ML20216D777

Annual SER 12 for Jan-Dec 1997, for Wolf Creek Generating Station

ML20216D777
Person / Time
Site:	Wolf Creek
Issue date:	12/31/1997
From:	Muench R WOLF CREEK NUCLEAR OPERATING CORP.
To:
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ML20216D753	List: ML20216D749 ML20216D777
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NUDOCS 9803170274
Download: ML20216D777 (240)

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i Attachment to ET 98-0014 Page i  !

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l WOLF CREEK NUCLEAR OPERATING CORPORATION i

I Wolf Creek Generating Station j

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Docket No.- 50-482 Facility Operating License No.: NPF-42 i

ANNUAL SAFETY EVALUATION REPORT j i

i Report No.: 13 i l

Reporting Period: January 1, 1997 through December 31, 1997 i

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s 9003170274 980311 2 DR ADOCK 0500

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i Attachment to ET 98-0014 Page 11

SUMMARY

This report provides a brief description of changes, tests, and experiments performed at Wolf Creek Generating Station pursuant to 10 CFR 50.59 (a) (1) . j This report includes summaries of-the associated safety evaluations that were i reviewed and found to be acceptable by the Plant Safety Review Committee for the perioc beginning January 1, 1997 and ending December 31, 1997 This -

report is submitted in accordance with the requirements of 10 CFR 50.59(b) (2) .

On the basis of these evaluations of changes, the following has been determined: ,

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• There is no increase in the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously  ;

evaluated in the Updated Safety Analysis Report (USAR).  !

• There is no possibility . that an accident or malfunction of . equipment i important to safety of a different type than any evaluated previously in the USAR may be created.

• The . margin of safety as defined in the basis for any Technical Specification is not reduced.

Therefore, all items reported herein are determined not to involve an l

unreviewed safety question. '

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s Attachment to' ET 98-0014 Page 1 oi 238 Safety Evaluation: 59 93-0021 Revisiona l Reactor Coolant System Leakage Monitoring Revision 2 of Plant Modification Request (PMR) 04540 provides for I evaluation of electrical separation distances that are.less than the distances required by documents E-1R8900 and E-11013. The revised criteria applies to the field-routed cables used for monitoring temperature on the Chemical and Volume Control System (CVCS) auxiliary spray injection to pressurizer piping. This temperature monitoring is recommended by NRC Bulletin 88-08. NRC Bulletin 88-08 recommends checking leakage from the Reactor Coolant System (RCS) into the connecting system piping. This leakage, if present, is judged to cause fatigue in safety-related piping.

This revision is necessary because the field-routed cables do not meet the electrical separation criteria, and they increase the amount of combustible loading described in the Updated Safety Analysis Report (USAR). The field-routed cables will be in place until Refuel 10.

Revising USAR Section 8.3 1.4.1.4 to list these cables as an exemption to the electrical separation criteria will exclude these cables from the requirements of USAR Section 8.3.1.4.1.1. The increase in combustible loading will be added to USAR Section 9.5B.7 for fire zones RB-1, 4, 6, 8, and 10.

The electrical separation criteria for permanent plant raceways was developed from.IEEE 384-1974 and NRC Regulatory Guide 1.75, Revision

1. The criteria is applied to Wolf Creek Generating Station (WCGS) in L.Ouments E-1R8900 and E-11013. Physical separation criteria is esssblished for the independence of circuits and equipment comprising or associated with Class 1E systems. This physical separation is provided to maintain the physical and electrical independence of a sufficient number of circuits and equipment so that the protective functions required during and following any design basis event can be accomplished. -The regulatory commitments are idertified in USAR Section 8.3.1.4.1. This section discusses a three foot horizontal and five. foot vr.rtical separation between safety-related raceways of different Separation Groups and between safety-related raceways and open non-safety related raceways. It also describes a one-inch separation between safety-related raceways and non-safety related

' conduits. USAR Section 8.3.1.4.1.4 describes exceptions from these physical separation requirements. The exceptions are based on individual analysis of the function of the cables contained within each safety-related raceway.

The electrical separation for these field-routed cables is less than the criteria specified in USAR Section 8.3 1.4.1. Specifically, the Auxiliary Spray Piping Monitoring cables do not meet the minimum

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Attachment to ET 98-0014 Page 2 of 238 l

l separation criteria for certain raceways. IEEE 384-1974 allows a departure from minimum separation distances by analysis. This analysis shall be based on tests performed to determine the flame retardant characteristics of the cable installation.

The Auxiliary Spray Piping Monitoring cables are used to connect resistance temperature detectors (RTD) to a data logger. These j cables are energized from an RTD card in the data logger. The current l produced is in the milliamp range. The cables are 16 AWG and can l easily handle the current should a short circuit occur. The energy level of a potential fault of these cables is self-limiting. If a fault condition (short circuit) was to occur, the amount of heat i

generated would not be sufficient to cause damage to the faulted cable or to surrounding cables.

Testing performed by Wyie Laboratories for Limerick Units 1 & 2 (Test Reports 4690-1 and 4690-3) reveal that cables 4/0 and smaller will not ignite when subjected to the maximum anticipated continuous current levels. These cables were energized with 660 amps in an open cable tray. Configurations with a one-inch vertical separation between i cable tray and zero separation between cable tray and enclosed conduit l were tested successfully without damage.

The Wolf Creek Auxiliary Spray Piping Monitoring cables are similar to those used at Limerick. The primary differences are the cable insulation and jacket material. Wyle tests were conducted with cables utilizing cross-linked polyethylene or neoprene jacketing material.

The insulation and jacketing material for the Auxiliary Spray Piping Monitoring cables is PVC. This difference can be reconciled as follows:

The Auxiliary Spray Piping Monitoring cables are U. L. Listed and meet the IEEE-383 70,000 BTU flame test.

The energy level of a potential fault of these cables is self-limiting. If a fault condition was to occur, the amount of heat generated would not be sufficient to cause damage to the faulted cable or to surrounding cables.

A test was conducted for these cables under DCP 05053 to verify the cables' resistive property to flame propagation. This cable was exposed to a flame from a propane burner. The estimated temperature was 1600 degrees Fahrenheit. The cable burned but self-extinguished when the flame was removed.

Based on the information provided above these cables will not degrade class 1E circuits. Therefore, the Auxiliary Spray Piping Monitoring cables do not need to meet the established separation criteria.

A field walk down was performed to obtain as-installed cable routing

Attachment to ET 98-0014 Page 3 of 238 and its support information. Based on the walk down, it was determined that the as-built tubing support configurations comply with existing design documents. Based on a review of the generic drawings, specifications, and generic calculations, it was determined that each field condition is enveloped by the existing design. Based on this review, it was determined that there is no impact to safety-related equipment.

There are no d?eign basis accidents discussed or referenced in USAR Chapters 2, 3, 6, or 15 that are impacted by reducing the electrical separation criteria for the Auxiliary Spray Piping Monitoring cables.

USAR Chapter 9 was reviewed for any adverse impacts to safe shutdown following a fire. The review concluded that the safe shut down cables in fire area RB-4 are nec affected by these field-routed cables since this fire area has a fire detection and suppression system. In other affected fire areas, the field-routed cables are not intervening combustibles that affect safe shutdown.

Based on the above discussion, this modification will not 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. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, The margin of safety as defined in technical specifications is not reduced by this modification.

Therefore, this modification does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 4 of 238 Safety Evaluation 59 93-0205 Revision: 1 Modification to Fire Protection Barriers for Auxiliary Feedwater Pump Supply Valve This evaluation addresses both physical fire protection plant changes and fire protection program changes being made as part of the Thermo-lag Resolution Project. There are four specific applications of the Thermo-lag fire barrier material which have been evaluated as a result of NRC Bulletin 92-01. A Fire Hazards analysis has been performed for all four applications following the guidance provide by Generic Letter 86-10. The four specific applications are:

1. Tendon surveillance access hatches in the Auxiliary building at elevations 2026' (two hatches) and at elevation 2047'-6" (two hatches), are composed of steel plate covered with the Thermo-lag fire barrier material. Based on the evaluation, the hatches meet the requirements of a non-rated feature and have been determined to provide fire protectica commensurate with the hazards present in the areas surrounding the hatches. The evaluation has determined that the original design bases and function of the hatches is unchanged, and that they continue to act as a fire barrier to prohibit the propagation of fire between the adjacent fire areas. These changes do not introduce any new failure modes. The Updated Safety Analysis Report (USAR) and applicable design documents are being updated to reflect this change.

2. Residual Heat Removal and Containment spray encapsulation access hatch covers in the Auxiliary building at elevation 2000' were covered with panels constructed from Thermo-lag material. These panels are being removed and replaced by 1/4" steel checker plate. The steel covers have been evaluated and meet seismic II/I requirements. The steel covers represent a non-rated feature and have been determined to provide fire protection commensurate with the hazards present in the areas surrounding the hatches. The USAR and applicable design documents are being updated to reflect this change.

3. The box type enclosure at elevation 1989' around motor operated valve ALHV0032, which supplies essential service water to the turbine driven auxiliary feedwater pump, was installed to meet Appendix R separation requirements by providing a three-hour rated fire barrier between redundant trains of circuits required for safe shut down located in the same fire area. This modification will install a wet pipe sprinkler system meeting seismic II/I requirements in rooms 1207 and 1208 in Fire Area A-33, and extend to full coverage the existing fire detection system in the two rooms. The existence of an automatic suppression system, full fire detection system, and the spatial separation of 20 feet free of intervening combustibles between ALHV0032 and its redundant valve ALHV0033 means the box-type enclosure will no longer be zequired. This modi.'.ication adds the suppression and detection systems and removes the box type enclosure. Flooding k

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Attachment to ET 98-0014 Page 5 of 238 hazards have been evaluated, and it is determined that existing floor drains, sized for the larger Auxiliary Feedwater piping breaks, will prevent flooding caused by any postulated fire suppression line break. An inadvertent localized sprinkler actuation will not affect safe shutdown capability. If valve failure is postulated due to water spray, the redundant valve will be available. Due to spatial separation between the redundant valves, any single sprinkler head actuation would not effect more than one valve. Internal conduit seals are installed in valves ALHV0030-36 via this modification. The automatic wet-pipe sprinkler system, in conjunction with 20-foot separation between safety related components with no intervening combustibles, the fire detection provisions, and at least one-hour fire wrap on the circuits associated with valve ALHV0032, provide the same fire protection function as the existing three-hour fire wrap enclosure on valve ALHV0033. The degree of fire protection is equal to or greater than that provided by the current design.

4. The Radiologically-controlled area personnel escape hatch located in the Auxiliary building at elevation 1989' is covered with Thermo-Lag material. Based on the hazards present and the mitigating factors available, a fire in either room 1125 (below the hatca) or room 1207 (above the hatch) is not expected to propagate to the other side and cause damage. The steel hatch covered with Thermo-lag represents a non-rated feature and has been determines to provide fire protection commensurate with the hazards present in the areas surrounding the hatches. The evaluation has determined that the original design bases and function of the hatch is unchanged and that it continues to act as a fire barrier to prohibit the propagation of fire between the ,

adjacent fire areas. These changes do not introduce any new failure 1 modes. The USAR and applicable design documents are being updated to reflect this information.

In all four applications, the requirements of 10 CFR 50, Appendix R, Section III G, for separation of safe shutdown circuits will be maintained by this design. The original design bases assumptions and inputs remain valid.

Based on this evaluation, this modification will not iverease the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. No Technical Specification limits are impacted by this change to the fire protection system. The equipment affected by this modification are not described in the Technical Specifications and their associated bases, and have no safety function. Therefore, the margin of safety as defined in the Technical Specification basis is not affected.

Therefore, this modification does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 6 of 238 Safety Evaluation 59 94-0060 Revision: 1-Thermal-Expansion Monitoring Program Noise Monitoring Equipment Installation j This modification provides for the installation of noise monitoring sensors and recording equipment inside the containment structure.

This modification is applicable to the Reactor Coolant System (RCS),

the RCS supporting' structures, and components. The sensors and the-recording equipment provide for data gathering only and perform no design function. The purpose of this instrumentation is to. provide for collection of data on the movement and vibration of the RCS'during thermal expansion and contraction. The supports to which the sensors are attached provide support functions for the RCS. The installation of sensors.is performed unobtrusively with an insignificant effect on the system. This instrumentation is installed in a manner that does not alter the design function or characteristics of the RCS.

No new failure modes are created by this installation. The capability of safety-related plant equipment to perform their required safety function is not affected by this modification. This modification does I not affect fuel parameters. This modification adds a negligible mass to the RCS components. The added mass poses no threat to the pressure boundary.

This modification will have no impact on accidents or' malfunctions )

evaluated as the licensing basis and there is no potential for the creation of a new type of unanalyzed event. There is no reduction in the margin of safety.

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Attachment to ET 98-0014 Page 7 of '238 Safety Evaluation: 59 95-0048 Revision: 1 Turbine Laboratory Eye-Wash Modification This modification provides for relocation _of the eye-wash station on the 2010' elevation of the turbine building chemistry' lab. The eye-wash station will be moved to the west side of the room next to the i stairway rail and slightly northeast of the emergency shower. The drain.for the eye-wash will be routed to drain with the emergency shower that is in the same area.

This change does not affect any procedures or administrative controls that govern work activities used for plant alignment, not will it affect the operation of any system, structure, or component as described in the Updated Safety Analysis Report (USAR). Only the ,

domestic water system and its associated drawings are affected. No

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other USAR descriptions or conclusions would be changed or made untrue as a result of this change. This modification does not affect any of I the design basis accidents identified or discussed in the USAR Chapters 2, 3, 6, 9, or 35.  !

This revision will not 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. This revision does not create a possibility for an accident or malfunction .

of.a different type than any evaluated previously in the safety j analysis report. The margin of safety as defined in technical j specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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Safety Evaluation: 59 95-0118 Revisions 1 Block Normal Exhaust Dampers in Equipment Rooms This modification was previously described and reported as Unresolved Safety Question Determination 95-0118 Rev. O. This is the same j modification as Rev. O except that in Rev. O only the exhaust register 1 in room 1512 was to be blanked off, while in room 1501 damper GK D329 was to be locked closed. This modification blanks off the registers in both rooms 1512 and 1501.

These rooms, while in the Auxiliary Building, are adjacent to the i Control Room and ventilated as part of the Control Building. Each room receives 400 cfm from the Control Building supply fan coil unit and has an out take of 350 cfm each to the control Building normal exhaust fans. By design, each. room should have a slight amount of  !

exfiltration to the Auxiliary Building.

l This modificati"n does not affect any safety-related flow balances and will not signif itly affect the Control Building normal supply or exhaust flow balances. These registers and associated ducts are not currently seismically supported and are not located over any safety-related equipment,. Adding a register blank does not create any new threat to any equipment important to safety.

Both the normal supply and exhaust to these rooms are isolated on a Control Room ventilation isolation signal so there is no safety I function affected by this modification. The rooms will be under a slightly more positive pressure than before, but this is  !

inconsequential.

i This modification is essentially equivalent to the design described in the USAR. The change has no significant affect on normal system operation and has no affect on post accident operation. The change has no affect on system failure mode effects.

There is no potential for any design basis accidents to be affected by this change. Neither the presence nor absence of exhaust from the Control Room air conditioning rooms during normal operations can initiate an accident. Blanking the exhaust from the Control Room air conditioning rooms does not change the failure modes or effects of any equipment or any single failure assumptions made in the USAR. These registers and associated ducts are net important to safety, are not currently seismically supported, and are not located over any safety-related equipment.

This revision will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety.previously evaluated in the USAR. This revision does not

Attachment to ET 98-0014 Page 9 of 238 create a possibility for an accident or malfunction of a different type than any evaluated previously in the USAR. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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Attachment to- ET'98-0014 Page 10 of- 238 Safety Evaluation: 59 96-0055 Revision: 0 Delete Essential Service Water Warming Line Valves This modification-removes manual isolation valves EFV0262 and EFV0263 from the Essential Service Water (ESW). warming lines'and replaces them with removable spool pieces. The purpose of this design change is to reduce the turbulence in each of the ESW warming lines to provide more' accurate flow measurement ultrascnic equipment. The accurate flow

. measurement will aid in confirmation of the ice prevention effectiveness effort for the ESW pump bays. Each of these warming lines is provided with two manually operated isolation valves. By removing EFV0262'and,EFV0263, one valve will remain-in each warming line to provide isolation. The function of the ESW warming lines and the ESW system.is not affected by this modification. . The valves are opened in the winter when the lake is cold to' warm the water entering the pump-house and closed during the other seasons.

Based on the above discussion, this modification will not 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. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of .i safety as defined in technical specifications is not reduced by this  !

modification. Therefore, this modification does not involve any l unreviewed safety question.

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l' Attachment to ET 98-0014 Page 11 of 238 Safety Evaluation: 59 96-0067 Revision: 2 Revision to Main Turbine Overspeed Protection Valve Testing Frequency This revision to the Updated Safety Analysis Report revises Surveillance Roquireme"; 16.3.2.1b, items a and b. This revision will

! change the freq cr , of testing the four high pressure turbine stop l valves, six low pressure turbine reheat stop valves,'and six low l pressure turbine reheat intercept-valves from once per 7 days to once per 92 days. These changes were reported in 1996 as USQD 96-0067, Revision 1. The changes to the Updated Safety Analysis Report evaluated by Revision 2 of USQD 96-0067 are identical to the changes evaluated by Revision 1 of USQD 96-0067. Revision 2 of USQD 96-0067 is a reevaluation of the same change.

This revision will not 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. This

, revision does.not create a possibility for an accident or malfunction l of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this L revision does not involve any unreviewed safety question.

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Safety Evaluation: 59 96-0075 Revision: 1 Deletion of Special Reporting Requirement from Relocated Technical Specifications

- This change to Updated Safety Analysis Report. (USAR) Chapter 16, Relocated Technical. Specifications, deletes the requirement to submit a Special Report to the NRC currently in. Operational Requirements 16.3.1.4, " Accident Monitoring Instrumentation," 16.3.1.5, " Loose-Part' Detection System," and'1 6 .7.4.1, " Area Temperature Monitoring."

Specifically, this change will eliminate the' requirement in these relocated. technical specifications to submit a Special Report to the NRC when selected Accident' Monitoring Instrumentation is out of service for more than 30 days, when the Unit Vent High Range Noble Gas Monitor is out of service for more than 7 days, when one or more Loose Part_ Detection System channels is inoperable for more than 30 days, and when the room temperature ~in certain monitored plant areas exceeds prescribed limits for more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or 30*F.

Associated with the chaage to 16.3.1.5 is a change to USAR Chapter 3,-

Appendix A, conformance to NRC Regulatory Guides, to reflect the exception to the special reporting recommendation in Regulatory Guide 1.133, Revision 1, Regulatory Position 5.b, and adding wording to USAR Section 4.4,6.4, Loose Parts Monitoring System, to clarify that the system complies with Regulatory Guide 1.133 with exceptions noted in USAR Chapter 3, Appendix A. In addition, words are added to USAR Chapter 3, Appendix A, to reflect that the USAR 16.3.1.5 does not provide the location of the sensors, as recommended in Regulatory Guide 1.133, Regulatory Position 5.a.

This change will not affect any design basis accidents described in Chapters 2, 6, 9 or 15 of the USAR, and will not affect any system or component operational requirements, design, surveillance requirements or safety limits. This change only deletes the reporting requirement-to the NRC, which is not discussed elsewhere in the USAR. Therefore, there is no impact on accidents and malfunctions evaluated as the

' licensing basis.

Since this change only eliminates a reporting criterion, and does not affect the operation or' design requirements of any system or component, there is no potential for the creation of a new type of unanalyzed event. Nor is there any impact on margin of safety. i

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Attachment to ET 98-0014-Page 13'of 238-Safety Evaluation: 59 96-0095 Revision O Clarification of.USAR Limits for Post Accident sampling System This change to the Updated Safety Analysis Report (USAR) involves the clarification of limits within~the USAR to coincide with the regulatory requirements of Regulatory Guide 1.97, Revision 2

" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs. Conditions During and Following an. Accident," and NUREG-0737, " Clarification of.TMI Action Plan Requirements." The analyses,to be performed by the Post Accident Sampling System were-clarified as.to the location for the analysis (i.e. on-site or off-site depending on dose rates). The change states that the containment hydrogen monitor is not a part.of the Post Accident Sampling System.

These-changes to the USAR will have no affect on the USAR as they only serve to clarify the current 'peration practice and philosophy.

There are no other activities performed in association with the Post accident Sampling System that would make the information contained in the USAR incorrect. The operation of the system is based on NUREG-0737 and Regulatory Guide.l.97, Revision 2, which the USAR will reflect when this change is incorporated.

A review of the accident scenarios in the USAR in chapter 2, 3, 6, 9, and 15 identified no scenarios in which the proposed revision will have any impact. The Post Accident Sampling system is used in post accident scenarios to help evaluate the extent of the accident.

There are no credible accident scenarios'that the proposed revision to the USAR could create. This statement is based on the nature of the change being a documentation only change.

The basis for modification for the limits and clarification of analyses locations is based upon a review of NUREG-0737 and Regulatory Guide 1.97, Revision 2.

Wolf Creek Nuclear Operating Corporation has issued a purchase order to have an off-site lab facility available to perform required analyses in accordance with NUREG-0737 if he dose rates preclude the analyses from being performed on-site.

The Post Accident Sampling System is not one of the initiating event or precursors for any accident evaluated in the USAR. The system is only used to evaluate the severity of an accident after the accident has occurred. As such, since it is not part of the initiation of any accident scenario, there will be no increased probability of an accident due to this revision to the USAR. The system will have no impact on the initiation of an accident, therefore, any radiological consequences would have already occurred as a result of the accident.

Attachment to ET 98-0014 Page 14 of 238 The Post Accident Sampling System panel is a non-safety related component and there 's no safety related nor important to safety related equipment in the location of the panel. There is no impact to f safety related nor important to safety related equipment. This revision to the USAR will have no impact on any malfunction of equipment important to safety. Therefore, there will be no increased radiological consequences associated with this revision.

The change to this non-safety related equipment will not change the j safety classification of.the system. The nature of_the change i combined with the safety classification of the equipment indicates that there will be no impact on any equipment important to safety.

This change to the USAR agrees with the current operating practice and philosophy of Wolf Creek Generating Station and will have no impact on the Technical Specifications.

Therefore, this revision will not increase the probability of l occurrence or the consequences of an accident or malfunction of I equipment important to safety previously evaluated in the safety l analysis report. This revision does not create a possibility for an  !

accident or malfunction of a different type than any evaluated  !

previously in the safety analysis report. The margin of safety as )

defined in technical specifications is not reduced by this revision. l Therefore, this revision does not involve any unreviewed safety j question. '

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Safety Evaluation: 59 96-0122 Revision: 1 l l

High Pressure Feedwater Heater Bypass Throttling Operations This evaluation provides the basis for implementation of the high pressure feedwater heater bypass through the end of plant life. l Procedure SYS AE-125, "HP FW Heater Bypass Throttling Operations" has  ;

been developed to facilitate long term implementation assuming a lower i bounding feedwater temperature of 400 degrees Fahrenheit at 100 percent power. The Updated Safety Analysis Report (USAR) Chapter 15 accident analyses assume initial conditions resulting in the worst case accident analyses. The operation with the feedwater temperature equal to 400 degrees Fahrenheit is bounded by the licensing basis analyses and does not create an unanalyzed condition. Therefore, no  !

Chapter 15 accident analysis is affected. This change does not affect j the ability of any safety-related systems, structures, or components '

to perform their safety-related function.

As a result of this change, the affected USAR section remains complete and accurate. However, a USAR change with regards to operating with a portion of feedwater bypass around the high-pressure feedwater heaters is in progress. This enhancement to the USAR will avoid a possible interpretation of discrepancy between this operational procedure and the USAR description of the feedwater system. The design basis accidents in USAR Chapter 15 have been reviewed. This change does not create the possibility of any new accidents of a type not previously considered in the USAR. l The primary plant operating parameters will only be slightly affected due to the feedwater temperature reduction resulting in a slight decrease in steam pressure. The elight decr vi in steam pressure will result in a slight decrease in steam flow, therefore ensuring that the pressure relieving capacity of the main steam safety valves is not impacted. The main feedwater isolation valves are required to be functional above Mode 4 where the secondary system temperature should be above 307 degrees Fahrenheit when feeding the steam generators. Therefore, the decrease in feedwater temperature will remain within the required operating specificatione since the system  ;

design conditions assume a feedwater temperstare of 446 degrees i Fahrenheit. The main steam isolatio- .alves are required to be operable abcva Mode 4 and thera',re will not be subject to conditions j outside the design basis due to slight changes in steam flow pressure. There are no credible malfunctions of equipment important to safety which may be directly or indirectly affected by this activity.

The current licensing basis analyses bound the proposed reduction in feedwater temperature. Therefore, plant safety analyses are demonstrated to continue to meet the acceptance criteria set forth in

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1 Page 16 of 238 i NUREG-0800 for the applicable ANS Condition event assuming a feedwater temperature of 400 degrees Fahrenheit. Therefore, this change does not result in a reduction in the margin of safety as defined in the bases for any technical specification.

The reduction in feedwater temperature has been evaluated with respect

~ i to the plant's safety analyses. The evaluation demonstrates that the I proposed condition is bounded by the licensing basis analyses. The redr:ed temperature has been considered with respect to plant equipment. Based on this evaluation, the proposed test will not result in the degradation of any plant equipment important to safety.

Therefore, the reduction in feedwater temperature will not increase the probability of occurrence of an accident previously evaluated in the USAR.

Based on the fact that the evaluation of the licensing basis analyses demonstrates that the test condition is bounded, the reduction if feedwater temperature will not increase the radiological consequences i of an accident previously evaluated in the USAR.

The feedwater temperature reduction will not result in the primary plant operating parameters exceeding their analyzed values and the feedwater isolation valves will continue to be operated within their designed temperature range. The feedwater system is designed such that it may be operated with the feedwater heaters either partially or fully bypassed. Therefore, plant equipment important to safety will continue to be operated within the design basis, and the probability of occurrence of a malfunction of equipment important to safety  ;

previously evaluated in the USAR will not be increased.

Based on the "act that the evaluation demonstrates that the probability of a malfunction of equipment important to safety does not increase, and that the analyzed primary plant operating parameters are not impacted, the operating condition will not increase the radiological consequences of a malfunction of equipment important to safety previously evaluated in the USAR.

Based on previous discussion, the reduced feedwater temperature will not create the possibility of an accident of a different type than any previously evaluated in the USAR. This condition will not result in a plant condition outside the designed operation conditions of plant equipment important to safety. Therefore, the reduction in feedwater temperature will not create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the USAR.

Based on previous discussion concerning the current licensing basis and acceptance criteria set forth in NUREG-0800, this condition does not result in a reduction in the margin of safety.

Attachment to ET 98-0014-Page 17 of 238 Safety Evaluation: 59.96-0138 Revision: 0 Revising USAR~ Discussion on Cable Derating This revision to Section 8.3 of the Updated Safety Analysis Report (USAR) - correctly identifies the standards that are used for cable ampacity evaluations and group derating factors. ICEA P-54-440 is used in conjunction with IPCEA P-46-426 for certain cable sizes in randomly filled trays. .This change is merely an editorial change to correct the USAR.

This revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical.

specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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i g _ . Attachment'to ET'98-0014 Page 18 of 238 l-Safety Evaluations. 59 96-0146 Revision 0 Revision.to USAR Table Indicate Flow Interpretation Several notes to USAR Tables 9.2-9, 9.i-10, and-9.2-11 are revised to incorporate additional guidance from design basis documents concerning performance of Component Cooling Water (CCW) system flow balancing, as follows:

1. Table.9.2-9, added Note 4 to clarify that the values ~in the table are nominal design flow rates and actual operating 1 flow rates may vary, i  :.-

2. Added Note 5 to Table 9.2-9, Note 7 to Table 9.2-10, and Note 3 to Table 9.2-11 to clarify acceptable flow rates for the PASS for cooling.

!' 3. Revised Note 3 to Table 9.2-9 and Note 6 to Table 9.2-10 to clarify acceptable minimum and maximum flow rates.

4. Added Note 6 to Table 9.2-9 and Note 8 to Table 9.2-10 to, clarify acceptable minimum flow rate.

5. Changed the CCW flow from 10 to 25 gpm for the SI pump on Tables 9.2-9, 9.2-10 and.9.2-11 to match field flows.

6. Re-totaled the flow and " Duty" totals on each of the tables.

L of the design basis accidents identified in the USAR, only the

' Decrease in Reactor Coolant. System Flow due to a Reactor Coolant Pump p

(RCP) Shaft Seizure' was considered to be potentially affected by this revision. This is because the CCW provides cooling to the RCP lube oil system. However, the clarification of the CCW flows during normal.

operation will not impact this accident because the required cooling

. flow rate to the RCP lube oil coolers will be maintained due to the additional flow margin of the CCW pumps.

This USAR revision will not 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. This l revision does not create a possib'lity for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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Attachment tci - ET 98-0014 Page 19 of 230 Safety Evaluation: 59 96-0151 Revision: 0 Revise Site Flood Analysis Revision to USAR Section 2.4.2.3 and USAR Figures 2.4-9 and 2.4-10 was made to incorporate the results of a reanalysis of the effects of site flooding on safety-related building due to the Prehable Maximum 1 Precipitation (PMP) . Some of tha changes evaluated in the reanalysis include additional obstruc*'onc to rainwater runoff flow paths from building additions to the plant site, past and future paving of areas around the site, and new PMP criteria published by'the NRC in Generic Letter 89-22. The reanalysis resulted in a slight increase in the

. max 4. mum calculated water level due to a PMP. However, the new maximum level remains lower than the original acceptance limit. Thus, the only affects of the reanalysis is to update applicable plant documents to reflect the evaluated changes, ihz;s are no adverse consequences to any safety-related structures, syttems or components from this change.

This USAR revision w il not 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. This revicion does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety j analysis report. The margin of. safety as defined in technical  !

specifications is not reduced by this revision. Therefore, this revision doen not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 20 of- 238 Safety Evaluation: 59 96-0163 Revision: O Circulating Water Pump and Discharge Valve Descrepancies The Updated Safety Analysis Report was. revised to reflect current piant configuration regarding the response of the Circulating Water pumps to & Condenser Pit High Level alarm. The discharge valves for Circulating Water pumps A and C are enabled to close on receiving a Condenser Pit High Level Alarm. Once these valves start to close, their respective pumps will trip. The remaining pump will remain energized and must be secured manually. The flooding analysis for the Turbine Building identifies that even if the Condenser Pit is fully flooded and water starts to spill out onto the 2000 ft. elevation of the Turbine Building, there would be no impact on any safety-related equipment. This'is due to the curbing that exists within the building to protect the safety-related equipment. All the equipment located within the Condenser Pit area in non- ;afety related. This USAR change does not change the Turbine Building flooding analysis.

This' revision will not increase the probability of occurrence or ti.e consequences of an accident or malfunction-of equipment important" o safety previously evaluated in the safety analysis report. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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Safety Evaluation: 59 96-0165 Revision 3 0 3ose Stream Demand Reduction  !

I Revised plant documents, including USAR Section 9.5.1 and USAR Table '

9.5A-1, to reflect a revision to the Fire Suppression System flow l requirement, from 3300 gpm at 80 psi to 2800 gpm at 80 psi, for the j water supply to the fire suppression systems in the power block. This j change in flow rate is based on a change in the flow rate for outside  !

hose stream demands from 1000 gpm to 500 gpm, in accordance with Branch Technical Position CMEB 9.5-1 and NFPA 850. This reduction in l hose stream capability does not constitute a change in the Operating l License or Technical Specifications, and does not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

This change will not 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. This j change does not create a possibility for an accident or malfunction of '

a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications j is not reduced by this revision. Therefore, this revision does not i involve any unreviewed safety question. I I

Attachment to ET 98-0014 Page 22 of. 238 Safety Evaluation: 59 9f-0168 Revision: 0 Auxiliary Steam condensate Recovery and Storage Tank Makeup USAR Section 9.5.9.2.1 was revised to reflect current plant design concerning the makeup water source for the. auxiliary steam condensate

. recovery and storage tank. The UEAR revision added the system's capability to allow the makeup water source for this tank to be provided by the demineralized water storage tank or the condensate storage tank. The USAR did not indicate that this makeup water could come from the demineralized water storage tank prior to this change.

This change updates the USAR to reflect system design and how the system is designed to be operated. There are no plant modifications associated with this USAR revision.

This is an administrative change to a non-safety related system. This change will not affect the operation of any structure, system or component. This administrative USAR revision'will not 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. This revision does not create a j possibility for an accident or malfunction of a different type than  !

any evaluated previously in the saf=ty' analysis report. The margin of safety as defined in technical specifications is not reduced by this j revision. Therefore, this revision does not involve any unreviewed i safety question.

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Safety Evaluation: 59 96-0170 Revision 2 l Containment Cooler "C"-

Revision 2 of this modification provides for the removal of up to I

three cooling coils from service in the containment cooler (with a  !

maximum of five coils per train). Revision 0 and Revision 1 were l previously reported as described below.

Revision 1 of Unreviewed Safety Question Determination (USQD) 59 96- '

0170 removes a maximum of three out of twelve cooling coils from service on containment cooler SGN01C. Revision 0 removed only one coil as is described below.

Revision C of this modification removes one coil of twelve from j service on containment cooler SGN01C. The inlet and outlet header connections will be blind flanged. To make room for the blind flanges, the inlet and outlet nozzles to the coil may be cut away.

The coil is abandoned in place by this modification, and is to be replaced no later than Refuel 9.

The function of the containment cooler is not affected by the replacement of one coil with blind flanges. The initiators of design basis accidents are not influenced by the presence or absence of one '

cooling coil, therefore there is no impact on accidents and malfunctions evaluated as the licensing basis.

The passive nature of the blind flanges precludes any possibility of their creating a new or different accident or malfunction. The margin of safety as defined in the basis for any Technical Specification is not affected.

This modification will not 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. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 24 off 238 Safety Evaluation: 59 96-0173 Revision: 0 Correct Component FJaber - Make-Up Domin System Drawing M-0025, sheet 2 (USAR Figure 9.2-5-02). is revised to correct component duplications pertaining to flow instruments on the makeup demineralizer system. Flow indicators 1FIWM0004, 0005 and 0007 were duplicated on the'M-0025. drawing and in the water treatment facility, causing confusion during both operation of and maintenance on these components. The duplicated numbers were assigned new numbers 1FIWM0267 (for the "A" precipitator recycle water line), 1FIWM0268 (for the "B" precipitator recycle water line), and 1FIWM0269 (for the dry chemical feeders dilution water line).

This is an administrative change to a non-safety related system. This change will not affect the operation of any structure, system or component associated with the water treatme't system.

This revision will not 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. This revision does not create a possibility for an accident or ma. function of a different type than any evaluated previously in the safety analysis report. The margin of cafety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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l Safety Evaluation: 59 96-0174 Revision: 0 j I

Aerofin Coil Replacement for Room Coolers  !

This modification provides for replacement of Carrier cooling coils with Aerofin cooling coils in various type "R" safety-related room coolers, including those in rooms housing the Safety Injection pumps, ,

Residt u % st Removal pumps, Component Cooling Water pumps,  !

Contaic >

Spray pumps, Auxiliary Feedwater pumps, Charging pumps, j and Spent Puel Pool pumps. Also evaluated is the use of AL6XN/ASME SB~

l 676/UNS NO 8367 material in the Aerofin cooling coils. In addition, the use of EPDM gasket material in association with the water box  ;

design for Aerofin type "R" room cooler coils is evaluated.

l The AL6XN/ASME SB-676/UNS NO 8367 tube tuterial used in the Aerofin cooling coils is stronger and thicker than the 90/10 copper-nickel tube material used in the existing Carrier cooling coils, but still l meets the heat transfer, seismic, and fit-up design requirements for the coolers. Also, the new material is expected to be more corrosion-resistant than the copper-nickel material. The EPDM gasket material is almost entirely enclosed by the water boxes on the ends of the room cooler coils. The room cooler components are in mostly mild environments, with a few in harsh environments following a LOCA event.

This material has been tested and found to be able to survive a 40-year service life and six months of post-accident service before failure. However, routine maintenance and cleaning is expected to lead to replacement of these gaskets on a fairly frequent basis, further assuring that the gaskets will not be significantly affected by environmental conditions.

