-

Attachment 1 ST ilt AE 4268 Unreviewed Safety Question I'valuatlon n92-011

Subject:

Special Nuclear Materials Control and Accountability D.scription: procedure revision deletes requirement to inventory fuel assemblies in the spent fuel pool during ret u. ling.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safet/ _

analysis report?

No. The SAR addresses several sit.gle failure accidents concerning fuel assembly drops an' fuel assemb;y misplacement. Removing the requirement to inventory fuel assemblies during refueling will only increase the time between misplacement and discovery of the misplacement. Misplacement of a fuel assembb in the spent fuel pool is a single failure analyzed accident. The m..g sis does not provide any time restrictior. for discovery of the misplacement. "

Criticality will be maintair. as long as 1000 ppm of soluble boron is present in the poc'

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a differ v* type 'han any evaluated previously in the safety analysis rtpo"t?

No. See (1).

D Does the subject of this evaluation reduce the margin of safety as _

defined in the basis for any tectnica. specification?

No. Technical Specification 5 (.1 requires a multiplication factor of less than 0.95 for the spent tuul pool storage racks when filled with unborated water. Only the actual misplacement (f a fuel assembly could compromise this margin cf safety, llowever, this change is not associated with placement of fuel assemblies.

Based upon the abovo, there is no unreviewed safety question.

Approved: 5/5/92 USQ\921M,001

l 4

l Attachment 1 ST HL AE 4268 Unreviewed Safety Question Evaluation w92 012

Subject:

Steam Line Break Core Response Report

Description:

The current licensing basis steamline break core response analysis assumed the initial reactor vessel upper head fluid temperature to be at the full power temperature. The hot- l zero power temperature should have been assumed. The PFSAR is to be revised to reflect this chang:,.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. The proposed change does not involve a physica1' change to the ,

plant or a change to proceduros used to operate the plant. The proposed change does not impact overall system design, performance, reliability, or availability, Since the DNBR acceptance limit for this accident continues to be met, fuel '

failute would not be expected to occur. The assumption of 5% i failed fuel in the dose analysis remains conservative and.the dose analysis is not impacted. The small increase in RCS cooldown rato does not challenge the structural design limits of the RCS.

2) Does the subject of this evaluation create 1the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

No. The change does not involve a: physical-change to the plant or a change to procedures used to operate'the plant. The small increase in RCS cooldown rate does not challenge the structural-design limits of the RCS. The proposed change _is for the DNB analysis and not the mass and energy release analysis,

3) Does the subject of- this evaluation reduce 1the margin of safety as defined in the basis for any technical specification?.

No. The proposed change does not result in the DNBR acceptance limit being exceeded, and does not result in the structural-design limits of the RCS being exceeded. <

i l

i Based upon the above, there is no unreviewed safety question.

Approvedi 6/25/92 l

5 USQ\92It 001 ,

w -

_w = -v

Attachment 1 ST ilL AE 4268 Unrevleired Safety Question Evaluation #92-014

Subject:

Refueling Water Storage Tank  !

Description:

This change revises UFSAR Section 6.3.22 and Figure 6.3 8 to l account for corrections to a calculation in order to i incorporate correct values for instrument error and instrument scaling. Neither the calculation nor the UFSAR  ;

(Section 6.3.2.2) accounted for the time for the mini flow  !

recirculation valves to close to show adequate margin from j initiation of recirculation to loss of suction for the pumps. .

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of i equipment important to safety previously evaluated in the safety i analysis report?

No. No changes are made to the physical facility as a result of the calculation correction, The changes result in an adequate  ;

volume of injection water available and adequate volume for '

transfer allowance to ensure that automatic transfer from-injection to recirculation occura. The ECCS will perform its functions as originally intended and designed.

2) Does the subject lof _this evaluation create the possibility for an f accident or malfunction of a different type than any evaluated ,

previously in the safety analysis report?

No, .There are no physical changes proposed to the-faellity, only  :

l- corrections for calculation errors.

• l L 3) Does the subject of this evaluation reduce the margin of safety as -

defined in the basis for any technical _ specification?