This modification will not affect any procedures, activities, administrative controls, sequences of plant operations, structures, systems, or any requirements outlined, summarized or described in the USAR, with the exception of the description of the tube material.

This modification will not 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. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

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Safety Evaluation: 59 96-0179 Revision 0 USAR Change to Update Fire Hazards Analysis The Updated Safety Analysis Report description of fire area A-18 is being revised totidentify that a separation zone does exist in the fire area, and that use of a local manual controller for valves ABPV002 ("B" steam generator atmospheric relief valve) and ABPV003

("C" steam generator atmospheric relief valve) may be required for fire safe shutdown.

The originai Bechtel fire analysis (electrical fire hazards analysis) l credits a separation zone in fire area A-1B for separation of opposite '

trains of safe shutdown equipment. However, specific discussion of i the separation zone was inadvertently omitted from the USAR fire .I hazards analysis.

As part of the.Thermo-Lag resolution effort, a local manual controller l

was installed for ABPV002 and ABPV003. These controllers allow local i control of the atmospheric relief valves for the "B" and "C" steam generators. This was deemed necessary due to control cables for "A" and "B" train decay heat removal components not meeting the 20 foot separation criteria for fire safe shutdown capability. Adding the manual controllers eliminates the need to protect the affected circuits. This is an administrative change to a non-safety related ,

system. This change will not affect the operation of any structure, l system or component.

This revision will not increase the probability of occurrence or the i consequences of an accident or malfunction of equipment important to i safety previously evaluated in the safety analysis report. This j revision does not. create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety I analysis report. The margin of safety as defined in technical '

specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question, 1

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Attachment to ET 98-0014 Page 27 of 238 Safety Evaluation: 59 96-0180 Revision: O Correct USAR Reference to Main Steam Isolation Valve Failure Mode This revision to.the Updated Safety Analysis Report (USAR) corrects Figure 6.2.4-1 to reflect that the Main Steam Isolation valves fail in the "as-is" position. This revision corrects the identification of the valve position to show the current failure position. There are no

' procedures, activtcies, administrative controls, or sequences of plant operations that are impacted by this change. This change will have no effect on any of the design basis accidents identified in Chapter 6 of the USAR. The acceptance limits defined in the bases for Technical Specifications are not affected.

Based on the above discussion, this revision will not 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. This revision does not create a possibility for an accident or malftnetion of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

Attachment'to ET 98-0014 Page 28 of 238 Safety Evaluation: 59 96-0181 Revision: O seal water Injection Line Drain f

This modification to safety related equipment provides for the addition of a new 3/4"' drain line to line number 158-BCA-2" for PBB01B seal inj :ss line. The existing 1 1/2" line will have a short section of pipe removed and a new reducing tee section added. The Tee section will have a 3/8" orifice for the 3/4" drain line. From the Tee section a new 3/4" pack-less globe valve, a short piece of 3/4" pipe, then a flange and a blind flange will encompass this modification. This new design will be in accordance with ASME Section III'regairements.

Design basis accidents in Updated Safety Analysis Report (USAR) chapters 6 and 15 were reviewed for potential impact by the modification The are no credible accidents created by this activity. New piping and support configuration meets all applicable design codes for ASME/ ANSI USAR section 3.6 and 3.7.

There are no. credible malfunctions of equipment important to safety which may be directly or indirectly affected by the modification. The reactor coolant pump main seals can be supplied by either seal injection water or reactor coolant water which is cooled by component cooling water as it passes up the shaft through the thermal barrier heat exchanger. The small bore break analysis will bound line BB-158- ]

BCA-2" for containment flood analysis. The new Tee section is installed with a 3/8" bore to meet the requirement stated in the USAR, section 5.4 and 10 CFR 50.55a (c) (2) . If the drain line were to shear at the Tee connection, the pressurizer level could be maintained at 2250 psia by the make up capability of a centrifugal charging pump.

There are no Technical Specification limits associated with the drain  ;

line in the reactor coolant syst.em. i Class 1 pipe stress and support analysis for modification show that  !

calculated stress results are less than allowable stress limits. The probability of occurrence has not changed for a design basis j accident. The modification meets all design requirements for Class 1  !

pipe stress and pipe support analysis. The small break LOCA analysis I is unchanged.

The probability of occurrence of a malfunction of equipment important to safety has not changed. The modification will not reduce the reliability of the seal water injection path to reactor coolant pump j PBB01B. The radiological consequences of a malfunction of equipment '

important to safety have not changed.

The modification meets all design reqairements for ASME/ ANSI standards

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.Page 29 of 238 l

and the USAR for piping support design. The modification will not create a different type of accident.

Reactor coolant pumps are designed to operate with or without' seal

. water, injection supply. .The reliability of seal injection line BB-158-BCA-2" has not' changed with this' modification. Therefore, a different t'ype of malfunction of equipment important to safety has not'been created.

Technical Specification 3.4.6.2 (identified and unidentified RCS leakage) and its. associated bases is unaffected by this modification.

Therefore, the margin of safety has not. changed.

This modification.will not 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. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is'not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 30 of 238 Safety Evaluation 59 96-0182 Revision: 0 Digital Control Modification This temporary modification provides for the installation of a digital controller with an I/P converter to provide automatic operation of equipment until a pneumatic control can be obtained. This modification is performed to the Steam Generator Blowdown Regenerative' Heat Exchanger. This heat exchanger cools the liquid flow fro.n the blowdown flash tank (tube side) and recovers heat back into the condensate /feedwater system (shell side). The degree of cooling is controlled by a valve that controls water flow on the shell side of this heat exchanger. The input to the controls is the blowdown water temperature at the outlet of the tube side. The current design uses a pneumatic controller. The pneumatic controller has reached the end of its life.

This temporary modification requires an electrical signal for a temperature input that Updated Safety Analysis Report (USAR) Figure 10.4.8 shows as hydraulic. This change does not alter the design

~ function of the equipment. Only the type of controller is changed.

The Steam Generator Blowdown System is used in diagnosing a Steam Generator Tube Rupture (SGTR). During a SGTR, this component will perform an isolation function to prevent a release to the environment. This temporary modification will not affect the diagnosis of the design basis accidents discussed or referenced in the USAR Chapters 2, 3, or 15.

The system functions are unchanged. This temporary modification will not affect the system failure mode or affect equipment important to safety. Neither are any acceptance limits affected.

Based on the above discussion, this modification will not increase the probability of occurrence or the consequences of an accident or malfunction of e saipment important to safety previously evaluated in the safety analysis report. This modification does not create a possibility for an accident or malfnnetion of a different type than any evaluated previously in the safec3 analysis report. The margin of safety, as defined in technical specifications, is not reduced by this modification. Therefore, this modification'does not involve any unreviewed safety question.

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Safety Evaluation: 59 96-0183 Revision 0 Residual Heat Removal System Description and USAR Description i Descrepancies j i

The Updated Safety Analysis Report ('US AR) and several design. documents '

are being revised to be consistent with each othsr concerning three )

items discovered during the WCNOC Engineering Itapection conducted by ]

the NRC in October, 1996. The specific issues are 1) time required for RHR cooling from 350 degrees F to 140 degrees F, 2) RHR pressure j during normal plant cooldown, and'3) the RCS design heat load and RHR l decay heat following reactor' shutdown calculations. Each of the items was reviewed, the discrepancies between the applicable documents and )

the USAR were resolved, and all documents were revised accordingly. J This change did not involve a change in any design basis. This was an  ;

administrative change to correct discrepancies in the associated documents and to make the design documents and the USAR consistent with each other. j i

These document revisions will not increase the probability of i occurrence or the consequences of an accident or malfunction of equipment important to eafety previously evaluated in the safety analysis report. These revisions do not create a possibility for an accident or malfunction of a different type than any evaluated i previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by these

. revisions. Therefore, these revisions do not involve any unreviewed j safety question.

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Attachment to ET 98-0014 Page 32 of 238 Safety Evaluation: 59 96-0198 Revision 0 Area Temperature Monitoring Surveillance Requirement Correction Updated Safety Analysis Report (USAR) Section 16.7.4.1.1 was revised to delete a reference to Technical Specification 6.8.5. This section is Technical Specification 4.7.12, which was relocated to the USAR in Technical Specification Amendment 89. The reference to 6.8.5 was added to the relocated specification in the original License Amendment Request, due to the intent to add a program reference for the Area Temperature Monitoring Program in 6.8.5. This program reference was deleted in a revised submittal to the NRC for the License Amendment Request, but the added reference to the relocated specification was not deleted as part of the revised submittal. This change is administrative only to correct the USAR section to reflect Technical Specification Amendment 89. This change does not affect any plant design or system or component operating parameters or procedures.

This document revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in  ;

technical specifications, is not reduced by this revision. Therefore, '

this revision does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 33 of 238 Bafety Evaluation: 59 96-0201 Revision: 0 Flood I9vels in Safety Related Rooms Up dated Safety Analysis Report ('US AR) . Table 3.6-6 is being revised to show flood levels of l' 4" instead of l' 9" above the floor for Rooms 1411 and 1412. .This change is to make the USAR consistent with the current flooding analysis (LF-FH-002, Rev. 1 and FL-04, Rev. 0) . The flood level for Rooms 1508 and 1509 is being reduced to "0" since they have open grating floors and any water (e.g., from a pipe break) would flow to the rooms below (Rooms 1411 and 1412). The '.aximum flood level is based on the feedwater line break in Rooms 1411 and 1412, which envelopes any pipe break in the rooms above them. There is no-change to the design or configuration of the plant. Only the USAR is affected to make it consistent with current plant design. This is an administrative change to a non-safety related system. Th'.s change will not affect the operation of any structure, system or component.

This document revision will not 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. This revision does not create a possibility for an accident or. malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as-defined in technical' specifications,~is not reduced by this revision. Therefore,-

this revision.does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 34 of 238 Safety Evaluation: 59 96-0202 Revision 0 Essential Service Water Strainers Differential Pressure CCP 07140 corrected USAR Table 9.2-5 and vendor and design documentation to reflect actual plant conditions with regard to ESW strainer pressure drops (differential pressure). Start-up test data for both ESW strainers (FEF02A/B) shows that the original clean i pressure drop was less than 1.0 psi. USAR Table 9.2-5 listed the clean pressure drop as 1.1 psi and the dirty pressure drop at the start of automatic backwash as 3.0 pai. Historical data from the strainers' computer data points, taken since 1993, shows that the .

clean pressure drop varies between 3.0 to 3.5 psid. The safety limit I setpoint, based on NRC Regulatory Guide 1.105, is 6 paid. To account for instrument uncertainty, the safety-limit setpoint was chosen as 5.56 paid. PMR 0903 was issued in 1985 to incorporate these setpoints into design documents, but did not identify that a USAR change was needed to correct Table 9.2-5. The vendor recommends that the backwash be started at 2 psi above the clean pressure drop. Assuming that the clean pressure drop is 3.0 psid, the maximum setpoint for initiating the backwash (5.56 psid) is 2.56 psi higher than 3.0 psid, The higher differential pressure, 2.56 versus 2 psi, was evaluated and determined to not be a concern due to the high backwash pressure for the strainers and because the backwash water is discharged directly to the lake through a 3-inch valve. In addition, all system openings (e.g., the heat exchanger tubes) are larger than the strainer slots, such that anything being forced through the strainers due to the higher pressure would be smaller in diameter than the system openings. Also, flow rates through the system heat exchangers and room coolers are maintained for MIC control, and periodic monitoring of the heat exchangers and room coolers is performed to identify performance degradation. Evaluation shows that the increased differential pressures do not adversely affect the performance of the {

ESW strainers or the ESW system.

This modification will not increase the probability of occurrence or l the consequences of an accident or malfunction of equipment important l to safety previously evaluated in the safety analysis report. This j modification does not create a possibility for an accident or j malfunction of a different type than any evaluated previously in the  !

safety analysis report. The margin of safety, as defined in technical j specifications, is not reduced by this modification. Therefore, this t modification does not involve any unreviewed safety question. l i

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Attachment to ET 98-0014 Page 35 of 238 Safety Evaluations. 59 96-0204 Revision 3 0 Change to Evaluation of Steam Generator Tube Rupture Radiological Consequences Up dated Safety Analysis Report (USAR) Tablec 15.6-4 and 15.6-5 were revised to include an additional radioactive release flow path to the atmosphere via the exhaust stack of the Turbine-Driven Auxiliary Feedwater (TDAFW) Pump. The analysis of radiological consequences resulting from a steam generator tube rupture (SGTR) accident was based on the worst-case scenario of a forced steam generator overfill with a safety valve stuck open. In the SGTR scenario, loss of offsite power is assumed to occur coincident with the reactor trip and no operator action is assumed for the first 16 minutes. The loss of offsite power causes valves to open that supply steam to the TDAFWf pump. This results in an additional radioactive release flow path to the atmosphere via the exhaust stack of the TDAFW pump for this 16 minutes. However, this additional release was not included in the Chapter 15 analysis. The increasc in the calculated offsite doses resulting from this additional radioactivity release during the postulated SGTR accident, is insignificant and the new calculated' doses remain significantly below the guidelines of Standard Review Plan (NUREG-0800) Section 15.6.3. The USAR change does not involve any plant design change, any plant hardware change, nor does it change the method by which any safety-related plant system performs its safetyifunction.

This document revision will not 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. This revision does not create a possibility for an accident  !

or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 36 of 238 Safety Evaluation

• 59 96-0206 Revision 0 Thermo-Lag Down Grade to Combustible Material Updated Safety Analysis Report (USAR) Chapter 9.5, Appendix B, the Fire Hazards Analysis, was revised to include Thermo-Lag 330-1 material as a fixed combustible material, per the guidance provided in l NRC Information Notice 95-27. Several fire areas are affected by this i change. This USAR section was also revised to correct a room number I in Section A.25.3.1 in Appendix B. Room 1322 was added to Section  ;

A.25.3.1 to reflect two drums of hoses stored in the room. The l fireload added by consideration of Thermo-Lag as a combustible l material is within the bounds of the existing fire barrier analysis. l This document revision will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment j important to safety previously evaluated in the safety analysis  ;

report. This revision does not create a possibility for an accident  !

or malfunction of a different type than any evaluated previously in l the safety analysis report. The margin of safety, as defined in j technical specifications, is not reduced by this revision. Therefore, I this revision does not involve any unreviewed safety question,

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Attachment to ET 98-0014 Page 37 of 238-Safety Evaluation: 59 96-0237 Revision 0 Fire Zone Safe Shutdown Capability During performance of the Thermo-Lag Resolution project an electrical fire hazards analysis was performed to validate areas of the plant

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containing fire wrap material for compliance with 10 CFR 50 Appendix R. Fire areas A-15, A-16, and A-33, described in U dated p Safety-Analysis Report (USAR) Section 9.5B, were found to have incorrect information with regards to design basis information and were revised. The revisions were editorial in nature and made to provide more accurate representation of the safe shutdown capabilities of the  !

equipment bounded by these fire areas. Area A-15 was corrected to I reflect that a raceway originally assumed to be routed through the area was not. Information concerning monitoring parameters found to be missing from Area A-16 was added. Fire area.A-33 had a wet pipe suppression system installed which was never added to the USAR description. This added system enabled a 3-hour rated fire wrap in the area to be downgraded to a 1-hour wrap. This.information was added to the.USAR. These changes are administrative changes to make j the descriptions of the fire areas more accurate and to reflect design I information.

This document revision is administrative in nature and will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question. l l

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Attachment to .ET 98-0014 Page 38 of' 238 Safety Evaluation: 59 96-0210 Revision 0 Engineering Organisation Change This Updated Safety Analysis Report (USAR) revision provides for combining the Nuclear Engineering organization with the Regulatory Services organization'to form an new organization called Nuclear Engineering Safety and Licensing. The Manager Nuclear Engineering is now the Manager Nuclear Engineering Safety and Licensing. The Manager Regulatory Services position has been eliminated. The Regulatory Compliance department has been combined with Licensing and placed under the direction of the Manager Nuclear Engineering Safety and Licensing.

The elimination of merging of Licensing with regulatory Compliance impacts USAR Section 13.1.1.2.2 which identifies Regulatory Compliance as a responsibility of the Vice President engineering. Also' Figure 13.1.1 will be revised to reflect this change.

These organization. changes are administrative in nature. The

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functions of Regulatory Compliance will not be eliminated by this organizational change, only the title will be eliminated.

No accident analyzed in the USAR relies on organization for mitigation. Equipment important to safety are not directly.or indirectly affected by this revision. Changes to these organizations will not create new accidents because the personnel and organizational ')

functions are not being changed. l l

Based on the above discussion, this revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously .1 the safety analysis report. The margin of safety, as defined in techaical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed l safety question. j I

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L Attachment to ET 98-0014 I Page-39 of 238 l

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safety Evaluation: 59 96-0212 Revision: 0 Fire Hazards Analysis Update A Thermo-Lag fire barrier had been installed on conduit IJ2027 in l Containment to meet the 3-hour fire barrier requirement from Appendix L A of BTP APCSB9.5-1 to ensure one train of pressurizer level i indication is preserved in the event of a fire. It was discovered during the Thermo-Lag Resolution program that this barrier was actually required to be a " noncombustible radiant energy shield."

Thus, due to the combustible material properties of the Thermo-Lag material-discovered by. industry testing, this material was replaced by CCP 07037 with a non-combustible material, Darmatt KM1. USAR Appendix 9.5B was revised to reflect the installation of the noncombustible radiant energy shield. This change does not affect system operation or the basis for any technical specification.

This revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety

. analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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Attachment to 'ET 98-0014 Paga 40 of 238 Safety Evaluations 59 97-0001 Revision 0 Refuel Machine Bridge Speed Increase

-This modification changes potentiometer / software for the refueling-machine to enable the bridge *;o operate in the range of zero to sixty feet per minute. In addition, the refueling machine trolley and refueling machine hoist will be able to operate in the range of zero to_ forty feet per minute.

This modification has been evaluated showing that the current acceleration / deceleration used by the refueling machine programmable logic controller (PLC) remains unchanged by the increased speed modification and therefore will result in no detrimental effects on the fuel assenblies. The PLC controls the allowable speeds of the refueling machine main hoist and allows the maximum speed of 40 fpm only when.the fuel assembly is being raised in open water (not in contact with other fuel assemblies or pressure vessel components). In the event the PLC fails, the main hoist reverts to a maximum speed of approximately ten feet per minute in the bypass mode. The maximum speeds in the bypass mode of operation are not being changed.

This modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysit report. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

i Attachment to ET 98-0014 Page 41 of 238 1

1 Safety Evaluation: 59 97-0002 Revision: 0 Fuel Transfer nyrtem Drive Upgrade The current shr.ft/ chain drive for the fuel transfer system will be ,

replaced by a cable drive system. The shaft / chain drive system, with l its associated components, will be replaced by a cable / winch system with associated sheave modules to direct the cable in its desired l pattern. The upender consoles in the fuel and reactor buildings will

.be replaced with new consoles and associated communication cables.

The current communication cable will be left between the two upenders, j but will be used only if BYPASS of the Programmable Logic i Controller (PLC) is desired to continue operation of the system. A fiber optic cable will be added to provide normal communication l between the reactor building and fuel building upender consoles. The upender console in the fuel building will be relocated on the opposite end of the upender skid (North end). The new console will use a bolt j arrangement to fasten it to the existing upender skid. The necessary I additional cable length will be obtained by installation of a terminal box and terminal strip. l Relocation to the North end of the skid will enable the operator to observe the upender and transfer components while standing at the-console. Additionally, it will allow work in the console to be done in a clean area without requirement for PC's during normal conditions.

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Current proximity switches and their support structures at both upenders will be replaced by dual contact proximity switches with underwater connectors. The current conduit between the proximity switches support structure below water level and the junction box on the operating floor will be removed (both reactor and fuel building).

The current proximity switch carbon steel actuators will be replaced by stainless steel encased magnetic actuators to. provide increased actuation response. The transfer tube gate valve proximity switch / actuator / bracket will be replaced by the same proximity switch / magnetic actuator as on the transfer system to provide gate valve position indication. Means will be added to the Spent Fuel Pool Bridge Crane and the Refueling machine to allow AUTOMATIC REMOTE initiation of the transfer system. Torque switch overload protection will be replaced with a load cell and PLC load monitoring components.

The PLC will control the operation of the upender, acceleration and deceleration of the transfer carriage, and the overload protection functions for the transfer system. The transfer system traverse speed

-will be increased to a maximum of approximately 50 fpm. Additional required instrument and power cables will be pulled in current conduit ,

along current cable routes. The VERSA four way valves, which orovide I means to control the upender cylinders, will be replaced with ASCO four way valves which have been shown more reliable in similar applications at other plants. The changes made to the system are l

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Attachment ~to. ET 98-0014 Page 42 of 238 expected to increase the reliability and ease of maintenance of the system.

The.USAR contains Figure 9.1-12 which shows the. current shaft / chain drive transfer system. This figure will be deleted because it provides no additional safety related information and provides no needed information with' regard to the transfer system. The USAR describes the transfer system on pages 9.1-39 to 9.1-41. Descriptions on these pages.will need to be updated to keep them current and accurate. The transfer system is classified as Non-Nuclear. Safety Relatti (NNSR). The changes being made to the transfer system do not change the structural components which support the fuel assembly, therefore, they will have no affect on preventing or mitigating the consequences of any accident. The speed increase of the system will not affect any previously analyzed accident because doceleration and acceleration is controlled by the PLC. Failure or Bypass of the PLC results in.the speed of the system reverting back to the current or lower maximum speeds. Accident analysis is based on the dropping of a fuel: assembly and not on the speed or accelleration/decelleration of the transfer system. The transfer tube gate valve proximity switch / actuator is for indication only and does not affect the ability to perform containment isolation.

There are no acceptance limits given for the transfer system in the bases for the technical specifications or any other licensing basis documents.

The transfer system is not designated as safety related. The proposed change doer not modify the means to clear the transfer tube of the transfer equipment. Failure of the proposed modification will not impact containment isolation requirements. Therefore, the proposed j

-modification cannot create the possibility of a different type of' malfunction of equipment important to safety than any previously

' evaluated in the USAR.

The transfer system is not described in the basis for any technical specifications. The proposed change does not impact the ability to achieve containment isolation.

This modification will not 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. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the

' safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 43 of 238 Safety Evaluation: 59 97-0003 Revision 0 Third stage Extraction to Heaters Piping Replacement

'This. modification provides for the replacement of a portion of piping

. lines in the Main Turbine System to mitigate abnormal pipe-wall

,. thinning because of Flow Accelerated Corrosion. This change consists of. replacing existing worn carbon steel piping components with low alloy steel which is more resistant to wear caused by Flow Accelerated Corrosion. Piping cross sectional properties, mechanical properties,

.and geometric configuration will remain unchanged. The change is an enhancement to the original design by providing an increased resistance to. flow accelerated corrosion.

This modification will not 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.. .This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 44 of" 23B

' Safety Evaluation: 59 97-0004 ~ Revision 0 High Pressure Turbine Extraction to Meater Piping Replacement This modification provides for.the replacement of a portion of piping.

lines in the_ Main. Turbine System to mitigate' abnormal pipeiwall thinning because'of Flow Accelerated Corrosion. This change consists

.of replacing existing worn carbon steel piping components with low alloy steel which is more_ resistant to wear caused by Flow Accelerated Corrosion. Piping cross sectional' properties, mechanical properties, and geometric configuration will remain unchanged. The change is an -

enhancement to the original design by providing an increased resistance to fisw accelerated corrosion.

This modification will not 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. This modification does not create a possibility for an accident.or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this modification. Therefore, this modification does not. involve any unreviewed safety question.

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' Attachment to ET 98-0014

'Page 45 of 238 i l

Safety Evaluation: 59 97-0005 Revision 0 I I

Clarification of Wording in USAR Table 7.5-2 '

i This revision to the Updated Safety Analysis Report (USAR) clarifies '

information in Table 7.5-2 (Safe Shutdown Information) . The existing I USAR information indicates that all three Auxiliary Feedwater Pump rooms have temperature annunciation in the main control room. Section 16 of the USAR requires operators to monitor room temperatures every j 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The design conditions do not provide this annunciation for the Turbine Driven Auxiliary Feedwater Pump room. Annunciation is provided in the two motor driven Auxiliary Feedwater Pump rooms via

.the balance of plant computer system which has no safety related or technical specification requirements. The actual determination of operability is accomplished by operator reading and logging readings ,

from caliorated thermometers mounted in each of the pump rooms.

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This revision will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to j safety previously evaluated in the safety analysis report. This l revision does not create a possibility for an accident or malfunction i of a different type than any evaluated previously in the safety 1 analysis report. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this

revision does not involve any unreviewed safety question. l l

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-Attachment to ET 98-0014  !

Page.46 of- 238 Safety Evaluation: 59 97-0006 Revision 0

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Radwaste Updated Safety Analysis Report Update This revision to the Updated Safety Analysis Report (USAR) provides for the following changes:

USAR Section'9.3.6.2.3: Removed steps a., b., and c. renumbered the remaining steps, and removed statements related to realigning valves in step f. from those activities observed when venting either of the recycle hold-up tanks (RHUT). This change will allow the flexibility to vent the RHUTs at any time without isolating them. The isolation of the tank and evacuating the RHUT gases is unchanged whether the tank is isolated or not. Monitoring of the evacuated gases is unchanged; therefore, there are no new expected effects from this  ;

change. '

USAR Section 10.4.10.2.3: Added the ability to drain the secondary liquid waste monitoring tank (SLWMI) to the dirty radwaste sumps and reprocess this fluid through the floor drain tank and liquid radwaste demineralizer skid (DTS). This process gives Operations personnel an alternative means of reprocessing water when it does not meet the release limit requirements. The addition of utilizing the DTS for drainage from the SLWMT does not affect the design basis or processing through this skid.

USAR Section 11.2.2.1: Added a statement that the floor drain tanks are processed through the DTS. In the past, the waste was sampled to determine the best route for processing. 'This is no longer required since the evaporators are no longer used and DTS resin skid is utilized for processing this fluid. The addition of processing floor drain tank fluids does not affect the design basis or processing through this skid.

USAR Section 11.3.2.1: Revised to clarify that a continuous purge of the volume control tank (VCT) is not performed and one tank in the gaseous radwaste system (GRWS) is used for shutdown gas decay. The effect of purging continuously or intermittently is unchanged by this

revision. The specification of the use of one gas decay tank for l shutdown does not change the design basis or functions previously l determined for this equipment.

1 USAR Section 11.3.2.3: Added a statement that the purge of the VCT is performed when directed by Chemistry and deleted the statement that the gas decay tanks are switched at on two day intervals. This change accurately reflects when the VCT is purged and removes specifics on the time frames of when the gas decay tanks are switched to distribute radioactive gas inventory in the system through all tanks to mitigate the possible consequences of a gas decay tank rupture. The function t

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' Attachment to ET 98-0014 Page 47 of- 238 and consequences'related to purging are unchanged. Likewise, the switching time between tanks is dependent on the gas input to the system.which does not change any functions on consequences.

USAR Sr?* don 11.4,2.3.1: Revised this section to describe that reactor mak2-up c Ater is used as the motive force for sluicing resins. The original design of the system utilized diaphragm pumps. These pumps are no longer used and instead,' reactor make-up water is the driver' for the resin to the waste disposal station. This overall process is unchanged as well as the effects and consequences for changing the specific drive of the_ resin.

These changes were made as a result of a new and improved technology in the industry for processing radweste and because the original equipment is not being utilized as designed. The affected procedures were_already r.ccurate in the operational description or have been reviseo to ;eflect these changes.

No new equipment is being added, removed, or modified by this change that would affect the design basis, functions, failure modes, sampling requirements, release limits or consequences related to the radwaste systems by these clarifications. In addition, all the components referenced in this change are either non-safety related or special scope (pressure boundary only) . Therefore, no safety related equipment or any nther equipment important to the safe and reliable operation of Wolf Creek Generating Station is affected by this change.

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Attachment to ET 98-0014 Page 48 of 238 Safety Evaluatfon: 59 97-0007 Revision 0 Turbine Driven Auxiliary Feedwater Turbine Gland Drains Modification Rev sion 0 of this modification was reported by USQD 93-0135 as follows:

This modification is being implemented to minimize the amount of steaming which occurs when the Turbine Driven Auxiliary Feedwater (TDAFW) Pump (PALO2) is in operation. This modification will install a steam trap on the associated TDAFW drain line that caused the steaming condition. This will eliminate the emission of live steam from the craining system. In addition, an atmospheric vent line will be added to assist in the elimination of backpressure on the steam leakoff lines and the turbine gland seal drains. This modification affects only non-safety related piping and components, which are seismically restrained such that they will have no adverse impact on safety related equipment.

l This modification will have no impact on accidents or malfunctions evaluated as the licensing basis and there is no potential for the l crettion of a new type of unanalyzed event. There is no reduction in j the margin of safety.

Revision 0 of this modification was not implemented and Revision 5 provides the following design changes:

l The new design disconnects the common turbine glands leak-off from common drain header FC065-HBD-2. The gland leak-offs are combined to form a common drain FC096-HBD-1 which penetrates the floor of Room 1331 and drainc into the existing equipment drain LE005-HCD-4 in the basement. The highest point of the drain is connected to a vent which is routed the same way as in the previous design. This design helps to improve the performance of the turbine glands and minimizes the amount of steam that will be released in the baseaent Room 1129.

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In order to further reduce steam in the basement of the Turbine Driven Auxiliary Feedwater Pump room, the new design requires that the steam ring train valve FCV013 be closed. This drain has been the source of significant steam release. Closing of the steam drain has no adverse impact on the turbine.

The amount of steam that is vented to the environment as a result of j this modification is insignificant compared to the steam that is l assumed to escape from a stuck safety valve in the accident analysis.

Therefore, this modification does not impact any accident analysis.

The closure of valve FCV013 has no credible failure modes. If the valve were to leak by, the condition created would be no different

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p Attachment to ET 98-0014 Page 49 of 238 than the current condition.

There is no adverse impact on any safety related equipment or equipment important to safety. There are no acceptance limits l contained in Technical Specification bases or other licensing basis  !

documents that apply to this modification.

Attachment'to' ET 98-0014 Page 50 of 238 Safety Evaluation: 59 97-0008 Revision O Clarification of Freeze Protection for Plant Tanks The Updated Safety Analysis Report (USAR) and system descriptions are not entirely accurate regarding the design and operation of the freeze protection provided for several plant tanks. Steam heating coils are provided by the design for the tanks to prevent freezing during cold-weather. The steam flow to each tank'was designed to be automatically controlled by temperature control valves. The valves modulate to maintain a nominal tank temperature.of 50 degrees Fahrenheit. During periods of cold weather, the heating coil condensate return lines on these tanks have frozen preventing proper operation of the system. To -

prevent this, the control valve bypass lines are opened during cold weather to provide a continuous steam flow. During 1990, the safety injection pump return line became blocked with ice due to inadequate freeze protection.

This change will clarify plant documentation regarding how the affected components are designed and operated. The clarification deals with the minimum temperatures for the Condensate Storage Tank, the Refueling Water Storage Tank, and the Reactor Makeup Water Storage Tank, and how they are maintained.

The tank temperature safety limits have not been changed. Therefore, the probability of occurrence of an accident is not increased by this change. Since the tank temperature safety limits have not been changed, the change to the USAR cannot affect the design basis accidents identified in the USAR. The function of the emergency core cooling system pumps and the auxiliary feed water system is not jeopardized. Therefore, consequences of an accident previously evaluated in the USAR have not been increased since assumptions previously made in evaluating radiological consequences of accident have not been affected.

Attachment to ET 98-0014 Page 51 of 238 l

l Safety Evaluation: 59 97-0009 Revision: 0' Radiological Emergency Response Plan Revision Revision 53 to the Radiological Emergency Response Plan provides for a l Change to the Emergency Action Levels which changes the response for l

'the "NO" path from ALERT to SITE AREA EMERGENCY. This change is being i made to correct an error made when implementing Revision 47 or the ,

Radiological Emergency' Response Plan. In addition, in Emergency I Action Level 8-SSFM5, the values for the Steam Generator narrow range

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were changed from 5% to 6% to be consistent with the Emergency operating Procedures. This is an administrative change. This change will not affect the operation of any structure, system or component.

This revision will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to l

safety previously evaluated in the safety analysis report. This j revision does not create a possibility for an accident or malfunction  !

of a different type than any evaluated previously in the safety I analysis report. The margin of safety as defined in technical I specifications is not reduced by this revision. Therefore, this l revision does not involve any unreviewed safety question.  !

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Attachment to ET 98-0014 Page 52 of 238 Safety Evaluation: 59 97-0009 Revision: 1 i

Radiological Emergency Response Plan Revision This evaluation is performed to document the performance and review of Revision 0 of this USQD by qualified personnel. Revision 0 was performed by unqualified personnel. Revision 0 is described below:

Revision 53 to the Radiological Emergency Response Plan provides for a change to the Emergency Action Levels which changes the response for the "NO" path from ALERT to SITE AREA EMERGENCY. This change is being made to correct an error made when (?plementing Revision 47 or the Radiological Emergency Response P?- In addition, in Emergency Action Level 8-SSFMS, the values for the Steam Generator narrow range were changed from 5% to 6% to be consistent with the Emergency Operating Procedures. This is an administrative change. This change will not affect the operation of any structure, system or component.

This revision will not 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. This revision does not create a possibility for an accident or malfunction  ;

of a different type than any evalutted previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 53 of 238 Safety Evaluation ' 59 97-0011 Revision 0 Organistion Change to Reflect New Chief Executive Officer Design Change Package 06001, Revision 0, was reported to the NRC in the Annual 50.59 Report dated, March 8, 1996. The evaluation was summarized as follows:

"This non-safety related modification provides for the installation of four relief valves 6n the Process Sample Panel (RM 172). As a result of a plant trip flow metets on the RM 172 Process Sample Panel were damaged. This modification will-provide pressure relief to preclude recurrence of this event.

'niis modification will. have no impact on accidents or malfunctions evaluated.as the licensing basis and there is no potential for the creation of a new type of unanalyzed event. There is no reduction in the margin of safety."

Design Change rackage 06001, Revision 1, expands this modification to install relief valves around the high energy sample flow line indicators RMFIOS69, 0570, 0571, 0572, 0573, 0575, and 0587 to prevent high pressure from causing damage to sample panel RM 172.

The Process Sample System is a non-safety related system. Failure of any instrument tube or gauge will have no impact on plant safety or a radiological release. This modification is being implemented for personnel safety and elimination of a potential non-radiological release path.

Locations of the process sample connection lines are down stream of isolation valves. During an accident event such as a steam generator tube rupture, the isolation valves will close, thus isolating the process sample system from any radiological release consequences.

Therefore, the operation or failure of these relief valves will not affect cnr create any credible malfunctions of any safety related system, structure, or component. There are.no acceptance limits contained in the bases for technical specifications or licensing basis documents that could be affected by the addition of these relief valves to the non-safety.related Process Sampling System.

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I Attachment to ET 98-0014 Page 54 of 238 I

Sa"ety Evaluation 8 59 97-0012 Revision 0 Addition of Relief Valves for Protection of RM 172 Panel Design Change Package 06001, Revision 0, was reported to the NRC in j the Annual 50.59 Report dated, March 8, 1996. The evaluation was i summarized as follows:

"This non-safety related modification provides for the installation of four relief valves on the Process Sample Panel (RM 172). As a result of a plant trip flow meters on the RM 172 Process Sample Panel were damaged. This modification will provide pressure relief to preclude recurrence of this event.

This modification will have no impact on accidents or malfunctions evaluated as the licensing basis and there is no potential for the creation of a new type of unanalyzed event. There is no reduction in the margin of safety. "

Design Change Package 06001, Revision 1, expands this modification to j install relief valves around the high energy sample flow line indicators RMFIOS69, 0570, 0571, 0572, 0573, 0575, and 0587 to prevent high pressure from causing damage to sample panel RM 172.

The Process Sample System is a non-safety related system. Failure of any instrument tube or gauge will have no impact on plant safety or a radiological release. This modification is being implemented for personnel safety and elimination of a potential non-radiological release path.

There are no tests or experiments not described in the USAR which may adversely affect the adequacy of the systems, structures, or components to prevent accidents or mitigate the consequences of an accident.

Locations of the process sample connection lines are down stream of isolation valves. During an accident event such as a steam generator tube rupture, the isolation valves will close, thus isolating the process sample system from any radiological release consequences.

Therefore, the operation or failure of these relief valves will not affect or create any credible malfunctions of any safety related system, structure, or component. There are no acceptance limits contained in the bases for technical specifications or licensing basis l documents that could be affected by the addition of these relief valves to the non-safety related Process Sampling System.

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Attachment to ET 98-0014 Page 55 of 238 Safety Evaluation *- 59 97-0013 Revision: 0 Drawing correction to Update for Previous Modification.

'This revision to the Updated Safety Analysis Report, Chapter 9.5B,

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provides a generic der:ription of hydraulic fluid for the actuators of the main steam and feedwater isolation valves instead of a trade name description. This is a drawing change only. No field work or change in material is required. This is an administrative change. This change will not affecc-the_ operation of any structure, system or component.

This revision will not increace the probability of occurrence or the l

consequences of an accident or malfunction of equipment'important to safety previously evaluated in the safety analysis report. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety an41ysis report. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this

/ revision does not involve any unreviewed safety question.

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U Attachment to ET 98 0014 Page'56 of 23' i

. Safety Evaluation: 59 97-0014 Revision 0 Radiological Emergency Response Plan Revision This revision to the Radiological Emergency Response Plan RERP provides for a change to Emergency. Action Levels (EALs). Block.1RER1,.

1RER2, and 1RER6 were revised such that'a lower value of radioactive release is needed to satisfy the blocks. This was required because of the fact that the original values were so'high that it would be possible to have a release rate which was not high enough to answer "Yes" to block 1RER2 but was higher than the values listed in 1RER3.

This could lead to a lower' Emergency classification than conditions required. This revision lowers the effluent release rates necessary for a Notification of Unusual Event'and lowers the effluent release rate necessary for an Alert classification.

This revision does not affect any plant equipment. The revision does not affect operator response to equipment malfunctions. Only EALs are affected. This revision will result in a more conservative classification of an emergency based on radioactive effluent release rates.

l Based on this evaluation, this revision will not increase the l probability of occurrence or the consequences of an accident or i malfunction of equipment important to safety previously evaluated in the safety analysis _ report. This revision does not create a l possibility for an accident or malfunction of a different type than  ;

any' evaluated previously in the safety analysis report. The margin of l safety as defined in technical spec'ifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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I Attachment to ET 98-0014 Page 57 of. 238 l

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Safety Evaluation: 59.97-0017 Revision: 0 , l Revise Drawings to Add Piping Enhancements and Clarify safety classifcations This change to the Updated Safety Analysis Report revises Piping and Instrument Diagram M-12AL-01, "P&ID Auxiliary Feedwater' System," to reflect the as-built configuration and to enhance references for Auxiliary Feedwater Pumps Seal Water Piping. In addition, this change j will clarify the safety classification of vents and drains on the ]

discharge side of the' Auxiliary Feedwater Pumps. This is a drawing l change only, and no field work is required.

.l No design basis accidents described in the USAR are affected by this change. No new types of accidents could be created by this change, l because this is a drawing change only. .No new types of credible {

malfunctions of equipment important to' safety could be created by this  !

change because this is a drawing change only. There are no bases for Technical Specifications that are affected by this' change. No accidents nor radiological consequences of an accident are affected by.

this change. No new components, modifications or changes to the operation of the Auxiliary Feedwater System are cluded with this .;

proposed change. No' equipment important to safety is affected by this

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i This change will not increase the probability of occurrence, nor the  ;

consequences of an accident, nor malfunction of equipment important to j safety'previously evaluated in the safety analysis report. This l change does not create a possib'ility for an accident or malfunction of j a different type than'any evaluated previously in the safety analysis j report. The margin of safety, as defined in technical specifications, i is not reduced by this change. Therefore, this change does not' involve I any unreviewed safety question.

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Attachment to ET 98-0014 Page 58 of 238 Safety Evaluation: 59 97-0018 Revision: 0 USAR Change to Update Figure 10.4-10 This revision to the U dated p safety Analys'is Report provides for a revision to Figure 10.4-10. This change revises component numbers and connections of the Turbine Driven Auxiliary Feedwater pump. (TDAFWP) speed control instrument loop, as it is shown'on plant drawings, and reflects the as-built condition. This revision is implemented for the purpose of correcting Figure 10.4-10 only. No field work will be performed. This is an administrative change. This change will not affect the operation of any structure, system or component.

This revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type _than any evaluated previously in the safety analysis. report. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 59 of 238 Safety Evaluation: 59.97-0020 Revision: 0 Accumulator Tanks Operating Temperature Limit This modification evaluates, determines, and documents the lower limit of the temperature range at which the Safety Injection Accumulator Tanks can safely operate. Evaluation includes-any potential impact on the existing Loss of Coolant Accident analysis. This modification ,

revises the Updated Safety Analysis Report (USAR) and other; design ~I documents to reflect this acceptable condition. In addition, the material description of the Accumulator Tanks is clarified in the Updated Safety Analysis Report Table 6.3-1 to be consistent with Updated Safety Analysis Report Table 6.3-4 This change lowers the minimum temperature for the Accumulators based on an analysis which shows these components'could perform safely at the revised temperature limit. Accidents previously evaluated in the USAR are independent of this change. Therefore, this change can not increase the probability of occurrence of an accident previously-evaluated in the USAR. l No functions.of any systems, structures or components are adversely affected by this change. Therefore this change does not affect any assuntptions made to radiological consequences of an accident pre 1 iously evaluated in the USAR.

Since the functions of SSCs are not adversely affected by this change, -j the probability of occurrence of a malfunction of equipment important i to safety are not increased. Since no malfunctions were identified, j the consequence 3 of a malfunctions in not affected by this change. l There is no potential for the creation of a new type of unanalyzed j event.

The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

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.Page 60 of 238-l 1

Safety Evaluation: 59 97-0021 Revision: 0 Drawing Corrections on Compressed Air System j i

This modification revises Piping and Instrument Diagrams and their j corresponding Updated Safety Analysis Report Figures which reflect the "

Compressed Air System. These revisions are for the purposes of l correcting component labeling duplications. These changes have no effect on any system, structure or component. This'is an' administrative change only, and requires no changes in the plant.

I-This modification will~not increase the probability of occurrence or l the consequences of an accident or r.alfunction of equipment important i to safety previously evaluated in the safety analysis report. T 21s

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j modification does not' create a possibility for an accident or i malfunction of a different type than any evaluated previously in the- I

, . safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this l modification does not involve any unreviewed safety question.

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I Attachment ~to ET 98-0014 Page 61 of 238 Safety Evaluation 59 97-0022 Revision: 0 Main Turbine Lubricating Oil Conditioner System Valve Replacement Equivalency This modification provides for replacement of inlet isolation valves and outlet isolation valves for the Main Turbin9 and Steam Generator Lubricating Oil Conditioner System and Steam Generator Feedwater Pump Lubricating Oil Conditioner System. The replacement valves are provided by a different vendor and are of a different valve type.

Evaluation of an equivalent replacement valve has concluded that the replacement valve is compatible with the working fluid and performs the same function as the original valve and, therefore, is acceptable.

The valve type does not change any work' activities nor will it affect the operation of any systcm, structure, or component (SSC) . This modification does not affect any procedures nor will it affect the operation of any SSC that are important to the safe, reliable operation of the plant.

! Both systems are located in the Turbine building. Neither system can fail in such a way that results in the failure of another safety

related SSC not performing its intended safety design function. These systems are non-safety related. Replacing the valves will not affect any safety related equipment. Therefore, no credible malfunctions of

equipment important to safety are identified.

l There are no acceptance limits contained in the Technical Specification bases or other licensing basis documents that are affected by this modification.

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Attachment to ET 98-0014 Page 62 of 238 Safety Evaluation: 59 97-0023 Revision 0 Spent Fuel Pool Domineralizer Exit Penetrations This modification provides the means to change drawings M-12ECO2, "P&ID Fuel Pool Cooling And Clean-Up System," M-11ECO2, "SFD Fuel Pool Cooling Cleanup System," to accurately reflect the Puel Pool Clean-Up Demineralizer FEC03 configuration. FECO3 is a 165 cubic foot demineralizer which contains 145 cubic feet of resin for mainly purifying the Fuel Pool but can also be used for purifying the Refueling Water Storage Tank. Changing the above documents will provide for assessment of information and benefit the radwaste process evolution. These document changes are editorial.only and are not the rer:tlt of any physical changes made to FEC03 by the modification process. These documents inaccurately depict the configuration of FEC03. This is a drawing change'only and does not affect plant equipment.

This drawing modification will not 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. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 63 of 238 Safety Evaluation 59 97-0024 Revision: 0 Motor-Operated Valve Differential Pressure Test During core off loaded conditions this new procedure will provide a test of system alignments and install test instrumentation to dynamically (differential pressure) and statically test motor operated valve (MOV) EM HV-8807B. EM HV-8807B is the Centrifugal Charging Pump (CCP)/ Safety Injection cross-tie isolation valve that is normally opened during the Emergency Core Cooling System recirculation phase of a Loss of Coolant Accident (LOCA). This open valve enables CCP and Safety Injection Pump (SIP) suction operation from the discharge of the running Residual Heat Removal (RHR) pumps while taking suction from the containment recirculation sump.

The test will cycle the MOV against the flow and pressure from RHR Pump B while it is taking suction from the Reactor Coolant System (RCS) with the electrical interlock to valve EJ HV-8804B defeated. EJ HV-8804B is the RHR Supply to SIP B Isolation Valve. The function of the interlock is to prevent over pressurization of the SIP and CCP suctions when the RHR loop suctions are open and to prevent post LOCA fluids being pumped to the Refueling Water Storage Tank (RWST) via the SIP mini-flow piping. This test is to be performed in Modes 6 or with fuel removed from the reactor, when the RCS is not pressurized and I

LOCA mitigation is not required. If the test is done in Mode 6, with a water level of less that 23 feet, Train A of RHR will be operable and in operation and Train B will be used to support the test. Stroking the MOV under these conditions will not degrade the operability of RHR Train B because the interlock protective function conditions are not present during the test.

The test installs a jumper to defeat the interlock of EJ HV-8804B so that the RER pump can operate in the closed loop configuration instead of an open loop configuration created by a LOCA. This test configuration is not described in the Updated Safety Analysis Report (USAR) for RHR shutdown operations. Additionally, temporary instrumentation for differential pressure measurement will also be installed for this test. This instrumentation is not reflected in USAR Figure 5.4-7.

The test allows, if the Normal Charging Pump (NCP) is providing seal injection, the removal of power from BG LCV-112B and BG LCV-112C when the valves are in the open position. BG LCV-112B and BG LCV-112C are outlet valves from the Volume Control Tank (VCT) and automatically close on low VCT level. Removing power from these valves in the open position prevents the NCP from potential loss of suction by adding 1 more operating margin to it. The test alignment separates the charging pump suction from the NCP suction via a manual isolation valve which also isolates the RWST as a suction source for the NCP.

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Attachment to ET 98-0014 i

Page 64 of 238' i Appropriate precautions on NCP operation, if testing is performed in

-this configuration, are placed in the procedure. Removal of the L automatic closure of the VCT outlet valves on low low VCT level makes the statement in the USAR no longer true.

The USAR was reviewed and the Fueling Handling and the Boron Dilution

-accidents were identified as potentially being impacted by the subject l test. The test procedure is not involved with fuel movement, procedures, or alterations of valving associated with dilution source interfaces. The test procedure requires that the operable CCP be aligned to the RWST as a borated water source if the test is performed in Mode 6. Thus no impact to the Fueling Handling and the Boron Dilution accidents exists as a result of the subject test.

There are no credible accidents that the proposed activity could create. The function of the RHR and Charging Systems are not introduced to any operating parameter that could create a credible accident. If the test is run during Mode 6, the possibility of creating a Fuel Handling accident as a result of the test is not credible due to the dissociation of the test with these activities.

Leakage from the temporary instrumentation would be minor,-contained in a radiologically controlled area, and isolatable.

Redundant methods of isolation of the subject MOV exist in the unlikely event that the MOV fails to isolate flow during the test.

Failure of the tested MOV in Mode 6 does not affect the function of L the RHR system during testing in Mode 6, but would affect the ECCS recirculation function if not corrected before entry into Mode 3.

Operation of the RHR System is not affected by the test. The RHR equipment will function as if it were in the LOCA recirculation configuration with the exception of taking suction from the RCS loop instead of the recirculation sump. Defeating the interlock to take

suction from the RCS loop does not introduce any new or unique factor l that would alter RHR equipment operation or evaluated failure effects. Spray effects due to leakage from the temporary instrument connections, should leaking occur, would be into the B SIP room and would not adversely affect any equipment operation in the room.

RHR trains must both be operable in Mode 6, with level less than 23 l feet above the Reactor Vessel Flange, reference 3.9.8.1. This test does not challenge the operability of the RHR system. The decay heat removal function of the RHR system is not impaired by the test.

l L When considering RHR system operation in Mode 6 the number of previously evaluated accidents are basically reduced to the Pueling Handling accident and RCS dilution event. The probability of occurrence of these accidents are not increased because the test does not affect fuel handling equipment or procedures nor create a source or system configuration for diluting the RCS from that previously evaluated.

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Attachment to ET 98-0014 Page 65 of 238 The consequences of the Fuel Handling accident and Dilution event remain as evaluated because the test does not change the fuel enrichment, containment structure, source range detection, boration l capabilities or dilution sources. The SIP mini-flow lines are j isolated (double isolation) during the test; thus pumping refueling _ i water back to the RWST is not possible should a Fuel Handling accident I occur in the reactor or fuel building (tube open) concurrent with this test.

The use of temporary test instrumentation is acceptable because it does not impact the seismic capability of the piping system. 1 Isolation using normal system valves to maintain pressure boundary '

integrity will be used when instrumentation is not in use. The CCP/ SIP suction piping is not challenged by the test because Mode 6 RCS/RHR suction pressures are low (basically static head conditions) and operator awareness and controls on RHR discharge pressure are addressed by the procedure to maintain the pressure below the CCP/ SIP suction relief valve set points. Piping integrity is maintained.

There is no increase in consequences of equipment malfunction because the equipment failure mcdes have not been changed by, or because of, the test and the utilization of the RHR equipment for the test is within the acceptable limits of its operation.

No new or different types of accidents are created by this testing because no new failure modes or adverse system configurations or modes j of operation are created by the test.

Leakage from the temporary instrumentation will be minor and into the B SIP room, which is already evaluated for high energy line breaks and spray effects. The RHR system is not a high energy system and the leakage will be isolatable. RHR pump and piping reliability is maintained.

The Technical Specification does not require the ECCS injection or recirculation function for plant operation in Mode 6. Defeating the EJ HV-BB04B interlock, which is an ECCS protective function, while in Mode 6 does not compromise the cooling function of the B or A RHR Train, both of which are required if level is less than 23'.

Therefore, there is no reduction in the margin of safety as defined in Technical Specification 3/4.9.8 Bases.

l l Based on the above discussion this test does not involve an unreviewed l safety question.

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Attachment to ET 98-0014 Page 66 of 238 Safety Evaluation: 59 97-0025 Revision: 0 Motor-Operated Valve Differential Pressure Test This test will monitor and trend the P.. rust capability of the Residual Heat Removal (RHR) pump suction isolation valve EJHV8701A. This test will satisfy WCNOC commitments with regards to the Motor Operated valve (MOV) program. The test will be conducted during cold shutdown (Mode 5). The test provides the system alignment and instrumentation to dynamically and statically test this MOV.

EJ HV-8701A is one of two in series MOVs that isolates the low prc.ssure RHR System from the high pressure Reactor Coolant System (RCS) when the RCS is at high pressure (Modes 1-4). In Mode 5, the RCS is in a low pressure condition. The RCS pressure will be controlled by the pressurizer bubble during the test. The idle RHR train will be used for the test, with the other train in operation.

Two steam generators with wide range level of at least lot will be used as heat sinks in place of the idle RHR train as allowed by the Technical Specifications. Normal CVCS letdown through the letdown heat exchanger w ll be unavailable for a short duration during this test. The test will cycle the subject MOV, that has an actual stroke time of less than a minute, against the flow and pressure from the RCS as it depressurizes by maximizing the RHR letdown flow to the VCT.

This evolution will result in a slight reduction in RCS pressure and level. A maximum flow of 200 gpm was observed flowing to the VCT during Refueling Outage 6 when this test was initially run.

The RCS/RHR pressure boundary will be affected by the use of temporary pressure instrumentation, located upstream and downstream of the MOV at small bore branch piping connections. This instrumentation is not seismically qualified. This instrumentation is not described in the Updated Safety Analysis Report (USAR).  !

i USAR Sections 5.4.7, RHR System, 5.4A.3.2, Achieve and Maintain Cold Shutdown, and 9.3.4, Chemical and volume control System (CVCS) were reviewed. The test will briefly make unavailable the RHR/CVCS low pressure letdown used for cleanup and pressure ccntrol while the MOV is cycled. This brief interruption does not make the information in j any of these sections no longer true or inaccurate nor does the I interruption violate any RCS/RHR requirements. The RHR/CVCS letdown function is not essential and may be interrupted, as the USAR does not state that the function is continuously present. The test is on the idle RHR train and does not adversely affect the operation of the other train of RHR. Closing EJHV8701A during the test does remove the  ;

idle train's relief valve; however, the RCS over pressure concerns are I satisfied by taking credit for a power operated relief valve (PORV) in its place, which is allowed by the USAR. The description of these plant features remains as described in the USAR.

Attachment to ET 98-0014 Page 67 of 238 The previously evaluated accident reviewed for impact, due to this test in Mode 5, were the normal CVCS 3" letdown line break outside containment at power (USAR 15.6.2), reactivity anomalies while in cold shutdown (USAR 15.4.6), and inadvertent opening of a pressurizer safety or relief valve at power (USAR 15.6.1').

A minor amount of leakape or loss of RCO/RER fluid from the temporary instrumentation or associated fittings could occur. This loss would be contained in a radiological controlled area, not impair decay heat removal and would be isolatable. A RCS pressure spike or over pressurization of the RCS will not be created because the RCS will not be in a solid condition due to the pressurizer bubble. The bubble will provide a cushion to absorb the chargfsg flow to the RCS when the maximum letdown is terminated by the closure of EJHV8701A. The pressurizer bubble is used to compensate for the brief mismatch between charging and letdown flows until these flows are restored to a balanced condition. Over pressure protection of the RCS is not challenged and is maintained by the PORVs or PORV and B Train RHR relief valve.

The potential to over pressurize the Volume Control Tank (VCT) is not possible due to the VCT relief valve that has the capacity to relieve the maximum letdown flow rate. The VCT high pressure alarm and automatic divert to the Recycle HM. dup Tank on increasing level functions, remain available during the test as does the capability to j stop the RCS de-pressurization by closing any one of several valves.

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The integrity of the VCT is maintained. The probability of equipment malfunction remains as evaluated because no equipment in the operable RHR train is affected nor are other system components subject to operational parameters beyond the acceptable limits of their design.

The test is performed during cold shutdown; therefore, the material properties limits associated with Technical Specification Bases 3/4.4.9 could be affected. Cold over pressure protection is maintained without reliance on the A Train RHR suction relief valve which is not available during the test because of closure of EJHV8701A. B train RHR suction relief valve is not affected by the i test. A PORV is a suitable replacement for an RHR relief valve per l Technical Specification Bases 3/4.4.9.3 It is acceptable by Technical Specification 3.4.1.4.1 to substitute the idle train of RHR in test for two Steam Generators; therefore, the Bases of Technical Specification 3/4.4.1 in regards to heat removal capabilities are maintained.

Because the RCS/RHR pressure boundary is not challenged by the test ,

nor is fuel enrichment or reactivity altered by it the radiological )

consequences remain as originally evaluated for all malfunctions that i can occur while in cold shutdown. Minor leakage from the temporary pressure instrumentation on the idle RHR train may occur but will not l

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Attachment'to ET 98-0014 Page 68 of 238 have an adverse' affect on the operating RHR pump nor result in a:

significant RCS-inventory loss. Leakage from the temporary instrumentation can be isolated and will be contained in a radiologically controlled area.

Utilization of the plant equipment and systems by the test lies well within the acceptable. limits of these plant features. A different or.

unique variable or condition needed to create a unique accident is missing from the test.

RHR equipment malfunctions listed in Table 5.4A-3 were reviewed and no

.different or unique malfunction possibilities were seen possible as a result of the test. The failure mode of all equipment remains as evaluated.

Review of the Bases of Technical. Specifications 3.4.1.4.1 Cold Shutdown RHR Operation, 3.4.9.3, Over Pressure Protection, were made and no resultant reduction exists.

Based on the above discussion, this test does not involve an unreviewed safety question.

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. Attachment t'o ET 98-0014' Page 69 of 238 Safety Evaluation: 59 97-0026 Revision: 0 anergency Makeup Water Requirement for Auxiliary Feedwater From Essential Service Water / Ultimate Heat Sink USAR Section 9.2.5.2.3 (Page 9.2-35) " Emergency Makeup Water Requirement" states in part that make up water is required to supply the auxiliary feedwater (AFW) system when'the condensate storage tank is unavailable ~or exhausted, as described in Section 10.4.9 and Figure 10.4-11. Plant Modification Request 04721 revised USAR Section 10.4.9 to delete the specific makeup rates and durations. In addition, l

Figure 10.4-11, which provided a graph of the same data, was deleted. j Performance Improvement Request 96-2698 identified this error and

-l recommended that the above makeup water requirement for AFW be '

described in USAR Section 10.4.9 and Figure 10.4-11 or the information provided in USAR Section 9.2.5.3 and referenced appropriately. j A Design Change Package (DCP 07225) was initiated to re-establish the~

makeup requirement design bases and correct the reference disparity.

Other editorial corrections were made to'this section under this DCF.

USAR Section 9.1.3.2.3.1 was clarified to indicate that the heat load from the Spent Fuel Pool (SFP), used in determining the maximum total head load to the Ultimate Heat Sink (UHS), "is based on the decay heat rate shown in Table 9.1-4," rather than "is shown" in the table. An unnecessary reference to USAR Section 9.1.3 in Section 9.2.5.2.2.2 was deleted, since the information regarding SFP heat rejection rate following normal shutdown using the UHS was already contained in Section 9.2.5.2.2.2.

Safety Evaluation One (providing necessary component cooling water for safe shutdown and continued reactor cooling) and Two (UHS reserve for emergency makeup water to the SFP and Closed Cooling Water Systems, as well as backup to the Auxiliary Feedwater System) in USAR Section 9.2.5.3 address UHS requirements. The UHS must be able to provide the required makeup to plant systems and be able to provide sufficient cooling water for 30 days at a maximum water temperature of 95 degrees F,' assuming maximum component heat load and the most severe meteorological conditions. The UHS is not an initiator of any design basis accident. Therefore, this change will not increase the probability of occurrence of previously evaluated accidents.

As indicated in.the DCP, the total makeup requirement of the UHS was not increased as a result of.the revised bases. Required makeup quantities to the AFW and SFP systems were increased; however, the required CCW quantity was reduced, resulting in no overall increase in required UHS makeup. The DCP documents that Safety Evaluation One and Two are unaffected by the proposed changes. The functions of the ESW and all systems relying on ESW are unaffected. Therefore, no assumptions made in evaluating radiological consequences of accidents

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-Page 70 of 238 are affected, and the consequences of previously evaluated accidents are not, increased.

Two credible malfunctions of equipment important to safety that_may.be

-directly or indirectly impacted by emergency makeup arei degraded ESW pump operation due to inadequate water level at the pump inlet, and

-degraded cooling of plant equipment supplied by ESW if water temperature of the UHS outlet exceeds the design basis temperature.

The described changes do not affect the minimum UHS water level or design basis temperature, since the total quantity of makeup has-not changed.. Therefore, the probability of a malfunction of equipment important to safety previously evaluated in the USAR is not increased.

The ability of the ESW system and other systems requiring EWS to mitigate accident consequences has not been impacted since the makeup l water quantity has remained the same and the minimum UHS water level and maximum temperature are not affected. Therefore, the change does j not increase the radiological consequences of a malfunction of equipment important to safety previously evaluated in the USAR.

The change does not result in any increase in makeup water requirements or heat load to the UHS; therefore, there is no effect on the' Safety Evaluations for this system and no new' accident or malfunction is created by the changes.

The change does not affect the quantity of makeup required or the amount of heat rejected to the UHS. Therefore, the ability of the UHS to provide the required makeup to plant systems and be able to provide sufficient cooling water for 30 days at a maximum water temperature of 95 degrees F, assuming maximum component heat load and the most severe meteorological conditions, is unaffected. No acceptance limits are affected, and no reduction in the margin of safety is introduced by the proposed changes.

Attachment to ET 98-0014 l

'Page 71 ef' 238 safety Evaluation: 59.97-0028 Revision: 0 Drawing Correction l

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-This revision to the Updated Safety Analysis Report adds a note to l Piping and Instrument Diagram M-12HE02, "P&ID Boron Recycle System,"- i alerting plant personnel of the requirement to keep one recycle holdup tank inlet isolation valve open at all times. The note is added to provide additional information in support of Control Room activities.

This document change does not affect any of the administrative.

controls or procedures that govern plant alignment, nor does it affect the operation of any system' structure or component. The revision will' not affect the safe and reliable operation of the plant.

This revision will not increase the probability of occurrence or the i consequences of an accident or malfunction of equipment important to 1 safety previously evaluated in the safety analysis report.

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This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety I analysis report. The margin of safety as' defined in technical specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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-~ Attachment to- .ET'98-0014 Page.72 of -238

Safety Evaluation: 59 97-0031 Revision 0 New Calculatiion to' Supplement Previous Calculation-

.The activity evaluated is the temperature profile for room 1331, Turbine Driven Auxiliary Feedwater Pump Room. Calculation GF-M-002 evaluated the four different' plant, conditions to determine the maximum room' temperature.for'each specific plant. condition. The maximum allowable temperature for room 1331 is 150 degrees Fahrenheit, which is the turbine driven auxiliary feedwater pump's environmental qualification temperature limit. None_of the calculated plant conditions exceeded the maximum allowable temperature limit of.150 degrees Fahrenheit.

Condition'1, normal plant condition, and Condition 3,-ventilation failure, did' increase above the values listed in the USAR. Condition 1 changed from a USAR value of 113 to 115 degrees Fahrenheit, and Condition 3 changed from 142 to 146 degrees Fahrenheit.

The design basis accident analysis is unchanged'due to the calculated maximum room temperature being below the environmental equipment qualification temperature limit for room 1331 of 150 degrees Fahrenheit. The Auxiliary Feedwater System is net analyzed as an initiator of any design basis accidents. Therefore, this change does not. create any credible accidents. This activity does not propose any.

credible malfunction of equipment important to safety, as calculation GF-M-002 has demonstrated that the maximum environmental temperature .

limit of 150 degrees Fahrenheit will not be exceeded. Technical I Specification ~16.7.2 (area temperature monitoring) uses a tolerance of j

+/- 3 degrees Fahrenheit and a value of 147 degrees Fahrenheit for the -'

surveillance room temperature value which would limit the maximum room temperature to 150 degrees' Fahrenheit. This change.does not adversely effect Technical Specifications and there is no reduction in the margin of safety.

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Attachment to ET 98-0014 Page 73 of 238 Safety Evaluation: 59 97-0032 Revision 0 Radiological Emergency Response Plan Revision This Unreviewed Safety Question Determination evaluates six changes being made to the Radiological Emergency Response Plan (RERP). The changes were evaluated under 10 CFR 50.54 (q) and were shown to not decrease the effectiveness of the RERP. Since the RERP is a part of the Updated Safety Analysis Report, the changes are also being evaluated in accordance with 10 CFR 50.59.

The following changes are being made to the Emergency Action Level (EAL) flow charts in the RERP:

In the Radiological Effluent Release Flow Chart, a "YES" answer to Box 1-RER5 led to declaration of an ALERT. A "YES" answer at this point in an event should more appropriately lead to an evaluation of the ongoing release at the site boundary, which would determine whether the event would be classified as an ALERT, SITE EMERGENCY, or GENERAL EMERGENCY.

In the Main Steam Line Break Flowchart, a "NO" response to Box 4-MSLB9 (For a faulted Steam Generator (SG), is SG faulted inside containment?) was redirected to Box 4-MSLB6 (event was previously directed to be classified as UNUSUAL EVENT), which more accurately further evaluates the event and classifies the event as either an ALERT or a SITE EMERGENCY. The change is consistent with other Emergency Action Level classifications.

Added radiation monitors to the reactor building and fuel handling building evaluation boxes in response to a fuel handling event. The additional monitors are already used in plant procedures governing this type event, and will further assist the operator in making an appropriate classification of a fuel handling accident.

Provided additional guidance in the Safety System Failure or Malfunction flow chart following a failure of the SSPS to trip the reactor. The guidance changes from " Unable to feed any SGs with any of the following systems... to " Unable to feed any SG at an adequate feed rate with any of the following systems .." The change provides clearer direction as to the basis for a decision in this box.

Corrected a typographical error ("progcess" to " process") in Box 13-l ADM1 on the Administrative EAL flow chart.

The changes noted above provide additional guidance and clarification to the operators during various emergencies. The revised guidance further ensures that the appropriate actions will be taken in response to an event. As such, the changes enhance accident mitigation. The changes in the flow charts do not introduce new or change existing I

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l Attachment to ET 98-0014 Page 74 of 238 accident initiator or equipment malfunction. Therefore, these changes , l will not increase the probability of occurrence of previously' evaluated accidents or malfunctions of equipment important to safety, nor will they ;reate the possibility of a new accident or malfunction. The clearer guidance further ensures that accidents and transients are appropriately mitigated and acceptance limits met. ]

Therefore, no reduction in any. Technical Specification margin of safety is introduced.

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Attachment to ET 98-0014 Page 75 of 238 Safety Evaluation: 59 97-0033 Revision 0 Drawing Corrections Performance Improvement Request (PIR) 96-2216 identified that USAR Figure 6.2.4-1, " Containment Penetratic:is," contains several errors.

JSAR Figure 6.2.4-1 contains 'N sheets, with each sheet showing a c ntainment penetration figure Several valves shown on the figure sheets are not currently installed in the field. Also, several valves are not shown in the correct position on the figure sheets. This USQD evaluates the changes to USAR Figure 6.2.4-1 which are necesstry to establish consistency with actual plant configuration.

The majority of the changes being made to USAR Figure 6.2.4-1 are administrative in nature (legibility improvemente, correction of symbology, typographical errcrs and valve numbering, and adding omitted valves and associated data). The administrative changes correct errors with the USAR figure. No physical changes have been made to the plant. The corrections reflect the plant as it was reviewed and approved by the NRC staff. As such, the change will not increase the probability of or consequences of any previously evaluated accident or malfunction of equipment important to safety, nor will any new accident or malfunction be ir' w . *d , or the margin cf safety by reduced.

With one exception, the substantive chant - cw +ed design changes made prior to commercial licensin ent. As such, the changes are not subject to 10 CFR 50.59. Tha e- ;cion involves a change to Sheet 66. Instrument isolation valve EMV0038, Safety Injection Test Line PI-929 Isolation Valve, is being revised on Sheet 66 to show the valve as a normally closed instead of a normally open valve. Sheet 66 is based on P&ID M-12EM01, "High Pressure Coolant Injection System," which was developed, in part, from a Westinghouse flow diagram. The Westinghouse flow diagram shows EMV0038 as a normally open valve. The valve is currently maintained locked closed, in accordance with plant procedurer, as the valve is considered a containment isolation valve. The locked closed status was established during ovaluation of an industry LER in 1992. An immediate action to this event was to lock closed EMV0038 and include it en the locked valve list that is verified per Technical Specification 4.6.1.1.a.

EMV0038 isolates local pressure indicator (PI) EMPIO929. This PI is used solely for the purpose of indicating pressure in '.ho ECCS test line during Reactor Coolant System (RCS) isolation cheen valve testing. Therefore, there is no reason to maintain this valve in the open position during normal operation. Maintaining EMV0038 in the closed position adds a barrier against leakage of fission products within the penetration envelope. The connecting piping, tubing, and isolation valve are constructed commensurate with the construction 4

Attachment to- ET.98-0014

-Page 76 of 238-requirements of the penetration.

The change to Sheet 66 to indicate the normally closed status of EMV0038 does not constitute an unreviewed safety question.

Containment isolation penetrations function to mitigate the consequences of lua accident. As such, no accident initiators are

-affected. Therefore, this change will not increase the probability of occurrence of previously evaluated accidents. The closed status of the valve will further ensure that the penetration achieves its mitigation function.

Therefore, no assumptions made in evaluating radiological consequences of accidents are affected, and the consequences of previously evaluated accidents are not increased.

EMV0038 is'not required to perform any active function. No new operational requirements are imposed on the valve. Its functions are unaffected by the redesignation. Therefore, the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR is not increased, and no new accident or malfunction is created by the changes. Acceptance limits are unaffected;.therefore, no reduction in the margin of safety is introduced by the proposed change.

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Attachment to ET 98-0014 Page-77 of--238 Safety Evaluation: 59 97-0054 Revision: 0 Removal of Table.7.3-7 of USAR This USQD evaluates eliminating response time testing of the Control

' Room Ventilation Isolation Control System Radiation Monitors. USAR changes include eliminating the sensitivities and response times located in Table 7.3-7 of the USAR.

Wolf Creek Technical Specifications 3.3.2, ESFAS Instrumentation, and 3.3.3.1, Radiation Monitoring for Plant Operation, have no response time requirements for the Control Room Ventilation Isolation Radiation Monitors GK-RE-04 and GK-RE-05. The bases for Section 3.3.2 indicates that for those ESEAS channels without response time requirements, no credit is taken for those channels in the accident analyses. A review of the accident analyses and discussions with the Safety Analysis Group confirms that no credit is taken for these monitors in the accident analyses. The minimum concentration requiring isolation and the limiting isotope data in Table 7.3-7 are. specific to response time testing. .Since response time testing is being eliminated for these monitors, this information is no longer necessary. The Monitor Range, Alert Alarm, and High Alarm setpoints for these monitors are provided in USAR Table 11.5-3.

The Control Roou Ventilation Isolation Control System Radiation Monitors are.not initiators of any accident. The response time of these monitors does not adversely affect the isolation function and control room dose analyses. No assumptions made in evaluating radiological consequences of accidents are affected. Therefore, this change will not increase.the probability of occurrence or consequences of previously evaluated accidents, nor will it increase the probability'or consequences of.a malfunction of equipment important to safety previously evaluated in the USAR.

There ie no effect on the Safety Evaluations for this system as a result of this change. Therefore, no new accident or malfunction is created, No acceptance limits are affected by the change, and no reduction in the margin of safety is introduced by the proposed changes.

Attachment to '.ET 98-0014' Page 78 of 238 Safety Evaluation 8 59 97-0035 Revision: 0 Assistant to the Chief Operating Office Organisation Change Chapter 13 of the USAR includes a description of the position of Assistant to the Chief Operating Officer and the resume of an

' individual who is no longer employed by WCNOC. This change to the USAR deletes both the resume and the position of Assistant to the Chief Operating Officer.

-No plant safety analyses, plant operation, or plant administrative controls' rely on this position. Therefore, this change will not impact the function of plant equipment important to safety, the safety analyses, or -lant operation. As such,. this change will not increase the probability of occurrence or consequences of previously evaluated accidents, or the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. No new accident or malfunction is created by the change. No acceptance limits are affected, thus no reduction in the margin of safety is introduced.

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Safety Evaluations 59 97-0036 Revision 0 Main Steam System Terminology This Unreviewed Safety Question Determination evaluates USAR changes being made to correct descriptions of the Main Steam System and the Atmospheric Relief Valves in several sections, and to provide consistency with the Technical Specifications, including: Section 7.7.1, which indicates that the main steam dump valves dump steam to the atmosphere (main steam dump valves only dump steam to the condensate condenser); Section 10.3.2.3, change to state that steam for the Auxiliary Feedwater (AFW) pump turbine comes from "two of four steam lines" instead of the current statement that indicates steam comes from "one of two steam lines"; Section 10.3.2.3, clarified that redundant check valves in the AFW steam feed lines also ensure that at least one steam generator is available to supply steam to the AFW pump turbine following a steam line break; Section 10.3.4.1, clarified that the lift-point of each ARV is verified by channel check and channel calibration, consistent with Technical Specifications; and, in numerous sections, changed the terminology of the power-operated  ;

atmospheric relief valves to eliminate possible confusion with the  !