No. Technical Specification 3.5.5 requires a total contained- , ,

l-  : volume'of 458,000 gallons.in the RWST. This was to: ensure an L injection volume of at least 350.000 gallons. The. actual net injection volume is 360,000 gallons. Technical Specification; 3/4.3.2, Table .3,3 4 Step 7b requires thatl transfer from injection to recirculation occur when RWST Low Low level reaches 11%. This value provides adequate margin to ensure that the transfer

~

allowance'is greater than required. .

^

Based upon the above, there is no-unreviewed safety question, Approved: 7/22/92 USQtO2i9.001

, - - - , _ , . - . . . - . , . . - , . u u ;m__. _ _ _ ..-__._ _ - _.,~.,0._.. . _ _ - _ . . -a. 2.... .- ;-

Attachment 1

! ST llL AE 4268 i

Unreviewed Safety Question Evaluation 992 015

Subject:

Essential Cooling Water (ECW) Systetu

Description:

This change deletes a reference that, under normal operations, one ECW train is cross connected to another train at the Essential Chillers. Procedures do not allow cross connecting trains duriig normal operation except during fill and vent, with an operator stationed at the valves continuously.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. The UFSAR does not analyze accidents or malfunctions involving loss of ECW. The change meets with the original design requirements and is consistent with the original design. If one ECW train is cross connected to another train of Essential Chillers, an ECW pump failure would Icad to two trains of essential chillers tripping. This change eliminates that possibility by eliminating cross connection of two ECW trains,

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

No. The Probabilistic Safety Assessment evaluates loss of ECW; _

however, this change will not pose any new failure types to the evaluation. Maintaining the cross-tie valves closed will not introduce additional malfunction mechanisms or affect the operability of any of the three trains of ECW or CR/EAB llVAC as described in Technical Specifications.

3) Does the subject of this evaluation reduce the margin of safety as defined in the basis for any technical specification?

No. Sufficient coolitig capacity will remain available for continued operation during normal and accident conditions. Three trains of ECW and CR/EAB llVAC will remain operable and available per Technical Specifications.

Based upon the above, there is no unreviewed safety question, Approved: 7/22/92 USQ\921W.001

Attachment 1 ST llL AE 4268 Onreviewed-Safety Question Evaluation n92 017

Subject:

New Fuel Shipping Containers

Description:

Westinghouse has upgraded new fuel shipping containers (NFC) to accommodate higher enriched fuel assemblics. As a result, the weight of a fully loaded _ container has increased from 9,500 pounds to 10,533 pounds. For lleavy load Program considerations, this safety evaluation justifies an increase in NFC weight to 12,000 pounds. This will provide a margin for rigging and future NFC weight increases.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. The equipment used to move the NFC is the singic failure-proof,15 ton FilB Overhead Crane and associated rigging. The revised NFC weight remains within the load capacity of the 15-ton FilB Over' ,ad Crane. The physical load path has not changed. Use of rigging continues to be in accordance with the previously approved lleavy Loads Program.

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

No. The NFC movement in the -Fuel 11andling Building (FilB) has been described-and evaluated in the Control of Ileavy Loads Program.

The FilB Overhead Crane and associated rigging are not operated differently than previously described or evaluated.- Loads heavier than the revised NFC have been previously approved for this crane.

3) Does the subject of this evaluation reduce the- margin of safety as defined in the basis for any technical specification?'

No. Technical specification 3/4.9.7 requires use-of the 15-ton-FilB Overhead Crane for moving heavy loads' cver the Sport Fuel Pool-(SFP). The.NFC change doe not cause any load to be moved over the SFP and does not affect the physical load path or consequences of

proviously evaluated accidents, l

I Based upon:the above, there.is no unreviewed-safety question, i

Approved: 7/29/92-USQT92iv.001-

Attachment 1 ST ilL AE 4268 l T

Unreviewed Safety Question Evaluation #92 018 Subj ec t: Spent Fuel Pool lleatup Rate Measurement

Description:

This test procedure determines the heatup rate of the_ Spent-Fuel Pool (SFP) at various fuel loadings when Component  ;

Cooling Water (CCW) flow to the SFP Heat Exchangers is_ ,

secured. The heatup rate obtained through this procedure is  !

used to calculate the maximum allowable time that CCW flow

~

to SFP Heat Exchangers can be secured. .j i

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. The CCW flow to the SFP lleat Exchangers is stopped for a- ,

-controlled duration when pool water remains below 128'F and pool-cooling can be restored by opening the_CCW valves at any time from the control room. Also, the heat load in the pool ouring this test is required to be significantly lower.than those analyzed in the UFSAR by means of the test pre-requisites. The pool temperature is maintained below analysis temperature such that no boiling occurs and doses remain below those analyzed. This test does not involve a physical change to plant equipment. The valves are designed to close by operator action or upon receipt of a safety injection signal .

2) Does the subject of this evaluation create the possibility' fL accident or malfunction of a'different type than any evaluated-

, previously in the safety analysis report? -'

1.

l- No. The test involves closure of CCW valves which does.not affect any system other than the SFP cooling system. .No other' accident, other than the LOCA to the SFP is postulated and evaluated in UFSAR Section 9.1.3.3. . Closure of the valves does_not have any impact on operation of CCW pumps, SFP cooling pumps 'or any other ESF equipment.

j 3) Does the subject of this evaluation reduce the margin of safety as -

defined in the basis for any technical specification?

No. Pool-temperature during this test does not'. exceed 131.2'F, L

which is the maximum. initial pool temperature in normal mode _in

! the UFSAR and 140*F, the acceptable limit in the SER.

l i

Based upon the above, there is no unreviewed safety question'.

Approved: 9/17/92 USQ\921M.001'

. l - . - .

Attachment 1 ST HL-AE.4268 Unreviewed Safety Question Evaluat. ion u92 019

Subject:

Quadrant Power Tilt Ratio (QPTR)

==

Description:==

The proposed change will add time delay pickup relays into the QPTR alarm circuitry. The new alarm circuitry will block nuisance QPTR alarms unless a real event occurs and lasts longer than the setpoint of an adjustable time delay.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of _

occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. Implementation of a time delay (froin 4 to 120 seconds) does not impact any of the input parameters to the accident analysis.

Tbc QPTR alarm provides a monitoring function only, with no automatic actions. No new performance requirements are imposed on the fuel such that design criteria are exceeded. The QPTR alarm will still perform its intended function to notify the operator of gross changes in radial power distribution. The failure inode of the QPTR alarm is not impacted. (Relays will fail to the clarmed state.) Since the peaking factor input assumptions to the safety analysis remain valid, there will be no additional mass releases, fuel failurea, or other fission product barrier degradation beyond that already modelled in the radiological dose consequences. The .

predicted offsite doses presented in the UFSAR remain bounding.

Opcrations must verify any valid QPTR alarm with a manual calculation. Technical Specifications provide specific _

requirements if the QPTR alarm is inoperabic.

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

No. Performance of the monitor and alarm functions does not generate any new failure mechanism or new credible accident scenario not already evaluated. Calculated QPTR is not used in any protection system. The QPTR alarm will still perform its intended function to notify the operator of gross changes in radial power distribution. This modification can only impact the QPTR alarm.

USQ\ 92W,001

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I Attachment 1 .l ST HL AE 4268  !

1 Unreviewed Safety Question F, valuation #92 019 (Con't) f

3) Does the subject of this evaluation reduce the margin of safety as defined in the basis for any technical specification?

No. The proposed change will enhance plant performance by minimizing nuisance QPTR alarms unless a real event occurs and ,

lasts longer than the added time delay, None of the limits or-  !

ACTION statetuents identified _in Technical Specification 3/4.2.4 i are affected by the change. In addition, the accident analysis results are unaffected.

Based upon the above, there is no unreviewed safety question.