Pressurizer power-operated relief valves (PORVs).

Neither plant safety analyses, plant operation, or plant administrative controls are affected by these clarifications. No plant equipment is being altered. The changes ensure that terminology is consistent between the USAR and plant Technical Specificaticns.

Therefore, this change will not increase the probability of occurrence or consequences of previously evaluated accidents, or the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. No new accident or malfunction is created by the change. No acceptance limits are affected, thus no reduction in the margin of safety is introduced, i

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i Safety Evaluation: 59 97-0037 Revision 0 Power Operated Relief Valve Block Valve Auto Open on High Pressure l

This Unreviewed Safety Question Determination evaluates reinstallation !

of the Pressurizer power-operated relief valve (PORV) block valve j automatic open function (Plant Modification Request 07195) . The 1 original block valve position control logic relied on an interposing j relay that acted similar to a toggle switch function. The relay would l either open or close the block valve automatically in response to  !

pressurizer pressure when the valve was placed in the " Auto" mode. I Following a 1987 unplanned reactor trip, both PORV block valves tripped their circuit breakers. The PORVs had opened and closed l during this transient, resulting in fluctuating pressure signals.

These signals caused a mid-cycle reversal of the block valves (valves  !

stopped closing and began opening). The reversal caused the motor operators to experience a large current, which exceeded the instantaneous trip settings on the block valve circuit breakers. A modification subsequent to his trip provided a seal-in circuitry design to ensure that the valve would travel its full stroke, thus precluding mid-travel reversal. The modification also deleted the valve automatic opening circuitry to prevent valve cycling due to )

pressurizer pressure fluctuations. Procedural controls were put in place to have the operator manually open the block valves when pressurizer pressure increases above 2185 psig.

Operations requested engineering to evaluate re-installing the l automatic open function. The safety function of the PORV block valves  ;

is to provide for isolation of the PORV inlets when excessive leakaga I in the PORVs occurs. The block valves are designed to automatically )

isolate when Reactor Coolant System (RCS) pressure drops below a i predetermined value indicative of a stuck open PORV (2185 psig  ;

setpoint). With pressurizer pressure increasing, if the operator does j not open the block valves, which allow the PORVs to operate, the  !

pressurizer pressure could increase and challenge the pressurizer safety valves, nullifying the safety design of the PORVs. Modification 07195 reinstalls logic to allow the PORV block valves to automatically open and provide a flow path to the PORVs. The modification will j preclude challenges to the prersurizer code safety valves if an l operator was unable to or inappropriately reacted to increasing pressurizer pressure. A time delay relay is provided to each of the block valve's control circuitry which will allow the valve's operator to close or open without tripping the associated breaker, to mitigate the original concern with mid-travel reversal of the valve.

This change is to the logic circuit of the PORV block valves. The PORV block valves's safety function is to isolate the PORVs from the Pressurizer in the event that the PORVs are stuck open or exhibit high j amount of leakage. The modification provides additional assurance of '

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Attachment to. ET 98-0014 Page 81 of 238 .

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the PORV function by eliminating a manual operator action. The block valves can fail open or closed without impacting any previously analyzed accidents. Therefore, this change will not increase.the

-probability of occurrence or consequences'of previously evaluated accidents.

If either or both of the block valves fail in the closed position, then the pressurizer safety valves would relieve RCS pressure. If either or both of the block valves were to fail in the open position, then the pressurizer.PORVs would control RCS pressure based on the differential presnure seen by the PORVs. The credible malfunctions of the block valves do not change as a result of this modification beyond what has been analyzed. The consequences of these malfunctions would still be mitigated by other safety-related equipment. The modification relieves operator burden by eliminating an operator action, which should enhance reliability of the function. Therefore, the probability or consequences of a malfunction of equipment important,to safety previcusly evaluated in the USAR are not '

increased. No new accident or malfunction is created by the cha'ge.

The design change does not affect the acceptance limits contained in the basis of the PORV block valve Technical Specifications (3/4.4.4, 3.4.4(d), 4.4.4.2). The setpoint to open or close the block valve-is below the hysteresis point (2193 psig) and is consistent with the instructions _in M-747-00025, " Precautions, Limitations, and Setpoints l for Nuclear Steam Supply Systems"; thus,'no. reduction in the margin of safety is introduced by the modification.

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Attachment to ET 98-0014 Page 82,of 238 Safety Evaluation: 59 97-0038 Revision 0 Auxiliary Steam System Description Change While reviewing the USAR for a proposed procedure revision,'two errors were detected in Section 9.5.9.2.1 of the USAR. An apparent inadvertent omission of words from a descriptive statement concerning the flowpath of feedwater within the Auxiliary Steam system has-resulted in an incorrect statement. Currently the USAR states, "The auxiliary steam feedwater pumps take a suction frem the auxiliary steam deaerator and reboiler, depending on which is in operation."

This sentence is incorrect and should read "The auxiliary steam feedwater pumps take a suction from the auxiliary steam deaerator and (feed the auxiliary _ boiler) and reboiler, depending on which is in operation." Also, in the next paragraph, the USAR states " water quality is maintained by periodic blowdown to an atmospheric blowdown tank." In actuality, current practice is to maintain a continual blowdown when the auxiliary reboiler is in service, and, when the auxiliary boiler is in service, blowdown would only be done when chemistry determines a need for blowdown.

The auxiliary steam system is designed to provide the steam required for plant heating and processing during plant startup, complete shutdown, and normal operation. The system consists of steam generation equipment, distribution headers, and condensate return equipment. Auxiliary steam is distributed throughout the plant. The auxiliary steam system has no safety function. There are no design basis accidents affected by this change. No change is being made to plant equipment. These changes correct an error in the system design description and amend the system operating description to reflect current practices with respect to maintaining water quality. No >

acceptance limits are affected by the change. The auxiliary steam system is not in any technical specification or technical specification bases.

Based on the above, these changes will not increase the probability of occurrence or consequences of previously evaluated accidents or malfunctions of equipment important to safety. No new accident or malfunction is created by the changes, and no reduction in the margin of safety is introduced.

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Attachment to ET 98-0014 Page 83 of 238 Safety Evaluation: 59 97-0039 Revision 0 welds for Normal Charging Pump Installation i During the welding process, a pyrometer is used to check the interpass temperature of the weld being made. The interpass temperature is limited procedurally to 350 degrees Fahrenheit. When pyrometer #WC 16878 was recalibrated, it was found out of calibration by-62 degrees Fahrenheit. Actual interpass temperature may have been as high as 412 degrees Fahrenheit for the weld that used the pyrometer. The USAR states'that heat input in all austenitic pressure boundary welds is controlled by limiting the maximum interpass temperature to 350 degrees Fahrenheit. Thirteen welds in question were performed during installation of the Normal Charging Pump PBG04 for Design Change Package 04590. One weld performed during implementation of Design Change Package 06110, was located downstream of normally closed drain line valve ECV027 which is located off of the fuel pool cooling heat exchanger ECC01B.

A defense in depth approach nas been taken to minimize the sensitization of austenitic stainless steels ~ joined by welding.

Stainless steels used in the piping system are in the solution annealed and quenched condition. Carbon content is low, with welding filler material carbon content of .03% or less. Welding filler metal materials are procured with elevated ferrite level. Pre-heat and interpass temperature controls are used to minimize the time that base metals are in the temperature range for sensitization. For the fourteen welds in question, the system operating temperatures are low and a non-harsh system chemistry is maintained. All of the factors j contribute to prevent the onset of intergranular stress corrosion 1 cracking at the weld heat affected zone. An adequate margin or protection still exists for the fourteen welds performed at an elevated interpass temperature. The activity did not create any credible accidents. No ASME/ ANSI code violations exist and the I elevated interpass temperature of 412 degrees Fahrenheit has no j detrimental effect to the weld joints or base metals around the weld area. )

No change has occurred to the mechanical properties of the pipe  ;

installed with the elevated temperatures, and the pipe meets all requirements of ANSI /ASME codes and the original design requirements.

Therefore, the change does not introduce a new failure mode and will  !

not increase the probability of occurrence or the consequence of an I accident previously evaluated. The change does not increase the  ;

probability of occurrence or the consequences of a malfunction of equipment important to safety than previously evaluated. No potential j exists for the creation of a new type of unanalyzed event. No ,

reduction in the margin of safety can result from this change, j

Attachment to ET 98-0014  ;

Page 84 of 238 Safety Evaluation 59 97-0040 Revision 0 Temporary Procedure TMP 96-060 Temporary Procedure 96-060 provides instruction to temporarily remove from service HVAC systems that exhaust into the unit vent and direct an exhaust flow path from the auxiliary building through the radwaste building. TMP 96-060 will secure the following systems:

decontamination tanks exhaust scrubbers access control ventilation, contair. ment purge ventilation, auxiliary building ventilation, and hot machine / instrument shops fan coil units. The activities of this procedure will not affect or disable any of the safety actuation systems, as described in the Updated Safety Analysis Report (USAR).

Automatic actuation of the auxiliary / fuel building emergency exhaust and/or control building emergency ventilation systems will still occur as designed. Having the boot removed from the unit vent will not increase the radiological consequences of an accident since ground releases are assumed in the dose analysis model.

M-12ALO1, USAR figure 10.4-9-00 is affected by this change because more detail is added to the P&ID showing seal water piping for the motor driven and turbine driven auxiliary feedwater pumps and this change will clarify the safety classification of vents and drains on the discharge side of the subject pumps.

I No field work or change to equipment is involved in this change. No new components, modifications or changes to the operation of the Auxiliary Feedwater System are included with this change. The change does not introduce a new failure mode and will not increase the l probability of occurrence or the consequence of an accident previously evaluated. The change does not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety not previously evaluated. No potential exists for the creation of a new type of unanalyzed event. No reduction in the margin of safety can result from this change.

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Attachment to ET 98-0014 Page 85 of 238 Safety Evaluatisnt 59 97-0041 Revision 0 )

Installation of Corrosion Sample Panel i l

The change being evaluated installs a temporary corrosion product  ;

sampling panel to monitor the Steam Generator Blowdown flow stream from "C" steam generator. The equipment will be connected to the flush connection, valve BMV0216, using 24" stainless steel tubing with <

cooling provided via hoses from the service water system. The service  !

water and the sample flow exiting the panel will be routed to the l turbine building sump. The subject equipment will be used by l Chemistry to determine the effectiveness of the secondary side water {

treatment and will be installed downstream of the Steam Generator l Blowdown Isolation valve. This location is on the non-essential i portion of the system, which is isolated from the essential portion following a Steam Generator Blowdown System Sample Isolation Signal (AFAS), blowdown, and sample isolation signals.

The Steam Generator (S/G) Blowdown System is used in diagnosing a S/G tube rupture. The system is isolated during a tube rupture event to prevent a release to the environment. The installed temporary equipment will not prevent the radiological monitoring equipment from providing the initiating signal to isolate blowdown.

a The S/G Blowdown Syctem also provides sampling capability, after isolation, to pin point which generator has ruptured. The I modification will not affect this diagnostic capability. l The modification will not create a new or different release path to the environment from those previously evaluated. The combined  ;

discharge will be monitored by radiation monitor LERE0095, The process fluid that is tapped off by this modification would still be contained within the power block by the oil Waste System radiation monitor (LERE0095) and isolation valve (LERV0059). In the case of a failure of the "C" Steam Generator Blowdown Isolation valve to close, concurrent with a tube rupture in that generator, no credit is taken I for the integrity of the nonessential piping. This assumption is not altered by the installation of the temporary equipment. The radiological consequences of a Steam Generator rupture will not be increased by this installation.

The failure of, or leakage from, the installed equipment will not {

affect safety related equipment due to its location in the Turbine Building. The possibility of a different type of Turbine Building flood will not be created because the maximum leakage from all the connections made is less than that from the rupture of a circulating Water System expansion joint.

The pressure boundary of the proposed sampling equipment, located in l

iAttachment to' ET 98-0014 Page 86 of '238

'the Turbine Building, will contain both S/G blowdown and service-fluid. These flow streams, after passing through sample equipment, will be directed to the building sumps. If the sample equipment'were to fail or. leak, the fluid or J eakage would'still end up in the Turbine Building sumps. As noted above, the amount of fluid flow or leakage.from the sample equipment would be significantly less than the bounding circulating water system leak used in the Turbine Building analysis.

No Technical Sper:ification, Technical Specification Bares,-or other Licensing Bano document acceptance limits are affectt f by the temporary-installation.

Based on the above, these changes will not increase the'probabil!a, of occuMrence or consequences of previously evaluated accidents or malfunctions of equipment important to safe.*.y. No new accident _or ,

malfunction is created by the changes, and no reduction in the margin of safety is introduced.

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safet- Evaluation: 59 97-0042 Revision: O shift Engineer / shift Technical Advisor Requalification Program Procedure AP 30B-006, "Shif t Engineer /Shif t Technical Advisor (STA) i Requalification Training program," was revised to bring the STA l requalification program ir4 line with the systematic approach to  !

training (SAT) process required by 10 CFR 55. The revision requires l STAS to participate in the License Operator requalification program to  !

maintain their qualifications. It also requires STAS to have  !

additional training on Transient and Accident Analysis and Mitigating 1 Core damage. Updated Safety Analysis Report (USAR) Section l 13.2.2.12.3, STA Requalification Training, currently describes I I

specific training topics ar.d their durations. This procedure change replaces specific training topics and their durations with a description of the use of the SAT process. No design basis accidents are identified which could be impacted by this proposed change because it does not impact SSCs and does not create any credible accidents.

No malfunctions of equipment important to safety are identified which )

could be affected by this change either directly or indirectly. This j change does not impact SSCs and does not reduce'the level of 1 qualification or ability of STAS. There are no acceptance limits  ;

identified which could be af fected and there is no reduction in the {

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Attachment to- ET 98-0014 Page 88 of 238 Safety Evaluation: 59 97-0043 Revision: O Licensed Operator Requalification Training Program USAR Section 13.2.1.2, Licensed Operator Requalification Training Program, provides a prescriptive training list ~for the program. A change is being made to the License Operator Requalification (LOR)

Training Program to bring the program in line with the Systematic Approach to Training (SAT) Requalification Process required by 10 CFR

55. This revision clarifies the SAT Process and what is required for the LOR program to meet the new requirements. Specific training topics and their durations are being replaced with a description of the use of the SAT process.

The changes being made are necessary to provide consistency with the SAT process for requalification. The purpose of the SAT is to focus training on the important aspects-of the Licensed Operator's job

'instead of a prescriptive training list. The SAT process evaluates, through the job analysis process, those tasks which are most important to the job. This evaluation forms the basis for both the initial license and the requalification training programs. The SAT process incorporates feedback from the Operations Department on' additional training areas that may be needed. Although certain low probability events or issues may not receive the level of training as in the past, by focusing on the most important tasks, the program increases the likelihood for the correct response being made to transients, thus maintaining or reducing the probability or consequences of an accident or malfunction of equipment important to safety. As such, no system, structure, component, or design basis accident scenario is adversely affected. No Technical Specification or Technical Specification Bases acceptance limits are not negatively affected.

Based on the above, these changes will not increase the probability of occurrence or consequences of previously evaluated accidents or malfunctions of equipment important to safety. No new accident or malfunction is created by the changes, and no reduction in the margin of safety is ittroduced.

Attachment to ET 98-0014 Page 89 of 238

-Safety Evaluation: 59 97-0044' A* vision 0

Condenser Vacuum' Pump Operation USAR Sections 10.4.2.2,1 and 10.4.2.2.3 state that the Condenser

-Vacuum Pumps are operated with two pumps running continuously, and a third pump is in standby during normal plant operations. Performance Improvement Request 97-0146 identified that this description does not agree with site procedures. The procedure SYS CG-120 establishes only one pump running continuously.with circulating water less than 80*F and air in-leakage less than 35 SCFM. This change revises the system description, USAR, and flow diagram to reflect the operational flexibility of running one condenser vacuum pump as in-leakage. flow rates permit.

The main condenser air removal system removes noncondensable gases and.

air from the main condenser during plant startup,.cooldown, and normal power operation. The system also provides an alternate vacuum source for the demineralized water storage and transfer system degasifier tank to facilitate deaeration of demineralized water supplied to the reactor makeup water system. The degasifier tank is normally lined up to the condenser vacuum pump suction header during normal power operation.

Initia? ,cuum is established prior to plant heatup by' evacuating air from the condensers using all three vacuum pumps. Once steam flow is initiated to the condenser, condenser pressure is established by steam flow rad circulating water heat removal rates. During normal opers.Laon , the vacuum pumps maintain the condenser at a vacuum (holding mode) by removing noncondensable gases and air from the condenser, since~the buildup of noncondensable gases can increase condenser pressure and inhibit heat transfer by forming air pockets within the tube bundles. Non-operating pumps are maintained in a standby mode.

Each pump rated for 35 SCFM flow at 1" HG absolute (Abs) suction pressure. Since actual pump suction pressures ate always greater than 1" HG Abs, one pump has ample capacity to maintai;: Jondenser vacuum whenever'in-leakage flow rates are less than 35 SCFM. Sjace flow '

-rates are within the capacity of one pump, running cdditional vacuum pumps does not lower condenser vacuum. Therefore, both energy consumption and maintenance costs will be reduced with fewer operating hours on the pump.

In the event air removal capability of a running pump (s) is lost, the standby pumps automatically start upon actuation of a low vacuum signal (25" Hg vacuum decreasing at the pump inlet manifold) . Each pump has its own independent vacuum switch. These features are unaffected by this change.

Attachment to ET 98-0014 Page 90 of 238 i

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A loss of condenser vacuum event is analyzed in USAR Section 15.2.3 and 15.2.5 as a Level II fault of moderate frequency. Failure to maintain air removal capability would cause a loss of condenser vacuum. Falling vacuum would result in a turbine trip at a setpoint of 21.3" Hg vacuum. The standby pump automatic pump actuation at 25"  :

.Hg vacuum decreasing ensures that condenser vacuum is maintained above I the turbine trip setpoint. Therefore, the probability of an accident is not increased. 1 The mitigating systems for a loss of condenser vacuum event analyzed in USAR Section 15.2.3 and 15.2.5 are unaffected by this change.

Therefore, no assumptions made in evaluating radiological consequences of accidents are affected, and the consequences of previously evaluated accidents are not increased.

The worst-case scenario introduced by the change would be the loss of one running pump, along with the failure of both standby pumps to i start. Since the automatic start feature and independent switch circuitry are unaffected by the change, multiple failures would have to occur for both standby pumps to fail to start. Therefore, the change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR.

Since the mitigating systems and bounding analysis for the turbine trip event are unaffected, no new accident or malfunction is created by the changes.

No acceptance limits are affected, and no reduction in the margin of safety is introduced by the proposed change.

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Safety Evaluation: 59 97-0045 Revision 0 Number One Seal Housing Bolting l

' Westinghouse Technical Bulletin ESBU-TB-96-02-RO identified the potential for seal housing joint leakage in Westinghouse model 93A-1 reactor coolant pumps (RCPs) . This technical bulletin identified problems associated with the No.1 seal housing bolts, P/N 1741C63H01, ,

that cause insufficient bolt pre-load. Performance Improvement Request 95-2488 was generated to evaluate the seal housing bolting problem experienced by Callaway that was later re-addressed in PIR 96-1810.

Since the seal housing bolting is to be removed during Refuel IX to facilitate RCP internals replacement, it is recommended that the existing bolts be replaced with studs, nuts and washers. As a result of this proposed change in configuration, an engineering evaluation would be required to evaluate and document the change in configuration. The configuration change has technical advantages over the original bolting, including: 1) Ability to achieve required torque and enhance bolt elongation; 2) Time to remove seal bolting is d creased; and 3) Stuck studs can be removed with less effort.

Configuration Change Package (CCP) 07109 was written to address this issue. The CCP provides the option of replacing RCP No. 1 seal housing lower bolts with studs, nuts, and washers. The No. 1 seal housing is secured to the thermal barrier flange with sixteen bolts and provides a pressure boundary for the reactor coolant. The bolts i are heated for pre-load stretch and removal. The process of heating l and cooling is cumbersome'and time consuming. The studs can be stretched using hydraulic tensioners. This design will save time and l reduce radiation exposure.

The original bolts were made of SA-540 Gr. B24 or B23 CI2 steel. The studs and nuts will be made of Inconel SB637 (N07718) . USAR Table 5.2-2 lists closure bolting material for the RCP, Inconel SB637 (N07718) is not included in this listing. Inconel is stronger and more corrosion resistant to borated reactor coolant that the current bolt steel. Inconel has a slightly different modulus of elasticity and therefore will require a slightly different stretch to provide the j same clamping force presently. employed.

The new design meets ASME' Boiler and Pressure Vessel,Section III, Class 1 requirements. The effects of additional weight due to the replacement of subject bolts with studs, nuts and washers have been reviewed by the pump vendor (Westinghouse). The existing seismic i qualification of the pump and the RCS loop analysis remains valid.

The studs and nuts are sized to require no modification of the RCP.

The change has no impact on the design basis function of the RCP. '

USAR Table 5.4-1 lists the weight of the RCP as 204,035 pounds. The

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change to studs, nuts and washers increases the weight by 101 pounds. j No new failure modes are being introduced. The change is considered an enhancement of the bolted joint as it is more convenient in ]

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installation and the higher corrosion resistant studs and nuts will J have longer life than the existing bolts.

The change was reviewed for potential impact on design basis accidents. USAR Section 15.3, Decrease in Reactor Coolant System Flow Rate, and Section 15.6.5, Loss of Coolant Accidents, are applicable to this change. The subject design has no impact the initiation of or i mitigation of these accidents. The studs and nuts perform the same function as the bolts and are designed to comply with the original design criteria. The material change enhances the life of the studs and nuts. The change has no affect on the ability of the pump to perform its safety functions or the pressure retaining capability of the bolted joint. Therefore, neither the probability or consequences i of a previously evaluated accident is increased. i l

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Safety Evaluation 8 59 97-0046 Revision: 0 Revision to Procedure AP 30E-002 USAR Section 13.2.2.10.3, Review for Effectiveness of Instructional l Techniques and Materials, indicates that "The educational specialist I reviews the critiques generated in-accordance with Section 13.2.2.10.2 and meets with individual trainees to determine areas for improved instructor performance." This statement implies that the trainees

! with feedback comments are interviewed by the instructional analyst for suggested improvements. While the instructional analyst may talk I with trainees to clarify comments, most often there is no need for an interview. In addition, the evaluation process conducted by the instructors many times involves discussing the comments with trainees. Therefore, the statement in Section 13.2.2.10.3 is deleted, since Section 13.2.2.10.2 adequately describes the trainee feedback evaluation process.

This change is considered to be a clarification to the USAR. The feedback evaluation process is not being modified. The change does not have an adverse impact on the trainir ; or qualification of Wolf l Creek personnel, nor does it affect the ability of personnel to

! perform their assigned duties. This change does not directly or indirectly affect the function of any structure, system, or component analyzed in the USAR.

Based on the above, neither the probability nor consequences of a

< previously evaluated accident-is increased, nor does the change

! increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. No new accident or malfunction of equipment important to safety is being introduced.

l There are no acceptance limits for this evolution in the Wolf Creek l Technical Specifications. The Technical Specifications and Bases are  ;

[ unaffected by this change. Therefore, there is no reduction in the l I

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Attachment to ET 98-0014 Page-94 of 238 Safety Evaluation: 59 97-0047 Revision: 0 Organization Changes

'The following organizational changes are being made to the USAR Section 13.1.l.3: 1) resumes are updated for new individuals assuming the roles of Chief Administrative _ officer and the Manager of Training, and 2) the resume for the Director Site support is being deleted due l

to the individual being assigned another management position.

l The changes being made reflect changes in personnel. Position responsibility requirements and qualification criteria are not being

• changed. The personnel meet the required qualifications. Design basis accident sequences and operation and design of the plant are '

unaffected by the changes- . The changes do not directly or indirectly affect the function of any structure, system, or component analyzed in the USAR.

Based on the above, neither the probability or consequences of a previously evaluated accident is increased, nor does the change increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. No new' accident or malfunction of equipment important to safety is being introduced.

Since qualification requirements and responsibilities are unaffected, no acceptance limits in.the Wolf Creek Technical Specifications are impacted. The Technical Specifications and Bases are unaffected by this change. Therefore, there is no reduction in the margin of safety l as a result of these changes.

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Attachment to ET 98-0014 Page 95 of 238 i

Safety Evaluation

• 59'97-0049 Revision:0 Change to USAR Table 9.5A-1 to Reflect Actual Configuration USAR Table 9.5A-1, Item D.1.4 (f) , states that WCGS power block stairwells which serve as escape routes and access routes for firefighting are protected with 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated fire doors. This statement is not accurate with respect to stairwells in the Turbine Builling, Radwaste Building, or Communications Corridor. Stairwell doors in these areas have a 1 1/2 hour rating. This change i establishes the acceptability of the current configuration and of revising the current USAR statement to indicate the following:

" Stairwell doors have a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rating in seismic Category I buildings. In non-seismic Category I buildings, etairwell doors have a 1 1/2 hour rating. Fire preplans and drills are performed to provide escape and access routes for all areas."

With respect to fire protection, Wolf Creek's licensing basis is compliance with NRC Branch Technical Position (BTP) APCSB 9.5-1 and Appendix A thereto. BTP APCSB 9.5-1, Section C.5.a(6) states 1

" stairwells outside primary containment serving as escape routes, I access routes for firefighting, or access routes to areas containing )

equipment necessary for safe shutdown should be enclosed in masonry or i concrete towers with a minimum fire rating of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and self-closing Class B fire doors." Class B fire doors have a minimum rating of 1 1/2 hours. In addition, compliance with Appendix A guidelines is demonstrated by fire preplans and drills performed by WCGS Operations and Fire Brigade personnel, using established administrative procedures, to provide escape and access routes for stairwells not enclosed with 3-hour rated fire doors (i.e., those in Turbine Building, Radwaste Building, and Communications Corridor).

The as-built Wolf Creek stairwell / door configuration complies with the j guidelines of the Branch Technical Position. Based on the above, neither the probability or consequences of a previously evaluated  ;

accident is increased, nor does the change increase the probability or l consequences of a malfunction of equipnient important to safety ]

previously evaluated in the USAR. No new accident or malfunction of i equipment important to safety is being introduced. The acceptance limits in the Wolf Creek licensing basis is being met by the change. I The Technical Specifications and Bases are unaffected by this change.

Therefore, there is no reduction in the margin of safety as a result of this change.

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Safety Evaluation 59 97-0050 Revision 0

-Installation of a spectacle Blind in the Essential Service Water System This modification provides for' spectacle blinds to be installed on the

' Emergency Service Water (ESW)' system piping just after the buried piping enters into the control building around elevation 1979 feet.

The function of these spectacle blinds is to isolate the water flow while performing a pressure srt to demonstrate the integrity of the pressure retaining components. During normal operation, the blinds are flipped by spectacles (spacers) to allow the water to flow and

' maintain the ASME pressure boundary. This change requires-adding a

. note to USAR Figure 9.2-2-02, indicating that'the spectacle blind flange is separated and stored in the adjacent area for future application.

CCP 07224 provides an evaluation of this change. The ESW system provides cooling water to plant components that require cooling for safe shutdown of the reactor following an accident. The ESW system is the emergency makeup / backup to the spent fuel pool, component cooling water, and auxiliary feedwater systems. This modification does.not change either the component or system design basis. The function of the blind flange is only during start-up and/or while performing a pressure test, to demonstrate the integrity of the pressure retaining boundary. At that time the blinds are flipped into the pipe to isolate the water flow. The flange installed on the pipe during normal operation has no function.

The spectacle blind'is the same as originally supplied by the vendor.

l The only change is separating the blind from the spectacle and adding

l. an eye bolt / lifting lug for maintenance / lifting purposes. No other l changes in the piping configuration or supporting system are being I made, and no significant impact on the piping stresses. No acceptance limits are affected by the change.

Based on the above, the ESW system will continue to function consistent with its design and licensing bases. Neither the probability or consequences of a previously evaluated accident is ,

increased, nor does the change increase the probability or I

( consequences of a malfunction of equipment important to safety previously evaluated in the USAR. No new accident or malfunction of equipment important to safety is being introduced. The acceptance i limits in the Wolf Creek licensing basis is being met by the change.

The Technical Specifications and Bases are unaffected by this change.

Therefore, there is no reduction in the margin of safety as a result j of this change. i i

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L Attachment to ET 98-0014 Page 97 of. 238-Safety Evaluation: 59 97-0051 J Revision 0

. Revision to USAR Table 8.3-4'to Correct Inconsistancies A revision is being made to USAR Table 8.3-4 to. reflect the location of.the undervoltage annunciation for the NB (Lower Medium Voltage System - 4.16 kV) and NG (Low Voltage System'- 480V) switchgear. In addition, redundant entries for NB0115 and NB0215 are being removed.

The information in the USAR regarding the undervoltage annunciation is being clarified to eliminate ambiguity as to the location of the

-undervoltage annunciation. The change makes it clear that the annunciation is made at the Main Control Board (MCB). The proposed

' revision to the USAR has no affect on other procedures,' activities, or items associated with the operation of Wolf Creek Generating Station.

The only document associated with this revision is the affected USAR table.

The clarification will not increase the probability'or consequences of any previously evaluated accident. The accident analyses have already evaluated undervoltage conditions for the NB and NO switchgear. The location of the undervoltage annunciation is not a factor in any of the accident analyses. There is no affect on the initiation of or any radiological consequence.of the analyzed accidents.

The location of the undervoltage annunciation does not affect the safety-related function of the switchgear. No plant equipment or operating procedures are affected by this change. The change merely clarifies a table'in the USAR to ensure that it agrees with the as-built condition of the plant. Therefore, the change does not increase the probability of or radiological consequences of a malfunction of' equipment important to safety previously evaluated in the USAR,' nor will it introduce a new accident or malfunction of equipment important to safety.  !

I The acceptance limits in the Wolf Creek licensing basis are being met ]i by the change. The Technical Specifications and Bases are unaffected by this change. Clarification of the location of the undervoltage annunciation, coupled with eliminating' redundant information in-the

' table will served to reinforce any bases information contained within

-the Technical Specifications. Therefore, there is no reduction in the margin of safety as a result of this change.

1 Attachment to ET 98-0014 Page'98 of .238 Safety Evaluation: 59 97-0052 Revision: 0 USAR Revision to Reference Procedures and Methods for the Fuel Building Emergency Exhaust Mode USAR Section 9.4.2.2.3 clarifies that operators can use either approved plant procedures or the fuel building ventilation isolation signal switch to manually initiate the fuel building emergency exhaust system prior to fuel handling activities. Presently, one area of the USAR states ' transfer from the normal HVAC operations to one train of emergency HVAC operations is manually initiated." This statement implies approved plant procedure can be used to align the system.

Another USAR section states "the *.31 building ventilation isolation-signal (FBVIH) is manually initiated, thereby starting one train of the emergency exhaust system and aligning the required dampers. Once this is accomplished, the FBVIS is manually cleared and a supply air handling unit is manually put in'a recirculation mode." This statement implies that the FBVIS switch is used to align the system.

This change will revise the USAR to clearly state that either approved plant procedures or the FBVIS switch can be used to align the fuel building ventilation ~ system into the emergency mode of operation.

The functionality of the FBVIS safety actuation system or other safety actuation systwems is not impacted. As discussed in the USAR, if.a FBVIS actuation is received, the fuel building normal supply fan (as aligned in the recirculation mode) will shut down. The use of a procedure to align the system in the emergency mode will not affect this function. The change has no impact on the parameters used to evaluate design basis accidents. The fuel building ventilation system is not an accident initiator and, per design, all fuel building exhaust flow will pass through the safety related filter absorber units. The change does not impact this assumption; thus, the radiological consequences of an accident will not be increased. Based on the above, this clarification will not increase the probability or i consequences of any previously evaluated accident.  !

The alignment and operation of the fuel building HVAC system and dampers are the name whether done per approved procedure or by using the FBVIS swit' . The alignment has been verified per drawing ]

configuration and during actual operating and test conditions. No i credible malfunctions are introduced by this clarification. Since i operation of the system is unaffected, the single failure analysis performed on the system is unaffected by this change. Therefore, the change does not increase the probability of or radiological consequences of a malfunction of equipment important to safety j previously evaluated in the USAR, nor will it introduce a new accident l or malfunction of equipment important to safety. l I

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l The Technical Specifications'and Bases are unaffected by this change.

Technical' Specification requirements for ensuring.that the system

' maintains W" negative pressure (Technical Specification 4.9.13) are

-not impacted by this change. This requirement ensures that a radiological release from the fuel handling building remains within allowable release limits contained in the CFRs. Since no acceptance i

limits are affected, there is no reduction in the margin of safety'as a result of this change.

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. Attachment to ET 98-0014 Page 100 of238:

Safety Evaluations' 59 97-0053 Revision 0 Temporary Modification to Trouble Shoot a Ground USAR Section 8.1.4.3 indicates that the non-class 1E PK02 system is provided with the following alarms in the control room: system ground;. battery imbalance; charger DC overvoltages. charger AC undervoltage; charger DC undervoltage; charger AC and DC breakers opens charger failure; loss of distribution board voltage; and loss of -l switchyard voltage. A Temporary Modification order (TMO) is being i inglemented to assist in troubleshooting grounds on the power block

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non class 1E DC power system associated with bus PK01. Specifically,  !

the TMO allows connection of test equipment to the non-safety related l equipment powered from PK02 and PK22 to assist in the troubleshooting effort.

During the troubleshooting efforts, at various_ times, not all of the  ;

alarms described in USAR Section 8.1.4.3 will be active This l Unreviewed Safety Question Determination evaluates the acc3ptability of these activities, j

The TMO. imposes a requirement that the equipment being used has an  ;

input impedance of sufficient magnitude (5000 ohms) to ensure no adverse affect on the system. At a maximum bus voltage of 140 VDC, the maximum current draw of a test instrument having 5000 ohms resistance would be 28 milliamps. The system has sufficient margin to j allow connection of multiple instruments having an input impedance of j at-least 5000 ohms to aid in troubleshooting. Although each parallel I

connection reduces the total effective impedance of the equipment, there is sufficient margin between 28 milliamps and an actual current which would noticeably affect system performance to accommodate the expected test equipment. l The subject TMO also allows disabling any or all of the PK02 local annunciation circuits as well as the circuitry associated with the remote annunciation in the control room, to allow proper connection of the DC ground fault detection equipment. Removal of the ground detection circuit has no adverse affect on the operation of the DC system with or without the ground fault. DC bus amperage indication will remain available to the control room operators. ]

The troubleshooting technique being used has been successfully used on DC systems at other nuclear stations. This technique does not degrade the performance of the DC system. These systems do not contribute any current to the system, they simply condition the current which is already present, as a result of the ground fault, into a form which can be detected.

The most likely accident probability or consequences which could be i

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I Attachment to ET 98-0014 Page 101 of .238 affected'by this temporary modification is the " Loss of Nonemergency- '

AC Power to the Station Auxiliaries."- The probability or consequences associated with this event will not change because the troubleshooting.

activities allowed by this TMO do not influence or disturb the power sources or distribution network. Therefore, this TMO does not increase the probability or consequences of any previously evaluated accident.

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-The PK02 bus is not a primary power source for any safety related equipment. A failure of the bus'could, however, affect support i systems for safety related equipment, and possibly' lead to a plant trip which would challenge safety systems. However, the controls identified in the TMO limit the' fault current to a low enough value that will. prevent any noticeable affect on system performance. If the troubleshooting is performed incorrectly, or if an accidental ground is caused while connecting or disconnecting equipment simultaneously with an existing' ground, the potential exists to challenge safe systems or caase a plant trip. The controls of this TMO are believed to be adequate to preclude such an occurrence. No seismic environmental, or separation concerns are created by this modification, as the allowed equipment will only be connected to non-~

l safety related equipment. Based on the above, the TMO will not-increase the probability of or consequences of a malfunction of equipment important'to safety previously evaluated in the USAR, nor will it introduce a new accident or malfunction of equipment important to safety.