Approved: 8/13/92 l,

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_usotsam.coi l:b

Attachment 1 ST Hb AE 4268

-Unreviewed Safety Question Evaluation #92 021 Subj ect: FHB HVAC Booster Fan Repair Description; The FHB HVAC (HF) System is to be temporarily modified by-isolating and removing the HF "C" train main exhaust booster fan from service. The exhaust booster fan will be removed from service to facilitate repair / replacement of the fan motor.

112 1%

A temporary valver of compliance to Technical Specification 3.0.3 was requested by letter. dated August 18, 1992 (ST HL AE 4185) . The waiver was necessary because all three i trains of the FHB ventilation system would be incapable of automatic initiation during the repair for longer than one hour. The NRC granted the waiver by letter dated August 19, 1992.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. The temporary modification of the.FHB HVAC (HF) System is within the bounds of the USFAR analysis for a single FHB' exhaust booster fan failure. HF exhaust system integrity will be maintained by blank. plates while the 210 booster-fan motor is being replaced. Mechanical-isolation (blank plates) will be-verified as air / leak-tight to ensure there is no adverse effect to the rest of the HF exhaust system. The FHB exhaust HVAC will' still be capable of operating at 100% capacity _with two. booster =

fans.

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated-previously in the safety analysis _ report?

p- No. Mechanical isolation (blankLplates) of'the affected exhaust booster fan will be with material equivalent to that-of the HVAC ducting it isolates, . Verification of satisfactory installation-will;be obtained by leak testing the plates to ensure they_present an air. tight seal-to the HVAC duct. The modification will" maintain the system integrity while isolating the fan for repair.

Even with one booster _ fan out of-service, the HF exhaust system I

still has the capability to operate at 100% capacity.-

o l

. UsQ\s2fW.001'

Attachment 1 ST llL AE 4268 Unreviewed Safety Question Evaluation v92 0?1 (Con't)

3) Does the subject of this evaluation reduce the margin of safety as defined in the basis for any technical specif1;ation?

No. Technical Specification 3.7.8 governs operation of the FilB exhaust air system in modes 1-4 and requires the system to operate at 100t capacity following a LOCA to ensure that radioactive raaterials leaking from the ECCS equipment within the FilB are filtered prior to reachin6 the environment. Loss of one exhaust booster fan was previously evaluated and per UFSAR Table 9.4-5.2 has no effect on system safety function capability. The Technical Specification LCO action requires that for less than three exhaust booster fans operabic, the Inoperable fan is to be restored within seven days or be in at least hot standby within the next six hours, This LCO requirement and any subsequent LCO actions are applicable to this temporary modification.

Based upon the above, there is no unreviewed safety question.

Approved: 8/17/92 L'SQ\ 021W. 001

Attachment 1 ST.llL AE 4268 i Unreviewed Safety Question Evaluation a92 022 j l

Subj ect: Containment Sump Water Leaks

==

Description:==

This change ostice updates the UFSAR for a change in design assumptions used to calculate radiological doses resulting '

from a LOCA. This update evaluates the doses resulting from-potential Icakage of containment sump water into the Refueling Water Storage Tank (RWST) located in the Mechanical Auxiliary Building (MAB).

Safety Evaluation:

1

1) Does the subject of this evaluation increase-the probability of 1 occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety 1 analysis report? 1 No. The proposed change adds an assumption to thoso.previously made in evaluating the radiological consequences of a LOCA described in the SAR, but the radiological consequences described ,

in this change are bounded by those set by 10CFR100 and SER sections 6.4 and 15.6.5.2.5 (dated April 1986). This chango evaluates the doses resulting-from potential leakage of containment sump water into the RWST located in the MAB The IACA analysis aircady assumes malfunction.of equipment important to safety.

2) Does the subject of this ovaluation create the possibility for. an  !

accident or malfunction of a different' type than any evaluated previously in the safety analysis report?

No. This change analyzes the effect of potential leakage of containment sump water into the RWST on.LOCA doses.

3) Does the . subject of this evaluation ' reduce the margin of safety as defined in the basis for any technical specification? .

No. -The-. maximum. dose due to RWST back leakage is 2.0 E+5 Rads,-

'which is well-bolov the: acceptance criteria value of;1.5_E48. Rads.

+

Based upon the above,-there is no unreviewed safety question.

Approved: 9/20/92 USQ)s2 W.001

i Attachment 1 j ST.!!L. AE 4268 Unreviewed Safety Question Evaluation u92 023

Subject:

Turbine Overspeed Reliability Program

Description:

The probabilistic inspection intervals for the low pressure turbine rotors have been revised to include all of the current rotors onsite with the current low pressure turbine locations. The method of calculating the probabilistic inspection interval was also revised. The change in method reduced the probabilistic intervals. The monthly test of the mechanical overspeed device via oil simulation will be  ;

changed to quarterly.  !

Safety Evaluation: i

1) Does the subject of this evaluation increase the probability of

• occurrence or the consequences of an accident ou malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. The revised probabilistic intervals follow the criteria in the Safety Evaluation Report, and therefore the probability of generation of a turbine missile as evaluated in the SER has not been increased. None of the changes affect the probability of a  ;

turbine missile striking a safety related structure, or the probability that a missile striking a structure penetrates and damages a required system or component.

Increasing the test interval of the mechanical overspeed device potentially affects the probability of malfunction of equipment important to safety previously evaluated in the SAR:lspecifically, the turbine overspeed protection system. Considering the simplicity and. inherent reliability of the mechanical overspeed trip mechanism, a quarterly test would be a better balance between the slight risk of a system separation while performing the test versus the slight risk of degradation in the mechanical overspeed trip mechanism during the quarter.

2) Soes the. subject of this evaluation create the possibility for an-accident or malfunction of a different type than any evaluated -

previously in the safety analysis report?

No. The inspection intervals for LP rotors and other items-related to the turbine overspeed reliability program do not affect any aspect of design or operation besides the probability of disc rupture through stress corrosion crack propagation or-destructive overspeed. These have been evaluated in the SAR;previously, s

-Usq W21v.001

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Attachment 1 ST.HL.AE.4268 Unreviewed Safety Question Evaluation #92 023 (Con't)

3) Does the subject of this evaluation reduce the margin of safety as defined in the basis for any technical specification?

No. The basis for Technical Specification 3/4.3.4 discusses ,

prevention of turbine missiles from destructive overspeed. No- j specific probability numbers are provided in the basis, o Volumetric examination of the turbine rotors is not discussed in- l the basis or other section of the Technical Specifications. The  ;

mechanical overspeed trip is not specifically discussed in'the Technical Specifications, j Based upon the above, there is no unreviewed safety question.

Approved: 9/20/92 ,

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Attachment 1 ST-HL AE 4268 Unreviewed Safety Question Evaluation #92 026

Subject:

Mixed Bed Resin in the CVCS Cation Domineralizer

Description:

The change will install an alternate mixed oed resin during 1RE04 in the CVCS cation demineraliter.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety _

analysis report? i No. Since the revised resin is identical to the resin now in use ^

in the mixed bed demineralizer, except for the-lithium form, lithium usage will be greater until the new bed is saturated. -No other interaction or consequence of this resin will occur to plant equipment, plant materials, or to plant chemistry. No_ change will occur in the CVCS, relative to its capability to act as a pressure boundary, and no new consequence or change in the effect of placing the unborated bed initially in service will occur. .The revised resin will not react differently or produce any new or altered output compared to these domineralizers now in service.

Procedural controls will ensure that boron concentration is not altered adversely by use of this new resin.

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

No. This resin is identical to that already in service, although not in a lithium form. Usage.of this resin will_be governed by procedures already in effect which monitor for full boration of the bed prior to placing it in service. If this bed were-left'in service mistakenly, it would'still act as a lithium removal domineralizer, although not as efficiently as the cation _ bed.

Since the other cation bed remains available and chemistry results are constantly monitored and evaluated, any-lack of response due to this resin being in service could be immediately rectified.