Technical Specification 3.8 and its associated Bases are unaffected by this change. Since no acceptance limits are affected, there is no reduction in the margin of safety as a result.of this change.

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Attachment to ET 98-0014 Page 102 of 238 Safety Evaluation: 59 97-0054 Revision 0 Change to Accuracies for Wind Direction Instruments on the Meteorological Tower This change corrects the wind direction instrument accuracy values in the Total Plant Setpoint Document to match the values in USAR Table 2.3-48 and Regulatory Guide 1.23, "Onsite Meteorological Programs."

In addition, USAR Table 7A-3, Data Sheet 17.5, and USAR Table 2.3-48 have conflicting accuracy requirements. The values listed in USAR Table 7A-3 are not correct and will be revised to be consistent with

.the values in USAR Table 2.3-48, Regulatory Guide 1.23, Regulatory Guide 1.97, and the achievable accuracy of the wind direction instruments. The specific discrepancies are described below.

USAR Table 2.3-48, " Operational Meteorological instrumentation on Tower (After March 5, 1980)," states that the wind direction sensors have a +/- 3 degree accuracy. USAR Chapter 16, " Wolf Creek Operational Requirements Manual," Section 16.3.1.2, " Meteorological Requirements," provides the limiting condition for operation, surveillance requirements, and Bases for the meteorological instrumentation at Wolf Creek. The Bases section states, " ...is consistent with the recommendations of Regulatory Guide 1.23, ... ".

Regulatory Guide 1.23 provides accuracy requirements for wind direction instruments of +/- 5 degrees. USAR Section 7A, " Comparison to Regulatory Guide 1.97," Table 7A-3, Data Sheet 17.5, identifies the Regulatory Guide 1.97 requirements for wind direction instrumentation as "O to 360 degrees (+/-5 degrees . . . ) " . WCGS Design Provisions for I this requirement are "0-540 degrees +/-2 degrees."

The Wolf Creek Generating Station wind direction instrument loop has tour active componants: sensor, transmitter (translator card), ,

computer (NPIS), and digital recorder. Normally, the NPIS computer will be used to determine wind direction. The chart recorder is used as a backup means for determining wind direction if the NPIS computer j is not available. To meet the requirements far the sensor and loop  !

accuracy, as stated above, sensor accuracy should be less than or i equal to +/-3 degrees and the instrument loop accuracy should be less l than or equal to +/-5 degrees. Loop accuracy is calculated by taking the square root of the sum of the squares for all components in the loop. Setpoint Change Request RD-97-027 changes the component accuracy to meet the USAR sensor requirements of +/- 3 degrees and loop requirements of +/- 5 degrees. Table 7A-3 is being revised to indicate an instrument loop accuracy of +/- 5 degrees. 1 i

l Meteorological instrumentation provides information to assist the j operator in making decisions which are intended to mitigate the consequences of nuclear accidents. As such, the instrumentation is not involved in the initiation of any design basis accidents. Since l

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i Attachment to ET 98-0014 Page 103 of-238 the requirements provided in Regulatory Guide 1.23 are being met, the assumptions in the accident analyses are unaffected. Therefore, this change does not increase the probability or consequences of any previously. evaluated accident.

The operation, design, function, and maintenance of these instruments is unaffected by the proposed change. Because the instruments will continue to meet the licensing requirements for Wolf Creek, this change will not increase the probability of or consequences of a r.alfunction of equipment important to safety previously evaluated in the USAR, nor will it introduce a new accident or malfunction of equipment important to safety.

1 As indicated above, USAR Section 16.3.1.2, " Meteorological l

Requirements," provides the limiting condition for. operation, surveillance requirements, .ind Bases for the meteorological instrumentation at Wolf Creek. The Bases section states, "

...is consistent with the recommendations of Regulatory Guide 1.23, ... . "

Since the requirements of Regulatory Guide 1.23 are being met, no acceptance limits are affected, and there is no reduction in the margin of safety as a result of this change.

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m Attachment to ET 98-0014 Page 104 of 238 Safety Evaluation: 59-97-0056 Revision 0 Revision of Documentation to Reflect Removal of Showers in Access Control Performance Improvement. Request (PIR) 97-0136 identified discrepancies between the plant as-built configuration for Access control Men's Shower Room 3215 modification and Plant Modification Request (PMR) 04627. The room originally had six shower facilities (fittings), on north and south walls of the room. Presently, the room only has three showers'on the south wall. The shower. facilities on the north wall had been removed and the-piping associated with the showers ha~e been capped during implementation of PMR 04627. USAR figure 9.2-17-02, and Table 1.7-2, Sheet 5, were affected by this change. However, the PMR did not discuss or approve removal of any of these' shower facilities in the room. The modification was performed outside of the scope of the PMR. This disposition is initiated to revise the documentation to maintain.the plant configuration and evaluated the acceptability of the modification to the shower facilities.

The showers have no safety related function. In addition, there is no safety related equipment in Room 3215. The removal of the showers does not affect the existing design bases of any safety-related system. No design parameters or accident analysis assumptions are affected by the change. Therefore, this change does not increase the probability or consequences of any previously evaluated accident.

The proposed change does not degrade performance of or seismic qualification et any safety related system's functional design. The change will not directly or indirectly affect equipment protective features, system redundancies, or frequency of operation of safety

-related equipment. The changes do not adversely affect the ability of any system to perform its safety related functions, nor does it affect assumptions in any radiological analysis. The change does not alter f ailure n:edes or result in requiring systems to perform outside of their design criteria. Based on the above, this change will not increase the probability of or consequences of a malfunction of equipment important to safety previously evaluated in the USAR, nor will it introduce a new accident or malfunction of equipment important to safety.

Since.the changes impact only non-safety related equipment that has no associated acceptance criteria, there is no reduction in the margin of safety'as a result of this change.

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i Attachment to. ET 98-0014 Page 105 of 238-8%fety Evaluation: 59 97-0058 ~ Revision 0

. Changing lthe Fire Protection Program Table to Address Radweste Building separation USAR Table 9.5A provides a comparison of the Wolf Creek Generating Station design to NRC Branch Technical Position (BTP) APCSB 9.5-1, GUIDELINES FOR' FIRE PROTECTION FOR NUCLEAR POWER PLANTS, and its Appendix A, GUIDELINES FOR FIRE PROTECTION FOR NUCLEAR POWER PLANTS

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l9 iDOCKETED PRIOR TO JULY 1, 1976. APCSB 9.5-1, Appendix A states "The Radwaste Luilding should be separate from all other areas of the plant

by. fire barriers having at least 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ratings. " USAR Table 9. 5A -

indicates, "The Radwaste Building is physically separated from the rest of the plant by approximately 100 feet." Actual design indicates that-there is approximately 30 feet between the Radwaste building ~and the fuel-building at their closest point, Additionally, there is a tunnel'from the Radwaste building which connects to the 1974' elevation of the Auxiliary Building. This Unreviewed Safety Question Determination evaluates the resolution of the discrepancy in USAR

' Table 9.5A versus actual Wolf Creek design.

i USAR Section 9.5B, Fire Hazards Analysis (FHA), contains correct information which was used to evaluated the impact of a radwaste building fire on safety-relatr,d areas of the plant. The FHA assumes j L -that 3-hour fire barriers provide adequate protection to safety- 1 related equipment from the effects of a fire external to the Fire I Area. With respect to separation of the Radwaste building and the Fuel building, the 30 feet of separation, along with the construction of the Fuel building walls (concrete construction), provides.

equivalent protection to that afforded by a three-hour barrier. With

. respect to the tunnel connecting the Radwaste building and the 1974' I elevation of the Auxiliary building, all cable and piping penetrations I

-are fitted with 3-hour rated penetration seals. Ducts penetrating the wall are fitted with 3-hour rated fire dampers. The tunnel is provided with automatic fire detection that alarms locally and in the Control Room. Manual fire suppression capability is provided by fire extinguishers and hose racks located outside of the Radwaste tunnel.

There are no safe shutdown circuits located in the Radwaste tunnel, j Therefore, the Radwaste tunnel is separated from the Auxiliary l building by 3-hour rated barriers.

Based on the above, .the Wolf Creek design is such that a fire in the Radwaste building will not spread to any safety related structures.

Therefore, this change does not increase the probability or consequences of any previously evaluated accident, j The proposed change does not degrade performance safety related-systems functional design. The change will not affect equipment protective features, system redundancies, or frequency of operation of

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Attachment to ET 98-0014 Page 106 of'238'

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safe shutdown' equipment. Therefore,

l. this change will not increase the probability of or consequences of a malfunction of equipment important safety previously evaluated in the USAR, nor will it introduce a new accident or malfunction of equipment important'to safety.

l Since the change being made'to the USAR reflects plant design and is consistent with the acceptance criteria of the NRC BTP and the Fire

, Hazards Analysis, no acceptance limits are affected. -Therefore, the l- margin of safety is not reduced.

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Attachment to ET 98-0014 Page 107 of 238 Safety Evaluation: 59 97-0061 Revision 0 Essential Service Water Pre-Lube Piping Modification Design Change Package (DCP) 05793 revised the P&ID for the Essential Service Water (ESW) prelube system to show new pipe lines, new line numbers, and new valves (Reference USAR Figure 9.2-2-03). The description of how the ESW system operates is also changed to conform with actual operating practices.

The changes to the piping and USAR description of the ESW prelube system are designed to eliminate the plugging of the lines which has plagued the system and establish conformance between system operation and the USAR. The prelube system design requires a continuous flow of water to the prelubt storage tanks and the ESW pumps. Th1 prelube system provides the lineshaft bearings and packing with water to moisten the packing and to prevent the bearings above the pit water level from running dry during pump start. Tank size is sufficient to supply a minimum of five minutes of water supply at six gpm with no makeup from the pump discharge line. When the pump is operating, the bearings are lubricated by the pumped fluid. The USAR and design basis documents indicate that the pump can be started with dry bearings if required to do so since the bronze bearings are capable of resisting any pump seizing. The manufacturer has indicted that the prelube system is not really needed as long as the lake level is at least 1085 feet. This level ensures that lubrication of the bearings is achieved within 4 seconds after pump start. However, the prelube  !

system is normally operating. Calculation EF-M-039, Revision 0, concluded that the prelube piping internal diameter needs to be at least 0.551" in order to deliver the 6 gpm of water stated in the USAR. The previously installed piping did not meet this criteria.

Operation of the ESW prelube pump is described on USAR page 9.2-7.

The prelube system is normally operating to lubricate the upper pump bearings. Per the design and licensing basis, the ESW pumps are capable of starting without pre-lubrication. The bronze bearings allow a dry pump start. Only the top three bearings would be running dry for approximately 4 seconds. The pump manufacturer has concurred that the pump can operate in this mode.

Since the ESW pumps are designed to withstand a dry pump start, operation of the prelube system is not necessary to achieve the safety function. The prelube system adds to the reliability of the ESW pump by continually lubricating the bearings. The changes being made enhance the ability of the prelube system to provide adequate bearing lubrication, and do not affect the capability of the ESW pumps to withstand a dry pump start. Since the fu tion of the ESW pumps is unaffected, the changes do not increase the probability or consequences of any previously evaluated accident.

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.Page 108 of 238

.The proposed. change does not degrade the performance of the safety related ESW system. The change does not affect equipment protective features'or system redundancies. .The changes will increase the reliability of the prelube system and further ensure that the ESW.

. pumps will be capable'of performing their safety function. Therefore,_

this: change will not increase the probability of or consequences of a malfunction of equipment important to safety previously evaluated in the USAR, nor will it introduce a new accident or malfunction of equipment important *,o safety.

The changes being made to the USAR are consistent with the design and licensing. basis of the plant. Since the capability of the ESW system

.is unaffected, no acceptance criteria are affected. Therefore, the margin offsafety is not reduced.

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Attachment to ET 98-0014 Page 109 of 238 Safety Evaluation: 59 97-0062 Revision 0 Auxiliary Boiler Modificktions Improving the Reliability, Capability and Availability of the Auxiliary Boiler This modification replaces the existing steam dump lines with large lines, a larger isolation valve, a pressure control valve, and a flow element. This combination of components will allow up to 20,000 LBM/hr flow to the atmosphere while maintaining the non-safety related Auxiliary Steam System pressure at 125 psig (140 psia), ceasuring that flow, and with later revisions of the design change package, feeding that information back into the system to allow for changes in load without tripping the auxiliary boiler. The modification affects the USAR with respect to the drawing depicting the system, Pigure 9.5.9-

1. There are no other affects on the contents'of the USAR. There are no safety related systems or components near to the changes being made by this modification package that can be affected by the changes.

The Auxiliary Steam System is designed to be separated by the maximum extent practicable and blocked by sufficient intervening obstructions to be of no concern to safety related SSCs. Additionally, the Auxiliary Steam System provides a natural barrier to the introduction of radiological influents because of the Auxiliary Steam being at a higher pressure then the system it serves. The changes to the system being made here do not alter those conclusions and will not' increase the probability of occurrence of an accident, radiological consequences, or the consequences of a malfunction of equipment important to safety previously ev sted in the USAR. The Auxiliary Steam System is not addressed in the Technical Specifications as it is a non-safety related system with strictly commercial concerns associated with its operation. Therefore, the modification does not reduce the margin of safety as defined in any Technical Specifications.

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' Safety Evaluation: 59 97-0062 Revisions 1 Auxiliary Boiler Modifications Revision 0 of this Unreviewed Safety. Question Determination (USQD) was reported as follows:

Improving the Reliability, Capability and Availability of the Auxiliary Boiler This modification replaces the existing steam dump' lines with large lin'eo, a larger isolation. valve, a pressure control valve, and a flow element, This combination of components will allow up to 20,000 LBM/hr flow to the atmosphere while maintaining the non-safety related Auxiliary Steam System pressure at 125 psig (140 psia), measuring that flow, and with.later revisions of the design change package, feeding that information back into the system to. allow for changes in load without tripping the auxiliary boiler. The modification affects the USAR with respect,to the drawing depicting the system, Figure 9.5.9-

1. There are no other affects on the contents of the USAR. There are no safety related systems or components near to the changes being made by this modification package that can be affected by the changes.

The Auxiliary Steam System is designed to be separate

• by the maximum extent practicable and blocked by sufficient intervening obstructions to be of no concern to safety related SSCs. Additionally, the Auxiliary Steam System provides a natural barrier to the introduction of radiological influents because of the Auxiliary Steam being at a j higher pressure then the system it serves. The changes to the system being made.here do not alter those conclusions and will not increase the probabilit/ of occurrence of an accident, radiological consequences, or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The Auxiliary Steam System is not addressed in the Technical Specifications as it is  ;

a non-safety related system with strictly commercial concerns j associated with its operation. Therefore, the modification does not i reduce the margin of safety as defined in any Technical Specifications.  ;

I Revision 1 of this USQD provides for evaluation of Instrumentation and  ;

controls associated with the flow element and pressure control valvi.

In addition, this revision provides improvements to the Foxboro control system. This includes modifications for drum level control, combustion control, steam header backpressure control and Oxygen Trim j control. A minor modification in Furnace Safeguard Supervisor System (FSSS) is necessary to provide inputs for the Foxboro Spec 2000 System.

l This modification affects the USAR with respect to the drawing depicting the system. There are no other affects on the USAR. The j accident scenarios described in Sections 2, 6, 9, and 15 are not I

Attachment to ET 98-0014 Page 111'of.238 affected by.this modification. There'are'no credible accident '

scenarios affected by making-the, improvements provided by this modification. The improvements in the non-safety related Auxiliary Boiler components are far enough removed fromfany safety related or

- special scope component to preclude.the chance of creating or

• affecting any credible malfunctions of equipment.important to safety.

The Technical specifications do not address the Auxiliary Boiler.

Based on this evaluation,.this modification will not 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. This modification does not create a possibility for an accident orfmalfunction of a.different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any 1 I

unreviewed safety question.

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Safety Evaluation Sb 97-0063 Revision 0 Low Total Disolved Solids Transfer Design Change Package (DCP) 05952 provides a cross-tie between the service water (EA) system and make-up demineralizer (WM) system. USAR Figures 9.2-1-01 and 9.2-5-02, Sections 4.6.2.1, 9.2.1.1.2.1, 9.2.3.1.2, and 9.2.3.2.1, and Table 9.2-1 will require updating as a result of the proposed change.

The original site design had the potable water and make-up demineralizer system water supply coming from the John Redmond Reservoir through the same pipe. The raw water pumps and auxiliary raw water pumps require high maintenance due to continuous service and the relatively muddy water they transfer from the John Redmond Reservoir. Also, the 8-inch WM line, which runs under the lake between the make-up discharge structure (MUDS) and the chlorine building is experiencing low flow due to a build-up of clams and silting. The installation of this cross-tie will allow the raw water pumps to be abandoned in place, along with all associated components and lines.

The site drinking water is now supplied by Rural Water District #3, which was installed per DCP 5586. This action allows the make-up demineralizer system to be supplied with water from the Wolf Creek Lake. The Wolf Creek Lake water has much lower mud and silt levels than the water being pumped from the John Redmond Reservoir. The use of Wolf Creek Lake water will result in an increase in time between backwashing of the sand and carbon filter beds in the WM system.

Trace amounts of nuclides currently exist in the Wolf Creek lake.

These nuclides would be backwashed back into the lake via the waste water treatment facility (WWT). The proceus will not introduce new nuclides or increase the levels of existing nuclides being introduced into the lake. The nuclides being re-introduced into the lake have already been accounted for in the annual radioactive effluent release report. Therefore, no radiation monitor for the pathway from the shop building to the WWT will be necessary. Daily grab samples performed for effluent released from the WWT to the Circulating water discharge satisfy the 10 CFR 50.55a requirements.

No adverse affect to the shop building equipment will result from the trace amounts of nuclides in the lake water. These trace amounts (including tritium) will be captured on the WM resins and backwashed to the WWT during filter regenerative cycles.

The Service Water system and makeup demineralizer system are non-safety related systems, and are not relied upon by any safety related systems to mitigate design basis accidents, nor are they considered

1 Attachment to ET 98-0014 Page 113 of 238 i

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'nitiators i of any design basis accidents. There will be no impact to any safety.related components or systems from this modification package. The improsements being made will result in ine demineralizer system becoming more reliable and improving availability by resulting in less frequently required maintenance, ultimately easing the burden on plant operators. Neither the quantity or type of radiological effluents is being affected. The WM and EA systems do not have any associated Technical Specifications, nor will the modification affect the Technical Specification or bases of other systems.

Based on the above, the changes do not increase the probability or consequences of any previously evaluated accident, nor will they increase the probability of or consequences of a malfunction of equipment important safety previously evaluated in the USAR, or )

introduce a new accident'or malfunction of equipment important to )

safety. Since the change being made to the USAR are consistent with d the design and licensing basis of the plant, no acceptance criteria are affected. Therefore, the margin of safety is not reduced.

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l Attachment to ET 98-0014 Page 114 of 238' l

Safety Evaluation: 59 97-0064 Revision: 0 I

Feedwater Chemical Addition System Revision The. Updated Safety Analysis Report is being revised to specify " oxygen i control chemical" instead of "hydrazine" in the condensate and  !

feedwater chemical addition system and "pH control chemical" instead of " ammonia" in the feedwater chemical addition system. The oxygen

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control chemical has been changed to carbohydrazide and the pH control i chemical has been changed to ethanolamine (ETA) . These chemical changes do not affect the systems' functions, failure modes or design i basis. The USAR changes to more general terms instead of listing the specific chemicals used allows the use of alternative pH and oxygen j control chemicals to enhance oxygen and pH control without affecting  !

the USAR. An instantaneous spill of 22 barrels of ammonium hydroxide i is analyzed in the USAR. ETA has a much lower vapor presrure and is j less volatile than ammonium hydroxide. Thus, less ammonia gas would reach the control room if ETA were spilled as compared to a spill of ammonium hydroxide. Similarly, carbohydrazide is classified as only a slight health and flammability risk and is, therefore, less of a '

hazard than hydrazine.

This change will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to I safety previously evaluated in the safety analysis report. This I change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications  ;

is not reduced by this change. Therefore, this change does not involve  !

any unreviewed safety question.

Attachment to ET 98-0014 i'

Page 115 of 238

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Safety Evaluation: 59 97-0065 Revision 0 Change in Accuracy of Hydrogen Monitors The USAR states that the Containment Hydrogen Analyzers have an

. accuracy of:S percent-full scalei which includes a 1 percent accuracy for the calibration gas. CCP 07322 changes this value to 6 percent.

.because'the calibration gas currently being.used has an accuracy of 2 l -percent instead of 1 percent This change will not affect the single-failure assumptions made in the USAR and will not affect the ability of the hydrogen' control system to keep containment-hydrogen concentration below 4 percent. The effect of the change in accuracy could result in'the hydrogen recombiners being initiated when containment hydrogen reaches a concentration as high as 1.4 percent, g which is higher than the 0.8 percent specified in station procedures.

However, this slight increase-would still cause the recombiners to be initiated within the "one-day following' accident initiation" assumed in the USAR,'and the recombiners would still prevent containment

hydrogen concentration from exceeding 4 percent.

I This change will not 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. This change does not create a possibii.ity for an accident or malfunction of

a different type than any evaluated'previously in the safety. analysis I report. The margin of safety as defined in technical specifications

is not reduced by this change. Therefore, this change does not involve l any unreviewed safety question.

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Attachment to ET 98-0014

'Page 116 of 238 Safety Evaluations 59 97-0066 Revision: O National Voluntary Laboratory Accreditation Program USAR Chapter 12 contains several statements regarding-the processing of TLDs (Thermo-luminescent Dosimeters) being conducted by the Wolf-Creek Nuclear Operating Corporation (WCNOC) onsite. WCNOC has contracted out the processing of TLDs to an outside firm. The firm used is a NAVLAP-accredited laboratory and is properly certified under

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the requirements of 10 CFR 20.1501. This change does not affect any

. plant system or component's operating procedures-or design basis, but this administrative change does make_information presented in USAR Chapter.12 no longer accurate or true.

This change will not 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. This-change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve

.any unreviewed safety question.

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Attachment to ET 98-0014 Page 117 of 238 i

Safety Evaluation: 59 97-0067 Revision: 0 Fire Protection Water Treatment Chemical Addition Tap USAR Table 9.5A-1 states that chemical treatment (to control organic fouling) of the fire protection water supply system way be accomplished by injection of chemicals into the (fire) pump suction area. CCP 07144 revises the USAR to also allow chemical agents to control corrosion and organic fouling to be' injected into designated injection points (other than the fire pump suction). This change will also designate system vents 1FP140 and 1WS021 as chemical addition taps. This change will also provide for chemically treating the fire protection main loop periodically and the jockey (pump) system continually, in order to control Asiatic Clams in the most active j portions of the system. While other plant water systems have been chemically treated for organic fouling since plant startup, Asiatic Clams have not been found in the fire protection water system until recently. This change does not affect the design function of the fire protection system nor does it affect the operation of the system.

-This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 118 of 238 Safety Evaluations' 59 97-0068 Revision: 0

--Deletion of Locked /open/ Closed valve Status from Piping and Instrument Diagrams

'This modification provides for. deleting notes referring to i locked /open/ closed valves from the Piping and Instrument Diagrams. ]'

Also, the "LO" designation of valves KCV307 and APV006 will be deleted. This will help to prevent confusion regarding locked valves because all locked valve information will be contained in Procedure AP 21G-001, " Control of Locked Component Status." The valve table for I USAR Figure 6.2.4-1 is also being changed. This'is a document change I only. There are no tests or experiments which may adversely affect the adequacy of systems structures or components to prevent accidents or mitigate the consequences of an accident.

This change does not affect any of'the design basis accidents

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discussed or referenced in Updated Safety Analysis Report (USAR)

Chapters 2, 3, 6, 9, or 15. This is a document change only. There are no credible malfunctions of equipment important to safety impacted by this change. Based'on a review of Technical Specifications and the USAR, no acceptance limits are identified that could be affected.

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l Therefore, based on this evaluation, this modification'will not 1 increase'the probability of occurrence or the consequences of an l accident or malfunction of equipment important to safety previously l evaluated in the safety analysis report. This modification does not i create a possibility for an accident or malfunction of a different i type than any evaluated previously in the safety analysis report. The l margin of safety as defined in technical specifications is not reduced  :

by this modification. I i

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I Attachment to ET 98-0014 Page 119 of 238 Safety Evaluationt 59 97-0069 Revision: 0 l

Radiological Emergency Response Plan Revision Revision 56 to the Radiological Emergency Response Plan (RERP) revises Emergency Action Levels regarding Safety System Failure or Malfunction. The value of 260 klbm/hr is being revised to 270 klbm/hr to be consistent with Emergency Operation Procedures. This change is a conservative change as the higher level of flow could require an emergency classification at an earlier point in an event.

There are no procedures, activities, administrative controls, or sequences of plant operations; or any plant structures, systems  ;

components or equipment; or any requirements outlined, summarized or described in the Updated Safety Analysis Report (USAR) which are no longer true or accurate as a result of this revision to the RERP. ,

This revision affect only the RERP response to a safety system failure or malfunction. I There are no tests or experiments that are not described in the USAR that may adversely affect the adequacy of systems, structures or components to prevent accidents or mitigate the consequences of an accident. This revision will provide for a possible earlier emergency classification during a safety system failure or malfunction. There ,

are no accidents discussed in USAR Chapters 2, 3, 6, 9, or 15 that are !

impacted by this revision. There are no credible accidents that this revision to the RERP could create. There are no credible malfunctions of equipment important to safety which may be affected by this revision. There are no acceptance limits contained in licensing basis documents that could be affected by this revision.

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' Attachment to ET.98-0014 Page 120 of 238 Bafety Evaluations. 59 97-0070 Revision: 0 USAR Corrections and Clarifications This revision to the Updated Safety Analysis Report (USAR).provides for multiple corrections in USAR Table 4.1-2 where " reload core,"

~1nformation is.in the " initial-core" sections of~the table. Excess-reactivity values in USAR Table 4.3-1 which are poorly defined and not mentioned in the USAR text are deleted. _ Figures 4.3-G and 4.3-46, which contain " typical" data are updated to the current cycle. The Section header and four and one half sentences of text, removed in an earlier revision, are restored. Sections 4.3.3.2 and 4.3.3.3'are revised to clarify which methods were used for initial core design and which metho's d are used for reload core design. Multiple minor editorial changes are incorporated to correct typographical errors, correct references to the correct table numbers, clarify the connections between values in tables and the associated texts and update. reference lists.

This' revision'has no effect on systems, structures, or components.

This revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 121 of 238 Safety Evaluation: 59 97-0071 Revision 0 Reactor Coolant Pump Motor Conduit Seals

. Two modifications have been made to the reactor coolant pump . tRCP) motors which affect the lube oil collection system. PMR 3702 added remote lube oil fill lines for the upper and lower bearing lube oil reservoirs for each RCP motor. The lower reservoir remote fill line consists of 2-inch pipe connected to the lube oil reservoir internal to the motor housing by welded connection at a point below the rtatic oil level line. The fill line then extends through a vent opening in the housing and extends upward to a point approximately 4.5 inches above the static oil level line. The upper reservoir remote fill line consists of 2.5-inch pipe connected to the motor housing by a welded connection at a point below the static oil level of the reservoir.

The fill line then extends upward to a point approximately 5 inches

.above the static oil level line. Both remote lube oil fill lines have been seismically analyzed and are designed to meet the applicable seismic requirements of Regulatory Guide 1.29. Lube oil leakage from the additional lines is-not postulated due to the location of these lines in the low pressure portion of the system, and the welded-joint construction, which meets the seismic design requirements of Regulatory Guide 1.29. Due to the location and design of the new lines, they do not introduce any new lube oil fire hazards and meet the safety objectives of Section III.O of 10 CFR Part.50, Appendix R.

'DCP 07280 added a conduit seal and leak-tight fittings to the resistance temperature detector (RTD) conduit boxes (three boxes on each motor). The RCP motor upper bearing RTD holders allow the motor upper bearing RTDs to penetrate through the wall of the upper bearing reservoir while providing an oil-tight' seal. The RTD holders are contained inside a conduit box and are located just above the upper bearing oil level with the motor operating. However, incidental oil may contact the RTD holders inside the reservoir due to normal oil turbulence and splashing while the motors are operating. The design of the RTD holders and the conduit boxes does not have any oil collection capabilities. In the past this has allowed minor oil leakage from the RTD holders to leak out of the conduit boxes and onto the exterior of the motors. The RTD holders utilized a combination of compression fittings with an o-ring to preclude oil leakage. However, failure of either the fitting or the o-ring would allow small amounts of oil to leak past the fittings, into the conduit boxes, and onto the motor casing. This modification added additional sealing features by installing EYS sealing fittings between the existing flex conduits and the RTD conduit boxes. These fittings have silicone foam seals

. installed to a depth of six inches and are installed such that the drain plugs in the fittings are not restricted. The existing conduit nipples, locknuts, and isolation unions are utilized in the new design, with the addition of sealing washers added internally to the

Attachment to ET 98-0014 Page 122 of 238 RTD conduit boxes to prevent oil leakage from around the pipe nipples. Gaskets are also added to the inside perimeter of the conduit box covers to prevent leakage from the boxes. All non-gasketed threaded connections have been coated with silicone sealant, as well. These measures are expected to minimize oil leakage to very low amounts. -These low amounts of oil leakage .is not considered to represent a fire hazard, due to the low amounts involved and the fact that the calculated surface temperatures for the component insulation in the expected areas of the leakage are well below the flash point of the oil, and there are no other sources of combustion in these areas.

1 Neither the addition of the remote lube oil fill lines nor the

. postulated oil leakage at the RTD conduit boxes represents a fire hazard, and the ability to achieve and maintain. safe shut down in the event of a lube oil fire are unchanged from previously accepted conditions. Additionally, no new' fire hazards are postulated during normal or accident conditions.

Based on this evaluation, the sealed conduit system will perform satisfactorily and will sufficiently contain any oil leakage from the upper bearing RTD holders. No lube oil leakage is postulated from the remote lube. oil fill lines. Thus, this modification'will not 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; does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report; and the margin of safety as defined in technical specifications is not reduced by this.

modification. Therefore, this modification does not involve any I unreviewed safety question. l l

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1 Attachment to ET 98-0014 Page 123 of 238 Safety Evaluation: 59 97-0072 Revision: 0 Revises USAR to Reflect Use of Fiber Optics Rather Than Micro Waves For Comminications This modification provides for updating the microwave communication system to a fiber optic based communications system. This will consist of eliminating the microwave stations located at the Wolf Creek Generating Station (WCGS) Meteorological (Met) Tower and several additional receiver / transmitters between the site and the Wichita i System Control Center. In their place will be a fiber optic backbone

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(Sonet Ring). This ring design includes the WCGS switchyard and Met Tower. The ring extends from WCGS in a west southwesterly direction.

The other side of the ring goes in a South Southeasterly direction, providing for significant spatial separation. Included in the ring )

are the Wichita System Control Center and the Western Resources i General Office in Topeka. This system remains an alternate to the  ;

four other systems described in Updated Safety Analysis Report (USAR)

Section 9.5.2. The old microwave system failure mode consisted of the failure of standard commercial equipment in any one of several towers between WCGS and Wichita. Because of the ring design the new failure mode will require the failure of the cable in two locations along the ring on opposite sides of the WCGS site, or the failure of the standard commercial equipment located at the Met Tower, similar to the I old microwave system. All existing microwave phone traffic will be rerouted via fiber optic.

There are no design basis accidents discussed in the USAR with respect i to the microwave communications system. This system provides no control functions for any plant hardware. Therefore, the system

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cannot create an accident. There can be no equipment malfunctions because this equipment provides no control functions for any plant equipment. Technical Specifications do not identify the microwave or j offsite communications. l Based on this evaluation, this modification will not 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. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

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. Attachment to ET 98-0014

. Page 124 of 238 Safety Evaluation: 59 97-0074 Revision: 0 USAR Revision to Correct Discrepancies Identified by Self Assessment Several discrepancies in the USAR were identified during the AFW Walkdown Self Assessment (SEL 97-11). These discrepancies include:

removing a statement from Section 3B.4.2,.which' states that part of I

the AFW system is located in the Main Steam / Main _Feedwater isolation valve compartments. correcting Table 7A-3 by replacing the, "ERFIS computer," with the, " Plant Computer;" deleting reference to NPIS computer points for several AFW parameters and correcting a typo in a figure number; correcting equipment room locations, fire zone and

- separation groups in Table 9.5B-2; clarifying in Section 10.4.9.2,2 E that' the Turbine Driven Auxiliary Feedwater Pump (TDAFWP) turbine exhaust exits over the auxiliary boiler building roof, and that.the TDAFWP discharge control valves are positionable, air-operated valves.

These changes are administrative changes to improve the accuracy of the USAR descriptions of the AFW system. These changes do not affect the AFW design, configuration, operation, or qualification and testing.

Based on this evaluation, these changes will not increase the l

probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in i the safety analysis report. These. changes do not create a possibility i for an accident or malfunction of a different type.than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by these changes. l Therefore, these changes do not involve any unreviewed safety question.

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' Attachment to ET 98-0014 Page 125 of 238 I

1 Safety Evaluation: 59 97-0075 Revision: 0 I Hot Machine Shop Reconfiguration This modification provides for the removal of Door 13332, the associated door frame, concrete blocks above the door frame, relocation of a conduit, and associated lighting panel QA48. This

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door is located in a eight foot tall concrete block partition wall that divides the hot machine shop (Room 1332)- from the decontamination 3 area (Room 1333). This door does not serve as a radiation, fire protection,.or flood barrier between Rooms.1332 and 1333.

The MVAC and fire' detection system for these rooms will be unaffected by this modification. The radiation zone designation for Rooms 1332 and 1333 will remain unchanged. The overhead crane controls will be able to move freely between Rooms 1332 and 1333.

There are no design basis accidents impacted by this modification nor are there any credible accidents created by this modification. There is no impact on equipment malfunctions.

Based on this evaluation, this modification will not increase the j

-probability of occurrence or the consequences of an accident or j malfunction of equipment important to safety previously evaluated in '

the safety analysis report. .This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this l modification. Therefore, this modification does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 126 of 238 Safety Evaluation: 59 97-0076 Revision 3 0 Temporary Procedure to Allow for Chemical Addition of Ammonium Hydroxide Temporary procedure TMP 97-015 was issued to allow for the addition of ammonium hydroxide to the condenser /hotwell at an injection point different than that described in the USAR. USAR Section 10.4.4.2.1 states that the condensate and feedwater chemicaliinjection system is provided to inject hydrazine and the pH control chemical (ammonium hydroxide) into the condensate pump discharge downstream of the condensate demineralizers. This procedure will allow ammonium i hydroxide to be injected into the condensate pump suction at drain valves ADV0198, ADV0200 or ADV0202. Under the temporary procedure, ammonium hydroxide will be injected at the new point one barrel at a time. Assuming a handling accident were to occur during use of the new procedura, only one barrel would be spilled, which is much less I than the 22 barrel spill. analyzed in USAR Section 2.2.3.1.8. No system design changes are required for this change, nor are any operating parameters or procedures modified for this change. j j

Use of the new procedure will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety a.Tlysis report. The new procedure does not create a possibility for an accident or malfunction of a different type than any evaluated previous,1y in the safety analysis report. The margin of safety ac defined in technical specifications is not reduced by the new ,

procedure. Thertfore, the new procedure does not involve any '

unreviewed safety question.

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, Attachment to ET 98-0014 Page 127 of_238 Safety Ryaluations.

. 59 97-0077 Revision:O Reactor Coolant Pump Main Flange Bolting CCP. 07126 modified the. reactor coolant pump (RCP) design to replace -!

the RCP main flange bolts.with studs and nuts. _The' studs and nuts are

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to be made of the same material as. the original bolts - (SA-540, Grade  !

B24, Class 4). 'The washer is a non-Code part and is made of AISI 4140-steel, The existing bolts have to be heated and cooled-for

- installation, which is cumbersome and time-consuming. The studs can be installed using hydraulic tensioners, which will save installation time and reduce personnel radiation exposure. Westinghouse has analyzed the studs, nuts and washers for'use and they meet the stress l

limits specified by ASME Section III Subsection NB 1971 Edition through Summer 1973 Addenda. Replacement of the bolts with studs and nuts will require a new No. 1 seal leak-off pipe extension to increase its vertical offset to avoid interference with the nut and washer combination. This change will also require the flex hose assembly BB- '

FH-012 to be modified to match the new elevation cf the pipe extension. These modifications have been evaluated and determined to

' be within Code limits. Similarly, the No. 3 seal leak-off pipe may require an extension to avoid nut interference, as well. All changes meet the original design criteria for the RCP and the ASME Section III modifications will be done in accordance~with the WCNOC ASME Section XI Repair and Replacement Program. These changes are external to the pump rotating elements thus pump performance will not be affected by-the change. The gec . cry changes to the No. 1 and No. 3 seal leak-off piping are minor and the pre-load stretch on the studs will ensure j the same clamping force is provided on the main flange joint that was j provided with the bolts. These changes do not affect _the flow rate  ;

and does not change the function of the seals. The weight added to l the RCP due to these changes is less than 1 percent of the pump I weight. The weight increase has been evaluated by Westinghouse and I the existing seismic qualification for the pump and the reactor )'

coolant system loop analysis remain valid. Thus, these changes will not affect pump operation or pump design basis functions.

This modification will not 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. This ,

modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the i safety analysis. report. The margin of safety as defined in technical l specifications is not reduced by this modification. Therefore, this l modification does not' involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 128 of 238

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l Safety Evaluation: 59 97-0080 Revision 0 Valve Operator Change j DCP 05006 modified valve BG8518, in room 1308, to replace the reach rod with a hand wheel on the valve. This valve had a reach rod and drive mechanism to enable manually cycling the. valve remotely.

However, mechanical difficulty in operating this mechanism resulted in the valve being over torqued, which can cause failure of the valve diaphram; or under torqued, which can cause. leakage past the valve, which in turn would alter the water chemistry of the reactor coolant system. Replacing this mechanism with a hand wheel will allow the valve to be operated locally'and thus ensure proper valve closure.

Radiation dose levels in room 1308 during normal plant operations will permit local operation of this valve, operation of this valve is not relied upon for any design basis accident. Making this change will require revision of USAR Figure 9.3-08-02 to remove the remote operator designation from the valve. This valve is normally closed during normal plant operation _and is manually opened and closed to adjust pH levels in the reactor coolant. This valve also serves as a pressure boundary for. radioactive fluid. However, this modification does not affect the pressure boundary function for the valve.-

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specificetions is not reduced by this change. Therefore, this change does not involve any unreviewed-safety question.

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-Attachment to ET 98-0014 Page 129 of 238 Safety Evaluation: 59 97-0081 Revision 0 Organisation Changes i This change is being made to USAR Section 13.1 to better describe the current organization and to reflect the persons filling certain-positions. The Accountability, Commitment and Excellence Officer (ACE) and the Director Site Support positions are being deleted. l Reporting relationships have been redistributed due to the deletion of j these titles. Electrical and I&C functions have been combined under j

4 one superintendent. Personnel changes (Manager Training and the Chief Administrative Officer) have been finalized and the resumes in the l USAR for these positions are being updated. In addition, .)

typographical errors are being corrected. These are administrative l

type changes and there is no change to any plant design or system  ;

operating parameters or methods. These changes do not. affect any j personnel qualification requirements for these positions.

'This document revision will not increase the probability of occurrence  ;

or the consequences of an accident or malfunction of equipment l' important to safety previously evaluated in the' safety analysis report. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysia report. The margin of safety as defined in technical specifications is not reduced by this revision.'Therefore, )

.this revision does not. involve any unreviewed safety question.  !

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Attachment to ET-98-0014 Page 130 of 238' safsty Evaluation: 59 97-0083 Revision 0 Removal of Reference to Strip Chart Recorders PMR 03989 removed 8 strip chart recorders from the main control board in the control room. USAR Section 7.7 is being revised tol remove the discussion of these recorders and add a reference to the NPIS. The

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recorders provided indication to the plant operators of output from the 8 detectors (4 upper and 4 lower) of the Nuclear Instrumentation (NIS) Power Range. This information is still readily available to the operators from the detector current meters and the NPIS computer. The NPIS also provides a record (non-QA) of detector output. The removed

. recorders are.not' utilized in any plant procedure, activity,.

administrative-control, or plant operation sequence. The recorders ]

were used for operator information, only, and provided no input to any i plant system, structure or component. The recorders are mentioned in the USAR only as an indication to the plant operators as detector  !

output, and there are no USAR analysis requirements to have these detectors. l 1

There are no tests or experiments that would be impacted by removal of these detectors, and there are no design basis accidents.which utilize the recorders as either an input to another piece of equipment or as an operator aid. Removal of the recorders would have an impact on the seismic response of the main control board. However, this impact was J evaluated in the PMR and determined to be negligible. The removal of these detectors has no impact on any other plant equipment.

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve  ;

any unreviewed safety question. l l

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o Attachment to ET 98-0014 Page 131'of 238 Safety Evaluation: 59 97-0084 Revision: 0 Minimum Decay Heat Removal Flow Correction USAR Table 10.4-17 was revised to clarify the comment concerning the

. failure of the turbine-driven auxiliary feedwater pump. The comment was revised to state that either of the motor-driven auxiliary feedwater pumps will provide 100 percent of the feedwater requirements for decay heat removal during normal plant cooldown. Also, USAR Section 10.4.9.2.1 was revised to clarify that 470 gpm is the minimum required auxiliary feedwater flow rate needed for decay heat removal during normal plant cooldown. These changes were madu to clarify that these statements are applicable to normal plant cooldown and not to plant accident conditions anal-z-r3 in USAR Chapter 15, which assume multiple auxiliary feedwater pumps operating to provide a higher minimum flow rate.

This editorial revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety es l

defined in technical specifications is not reduced by this revision.

Therefore, this revision does not involve any unreviewed safety l question.

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1 Attachment to ET 98-0014 Page 132 of 238 Safety Evaluation: 59 97-0085 Revision 0 Information Added to the USAR Concerning Residual Heat Removal and Steam Dumps i

Operating procedure GEN 00-006, " Hot Standby to Cold Shutdown," allows i the use of both the Residual Heat Removal (RER) System and the Main Steam Dump Valves (MSDV) for normal cooldown of the reactor coolant i system (RCS) from Hot Standby to Cold Shutdown. USAR Section 5.4.7.2, l however, does not specifically allow using both of these nyfe,* at  !

the same time for this purpose. Thus, the USAR was revis*d O? attx-the use of both methods at the same time, as specified it ar - to

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The USAR was also revised to use the same terminology for the Atmospheric Relief Valves used in the WCGS Technical Specifications.

This change also clarifies any confusion between the Atmospheric Relief. Valves and the Pressurizer Power-Operated Relief Valves in this section of the USAR. This USAR change was made to reflect current 1 plant procedures, practices and terminology, and does not involve r l

. plant modification. The maximum allowable cooldown rate of the RCC, '

as specified in the technical specifications, and the methods used to  !

achieve that cooldown have not toen changed. I This revision will not increase the probability of occurrence or the  !

consequences of an accident or malfunction of equipment important to l safety previously evaluated in the safety analysis report. This l revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety I analysis report. The margin of safety as defined in technical i specifications is not reduced by this revision. Ther+ fore, this  !

revision does not involve any unreviewed safety quet".an.  !

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. Attachment to ET 98-0014 Page 133 of 238 Safety Evaluation: 59 97-0085 Revision: 1 Information Added to the USAR Concerning Residual Heat Removal and

-Steam Dumps Operating procedure GEN 00-006, Hot Standby to Cold Shutdown, allows the use of both the Residual Heat Removal (RHR) System and the Main

. Steam Dump Valves (MSDV) for normal cooldown of the reactor coolant system (RCS) from Hot Standby to Cold Shutdown. USAR Section 5.4.7.2, however, does not specifically allow using both of these systems at the same time for.this purpose. Thus, the USAR was revised to allow the'use of botn methods at the same time, as specified in GEN 00-006.

The USAR was also revised to use th's same terminology for the Atmospheric Relief Valves used in the WCGS Technical Specifications.

This change also clarifies any confusion between the Atmospheric Relief Valves and the Pressurizer Power-Operated Relief Valves in this section of the USAR. This USAR change was made to reflect current plant procedures, practices and terminology, and does not involve a plant modification. The maximum allowable cooldown rate of the RCS, as specified in the technical specifications, and the methods used to achieve that cooldown have not been changed. Revision 1 to USQD 59 97-0085 was written to add an additional change to another paragraph in USAR Section 5.4.7.2 to reflect the above changes for consistency.

This additional change was not identified in the original USQD. i Revision 1 also provided some additional clarification to *he above changes for a better understanding of how the Main Steam bump Valves ,

are used for RCS cooldown.- l l

l This revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis repc rt. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

Attachment to ET 90-0014 Page 134 of 238 Safety. Evaluation: 59 97-0086 Revision: 0 Fiber Optic Containment Penetration DCP 07065 added a Fiber Optic Containment Penetration (Penetration ZSE251, spare nozzle E-251) in Room 1409 of the Auxiliary Building (South Electrical Penetration Room, El. 2026') . This penetration provides a raceway for 48 fibers to be used in a variety of non-safety related applications (computer and telephone communications, video camera connections, and eddy current testing). .The penetration closure assembly is classified as a safety-related component, with its sole safety-related function being to serve as a pressure boundary for containment. Therefore, the closure assembly is designed, manufactured, tested and installed in accordance with ASME Section III code requirements for Class MC components. The function of this assembly is equivalent.to existing containment electrical penetration assemblies with regard to maintaining the containment pressure boundary. The fire loading for fire zones A-17 and RB-3 is being

. increased slightly by the fiber optic " pigtails" which are being routed in Separation Group 6 cable tray. This loading change affeci USAR Section 9.5B.7. Also, the penetration assembly is being furnished under the 1983 revisions of IEEE-317 and IEEE-383 standards, which are later revisions than listed in USAR Table 3.2-1.

This modification will not 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. This modification does not create a possibility for an. accident or malfunction of a different type than any evaluated previously.in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 135 of 238 Safety Evaluation: 59 97-0086 Rmrision: 1 Fiber Optic Containment Penetration Dcr 07065 added a Fiber Optic Containment Penetration (Penetration ZSE251, spare nozzle E-251) in Room 1409 of the Auxiliary Building (South Electrical Penetration Room, El. 2026') . This penetration provides a raceway for 48 fibers to be used in a variety of non-safety related applications (computer and telephone communications, video camera connections, and eddy current testing). The penetration closure assembly is classified as a safety-related component, with its sole safety-related function being to serve as a pressure boundary for containment. Therefore, the closure assembly is designed, manufacture 3, tested and installed in accordance with ASME Section III code requirements for Class MC components. The function of this assembly is equivalent to existing containment electrical penetration assemblies with regard to maintaining the containment pressure boundary. The fire loading for fire zones A-17 and RB-3 is being increased slightly by the' fiber optic " pigtails" whici. are being routed in Separation Group 6 cable tray. This loading change affects USAR Section 9.5B.7. Also, the penetration assembly is being furnished under the 1983 revisions of IEEE-317 and IEEE-383 standards, which are later revisions than listed in USAR Table 3.2-1.

Revision 1 of this USQD addresses additional changes.needed to the USAR that were not specifically identified in the original USQD.

Specifically, Section 8.1.4.3 needs to be revised to reflect the later (1983) revision of IEEE-317 that applies only to the fiber optic penetration and to discuss the additional Appendix J testing done for the penetration that is in addition to that tnsting specified in IEEE-317 (1983). The USAR is also revised to discuss that the fiber optic penetration was qualified without the need for constant nitrogen pressurization.

This modification will not 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. This J modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this  !

modification does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 136 of 238 Safety Evaluation

• 59 97-0086 Revision: 2 Fiber Optic Containment Penetration The USAR is being changed to provide additional information on the

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routing of " free-air" fiber optic and computer cables in the Power Block structures. The cables are being routed in accordance with new guidance developed in DCP # 07065. 7n.4 guidance differs from the USAR prescribed separation criteria. USAR Section 8.3.1.4.1.4 identifies exceptions to:IEEE-384, but does not address " free-air" c' ' O e routing. This section will be changed to identify'a relt.:. ion of I

-separation requirements, while still maintaining Safety Train independence.

This revision will not 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. This revision does not create a possibility for an accident or malfunction of a ditferent. type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 137 of 238 Safety Evaluation: 59 97-0087 Revision: O Procedure Revision to Change Electro-Hydraulic Control Pressure Tolerance Settings Procedure MCM-CH-004, Revision 2, specifies an operating pressure range of 1600 to 1650 psi for the EHC hydraulic pumps. USAR Sections 3.5.1.3.2 and 10.2.2.2 state that turbine valve opening actuation is provided by a 1600 psi fluid. This operating pressure is based'on a vendor drawing which specifies 1600 psi as a nominal value. Revision 3 of MCM-CH-004 changed the operating range from its current values to an operating range of 1600 to 1800 psi. This will provide more stable

. operation during pump testing and allow more operating margin. Each EHC hydraulic pump is equipped with a discharge relief valve set at 2000 psi for overpressure protection. The standby EHC automatically starts when system pressure decreases to 1300 psi, and a turbine trip is initiated when system pressure drops to.1100 psi. Therefore, the proposed change is within system operating limits. The EHC operating pressure has no effect on turbine stop and control valve closure times and does'not affect the fail safe function of these valves.

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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Attachment to ET 98-0014.

Page'138 of 238 Safety Evaluation: 59 97-0089 Revision: 0 Change to USAR Table 3.11(B)-1 USAR Table 3.11(B)-1 was revised to specify that certain rooms in the Auxiliary Building containing tanks, piping or components with 4 WTt or higher boric acid solution may experience boron precipitation at

. temperatures.below the stated minimum normal operating temperatures given in the table, and that the minimum room temperatures listed for these rooms should be treated as minimr .ormal operating limits.

These minimum temperatures do not repr ut operability or equipment

. qualification limits but this change r a the concern for the possibility of boron precipitation in .ese rcoms. This change does not impact any of the design basis acc lents discussed in the USAR and brings the USAR table into conformance with design and safety analysis assumptions as stated in the USAR.

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously .n the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 139 of 238 Safety Evaluation: 59 S7-0090 Revision 0 Fuel Assembly Verification USAR Sections 9.1A.7 and 15.4.7 are revised to remove the requirement to positively identify fuel assembly ID numbers leaving the cavity for the spent fuel pool, to replace the requirement to positively identify fuel assembly ID numbers prior to entering the reactor cavity with the performance of an inventory when the core has been completely loaded, to remove the requirement to perform a spent fuel pool inventory of

. fuel assemblies located in Region 1 of the spent fuel pool, and to replace the requirement to perform an annual spent fuel pool inventory with an inventory prior to each refueling outage.

The purpose of verifying fuel assembly identification numbers being placed in the spent fuel pool is to prevent a fuel assembly from accidentally being placed in a Region 2 configuration prior to achieving the required burnup. .off-loaded fuel assemblies are required to be placed in a Region 1 configuration during refueling operations; therefore, it is not possible for fuel assemblies to be accidentally placed in a Region 2 configuration. Also, each fuel assembly identification number is verified immediately prior to moving it to Region 2. Therefore, number verification of fuel assemblies traveling from the core to the spent fuel pool can be eliminated.

Fuel assemblies being moved to the reactor cavity are positively identified after the core is fully-loaded during a core map. This ,

verification replaces the assembly identification from the spent fuel pool to the reactor cavity. Since fuel assemblies are being loaded into the core in a critical configuration, inadvertent criticality in the reactor cavity is prevented by refueling pool boron concentration. Fuel assembly identification is performed to prevent reactor operation with fuel assemblies in the wrcng location. Fuel assembly identification during fuel movement could still lead to a fuel assembly being placed in the wrong location. Thus, a verification of fuel assembly identification numbers after the core is fully loaded adequately assures that fuel assemblies are loaded per design documents.

The spent fuel pool Region 1 inventory prior to Region 2 consolidation is performed to identify which fuel assemblies.are located in Region 1 so that burnup calculations for Region 2 considerations can be performed. The burnup calculation identifies which assemblies have achieved sufficient burnup to be moved to a Region 2 configuration. A i Region 1 spent fuel pool inventory is not necessary because SNM l' records are available to identify the location of each fuel assembly in the spent fuel pool. Since fuel assembly identification number verification is required immediately prior to moving a fuel assembly )

to Region 2, a Region 1 spent fuel pool inventory is repetitive and  !

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~ Attachment to, ET 98-0014 Page 140 of . 238 unnecessary.

'The annual spent fuel pool and'new fuel storage-inventory required in-the'USAR assured that all fuel assemblies located in the spent fuel

pool and new-fuel storage were consistent with SNM records. The l annual inventory was based on the original annual' refueling outage schedule. Since refueling outages'are performed on an 18 month cycle this requirement should also be changed. Performing multiple spent 1 fuel pool and new fuel storage inventories during a cycle is repetitive and does not provide any additional benefit.

The inadvertent loading and operation of a fuel assembly in improper j j position accident analysis, Section 15.4.7 of the USAR, assumes that the probability of core loading errors is reduced because of fuel assembly number verification prior to loading the assembly into the core. .This USAR change will not remove this assumption, it will change the method by which fuel assembly identification number and t position verification is performed. Higher confidence is gained by l verifying the fuel assembly numbers while they are not in' transit.

l This change does not effect the various fuel handling accident analyses.

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications I

l is not reduced by this change. Therefore, this change does not involve any'unreviewed safety question.

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Attachment to -ET 98-0014 Page 141 of 238 Safety Evaluation: 59 97-0091 Revision: O Change in Hydrogen Recombiners Start time i

USAR Sections 6.2.5 and 7A are revised to reflect how the containment  !

hydrogen recombiners are operated. The current USAR description states that the hydrogen recombiners are started one day following a Loss of Coolant Accident (LOCA). However, current emergency operating.

procedures require the hydrogen recombiners-te start when the analyzers detect hydrogen. Using the hydrogen analyzers to start the recombiners could allow the containment hydrogen concentration to

. reach 1.4%. The hydrogen concentration assumed in'the hydrogen-analysis for starting the recombiners one day following a LOCA is 2.25%. Therefore, current operating. practice is more conservative  !

than the hydrogen analysis in the USAR. Starting the hydrogen i recombiners when hydrogen is detected will make the detection of 'j hydrogen a Type A Variable per Regulatory Guide 1.97. However, this classification change does not t 'fect plant operation nor the operation of the hydrogen analyzers.

This change will not increase the probability of occu e rence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This change does not create a possibility for an accident or malfunction of {

a different type than any evaluated previously in the safety analysis l report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any-unreviewed safety question.

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Attachment to ET 98-0014 Page 142 of 238 Safety Evaluation: 59 97-0092 Revision: O USAR Change to Reflect As-Built Configuration of Solid State Protection System USAR Section 7.2.2.2.3.j was revised to correct information concerning the Solid State Protection System (SSPS) General Warning alarm reactor trip subsystem. Specifically, following changes were made to the list of conditions that will cause the general warning alarm circuits to

- automatically trip the reactor: 1) item a was split into two items, one for the loss of either 48 volt de power supplies, and one for the 15 volt de power supplies, 2) item g was revised to state that it is that-train's bypass breaker that causes the signal, not the opposite i train's bypass breaker, 3) item h was split into two items, one for 6- the permissive test switch not in the off position and one for the memory test switch not in the off position, 4) item i was changed to specify the Logic A test switch not in the off position, and 5) a new

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item was added for the Master Relay Selector Switch not in the off position. These changes correct the USAR to' reflect the as-built plant configuration of the SSPS. These documentation changes to the USAR will not affect any plant, procedure or system design information, These USAR corrections will not change any of the j functions or failure modes of the SSPS or any plant system, structure I or component.

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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Attachment to- 'ET 98-0014

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Safety Evaluation: 59 97-0094 Revision: 0

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Reactor Vessel Closure Stuck Stud CCP 07032 evaluates revising the USAR and refueling and inspection procedures to allow stuck reactor vessel studs to remain in the vessel flange during refueling activities rather than cutting out the stuck stud and replacing it. The existing USAR requirements indicate that all studs are removed from the reactor' vessel flange and the stud holes ~are covered, to protect the stud and stud hole from the corrosive effects of borated water in the refueling pool. The procedure revisions would allow the stuck stud to remain in the vessel flange during flooding of the refueling cavity, protected from the borated water by a temporary cover. Additional administrative controls would be added to refueling procedures to prevent inadvertent contact with a stuck stud during fuel movement activities. In addition, the USAR would be revised to allow the studs to be removed to a location other than the containment operating deck for cleaning ,

and storage. i l

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of~ safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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Page 144 of 238 Safety Evaluation: 59 97-0094 Revision: 1 Reactor Vessell Closure Stuck Stud Revision 1/of Unreviewed Safety Question Determination 97-0094 changes Revision'0 as follows:

This modification will allow the option of having a reactor vessel stuck stud or a cut off stuck stud to remain in the vessel flange during refueling activities, with a protective cover installed. The l cover will protect the stud and the stud hole from the corrosive

-affects of borated water.

In addition, repair methods for damaged reactor vessel stud hole threads.will be changed. The repair methods slightly change the reactor vessel stud hole configuration. The methods include removal of damaged threads and possible installation of a sleeve' insert to provide new sound threads. The installation of a threaded insert will I

involve boring a larger diameter hole in the reactor vessel flange.

l This configuration change will not invalidate any USAR statements and i' there is no USAR change being proposed for this repair. The repair l activity will be performed in accordance with the ASME Code Section l XI, 1989.

i Additionally, the wording of one section of the USAR is changed to allow studs to be removed to their cleaning locaLfon, which may be a location other than the containment operating deck.

l l This modification will not increase the probability of occtrren.~e or the consequences of an accident or malfunction of equipment important l to safety previously evaluated in the safety analysis report. Tnis l modification does not create a possibility for an accident or l malfunction of a'different type than any evaluated previously in the l safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

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Attachment to ET 90-0014 Page.145 of 238 Safety Evaluation: 59 97-0094 Revision: 2 Reactor Vessell Closure. Stuck Stud Revision 2 of Unreviewed Safety Question 97-0094 changes Revision 0 and Revision 1 as follows:

CCP'07032 Revision 2 evaluates revising the USAR and refueling and inspection procedures to allow a stuck or a cut off stuck reactor vessel stud to remain in'the vessel flange during refueling activities, with a protective cover installed, rather than cutting out the stuck stud and replacing it. It also evaluates repair methods for damaged reactor vessel stud hole threads. These methods include removal of damaged threads and possible installation of a sleeve insert to provide new sound threads. However, this configuration would not invalidate any current USAR statements, and the repair activity would be performed in accordance with the ASME Code Section XI, 1989, requirements. This type of repair will provide a joint with I equal or greater strength than the original configuration, will I satisfy the design basis requirements for the threaded joint, and will not require any modification to the stud. The existing USAR requirements indicate that all studs are removed from the reactor vessel flange and the stud holes are covered, to protect the stvd and stud hole from the corrosive effects of borated water in the refueling .

pool. The. procedure revisions would allow the stuck stud to remain in )

the vessel flange during flooding of the refueling cavity, or to be cut off but not removed from the vessel flange, protected from the borated water by a temporary cover. Depending on the repair method chosen, additional administrative controls would be added to refueling procedures to prevent inadvertent contact with a stuck stud during fuel movement activities.

This change will not 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. This change does not create possibility for an accident or malfunction of a different type than auf evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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Page 146 of 23P Safety Evaluation: 59 97-0095 RevisionsO 1

Make-up to Spent Fuel Pool fram the Recycle Holdup Tank Procedure SYS HE-206, " Boron Recycle Holdup Tanks Transfer," was revised to allow Recycle Holdup Tank (RHUT) makeup water to be added to the Spent Fuel Pool (SFP) without using a temporary hose connection. Currently, when the Fuel Transfer Canal (FTC) gate is installed between the $FP and the FTC, the SFP and FTC waters are

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'aeparated from each'other. The addition of makeup water from the RHUT-to the SFP then has to be routed through a temporary hose connection from the Boron Recycle System (BRS) FTC flange connection, over the west side of the SFP, between the FTC and SFP, to the SFP. This procedure change provides an alignment configuration for makeup to the SFP from the RHUT using the Recycle Evaporator Feed Pump of the BRS to j the Solid Radwaste System (SRS)' sluicing header then into the spent l Fuel Pool Cooling and Cleanup System (SFPCCS) piping to the SFP. This configuration allows for makeup to the SFP with the FTC gate installed 'j and eliminates the need to use the temporary hose connection when the l FTC gate is. installed. This new configuration is not explicitly l

' described in.the USAR. I I

Evaluation of this revision has concluded !ba.t this change will not 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. This change does not create a possibility for an accident.or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of j safety as defined in technical specifications is not reduced by this  !

change. Therefore, this change does not involve any unreviewed safety l question. 1 i

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Attachment to ET 98-0014 Page 147 of 238 Safety Evaluation: 59 97-0096 Revision: 0 i New Procedure to Place a Load on the Auxiliary Boiler to Allow Instrument Adjustment j i

Temporary Procedure TMP 97-013, " Auxiliary Steam Instrument '

Adjustment," provides for using the Auxiliary Steam Inlet valve (FB PIC-029) to place a load on the Auxiliary Boiler to allow instrument 1 adjustment. This procedure involves removal of the Auxiliary Steam j Header Relief Valve (FBV0130) and replacing it with a spool piece to 1 obtain enough load on the Auxiliary Boiler for main gun operation.

This process is necessary to provide post maintenance testing after implementation of Design Change' Package 05711. The Auxiliary Steam System has no safety function.

No procedures, activities, administrative controls, or sequences of  !

plant operations nor any plant structures, systems or equipment were identified that would be impacted by performance of this temporary procedure. There are no requirements summarized or described in the Updated Safety Analysis Report (USAR) that would become no longer true j or accurate by performance of.this procedure nor would any USAR l requirement be violated. Performance of this procedure will not adversely affect the adequacy of any systems, structures, or l components to prevent accidents or mitigate the consequences of an accident. There are no design basis .idents impacted and there are no credible accidents that could be created by performance of.this l procedure. No credible malfunctions of equipment important to safety are identified by this evaluation.

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No acceptance limits were identified that could be affected, and therefore, the margin of safety is not affected by this prceedure.

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Attachment to ET 59-0014 Page 148 of'238 Safety Evaluation: 59 97-0097 Revision: 0 Operations Requalification Grace Period This revision to Procedure AP 30B-002, " Nuclear Station Operator (NSO)

Requalification Training," provides for the addition of a step which will allow a grace period for completing on-the-job training (OJT) requirements. Procedure AP 30B-002 requires an NSO who has been fully qualified to demonstrate their qualification on a biennial cycle. No grace. period is stated for biennial requalification. This revision will allow a 30 day grace period to complete form APF 30B-002-01, which will complete the requalification process. This revision will affect section 13.2.2.2.4 of the Updated Safety Analysis Report (USAR).

All chapters of the USAR have been reviewed and no accidents were identified that could be potentially impacted by this activity.

Allowing a 30 day grace period for an NSO to complete their requalification can not be related to any credible accidents. Neither are any credible accidents associated with this revision. Technical Specifications require a training program that meets the requirements of ANSI /ANS 3.1-1978 Section 5. This document was reviewed for time requirements on requalifications. ANSI /ANS 3.1-1978 requires that a requalification be performed on a periodic basis.

Base on this evaluation, this revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 149 of 238

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Safety Evaluation: 59 97-0098 ' Revision: O Security Plan Revision 29

,The following changes are being made to the Security Plan by Revision

29. ' Change 1 incorporates a change in the Wolf Creek Nuclear-

' Operating Corporation- (WCNOC) management organization. The.

Superintendent Security now reports'directly to the Chief Operating Officer. Change.2-incorporates the addition of a permanent structure in the vehicle search area.' -Change 3 reflects the upgraded communications system and changes the name of the local telephone provider. Change 4 provides for welding the roof hatch of the Refueling Water Storage Tank valve house being welded shut.

Additional changes incorporated by Revision 29 of the Security Plan

.are evaluated by Unreviewed Safety Question Determination 59 97-0153.

. Details of.this Unresolved Safety Question Dee-rmination are classified as Safeguards 'information and theto< ore are not included in this summary. The details are available for review at the site.

This revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical'

. specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 150 of 238 Safety Evaluation: 59 97-0099 Revisions 0 Issential Service Water Flow Indication USAR Table 7-.5-1 was revised by CCP 07369 to reflect the correct as-built Essential' Service Water (ESW) flow indication scale of 10.6 X 106 lb/hr, and to correct the descriptions of ESW flow and pump

' discharge pressure, which were described incorrectly as Essential

-Cooling Water. Plant design documents reflect the correct value for the. flow indication scale and this change only corrects the USAR to match plant design documents. There are no plant hardware changes associated with this change.

This change will not 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. This change does not create a possibility for an accident or malfunction of i a different type than any evaluated previously in the safety analysis' i report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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Attachment to ET 98-0014 page 151 of 238 Safety Evaluation: 59 97-0101 Revision: 0 Revise USAR to Incorporate Generic Letter 97-02 Guidance This revision to the Updated. Safety Analysis Report (USAR) replaces the current discussion of Regulatory Guide 1.16, " Reporting of operating Information - Appendix A Technical Specifications," in Appendix 3A with the sentence, "The content of the_ monthly operation report is in accordance with Technical Specifications and meets the guidance of Generic Letter 97-02, ' Revised Contents.of the Monthly Operation Report.'" In Section 12.5.3, the third sentence of the last paragraph is revised to state, " Compliance with Regulatory Guide 1.16 is discussed in Appendix 3A, and report content is also discussed in the'WCGS Technical Specification Section 6.9." These administrative changes to these sections of the USAR are being made to allow implementation of monthly operating report format and content changes allowed by the NRC in Generic Letter 97-02. This is an administrative change in reporting requirements only. No plant system, structure, or component is affected by this revision.

Therefore, this revision will not 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. This revision does not create a possibility for an

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accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this revision.

Therefore, this revision does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 152 of 238 I i

Safety Evaluation 59 97-0102 Revision 0 l Recycle Hold Up Tank (RHUT) to RHUT Level Equalization j l

This Unreviewed Safety Question Determination evaluates the use of I Temporary Procedure TMP 97-025, which allows Operations to transfer a j portion of the contents of Recycle Holdup Tank (RHUT) A to RHUT B or '

vice-versa by cross connecting their discharge piping. This transfer will be accomplished by opening each respective tank outlet isolation valve and allowing the level in the tanks to equalize. This configuration conflicts with Updated Safety Analysis Report (USAR)

Figure 9.3-11 and Section 9.3.6.2.1, which reflect that only one of the discharge valves is to be normally open as opposed to having both of the valves open. A precaution / limitation has been added to the procedure, requiring that the total volumetric contents of both tanks shall at no time during the performance of this procedure exceed the full capacity of one tank. In this manner, all flood, tank failures, and system leakage analyses in the USAR for one tank will remain valid.

I Portions of USAR Sections 2.4 and 15.7 were reviewed for potential l impact by this procedure. These USAR Sections and associated analyses remain valid because of the precaution / limitation statement in the procedure requiring the total volume in both tanks to remain less than i the full capacity of one tank. USAR Section 15.7.2.5 discusses the failure of a single RHUT. The tank failure is assumed to occur when I the contents of the tank are at a maximum. The limitation requirement of maintaining the total volume in both tanks to remain less than the 2 full capacity of one tank is bounded by this previous evaluation.

This change does not alter the original single tank failure assumption. However, the fluid volume loss can now come from both tanks due to a single failure.

The RHUT room (dike) is designed to contain any possible leakage from the tank (s) or its associated piping. There is no safety related '

equipment within the vicinity of these two tanks. Therefore, any water from a potential rupture of a tank or piping will not damage any equipment important to safety. Operations will perform periodic inspections of the tank and piping to det ect any possible leakage.

This procedure uses existing plant equipment; therefore, no credible accidents that could be created are identified.

There is no equipment important to safety in the Radwaste Building where the tanks are located. Therefore, no malfunctions of equipment important to safety will be affected by these activities. The RHUT discharge piping and the outlet valves are special scope "D" Augmented. Therefore, failure of these components will not increase the probability of a malfunction of equipment important to safety.

There are no acceptance limits associated with the RHUTs in the Technical Specifications or licensing basis documents. Therefore, no

Attachment to ET 98-0014 Page 153 of 238-

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Lacceptance limits can be affected.

Based on the above discussion, this modification will not 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. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously'in the safety analysis report. The margin of safety as defined _in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

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l l Attachment to ET 98-0014 Page 154 of 238 Safety Evaluation: 59 97-0103 f.evision 0 l ' Replacement of Steam Generator Feedwater Pump A&B Warmup Lines Due to Flow Accelerated Corrosion This modification provides the guidelines for the replacement / modification of the affected pipe sections on Main Feedwater System lines AE-104-ABD-1 and AE-105-ABD-1 with low alloy i- steel (2 1/4 Cr-1 Moly; piping class AAD), which is more resistant to wear caused by Flow Accelerated Corrosion (FAC). Piping cross sectional properties, mechanical properties and geometric '

configuration will remain unchanged.

To reduce the high thermal stresses resultant from dissimilar metal welds currently existent at the affected pipe and orifice interface, the replacement of the stainless steel flow orifices (FO-3 and FO-4) with a low alloy Cro-Moly material is also required. The replacement is a programmatic enhancement.

l The piping replacement will not change the cross sectional properties ]

(section modulus, moment of inertia), or the mechanical properties )

! (yield or tensile strength, and/or code allowable stresses) or the J existing geometric configuration. Therefore, the change does not adversely affect the existing safety margins or structural integrity

-of_the affected piping system. The piping stresses will remain acceptable within code allowables.

l The proposed replacement does not adversely affect any system, j component, or procedures required to mitigate the consequences of an {

accident previously evaluated in the Up dated Safety Analysis Report l (USAR). The modification will restore a degraded section of the

.aff.ected piping system to its original design configuration.

l Ductile fracture, corrc:lon, erosion / corrosion, loss of mechanical l l properties, excess strain, mechanical creep etc., are credible failure

! modes for which the piping replacement has been evaluated, through a l

critical characteristics comparison to the existent piping system design. Essed on the evaluation, it was concluded that a new credible  ;

failure mode is not introduced. Therefore, there are no malfunctions of equipment important to safety.  !

l This modification will restore a degraded section of the affected 1 piping system, to perform its original design intent. The proposed replacement does not involve or affect any safety related system or component. All system functions will continue to be performed as designed.

This modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important

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Attachment to ET 98-0014 Page 155 of 238 I

to safety previously evaluated in the safety analysis report. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

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' Attachment to ET 98-0014 I Page 156 of 238 l l

Safety Evaluation: 59 97-0104 Revision: O l Low Pressure Heater Seal Water Vent Replacement This modification provides the guidelines for the i replacement / modification of the affected pipe sections on Condensate  ;

System line AD-104-HBD-3, Feedwater Heater Drains and Extraction j System lines AF-287-HBD-3, AF-284-HBD-3, AF-284-HBD-1, and AF-285-HBD- )

2 to mitigate abnormal pipe-wall thinning due to Flow Accelerated Corrosion (FAC), with low alloy steel (2 1/4 Cr-1 Moly) , which is more resistant to wear caused by FAC. Piping cross sectional properties, ,

mechanical properties and geometric configuration will remain l unchanged.

The piping replacement will not change the cross sectional properties (secsion~ modulus, moment of inertia) , or the mechanical properties (yield or tensile strength, and/or code allowable stresses) or the existing geometric configuration. Therefore, the change does not  ;

(. adversely affect the existing safety margins or structural integrity of the affected piping system. The piping stresses will remain acceptable within code allowables. The piping replacement does not adversely affect any system, component or procedures required to mitigate the consequences of an accident previously evaluated in the Updated Safety Analysis Report (USAR). The modification will restore 1 a degraded section of the affected piping system to its original design configuration.

l Ductile fracture, corrosion, erosion / corrosion, loss of mechanical properties, excess strain, mechanical creep etc., are credible failure 1

modes for which the piping replacement has been evaluated, through a l- critical characteristics comparison to. the existent piping system design. Based on the evaluation, it was concluded that a new credible failure mode is not introduced. Therefore, there are no malfunctions of equipment important to safety.