Since the revised resin is chemically and physically similar to that already in use,-it will not affect the CVCS, reactor coolant,

'or coolant chemistry in a unique _way.

USQ\92Iw.001

1 j

l Attact. ment 1 ST HL AE.4268 Unreviewed Safety Question Evaluation #92-026 (Con't)

3) Does the subject of this evaluation reduce the margin of safety as defined in the basis for any technical specification? -l l

No. Since only one bed of cation resin is needed to remove l lithium and other tonic isotopes (associated with failed fuel) and the second cation bed is fully available, this change of resin will not alter the capability of the CVCS to perform its function.

The revised-resin is similar to that already in service, and no l change in its behavior or impact on the chemistry or coolant will -

occur. <

Based upon the above, there is no unreviewed safety question.

Approved: 9/17/92 l

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~ USQ\92 N,001' '

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Attachment 1 ST.HL.AE.4268 i

l Unreviewed Safety Question Evaluation #92 027

Subject:

Operator Response Times for SG Tube Rupture Overfill Scenario

Description:

The evaluation extends the allowable operator response time for identification and isolation of the ruptured steam generator from 12 minutes to 19 minutes $0 seconds. ,

i Safety Evaluation:

1) Does the subject of this evaluation increase the probability of '

occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report? i

+

No. The proposed change reflects the additional analysis used to support extendinS the allowable _ operator response-times. The i revised results show that the steam generator does not go water solid. Therefore,- the acceptance criteria for this accident is not violated. The revised analysis is or.ly for the steam generator _ overfill analysis. The dose analysis is not impacted.

The proposed change does not involve a physical change-to the plant. No actions are-involved that change the operation of equipment.

2) Does the subject of this evaluation create the. possibility for an accident or malfunction of-a different type than any evaluated previously in the safety analysis report? -'

No. The results of the analysis show that the steam generator does not go water solid.- The steam lines at STPEGS are designed to support water filled lines up to the^ main steam isolation-  ;

valves. Therefore,' filling the steam generator lines with water i would not create a steam line break. However, the NRC acceptance limit is that the steam generator does not overfill. The proposed change does not involve a physical change to the plant. No changes to operation of equipment considered in the safety analysis are proposed.

3) Does the subject of this evaluation' reduce the margin of safety as defined in the basis for any technical specification?

i-No. The acceptance limit for the steam generator overfill-analysis is that the steam generator-.does not_ overfill. ~The

-analysis shows that.the steam generator does not overfill. i Based upon'the above, there.is no unreviewed safety question.

Approved: 9/21/92 USQ\921W.001 A .- ,. .._;.._,.m__.._-..,__.-_ .__ ___.._-___ u. _-_-...u , - ....a..-_-..w---- _ _ _ ,. -,

Attachment 1 ST ItL AE-4268 I

I Unreviewed Safety Question Evaluation #92 032 Subject. Configuration Management Program

Description:

The level oi detail in the UFSAR and Envirotunental Report is to be reduced regarding the River Makeup Pumping Facility, Potable Water System, Sanitary Water System, Miscellaneous llVAC, and Maintenance Operations Facility Fire Protection System.

Safety Evaluation:

1) Does the subject of this evaluation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report?

No. There is no change in the physical facility. No essential detail cited in the Safety Evaluation Report (SER) or in the Final Environmental Statement - Operating License (FES-OL) has been deleted. All areas affected by this change are unrelated to systems required for mirigation of accidents descri'ved in Chapter 15 of the UFSAR.

2) Does the subject of this evaluation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

No. The physical facility is not changed. No essential detail cited in the SER or in the FES OL has been deleted. All areas affected by the change are unrelated to systems required for mitigation of accidents described in Chapter 15 of the UFSAR.

3) Does the subject of this evaluation reduce the margin of safety a::

defined in the basis for any technical specification?

No. The physical facility is not changed. No essential detail cited in the SER or in the FES OL has been deleted. The changes are unrelated to Technical Specification Limiting Conditions for Operation.

Based upon the abovc, there is no unreviewed safety question.

Approved: 9/17/92 USQ\921R 001

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