This modification will restore a degraded section of the affected l piping system, to perform its original design intent. The proposed replacement does not involve or affect any safety related system or component. All system functions will continue to be performed as designed. This modification will not 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. This modification does not create a possibility for I an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any j unreviewed safety question.

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Attachment to ET 98-0014 Page 157 of 238 Safety Evaluation: 59 97-0105 Revision: 0 Main Turbine Moisture Separator Reheater A Drain Replacement This modification provides the guidelines for the^

replacement / modification of the affected pipe sections _on Main Turbine System line AC-091-GBD-02 to mitigate abnormal pipe-wall thinning due to Flow Accelerated Corrosion (FAC), with low alloy steel (2 1/4 Cr-1

. Moly), which is more resistant to wear caused by FAC. Piping cross sectional properties, mechanical properties and geometric configuration will remain unchanged.

The' piping replacement will not change the cross sectional properties (section modulus, moment of inertia), or the mechanical properties (yield or tensile strengthi and/or code allowable stresses).or the existing geometric configuration. Therefore, the change does not adversely affect the existing safety margins or structural integrity of the affected piping system. The piping stresses will remain acceptable within code allowables.

The piping replacement does not adversely affect any system, component or procedures required to mitigate the consequences of an accident previously evaluated in the Updated Safety Analysis Report ('USAR) .

The modification will restore a degraded section of the affected piping system to its original design configuration.

Ductile fracture, corrosion, erosion / corrosion, loss of mechanical properties, excess strain, mechanical creep etc.., are credible failure modes for which the piping replacement has been evaluated, through a 1 critical characteristics comparison to the existent piping system i design. Based on the evaluation, it was concluded that a new credible failure mode is not introduced. Therefore, there are no malfunctions of equipment important to safety.

This modification will restore a degraded section of the affected piping system, to perform its original design intent. The proposed replacement does not involve or affect any safety related system or component. All system functions will continue to be performed as designed.

This modification will not 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. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the  ;

safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

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c-Attachment to .ET 98-0014 Page 158 of - 238 Safety Evaluation: 59 97-0106 Revision 0 Auxiliary Feedwater Pump Turbine Drain Replacement This modification provides the guidelines for the .

replacement / modification of the affected pipe sections on Auxiliary i Turbine System lines FC-005-DBC-1, FC-006-DBC-1, and FC-075-HBD-1 to mitigate abnormal pipe-wall thinning due to Flow Accelerated Corrosion j (FAC), with low alloy steel (2 1/4 Cr-1 Moly) , which is more resistant

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to wear caused by FAC. Piping cross sectional properties, mechanical

' properties and geometric configuration will remain unchanged.

The piping replacement will not change the cross' sectional properties (section modulus, moment of inertia) , or the mechanical properties (yield or tensile strength, and/or code allowable stresses) or the

. existing geometric configuration. Therefore, the change does not adversely affect the existing safety margins or structural integrity of the affected piping system. The piping stresses will remain 1 acceptable within code allowables.

The piping replacement does not adversely affect any system, component or procedures required to mitigate the consequences of an accident j previously evaluated in the Updated Safety Analysis Report (USAR). l

.The modification will restore a degraded section of the affected  !

piping system to its original design configuration.

Ductile fracture, corrosion, erosion / corrosion, loss of mechanical . j properties,~ excess strain, mechanical creep etc., are credible failure modes for which the piping replacement has been evaluated, through a critical characteristics comparison to the existent piping system design. Based on the evaluation, it was concluded that a new. credible failure mode is not' introduced. Therefore, there are no malfunctions of equipment important to safety.

This modification will restore a degraded section of the affected piping system to its original design intent. The proposed replacement does not involve or affect any safety related system or component.

All system functions will continue to be performed as designed. 1 This modification will not 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. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical j specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 159 of 238 Safety Evaluation: 59 97-0107 Revision: 0 Replacement of Steel Manhole Cover With Plexiglass This modification allows for the replacement.cf the 24" diameter, one fourth inch thick austenitic stainless steel man-way located on top of Recycle Holdup Tank (RHUT) THE02A and B with a plexiglass man-way of

• he same dimensions. This configuration conflicts with Table 9.3-13 of the Updated Safety Analysis Report (USAR) which lists the tank as being austenitic stainless steel. The functions of.the man-way, which are to provide access to the tank internals and foreign material exclusion, have not changed. THE02A and B are vented and the man-way serves no pressure retaining function. No other USAR descriptions or conclusions are affected by this modification.

USAR Section 15.7.2 discusses the failure of a single RHUT. The tank failure is assumed to occur when the contents of the tank are at a maximum. This analysis is based on the volume in the RHUT. Changing the materials on the top man-way has no impact on this failure analysis.

This modification does not change the function of the man-way or the tanks. The RHUT room (dike), ir designed to contain any possible leakage from the. tank (s) or its associated piping. There are no I safety related liquids within the vicinity of the RHUT tanks. 'Any water from a potential . rupture of a tank or piping will not damage any equipment important to safety. Therefore, no credible accidents that could be created have been identified.

I There is no equipment important to safety in the Radwaste Building, where the RNUTs'are located. Therefore, no malfunctions of equipment important to safety.will be affected by the proposed. activities.

This modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equ!pment important ,

to safety previously evaluated in the safety analysis report. This  !

modification does not create a possibility for an accident or j malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 160 of 238 Safety' Evaluation: Sr 97-0108 Revision: 0 New Temporary Procedure TMP 97-028 Temporary Procedure TMP 97-028 establishes post modification test instructions to verify operation, in accordance with design, of Swing Battery Charger NK25 and'its associated transfer switches. Test requirements include:

1) Verification that Transfer Switch NK77 will select either 480 VAC power source, NG01 or'PG19, in accordance with design to power spare charger NK25.

2) Verification that Transfer Switches NK71, NK73 and NK75 will transfer the 125 VDC output from spare charger NK25 to bus NK01 or NK03 and disconnect the output from the corresponding preferred charger NK21 or NK23 from the applicable bus.

The expected effects of the test are that NK77 will connect spare charger NK25 to the selected 480 VAC bus, and Transfer Switches NK71, NK73, and NK75 will transfer the 125VDC output from spare charger NK25 to the selected NK01 or NK03 bus and disconnect the output from the corresponding preferred charger. All interlocks and annunciation are expected to function according to design.

The Swing Charger modification described in Design Change Package 05248 and revisions to Technical Specifications 3/4.8.2.1, 3/4.8.2.2, 3/4.8.3.1 and 3/4.8.3.2 have been previously approved by the NRC.

Performance of this temporary test procedure conforms to the revised USAR sections and will not invalidate the USAR information. None of the design basis accidents discussed or referenced in UAAR Chapters 2,2,6,9, or 15 are impacted by the swing charger ter , y test procedure. There are no credible accidents which can L. reated by this tect procedure.

The post modification temporary test procedure will identify equipment malfunctions associated with the 125 VDC battery chargers and transfer switches. If the class 1E 125 VDC buseo, NK01 and NK03 cannot be powered by the preferred chargers of the swing charger, then the inoperable charger (s) must be restored to operable status within two hours or else the plant must be in at least Hot Standby within the next six hours and in Cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Since none of the accidents previously evaluated in the USAR are affected by the temporary test procedure, the probabilities of those accidents occurring are not affected by the temporary test procedure.

This temporary test procedure has no affect on the radiological consequences of accidents previously evaluated in the USAR.

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Attachment'to ET 98-0014 Page 161 of 238 ,

The temporary test procedure will confirm that the Swing Charger modification is operational and does not malfunction or it will identify the existence of a malfunction. Performance of the test will not cause an equipment malfunction.

,This modifi ation will not' increase the probability of occurrence or the conseque ces of an accident or malfunction of equipment important to safety prev.ously. evaluated in the safety analysis report. This modification d,=s not create a possibility for an accident or

malfunction of a different type than any evaluated previously. in the safety analysis report. The margin of safety as. defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 162 of 238 Safety Evaluation: 59 97-0109 Revision: 0 New Temporary Procedure TMP 97-029 Temporary Procedure TMP 97-029 establishes post modification test instructions to verify operatien, in accordance with design, of Swing Battery Charger NK26 and its associated transfer switches. Test requirements include:

1) Verification that Transfer Switch NK78 will select either 480 VAC power source, NG04 or PG20, in accordance with design to power spare charger NK26.

2) verification that Transfer Switches NK72, NK74 and NK76 will transfer the 125 VDC output from spare charger NK26 to bus NK02 or NK04 and disconnect the output from the corresponding preferred charger NK22 or NK24 from the applicable bus.

The expected effects of the test are that NK78 will connect spare charger NK26 to the selected 480 VAC bus, and Transfer Switches NK72, NK74 and NK76 will transfer the 125VDC output from spare charger NK26 to the selected NK02 or NK04 bus and disconnect the output from the corresponding preferred charger. All interlocks and annunciation are expected to function according to design.

The Swing Charger modification described in Design Change Package 05248 and revisions to Technical Specifications 3/4.8.2.1, 3/4.8.2.2, 3/4.8.3.1 and 3/4.8.3.2 have been previously approved by the LRC.

Performance of this temporary test procedure conforms to the revised USAR sections and will not invalidate the USAR information. None of the design basis accidents discussed or referenced in USAR Chapters 2, 3, 6, 9, or 15 are impacted by the swing charger temporary test procedure. There are no credible accidents which can be created by t'.is test procedure. l The post modification temporary test procedure will identify equipment I malfunctions associated with the 125 VDC battery chargers and transfer switches. If the class 1E 125 VDC buses, NK02 and NK04 cannot be l powered by the preferred chargers of the swing charger, then the  !

inoperable charger (s) must be restored to operable status within two hours or else the plant must be in at least Hot Standby within the next six hours and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Since none of the accidents previously evaluated in the USAR are affected by the temporary test procedure, the probabilities of those accidents occurring are not affected by the temporary test procedure.

This temporary test procedure has no affect on the radiological consequences of accidents previously evaluated in the USAR.

1 Attachment to ET 98-0014 Page 163 of 238 The temporary test procedure will confirm that the Swing charger j modification is operational and does not malfunction or it will J identify the existence of a malfunction. Performance of the test will not cause_an equipment malfunction.

This modification will not 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. This j modification does not create a possibility for an accident or 1 malfunction of a different type than any evaluated previously in the l safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

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i Attachment to ET 98-0014 Page 164 of 238 Safety Evaluation: 59 97-0111 Revision: 0 Heater Drain Tank Dump Valve Replacement This modification provides for a change in the Feedwater Heater Extraction-Drains and Vents System valves AFLV0047A and B, Heater Drain Tank Dump Valves to the Condenser. The original valves are Masonellan Camflex design and are shop:n on Piping and Instrument Diagrams (P&ID) as globe valves. The replacement valves are Fisher V-Ball design and shall be shown as a ball valve on the P& ids. The function of these non-safety related valves, which is to control high level in the Heater Drain Tank, has not changed.

Up dated Safety Analysis Report (USAR) Section 15.1.1, "Feedwater System Malfunctions That Result In A Decrease In Feedwater Temperature," assumes the loss of both heater drain pumps. Changing the Heater Drain Tank high level control valves from globe valves to ball valves does not affect the valves' ability to perform this high level control function. Since the function of the valves has not changed, changing to a ball valve will have no affect on this accident.

There is no equipment important to safety in the Turbine Building where these valves are located. Therefore, no malfunctions of equipment important to safety have been identified by this modification.

There are no acceptance limits associated with the heater drain tank high level control valves in the technical specifications or licensing basis documents. Therefore, no acceptance limits can be affected.

This modification will not 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. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, thin modification does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 165 of 238 Safety Evaluation: 59 97-0113 Revision: 0 Drawing Correction 1 This modification provides for the ' correction of Drawing M-K2GD01 Revision 6, "P&ID Essential Service Water Pumphouse HVAC." This

- drawing incorrectly identifies a Hand Isolation switch for the inlet air damper GDTZ11A as HIS 11A. This isolation switch is Electroswitch l Type 24206B, which.is not an indicating switch (HIS). This revision

'is a drawing correction only. There is no change to the physical plant operation or any safety boundary. The Hand Switch that is.

incorrectly labeled is correctly labeled on supporting drawings.

There are no design basis accidents affected by this administrative drawing correction.

This modification will not 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. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 166 of 238 Safety Evaluation 59 97-0114 Revision: 0 i Temporary Procedure TMP 97-031 This Unresolved Safety Question Determination evalaates the use of Temporary Procedure TMP 97-031. TMP 97-031 will be used to operate one train of the fuel building emergency exhaust system in its emergency exhaust line-up, except the Auxiliary Building isolation dampers GGD019 (train B) and GGD026 (train A) will be placed in their open position. The purpose of this test is to determine if the associated emergency exhaust train can maintain a negative pressure, equal to or greater than one fourth inch water gauge in the fuel building relative to the outside with the dampers open. Past

. surveillance testing has been performed with the dampers closed. 'The

' duration of'this line-up will be approximately one to two hours. The line-up will only need to be maintained long enough to measure the fuel building negative pressure relative to the outside environment.

After the pressure measurement is complete, the system will be

' returned to its normal line-up.

The Fuel Building is designed with'two redundant cmergency exhaust trains. The USAR states that either train is capable of maintaining the fuel building at the required negative pressure. When a Fuel- )

Building Ventilation Isolation Signal (FBVIS) is received, both trains automatically start.

Fuel handling activities or load movement over the Spent Fuel Pool will not be allowed during the performance of this test. Therefore, the conditions for a Fuel Handling Accident (FHA) are not present.

Thus, when the test is being performed on a specified train the conditions that would give a valid FBVIS will not be present.

Updated Safety Analysis Report (USAR) Figure 9.4-2 shows the normal position for dampers GGD019 and GGD026 as being closed. TMP 97-031 will open these dampers. The dampers are normally controlled by their associated MCB hand switches. If a Safety Injection Signal (SIS) is received, these dampers will automatically open for their ECCS function.

Because of the test prerequisites the conditions for a FHA in the Puel Handling Building (FHB) are not present during this test. The design basis FRA in the FHB can not occur during the performance of this procedure because of the test prerequisites. The fuel building emergency exhaust system is not an initiator or contributor to any of the design basis accidents discussed in the USAR.

There are no credible accidents introduced by the performance of this temporary procedure. There will be no fuel handling activities taking place during the test and the control room can place the isolation

Attachment to ET 98-0014 Page 167 of 238 dampers (s) in the normal and/or test positions if desired. Plant i equipment will not be altered in a way that would change the design parameters of the plant.

The failure mode of these dampers is "as is" and conditions governing this mode have not been changed. These dampers would automatically open if an SIS were received. Therefore, for the Emergency Core Cooling System mode of operation, the damper function for an SIS has not been changed. This test will not manipulate the dampers in a manner outside their design capabilities.

Technical Specification 4.9.13.g.2 requires the emergency exhaust system to be capable of maintaining the Fuel Building at a negative pressure relative to the outside. The ability to maintain this acceptance limit will not be affected by the performance of this test.

Technical Specification 3.9.13 states: "With one emergency exhaust ,

system inoperable, fuel movement within the fuel stcrage areas or )

crane operation with loads over the fuel storage are may proceed '

provided the OPERABLE emergency exhaust system is in operation." No fuel handling activities or loads over the SFP will be allowed during this test. Therefore, this limit will not be affected. i l

Performance of this temporary procedure will not 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. This temporary procedure does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this temporary procedure. Therefore, performance of this temporary procedure does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 168 of 238 l

Safety Evaluation: 59 97-0115 Revision 0 l

Evaluation of Performance Improvement Request i

The use of Armaflex closed cell rubber anti-sweat insulation is limited to areas outside of those identified in U pdated Safety Analysis Report (USAR) Section 9.5B, " Power Block Fire Hazards Analysis," unless specific engineering approval is obtained prior to installation. This is because in some areas credit is taken for greater than 20 feet between trains with no intervening combustibles in order to satisy 10 CFR 50 Appendix P. This precludes the 1 installation of Armaflex in these areas. Armaflex should be acceptable in other areas analyzed in USAR Section 9.5B, but the combustible loading must be calculated by Engineering. ,

Armsflex insulation is currently installed in the motor driven auxiliary pump rooms 1325 and 1326 but the appropriate Engineering evaluation and approval have not been performed. This USAR revision is to incorporate the required update for the combustible loading in USAR Section 9.5B based on fire loading calculated in Calculation XX-M- ,

039, Revision 2. The combustible loading for Fire Zones A-13 and A-14  !

will be increased by the inclusion of Armaflex insulation.

None of the design basis accidents identified in the USAR will be impacted by this revision. Based on Table 7-9B of the Fire Protection j Handbook, the combustible loading for the fire zones can be as high a i 80,000 BTU /sq.ft. per hour of fire rating assuming the fire hazard I will follow the standard time temperature curve shown in Figure 7-9A of the Handbook. The amount of combustible being added by the Armaflex insulation is less than 500 BTU /sq.ft., which is well below I this limit. Thus the conclusions of the Fire Hazards Analysis are not  ;

affected.

This revision will not 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. This i revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical l specifications is not reduced by this revision. Therefore, this I revision does not involve any unreviewed safety question.

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Attachment to ET 90-0014 Page 169 of 238 Safety Evaluation: 59 97-0116 Revision: 0 Feedwater Heater Operation This revision to the Updated Safety Analysis Report (USAR) corrects references to plant operations with feedwater heaters out of service.

This correction is to establish consistency and reflect the current operation philosophy in accordance with guidance provided by the Turbine Technical Manual.

USAR Section 3.9(N).1.1 lists feedwater heaters out of service as being a, " normal condition," as defined in Section III of the ASME Code. " Normal conditions," are defined as any condition in the course of startup, operation in the design power range, hot standby and system shutdown, other than set, emergency, faulted or testing conditions. This section describes the process of taking feedwater heaters out of service as follows:

Reducing load for commercial issues on the turbine is conservative with respect to this original assumption. The statement in USAR Section 3.9(N).1.1 regarding taking feedwater heaters out of service will not be changed to coincide with USAR Section 10.4, as the reference in 3.9(N) is considered to be a conservative assumption used for equipment fatigue evaluation.

Since feedwater heaters out of service is considered a, " normal condition," as defined in Section III of the ASME Code, there are no credible accidents that this administrative change could create.

Since this activity still falls within the realm of " normal conditions," as defined above, and it involves non-safety related components, there are no credible malfunctions of equipinent important to safety affected by this change.

There are no acceptance limits associated with isolating feedwater heater strings in the. technical specifications or licensing basis documents. Therefore, no acceptance limits can be affected.

This revision will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to sa!'*y previously evaluated in the safety analysis report. This revisica does not create a possibility for an accident or malfunction of a dif t arent type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this revision oces not involve any unreviewed safety question.

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Attachment to ET-98-0014 Page 170 of 238 Safety Evaluation: 59 97-0117 Revision 0 Technical Support Center and Emergency Operations Facility Area Radiation Montitor Clarificaion This revision to the Updated Safety Analysis Report (USAR) corrects information in USAR Section 12.3.4 1. Area Radiation Monitors (ARMS) are located.thrnughout the plant to alert plant personnel and control Room operators of increasing radiation levels in areas which generally have low radiation fields. The Technical Support Center (TSC) monitor and Emergency Operating Facility (EOF) monitor are included as ARMS but are stand alone units and are not connected to the other ARMS which have alarms and read out modules in the Control Room. The USAR currently does not identify that the TSC and EOF monitors are not connected to the control Room. This revision will clarify that the -)

TSC and EOF monitors have no alarms, reedout module, or remotely l actuated check source in the Control Room.

This revision only affects USAR Section 12.3.4.1. There is no other information in the USAR that this revision would make no longer true j or accurate. This change does not affect any proce cres, activities, administrative controls or sequence of operations or any plant systems, structures or components.

This revision does not affect any design basis accidents described in ,

the USAR. The ARMS are not accident initiators. This change will not

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create any new type of accident.  :

This revision does not alter the functions or failure modes of ARMS of any other system, structure, or component. This revision does not affect any of the credible malfunctions of equipment.important to safety previously evaluated in the USAR nor does it create any new

. malfunctions of equipment importent to safety. i l

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Attachment to ET 98-0014 i

Page 171 of 238 Safety Evaluation: 59 97-0118 Revision: 0 USAR Revision to Correct Inaccuracy USAR Section 10.4.6.2.3 was. revised to correct an incorrect reference l to another USAR section for a description of the wastewater treatment system,-and to correct the system name from wastewater treatment system to. wastewater treatment facility. Section 10.4.6.2.3 was corrected to indicate,that the wastewater treatment facility is shown

-in USAR figures 9.2-24 and 9.2-25, not.USAR Section 10.4'.11. This document change'affects the USAR section only. No plant procedures or

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design documents are affected.

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The nargin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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Safety Evaluation: 59 97-0120 RevisionsO Minimum Cable separation Requirements Revision 4 to Design Change Package 05248 authorizes less than the six inch minimum separation, inside switchboards NK01, NK02, NK03 and )

NK04, between class 1E electrical. cables and the-non-class 1E Maintenance bus tie cable 5NKK01CA, 5NKK01CB, 6NKK02BA and 6NKK02BB where the minimum distance cannot be maintained. In accordance with I the requirements of Regulatory Guide 1.75, USAR Section 8.3.1.4.1.4 is being revised to document the exception and to include the evaluation. Drawings WIP-E-11NK01-01-A and WIP-E-11NK02-01-A are being_ revised to note the authorized deviation from the minimum separation requirements and to list the USAR reference. The swing charger modification has been approved by the NRC under Technical Specification Amendment 104. A USQD was performed because of the electrical cable minimum separation distance issues inside switchboards NK01, NK02, NK03, and NK04.

The Maintenance bus ties permits connection of Class 1E, 125 VDC, buses NK01 to NK03 and NK02 to NK04. Connection of the buses is allowed only during maintenance operations while the Class 1E buses are out of service. With the two NK buses within a load group connected, both Class 1E, 125 VDC batteries within a load group can be charged by a single spare swing charger powered from a nonsafety-related bus. The maintenance bus ties cable are routed Non-Class 1E, because they serve no safety function and do not connect to any Class 1E circuits during plant operation. When the NK buses are in service, the manually operated disconnect switches, at each end of the bus tie i are administratively controlled to be locked in the open position with the fuses removed to prevent connection of the buses. In this configuration, cables SNKK01CA, SNKK01CB, 6NKK02BA and 6NKK02BB are completely isolated with no electrical connection. The isolated Non-Class 1E cables cannot degrade the Class 1E circuits.

Based on the above evaluation, the change does not introduce new failure modes. This change.will not increase the probability of occurrence or the consequence of an accident previously evaluated.

The change does not increase the probability of occurrence or the consequences of a malfunct4on of equipment important to safety than previously evaluated. No potential exists for the creation of a new type of unanalyzed event. No reduction in the margin of safety can result from this change.

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Attachment to 'ET 98-0014 Page 173 of 238 Safety Evaluations 59 97-0121 Revision 0 Description Clarification for Component Cooling water Radioactivity Monitors This revision to the Updated Safety Analysis Report (USAR) provides for changes to Section 11.5.2.2.2.1, " Component Cooling Water Monitors," and USAR Table 11.5-2, " Liquid Process Radioactivity Monitors." This revision clarifies the function.and operating parameters of the radiation monitors EG-RE-09 and EG-RE-10. No hardware changes will be performed, and design documents have not been changed. The installed equipment is performing its design function.

The Component Cooling Water (CCW) radiation monitors measure the radiation levels in the CCW sample flowing through them. The sample flow through the monitors is proportional to the differential pressure across the CCW Heat Exchanger because the monitors are piped parallel to the heat exchangers. The CCW pumps provide the system flow and thereby the motive force for the sample flow. The radiation monitor skids do not contain a sample pump.

As described in USAR Table 11.5-1, the radiation monitors detect gamma radiation. The purpose of the monitors is to indicate the possible j i

existence of radioactive in leakage into the CCW System, and to isolate the surge tank in conjunction with a high alarm.

The monitor output triggers an alert alarm and a high alarm. The radiation monitors on high alarm signal will also isolate the CCW surge tank vent and makeup valves thus allowing time for the operators to find and isolate the source of the radioactive in leakage. This is not a safety related function as the CCW system will function satisfactorily irrespective of the position of the surge tank vent and makeup valves. I The operation or failure of the radiation monitors and/or the surge tank vent valves is not a 10 CFR 100 issue because any contaminated ,

air or water venting or leaking from the CCW system is contained l within the Auxiliary Building.

This USAR revision does not require the taange of any plant activity or hardware. This change is being made to make the USAR more accurate. Th.s change in USAR wording does not reflect a change in system or component function.

This revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety

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Attachment to ET 98-0014 Page 174 of 238 analysis report. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 175 of 238 Safety Evaluation: 59 97-0122 Revision: 0 Revision to USAR to Correctly Reflect Ultimate Heat Sink Monitoring system This revision to the Updated Safety Analysis Report (USAR) provides a correction to Section 2.3.6.8.4. The second paragraph of this section indicates.that a method to monitor horizontal and vertical movements of the Ultimate Heat Sink (UHS) dam, used only before it was submerged, is still in use. The actual monitoring method now used, since submergence of the dam, is described in a later paragraph-within USAR Section 2.5.6.8.4. This USAR revision corrects this discrepancy and is editorial in nature.

This change does not affect any system, structure, or component or administrative control, nor does it~ change the performance of activities important to safe and reliable operation of the plant.

This revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical -

specifications is not reduced by this revision. Therefore, this revision does not in.'olve any unreviewed safety question.

Attachment to ET 98-0014 Page 176 of ' 238 Safety Evaluation: 59 97-0123 Revision: 0 Removal of Support on Chemical ta Volume Control System This modification provides a design change to eliminate two sn'ubbers (Tag No. BG24R510) from Chemical and Volume Control System drain line BG-194-HCD-3/4". These two horizontal snubbers are located in the Containment Building, near the floor (Elevation 2002 feet 5 and 1/2 inches) in a high traffic area. They are located in a position where field workers may step on and damage the component during outage work. Elimination of these snubbers will prevent component damage.

Per new analysis performed under Calculation XX-M-046, these two snubbers, located at Node Point 763, can be eliminated from the piping system. This stress calculation demonstrates that the pipe stresses are within the ASME Code allowable limits for seismic and transient load conditions, without the two snubbers in place. No additional supports-or modification to the existing supports are required.

The Updated Safety Analysis Report (USAR) has no direct reference to these snubbers since the piping system within the boundary of this problem is not a high energy line.

There are no accidents identified in USAR Chapters 2, 3, 6, 9, or 15 that will be impacted by this modification.

The new analysis (calculation XX-M-046) uses as-built 3 site spectra and Code Case N-411. This methodology is within our licensing commitments. Therefore, this modification will not create any credible accidents. Neither will this modification create any credible malfunction to equipment associated with the Chemical and volume Control System because the new analysis was performed which  !

meets all of our commitments. Acceptance limits as identified in the USAR are not affected by this change.

Based on the above discussion, this modification will not 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. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this I modification. Therefore, this modification does not involve any l unreviewed safety question. J l

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Attachment to ET 98-0014 Page 177 of 238 Safety Evaluation: 59 97-0124 ~ Revision: 0 Rvaluation of Trevitest Procedure This Unreviewed Safety Question Determination provides evaluation for the use of Vendor Procedure TT-97002. This procedure _ verifies the set' pressure of various relief valves by lifting them slightly off their seats. The relief valves to be tested by this procedure consist of various secondary plant non-safety relief valves located in the Main Turbine (AC); Condensate ( AD) ; Feedwater (AE); Feedwater Heater j Extraction, Vents and Drains ( AF) ; Steam Seals (CA) ; Auxiliary Steam I (FB); and Plant Heating (GA) systems. All of the valves to be tested are located downstream of the Main Steam Isolation valves and upstream of the Main Feedwater Isolation Valves. This test will not adversely affect any equipment important to safety. ,

This procedure will utilize the Trevitest method, which has been used l

industry-wide on systems during normal plant operation. This test j method does not prevent the relief valve from protecting its system i from an over-pressure event. During the test, a vendor-supplied  ;

hydraulic test unit slightly lifts the valve from its seat and then restored. In the event a valve fails to re-close, the hydraulic unit could be used to re-seat the valve. The amount of steam released through the valve during this test is insignificant. This test method ,!

is considered to be a test not described in the USAR. I Chapter 15 of the USAR was reviewed for potential impact by the test using two conservative assumptions: First. that the test would cause the tested relief valve to fail open; and second, that the test would I prevent-the tested relief valve from opening during a simultaneous over-pressure event. It was concluded that this test would have no impact on the USAR Chapter 15 analyzed events. There are no relevant' ]

acceptance limits specified in the bases for Technical Specification that are impacted by this test.

l Based on this evaluation, this test will not 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. This test does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this test.

Therefore, this test does not involve any unreviewed safety question.

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Attachment to ET 98-0014

.Page 178 of 238 Safety Evaluation 59 97-0125 Revision: 0 Drawing Correction to Reflect As-Built Condition -

This modification provides for incorporation of the as-built condition of the Makeup Demineralizer System (WM)into the Piping and Instrument Diagram. The raw water to chlorinated water retention tank vent valve (1WM0315) and pipe (1WM150A-3/4) were not installed during plant construction. The P&ID will be updated to reflect the as-built condition.

There are no design basis accidents identified or evaluated for the non-safety related WM system. Since the systems functions remain unchanged, no credible accident that could be created have been identified. Since the modification would not affect the system's failure modes of affect equipment important to safety, no credible malfunctions of equipment important to safety are identified.

The WM system is not included in the bases of technical specifications. Therefore, no acceptance limits are identified that

.could be affected.

This modification will not increase the probability of occurrence or the co'nsequences of an accident or malfunction of equipment important to safety.previously evaluated in the safety analysis report. This modification does not create a possibility for.an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 179 of 238 Safety Evaluation: 59 97-0126 Revision: 0 Temporary Installation of Blind Flanges CCP 07458 allows for the temporary replacement of one of the Essential Service Water (ESW) System cross-tie valves with temporary blind flange (s) for the purpose of underground leak testing. This method of testing will be used only when leak testing criteria is not met with the two ESW cross-tie valves in place. The valve corresponding to the train being tested can be removed and blind flanged while the other train is being operated. This temporary under ground leak testing method can be used in Modes 5 and 6. The operating train's pressure integrity is maintained by the closed boundary cross-tie valve. This new configuration has been evaluated and justified in calculation EF-S-028. This stress calculation demonstrates that the pipe stresses are i within the ASME Code allowable limits for seismic and dead load conditions, without one cross-tie valve in place. Additional dead weight support will be used prior to removing a cross-tie valve. The ,

line is not a high energy line and the valves (EFV0093 and EFV0094) l are ASME Class 3 valves.

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 180 of 238 Safety Evaluation: 59 97-0127  : Revision: O. _;

Fire Protection Test Header Valve Removal CCP 07460 is issued to remove twelve. Fire Protection System test header valves from the Fire Protection water suppression system test  ;

header: located at the CWSH, This test header is isolated from the fire suppression system in the normal system lineup, and aligned to the fire protection system during the annual pump. testing to provide a discharge point at which the discharge flow rates can be measured.

The subject valves are 2-1/2 inch; 250 psi hydrant discharge type gate valves attached to the twelve discharge connections on the test header. The test header and valves are shown on USAR Figure 9.5 01. .These valves are leaking and' degraded due to wear and exposure to the weather. These valves will be removed, and the existing pipe cap installed on_the valve discharges will be installed on the test header discharges. Annual fire pump test procedures STN FP-204 and STN FP-209 will be revised to include the installation of quarter turn ball.

type hydrant discharge valves designed for this type of application.

These valves will be stored with the other pump test equipment and thus not subject to degradation from exposure to the weather. USAR Figure 9.5-1-01 will be revised to delete the existing valves.

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 181 of 238 Safety Evaluation: 59 97-0128 Revision: O e

Organization Change for Shift Supervisor A personnel change was made to assign a new individual to the position of Shift Supervisor. This is a personnel change only and has no affect on the plant. organization since the new individual is fully qualified for the position and is a licensed operator. This change only affects the resumes for Shift Supervisors in the USAR by adding a new one. .This change will not affect any equipment, procedures, tests or experiments. The functions of the position will continue to be performed and the new individual meets the ANSI qualifications for the position.

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 182 of 238 Safety Evaluation: 59 97-0129 Revision 0 Replacement of Secondary Liquid Radwaste System Ps.mps CCP 06225 replaced the low total dissolved solids (TDS) collector tank pumps and their associated motors. These pumps are part of the Secondary Liquid Waste System (SLWS). The existing pumps are manufactured by Ingersoll-Rand and are equipped with a General Electric motor. The new pumps are manufactured by Goulds Pumps and i equipped with a motor manufactured by the Emerson Electric Company, U.S. Electrical Motor Division. The existing pumps have had repeated l bearing and seal failures and were replaced to increase  !

component / system reliability.

1 Low TDS wastes are wastes that result from the resin washing,  !

flushing, and sluicing operations that are a part of the condensate  !

demineralizer regeneration process. These wastes flow by gravity from the demineralizer regeneration system to the low TDS transfer tank in the condenser pit of the turbine building. These waste fluids are then pumped by either or both low TDS transfer tank pumps to the two low TDS collector tanks. The two low TDS collector tank pumps are provided for pumping the waste to processing equipment. The form and fit of the new pumps, and some pump parameters (horsepower, required net positive suction head, rated head, etc.) are different than the existing pumps and will require some minor installation modifications (supports, piping, etc.) These differences were evaluated and although the USAR will need to be updated to reflect the new ]

parameters, the differences will not adversely affect the pump's functional requirements. USAR Section 10.4.10.1.1 states that "the SLWS is not a safety-related system, and its failure does not compromise any safety-related system or prevent a safe shutdown of the reactor."

This modification will not 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. This modification does not crt'te a possibility for an accident or malfunction of a differc.it type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 183 of 238 Safety Evaluations- 59 97-0130 Revision 0

-Organization Change A new individual has been assigned to the-position of Manager Licensing'and corrective Action reporting to the Chief Operating Officer. This is a personnel addition only and has no affect on the organization since this person is an experienced individual. This change only affects organization description, and adds the resume for the Manager Licensing and Corrective Action in the USAR and the organization chart (USAR Figure 13.1-1). This is a personnel change only and does not involve any plant hardware modifications ot- plant design changes.

This change will not 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. This change does not create a possibility for an accident or. malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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Attachment to- ET 98-0014 Page 184 of 238-safety Evaluation: '59 97-0131 Revision 0 l Addition of shift supervisor A personnel. change was made to assign a new individual to the position

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of Shift Supervisor,' .This is.a personnel change only and has no affect on the plant organization since the new individualt is fully qualified for the. position and is a licensed operator. This change only affects the resumes for Shif t Supervisors in the USAR by adding a new one. This change will not affect any equipment, procedures, tests.

or experiments. The functions of'the position will continue to'be-

. performed and the new individual meets the ANSI qualifications for the position.

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. .The margin of safety as defined in technical specifications is not reduced by-this change. Therefore, this change does not involve any unreviewed safety question, j

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Attachment to ET 98-0014.

Page 185 of 238 Safety Evaluation: 59 97 0132 Revision: 0 meergency Plan Procedure Change A new numbering scheme was implemented for Emergency Plan Procedures.

Changing existing Emergency Plan Procedure numbers requires the Radiological Emergency Response Plan (RERP) to be revised, as the RERP references these procedures by number. This is an administrative change only to change the procedure numbers. This change will have no effect on the implementation of the RERP. This change does not involve any plant hardware or design changes.

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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Attachment to- ET 98-0014 Page_186 of 238 Safety. Evaluation: 59 97-0133 Revisions.0-Technical specification Bases Change This revision to Technical Specification (TS) Bases.3/4.9.2,

" Instrumentation" adds information concerning OPERABILITY: requirements associated with the source range neue 'n flux monitors during MODE 6.

The following wording would be adde the TS Dasess "When any of the safety-related busses supplying y <er one of the detectors (SE-NI-31 or 32) associated with the Source Range Flux Monitors are taken out of service, the corresponding Source Range Flux Monitor may be considered OPERABLE when its detector is powered from a temporary non-

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safety related source of power, provided the other Source Range

' Monitor is powered from its normal power source." The proposed change to the TS Bases would allow powering a source range neutron flux q

monitor from a non-safety related power scurce (e.g., providing i temporary power to either SE-NI-31 or SE-NI-32 from a non-aafety l related 480 volt power _ source (welding receptacle)).

Powering one of the source range neutron flux detectors from a non-safety related power source is not a safety issue, as a source range channel in Mode 6 provides monitoring capability only. In MODE 6, the reactor trip breakers are normally open (may be closed for short periods for maintenance or testing). During most of MODE 6, the reactor vessel head is. removed and the control rods are not capable of being automatically withdrawn. The source range monitors do not provide a reactor trip function in MODE 6. Powering one of the source range neutron flux detectors from an alternate power source does not affect the source range monitors capability to detect changes in core reactivity.- The normal safety-related power source to the detector is not.necessary, in MODE 6, for the detector to perform its specified function (provided the detector can be powered from an alternate source). Therefore, the OPERABILITY of a source range detector in MODE 6 is not affected by the source of its power supply.

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The normal source of power to the source range neutron detectors comes

.off the 120-VAC vital busses NN01 and NNO2 (one of the Gamma-Metrics wide range neutron flux monitors is powered from NN04). These busses are powered from safety-related 480 VAC busses NG01 and NG02, respectively, either directly through a 480/120 VAC'SOLA transformer or their associated 125 VDC batteries / chargers (NK11/NK21 or j NK12/NK22, respectively) through an inverter, j I

The source range neutron flux monitors function to provide monitoring j capability to detect changes in' core reactivity in MODE 6 is not j credited in any design basis accident for accident mitigation l activities. There are no credible accidents that the proposed i activity could create since the monitors and their alternate power source are not related to the events or conditions associated with I

Attachment to ET 98-0014 Page 187 of 238 accident initiation. USAR Chapters 2,3,6 and 15 have been reviewed.

The proposed change has no impact on the accidents discussed in these chapters.

Applying the WCGS Technical Specification definicion of OPERABLE, a source range neutron flux monitor is considered OPERABLE when it is capable of performing its specified function and all necessary support sya ,er/ it needs to perform that function are capable of performing thesr slated support functions. The only specified function the sour ^<. range neutron flux monitors have in MODE 6 is to provide redundant monitoring capability to detect changes in the reactivity condition of the core. In order to perform this function, the' associated detectors have to be powered, and their normal source of power is from the 120 VAC Vital AC busses (NN01 and NN02). However, in MODE 6, as described above, one safety-related power division may be removed from service as allowed under Technical Specification 3.8.3.2 (also 3.8.2.2 allows one safety-related 125 VDC battery bank and associated charger to be out of service in MODE 6). The action required by both specifications in the event both divisions are lost includea immediately suspending all operations involving core alterations or positive reactivity changes. This is the same action required by Specification 3.9.2 for the loss of one source range neutron flux monitor. Removing one power division under Specification 3.8.3.2 or one 125 VDC train under Specification 3.8.2.2 would affect the normal power supply to the associated source range neutron flux detector, since they supply power to the 120 VAC Vital AC bus powering the source range neutron flux detectors. Under these conditions, power may be supplied to the aftected 120 VAC Vital AC bus from a non-safety related power source. This situation would provide the power needed by the affected source range neutron flux detector in order to perform its specified function while complying with all requirements and allowances of Specifications 3.8.2.2, 3.8.3.2, and 3.9.2. In this particubir situation the af fected source range neutron monitor may be declared OPERABLE when it is poteted from a non-safety related power supply in MODE 6, based on several considerations:

1) In MODE 6 the monitors provide only a monitoring function, no reactor protection or accident mitigation actionc.

2) The remaining detector will continue to be supplied from its normal safety-related power source.

3) In MODE 6 only one onsite power distribution division and one 125 l VDC batter / charger train is required to be OPERABLE; the other onsite power distribution division and 125 VDC battery / charger train is not necessary to meet MODE 6 power requirements.

4) The source range neutron monitors are not credited in any design basis accident for accident mitigatien activities,

5) Surveillance testa specified in Technical Specification 4.9.2 will continue to verify that the affected monitor is functioning properly while being powered from its alternate power supply.

6) The Gamma-Metrics wide range neutron flux monitors would be l

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Page 188 of 238 available to provide backup core flux monitoring capability in the event of a loss of the temporary power supply to the source range neutron detector.

This change will not 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. This change does not create a possibility for an accident or malfunction of )

a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve ,

any unreviewed safety question. )

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Attachment to ET 98-0014 Page 189 of 238 Safety Evaluation 8 59 97-0134 Revision 0 Fuel Transfer Tube Valve This is a change to the Updated Safety Analysis Report (USAR) description of the status of the fuel transfer tube valve during refuelin3 operations. The USAR currently describes the fuel transfer tube valve as being locked open when the refueling pool water level is at the same level as the spent fuel pool prior to initiating fuel movement. The change to the USAR is to delete locking the fuel transfer tube valve open as this activity is not essential for safe handlind of the fuel assembles.

The fuel transfer tube valve has no safety related or special scope design, and its sole purpose is to enable an operator to gain access to the blind flange on the reactor side when the Fuel Transfer Tube is flooded and the Refueling Pool empty. The fuel transfer tube valve must be open to enable an electrical interlock to make up such that traverse of the transfer cart, also a component that has no safety related or special scope design, can begin. The change or deletion of naving the fuel transfer tube valve locked open will make the USAR in agreement with the fuel handling procedures, namely F9P 02-001.

Removing the description of having the fuel transfer tube valve locked open during refueling operations does not violate any requirements contained in the USAR or Technical Specifications. Administrative controls in fuel handling procedures are taken credit for in the USAR for keeping the possibility of a fuel handling accident remote (reference section 15.7.4.5.1.1). The reason for locking the valve open is for non-vital equipment protection, not for any nuclear safety concern. This form of control is not essential for safely handling the fuel Removing the description that the fuel transfer tube valve is locked open is an insignificant change that does not have any affect on the likelihood of a fuel handling accident.

The Fuel Handling accident evaluated in USAR Section 15.7.4 in both Containment and Fuel Handling Buildings was reviewed for potential impact. All other evaluated accidents in USAR Section 15 are seen as l not applicable because of the nature of the USAR change to fuel handling activities during refueling. Reviewed also the loss of Spent Fuel Pool level due to pool gate failures in USAR section 9.1.2.2.

The proposed change to the USAR has no affect on the Fuel Handling accidents or pool gate failure events as evaluated in the USAR. This conclusion is reached based on the fact that these evaluated occurrences did not take any credit for the locked or unlocked st(tus of the fuel transfer tube valve.

I The act of closing the transfer tube valve while a fuel assembly is in transit through the tube is not likely due to the fact that the i I

Attachment to ET 98-0014 Page 190 of 238 transfer cart control console on the fuel building side is manned while fuel is in movement. This console is in close proximity to the transfer tube valve hand whael which is readily visible from the console. It is not likely that a manipulation of this type would occur directly in view of.the fuel team cperators. The traverse of the transfer cart would stop, or not begin if not in motion, as soon as the transfer tube valve open interlock was not made. 'lhis would alert the fuel team operators at the console monitoring the movement of the fuel assemble if they did not see the unlikely inadvertent closing of the transfer tube valve.

It is not credible that inadvertent manual operation of the transfer tube valve, if it did occur, would lead to shearing a fuel assemble in half because the fuel assemble is inside the stainless steel fuel transfer container which would block the possibility of such an accident.

Not locking the valve open only prevents inadvertent closure of the j valve and would have no adverse affects on any equipment important to l safety. The fuel transfer cart is not important to safety. Spent Fuel Pool level indication or cooling pump trip on low level is not l affected by the locked or unlocked status of the transfer tube valve. ,

The fuel storage design features addressed in Technical Specification 5.6 of criticality, drainage, and capacity have not been affected by this USAR change. The USAR change does not affect the fuel enrichment or spacing between assemblies nor drainage of the fuel pool below the 2040' elevation nor increase the storage capacity of the fuel pool as identified by this Technical Specification.

The limit of 23' of water over the irradiated fuel assemblies seated in the storage racks addressed in Technical Specification 3/4.9.11 has not been affected by the proposed USAR change. The water depth requirement in the Technical Specification maintains the assumptions and conditions valid in the USAR for the evaluation of the Fuel Handling accident. The proposed USAR change does not affect these assumptions or conditions.

The position or status of the fuel transfer tube valve has no affect on the probability of a Fuel Handling accident as previously evaluated because it is not directly or indirectly associated with the factors that control the frequency of the occurrence of this ccident. The change does not affect any previously evaluated accidents therefore no change of consequences can occur.

Not locking the fuel transfer tube valve open does not affect the seismic ability of the fuel transfer tube to safely transport fuel through it. No other equipment important to safety is associated with i the proposed change. The malfunction probability of the fuel handling l cranes is not affected by the proposed change. The consequences of j l

Attachment to ET 98-0014 Page 191 of 238 dropping a fuel assembly by a crane has not been altered by not locking the fuel transfer tube open. The weight limitations of loads over spent fuel have also not been changed.

A spent fuel pool drainage accident down to the bottom of the transfer canal gate due to a gross reactor cavity seal ring failure with the fuel transfer tube open has not been created by not locking the fuel transfer valve open. This accident, which has not been previously evaluated in the USAR, will be terminated sooner if the fuel transfer valve is not locked.

This USAR change does not create any new or unique conditions for the transfer tube, spent fuel pool cooling equipment, pool level instruments or structural integrity of the pools themselves. No malfunctions have been created by the proposed change.

Review of the basis for Technical Specifications 3/4.9.11 and 5.6 found no margins affected. Based on the above considerations, this change does not involve an unreviewed safety question.

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Attachment to ET 98-0014 Page 192 of 238 Safety Evaluation 59 97-0135 Revision

0 Cycle Ten' Reload Design This evaluation is for CCP 07015, which covers the core reload design for Wolf Creek Generating Station Cycle 10. The change package involves a physical change to the core loading plan and associated changes to the Updated Safety Analysis Report (USAR). All methodologies used to evaluate the reload design have been reviewed and approved by the-NRC. The complete Safety Evaluation, as documented in the Cycle 10 Reload Safety Evaluation and the Cycle 10 Core Operating Limits Report, demonstrates the acceptability of the Cycle 10 reload design and the introduction of Region 12 fuel assemblies, Demonstrated adherence to applicable standards and acceptance criteria preclude any challenges to equipment important to i safety.

It is demonstrated that the mechanical and hydraulic response of the WCGS Cycle 10 core is not adversely affected by the revised loading pattern or the insertion of the Region 12 fuel; including the 8 Robust Fuel Assemblies (RFA). The hydraulic and mechanical responses of the fuel assemblies with the RFA design have been shown to be not significantly changed from.the non-modified design previously used in reload designs. The functions of the reactor coolant system, reactor internals, and the reactor fuel system are maintained.

The modifications associated with the Cycle 10 reload design do not cause the initiation of any accident nor create any new credible limiting single failure, nor result in any event previously deemed incredible being made credible. The existing separation of control and protection functions are not adversely impacted. In addition, the safety functions of safety related systems and components related to accident mitigation have not been altered. Therefore, the probability ,

of an accident previously evaluated in the USAR will not be increased.

The mechanical and hydraulic characteristics of the Cycle 10 reload fuel, including the RFA design, are not significantly changed from the l current design. The fuel assembly pressure drop is unchanged and is compatible with resident fuel. The modifications do not affect the integrity of the fuel or reactor internals such that their function in the control of radiological consequences is affected. In addition, the' proposed modifications do not affect any fission product barrier. l The proposed modifications do not change, degrade, or prevent the l 1

response of safety related mitigation systems to accident scenarios described in the USAR. Therefore, the consequences of an accident previously evaluated will not be increased.

The proposed modifications do not result in an increased probability  ;

of scenarios previously deemed improbable. The modificatiens do not j

Attachment to ET 98-0014 Page 193 of 238 create any new failure modes for the safety related equipment. The modifications do not result in any original design specification, such as seismic requirements, electrical separation requirements, and environmental qualifications, being altered. The proposed modifications do not result in equipment used in accident mitigation being exposed to an adverse environment. Therefore, the proposed modification will not increase the probability of a malfunction of equipment important to safety previously evaluated in the USAR.

This evaluation concludes that the Cycle 10 reload design will not result in an unreviewed safety question, as defined in 10CFR50.59, since it does not increase the probability of occurrence or the consequences of an accident in the USAR. No mechanism for an accident or malfunction, which has not been previously evaluated, has been identified. Finally, the Cycle 10 reload design does not decrease the margin of safey as identified in the Bases for any Technical Specification.

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Attachment to ET 98-0014 Page 194 of 238 Safety Evaluation: 59 97-0136 Revision 0 Reactor Reassembly in Phase V In section 9.1.4.2.3.1 of the Updated Safety Analysis Report ('USAR),

the Reactor Reassembly in Phase V of the fueling operations is outlined. Items e and g are to be revised to more accurately, and in some cases more explicitly, describe the subject activities.

One of the tnanges to item e is to distinguish that a single Residual Heat Removal (RHR) Pump can be usr.d to drain the Refuel Pool (RFP) to enable Reactor Vessel (RV) head reassembly. The current USAR text states that the RHR Pumps are used. The proposed change will allow flexibility for draining the RFP by using only one RHR Pump if desired. This change is considered more conservative because it allows the other train of RHR to be dedicated to its cooling function. The rate of draining when using only one RHR pump is not affected because this is controlled by throttling the valve in the return to the Reactor Later Storage Tank (RWST). Additional changes to item e include allowing for the use of the standby Spent Fuel Pool cooling Pump or cleanup Pump (s) to be used initially if desired to lower RFP level. These pumps are only capable of lowering RFP level to just above the 2040' elevation. The utilization of these pumps for this evolution adds more flexibility.

In item g, the wording will be changed to reflect that after the RV head is brought close to the Reactor Vessel (RV) flange, the RFP is further drained by Low Pressure Letdown, the Reactor Coolant Drain Tank (RCDT) Pump (s) , or gravity. Currently, item g only describes using both RCDT Pumps taking suction from the piping connection to the Rod Control Cluster (RCC) changing fixture pit to enable this task.

The RCDT pump or pumps are smaller capacity pumps and going to a single RCDT pump will add more flexibility and greater control for this evolution. Additionally, the change adds the ability to lower the RFP level at this point by using Low Pressure Letdown or gravity drain. These methods will accomplish the same effect as the RCDT pump or pumps as they allow for greater control in slowly reducing RFP level.

The above changes to the USAR text have no effect on any previously evaluated design basis accidents and will not create any new credible accident. These changes will have no impact on any nelfunctions of equipment important to safety and the change also has no potential for the creation of a new type of unanalyzed event. No reduction in the margin of safety as defined in the basis for any Technical Specifications will result from these changes.

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Attachment to ET 98-0014 Page 195.of 238 i Safety Evaluation: 59 97-0137 Revision: 0 organisation Changes This change moves the reporting of the Plant Trending & Evaluation (PT&E) group from Performance Improvement & Assessment to Licensing &

Corrective Action. This is an administrative change in reporting. No responsibilities are:being changed or deleted. Therefore, no equipment is being affected, and no accidents are being affected or created.

Based on the above, this change will not 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. This change does.not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. .The margin of safety as

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defined in technical specifications is not reduced by this change.

Therefore, this change does not involve any unreviewed safety question.

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Attachment.to ET 98-0014 Page 196 of 238 Safety Evaluation: 59 97-0137 Revision: 1 Organization Changes Revision 0 of this Unreviewed Reviewed Safety Question Determination (USQD) evaluated an organizational change in which the reporting of the Plant Trending & Evaluation (PT&E) group was moved from Performance Improvement & Assessment to Licensing & Corrective Action.

Revision 1 adds the following organizational changes, m.3de after Refueling Outage IX, to this evaluation:

1. Change in reporting of System Engineering from Operations to Engineering

2. Change in reporting of Materials Engineering from Chief Administrative Officer to Engineering (Support).

3. Change in reporting of Licensing from Engineering (Nuclear Engineering, Safety and Licensing) to Operations (Manager Licensing and Corrective Action)

4. Change in title of Manager Nuclear Engineering, Safety and Licensing to Manager Nuclear Engineering.

5. Create the position of Manager Resource Protection reporting to the Chief Operating Officer and responsible for Security, Emergency Planning, Fire Protection, Environmental and Industrial Hygiene.

6 Change in reporting of Training to directly to the Chief Operating Officer

7. Combine Environmental and Fire Protection under a supervisor rather than a manager. .

8. Change in the reporting of Emergency Planning group to a superintendent

9. Create the position of Supervisor Industrial Hygienist.

10. Create the position of Manager Corporate Services with Corporate Communications, Corporate Deve77pment, Public Info.mation and Low Level Waste Issues reporting to that position.

These are administrative changes that reflect organizational changes in reporting chains, titles, and the creation of new positions. The Updated Safety Analysis Report is revised to reflect changes in resumes presented in Chapter 13. No responsibilities or qualifications are being deleted. No equipment is being affected, and no accidents are being affected or created. Based on the above these changes will not 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. These changes do not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety es defined in technical specifications is not reduced by these changes. Therefore, these changes do not involve any unreviewed safety question.

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Attachment to ET 98 - 0014 -

Page 197 of 238 Safety Evaluation: 59 97-0139 Revision: 0 USAR Changes Associated With Refueling Activities This is an evaluation of proposed' Updated Safety Analysis Report (USAR) changes being made-to clarify where new fuel is unloaded from the off-site carrier and where new fuel inspections are done at Wolf Creek Generating Station (WCGS). The U3AR is being revised to clarify hi

• refueling is done at WCGS by performance of complete core off-loa w and how the Rod Control Cluster (RCC) change fixture is used d' ,Ang refueling at WCGS. The USAR is being revised to clarify the ~i refueling machine safety interlock regarding hoist position and

+ovement between. refueling areas. The revisions are expected to have ao effect on safety of equipment or personnel at WCGS nor on the rafety of the general public.

Th* changes will improve the accuracy of the USAR and will make procedures and controls (FHP 01-001, FHP 02-011, and FHP 03-001) used at WCGS during refueling consistent with regard to current practices.

The accidents reviewed for potential impact include those listed in chapter 9 of dropping of a fuel assembly and Spent Fuel Pool Cooling capabilities for a complete core off-load. Additional impact is examined with regard to inadvertent criticality of the core during refueling activities.

The above proposed USAR changes will not create any credible accidents which have not already been addressed.by the original FSAR, USAR, or revisions already addressed in the USAR.

The above USAR changes do not contain any acceptance limits which are contained in the bases for the technical specifications or license bases documents that could be affected by the above proposed changes  ;

to the USAR.

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Attachment to 'ET 98-0014 Page 198 of 238 safety Evaluation: 59 97-0140 Revisions 0 USAR Changes Associated With the Description of the Spent Fuel Pool Cleanup system These changes to the Updated Safety Analysis Report (USAR) add flexibility to the description of the operation the Fuel Pool Cleanup System in USAR Section 9.1.3.2.3.2. Currently, the text in this section reads that both cleanup pumps and filters are used'to cleanup the water.in the refueling pool after the refueling pool is filled from the Refueling Water Storage Tank (RWST), and also to cleanup the RWST after the refueling pool water is transferred back to the RWST.

The wording changes allow for one pump and_one or both filter operation if desired. The current text description is for achieving the design flow rate for the Cleanup System. The design flow rate enables processing one Spent Fuel Pool or RWST volume in about a day and the volume of the refueling pool in slightly less than a day. The pumps and filters are designed for handling half the system's capacity unlike the demineralizer and strainer that are sized for full system capacity. Reducing the flow rate through one or both filters by running only one pump doubles the time to process the volume to achieve the same quality of water, q i

l The failure modes and effects of the malfunctions of equipment important to safety remain as evaluated in the USAR because the changes made to the USAR are confined to the Fuel Pool Cleanup System, a subsystem of the Fuel Pool Cooling System that when the containment penetrations are excluded, does not contain any equipment important to safety. The failure evaluation for the cleanup pumps due to a seismic event is not affected by the proposed changes. Revi?w of the plant configuration and conditions when the associated USAR changes can take place and the significance of the change, has led to the conclusion that the subject USAR changes will not 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, will not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report, and will not reduce the margin of safety as defined in technical specifications. Therefore, this change does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 199 of 238 Safety Evaluation: 59 97-0141 Revision 0 Testing of the Main Steam Isolation valves This Unreviewed Safety Question Determination provides evaluation for j the use of Temporary Procedure TMP 97-0037. In Mode 4, with Residual {

Heat Removal (RHR) in service and steam flow stopped, the Main Steam I System / MSIV valve fast stroke test is to be performed under temporary procedure TMP 97-0037. This procedure will set up and fast f close the main steam isolation valves one at time using one train j accumulators. The intent of this procedure is to verify that the {

valves will perfore their safety function at the new accumulator low l

pressure alarm limM issued under Plant Modification Request 04300. l This test is not described in the Updated Safety Analysis Report I (USAR).

The train of accumulator to be used, will be depressurized to at or below the new alarm set point. The controls for the valves will be configured to prevent three of the valves from closing, and the remaining valve will be fast stroked close. The tested valve will be reopened, and the test will continue to cycle the remain valves. At no time will there be more than one closed MSIV so the plant cool down j can be maintained and controlled. In Mode 4, with Residual Heat 1 Removal - (RHR) in service (Tavg at or below 350 Deg. F and above 200 Deg. F), the RHR system provides the heat sink for the reactor decay heat. The main steam isolation valves are not required to be operable (Technical Specification 3.7.1.5 and Technical Specification Table 3.3-

3) in this plant condition. Although the steam generators are not required, they will remain available, if necessary, to support decay heat removal with three generators available to the condenser and the atmospheric relief valve available on the steam line with the MSIV closed.

This test involves fast closure of one Main Steam Isolation Valve (MSIV) at a time and is not identified in the USAR. The test described closes all MSIVs at the same time. This test intentionally prevents closure of three of the four MSIVs.

This test will temporarily alter the controls so that one of the system level switches will not perform its complete design function.

This condition will be restored after testing. Chapter 15 of the USAR was reviewed for potential impact by the test and it was concluded that this test would have no impact on the system or components that are capable of initiating these events. Based on this evaluation, no new credible accidents are created. There are no relevant acceptance limits specified in the bases for Technical Specification that are impacted by this test.

This test will not increase the probability of occurrence or the

Attachment to ET 98-0014 Page 200 of 238 consequences of an accident or malfunction of equipment important to l safety previously evaluated in the safety analysis report. This tent i does.not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this test. Therefore, this test does not involve any unreviewed safety question.

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Attachment'to ET 98-0014 Page 201 of 230 Safety Evaluations 59 97-0141 i 4.1r Testing of the Main steam Isolation Valves Revision 1 to Unresolved Safety Question Determination provides the following evaluation not included in Revision O.

In Mode 4, with steam flow through the Main Steam Isolation Valves (MSIV) stopped (steam dump valves closed and all other downstream loads isolated), a Main Steam System / MSIV valve fast stroke test is to be performed under' temporary procedure TMP 97-0037. This procedure will set-up and fast close the main steam isolation valves one at time using one train of accumulators. The intent of this procedure is to verify that the volves will perform their safety _ function at the new accumulator low pressure alarm limit to be implemented this outage under PMR 04300.

The train of accumulators to be used will be de-pressurized to at or below the new alarm setpoint. The controls for the valves will be configured to prevent three of the valves from closing, and the remaining valve will be fast stroked close. The tested valve will be reopened, and the test will continue to cycle the remaining valves.

In plant Mode 4 (Tavg at or below 350 Degrees. F and above 200 Degrees. F), the Residual Heat Removal system and/or the steam generators (though the associated Atmospheric Relief Valves) provides the heat sink for the reactor decay heat. Also, main steam pressure

(< ~ 120 psig) is to lo" to cause a rapid cool down of the RCS. The main steam isolation valves are not required to'be operable'in these plant conditions (Tech. Spec. 3.7.1.5 and Tech. Spec. Table 3.3-3).

Therefore, the testing of MSIVs during Mode 4 is not a safety issue as the MSIVs are not required for any active safety related function during Mode 4.

The main steam isolation valves are not required to be operable in these plant conditions (Tech. Spec. 3.7.1.5 and Tech. Spec. Table 3.3-3). Therefore, the testing of MSIVs during Mode 4 is not a safety issue as the MSIVs are not required for any active safety-related function during Mode 4."

This test will not increase the probability of occurrence or the consequences of.an accident or malfunction of equipment important to sa'aty previously evaluated in the safety analysis report. This test

< se not create a possibility for an accident or malfunction of a C fferent type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this test. Therefore, this test does not involve any unreviewed safety question.

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l Attachment to ET 98-0014 Page 202 of 238 Safety Evaluation: 59 97-0142 Revisionio l

Inservice Inspection Program Update i The change is revising Updated Safety Analysis Report (USAR) Section 5.4.2.4.1 " Compliance with Section XI of the ASME Code" to read as follows- I

" Eddy current examinations of steam generator tubing are performed in accordance with the 1989 Edition of the ASME Code,Section XI, per )

10CFR50. 55a (g) with certain exceptions whenever specific written i relief is granted by the NRC per 10CFR50.55a(g) . l l

Other Class 1 and 2 components of the steam generators are examined in ,

accordance with the Inservice Inspection Program....."  !

Similar changes were made under USAR CR 95-111 to Sections 5.2.4.5, 6.6.1.3, 6.6.3, 6.6.6. However, this Section 5.4.2.4.1 was overlooked. The affect of this USAR change will be to formally document in the USAR WCNOC's commitment to the latest code and/or edition as required by the NRC in 10CFR50.55a (g) .

The only credible accident associated with this change is a Steam Generator Tube Rupture. The enhanced technology for identifying tube "efects will not increase the probability of an accident evaluated in the USAR.

The enhanced technology for identifying tube defects will not increase the radiological consequences of an accident previously evaluated in the USAR.

The Eddy current testing will not affect any credible malfunctions of equipment important to safety. Eddy current testing reduces the probability of a malfunction of a Steam Generator Tube Rupture by identifying degraded tubes.

This is an administrative change and will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this change.

Therefore, this revision does not involve any unreviewed safety question. Based on this evaluation, this change does not involve an unreviewed safety question.

i Attachment to ET 98-0014 1 Page 203 of 238 Safety Evaluation: 59 97-0143 Revision: 0 I Battery Replacement Round Cell Batteries DCP 0$846, Rev. 11, captures additional documentation regarding the ,

Class 1E, AT&T round cell batteries installed during Refuel 8 as i components NK11, NK12, NK13 and NK14. This revision to the DCP does not alter the previously approved design. The battery modification requires that USAR Table 3.11(B)-3 (Sheet 151) be revised to show that the new battery specification, E-050A, is applicable, rather than E-050, the previous battery specification. USAR Section 9.5B.7 must also be revised to reflect changes in combustible loading due to replacement of the previous batteries with the AT&T round cells. DCP i

05846, Rev. 11, is administrative in that it captures vendor documents J which add to but do not change any hardware or design basis.  ;

I Therefore, the DCP revision does not impact the Design Basis Accidents I listed in the USAR. No credible accidents will be created by the capture of vendor documents which support the existing battery design. No malfunctions of equipment important to safety will be affected by the capture of vendor documents which support che existing battery design. Since DCP 05846, Rev. 11, does not alter the previously approved design, the acceptance limits contained in the bases for the technical specifications are unaffected.

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' Attachment to ET 98-0014 Page 204 of 238 Safety Evaluation: 59 97-0144 Revision: 0 Temporary Power.for Distribution Temporary procedure.TMP 97-046 provides instructions for bypassing the nornal power feeds to Distribution Panels NK41, NK51 and NK43 and batteries NK11 and NK13. The temporary configuration will allow the identified Train A equipment to be temporarily powered in a manner which isolates Buses NK01 and NK03.. The procedure is limited for use only when the plant is in Operational Modes 5 or 6 and the opposite Train (Train B) is operable. Operational Modes 5 and 6 require only one train to be operable so the opposite train may be taken out of service for maintenance. The USAR describes the electrical distribution system. That description will not be' accurate when the temporary power configuration is implemented. This temporary configuration is acceptable because no credit is taken for the proper functioning of this equipment while the train is inoperable.

This change will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important te safety previously evaluated in the safety analysis report. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety.as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any.unreviewed safety question,

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Attachment to ET 98-0014 Page 205 of 238 Safety Evaluation: 59 97-0145 Revision: 0 Essential Service Water Underground Piping Leakage Test Procedure STS PE-049C, "A Train ESW Underground Piping Leak Test," is an ASME Section XI inservice inspection piping pressure test. It demonstrates the integrity of the A train ESW, underground supply piping from the pump house to the auxiliary building. The test is required by Technical Specification 4.0.5.

Train A is out of service for this test, and the test is to be performed only during operational Modes 5, 6 or E (defueled). The active ESW' train, B, is used as the test and makeup water source. The amount of makeup water needed to maintain pressure in the A train pipeline will equate to the leakage rate of-that pipeline. The trains are cross connected for testing by a hose between valves EFV0337 (B train) and EFV0336 (A train). EFV0337 and EFV0336 are the fire protection hose station control valves. The hose station in the B pump room will be fitted with a wye, in accordance with AP 10-100, to maintain the hose station operable. Both pump room personnel doors are blocked open, in accordance with AP 10-104, to allow for the paesage_of hoses.

Using the operable Essential Service Water (ESW) train to supply water at operating pressure to the opposite train will not challenge the capacity of that train. The maximum expected makeup flow to initially fill the idle ESW train piping is 150 gpm. This flow rate is typical for a 1 Ma fire protection hose station. The flow rate is restricted by the head loss across the piping and valves and hose. This additional demand on the system does not challenge the operable ESW pump capacity.

In Operational Modes 5, 6 and E, in which this testing is to be performed, emergency makeup water to the auxiliary feed water system is not required. Procedure GEN 00-006, Hot Standby to Cold Shutdown, closes and tags closed the ESW manual isolation valves to the auxiliary feed water pump suctions. This amount of lessened ESW system demand more than offsets the demand presented by the test.

This procedure revision is acceptable within the 10 CFR 50.59 process because the use of water from the operable ESW train does not compromise its function in Operational Modes 5, 6 and E.

This USQD also applies to procedure STS PE-049D which performs the same inservice test on ESW B train.

This change will not 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. This i

Attachment to ET 98-0014 Page 206 of 230 change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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LAttachment to ET 98-0014 Page 207 of 238 Safety Evaluation: 59 97-0146 Revision: 0 Reactor Coolant System Venting The description of the preparation for refueling operation in Updated Safety Analysis Report (USAR) Section 9.1.4.2.3.1 is being revised to more accurately describe the evolution used to drain the Reactor Coolant System (RCS) from solid conditions to the desired vessel level. The process is started with the pressurizer solid and Reactor Coolant System (RCS) pressure less than 30 psig. The reactor vessel head is then vented to the pressurizer. The tygon level hose from the RCS loop 1 cross-over leg to the temporary manifold connection is placed in service. Nitrogen is introduced into the pressurizer via a level tap to help purge hydrogen. The purged effluent gas is directed to the containment shutdown purge system or the containment mini-purge systen. Drain-down begins while monitoring the effluent of the purge system. After level has been decreased a small amount, draindown is stopped to drain the pressurizer safety line loop seals to the Pressurizer Relief Tank PRT and to sample the. pressurizer for combustible levels of hydrogen. If hydrogen levels are too high, purging and sampling continues until acceptable levels are reached.

Nitrogen to the pressurizer is then isolated. ' Draining of the RCS is reestablished and continues until the desired level is reached.

The effect is a safer, more efficient draindown of the RCS. Draining the RCS using the low pressure letdown from the Rasidual Heat Removal (RHR) is generally faster than using the Reactor Coolant Drain Tank.

Applying nitrogen to the pressurizer before the draindown begins helps to remove the hydrogen and minimizes the risk of a hydrogen burn. The hydrogen presents an ignition and explosion hazard.

A sentence is also being added to USAR Section 9.4.6 to state that during RCS draindown activities, hydrogen from the pressurizer can be directed to the containment shutdown purge system or the containment mini-purge system via temporarily installed hoses.

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page'208 of 238 Safety Evaluation .59 97-0147 Revision 0 Breathing Air system operations Procedure SYS-KB-200, " Breathing Air System Operations" is being revised to provide an alternate source of cooling water to the aftercoolers, EKB01A and EKB01B. This change will allow demineralized water from the AN (Demineralized Water) system to be used-for cooling the heat exchangers in the event normal chilled water is unavailable.

Unavailability of the chilled water (GB) system will usually coincide-with a service water system outage. The presence of two chiller units precludes unavailability of chilled water by common cause at any other s time.

The demineralized water storage and transfer system (DWSTS) stores water for use upon demand for makeup within the plant. The DWSTS serves no safety function and has no safety design basis. The breathing air system skids and heat exchangers EKB01A and B are not safety related.

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 209 of 238 Safety Evaluation: 59 97-0149 Revision: 0 l I

Revision to Procedure SYS EJ-110 This revision to procedure SYS EJ-110, "RHR System Fill and Vent Including Initial RCS Fill," allows the operation of Chemical and Volume Control System (CVCS) Charging Pumps (Centrifugal Charging Pumps / Normal Charging Pumps) to flood up the refueling cavity in lieu of the Residual Heat Removal (RHR) Pump (s). The Residual Heat Removal (RHR) System will still be operating normally to recirculate and cool the water in the reactor vessel and refueling cavity, as required by Technical Specification Section 3.9.8.2.

The flow path is only slightly different using a Charging Pump. The original flow path was to take suction from the RWST and deliver to the RCS/ refueling cavity via a RHR pump. The source and point of delivery remain unchanged when using a Charging Pump.

The use of the lower flow rate Charging Pump will help maintain water clarity. With the high flow RHR pumps operating to fill the refueling cavity, water clarity becomes an issue and interferes with fuel movement. The issue of water clarity is of greater importance during 3 this Refueling Outage than in the past based on observations by l Operations / :hemistry personnel of the RCS fluid and on past i experience. It is estimated that if we use the RHR pumps to fill the refueling cavity we will be delayed while waiting for the water to clear sufficiently to allow fuel movement.

The use of a Charging Pump to fill the refueling cavity will not effect any of the design basis accidents in Updated Safety Analysis Report (USAR) Chapters 2, 3, 6, 9, or 15. During refueling, there are ,

only three basic accident scenarios: loss of decay heat removal, a l pipe break that would essentially result in either loss of decay heat I removal or flooding, and a boron dilution accident The possibility of a pipe break as a result of operating a Charging Pump under the these conditions is less likely than during normal operation. Therefore, there are no potential detrimental effects on design bases accidents.

The Boron Dilution Accident as described in Section 15.4.6.2 is affected with respect to wording, but not technically, when we substitute a Charging Pump.

i The kinds of design operating conditions required to fill the l Refueling Cavity are lower than the normal / accident conditions the Charging Pumps are designed to mitigate. The design flow rates from the Residual Heat Removal (RHR) pumps are higher than the design flow rates of the Charging Pumps. Therefore, during the refueling cavity filling operation any hypothetical RHR pump accident scenarios (e.g.,

boron dilution, pipe break, flooding) easily bounds the lower flow Charging Pumps. j l

Attachment to ET'98-0014 Page 210 of 238 There are no acceptance limits in the bases!for the technical specifications that could be affected by operating a Charging Pump versus the RHR pumps to fill the refueling cavity. A Charging Pump will fill the refueling cavity so much more slowly than the RHR pumps that it can not possibly exceed the envelope as defined for the RHR pumps.

The use of the relatively low flow rate Charging Pump in place of the high flow rate RHR pumps.is bound by all considerations given to the RHR pump operation. The-fact that the Charging Pump is such a low flow pump in comparison to the RHR will more probably allow for any concerns to be readily identified and resolved before they become problems. Therefore, the probability of occurrence of an accident previously evaluated in the USAR will not be increased, and may actually be reduced.

With respect to any radiological consequences of any accidents previously evaluated in the USAR, the use of a Charging Pump instead of the RHR pumps will have no'significant affects.

The only affect of this change is to provide more run time for a Charging Pump. With the existing programs of monitoring these pumps and systems, the additional run time is considered insignificant.

Therefore, the proposed change will not increase the probability of occurrence of a malfunction of equipment important to safety as evaluated in the USAR.

The radiological consequences of a malfunction of equipment important to safety as previously evaluated in the USAR will not be affected.

The system is open to the atmosphere and filling the refueling cavity. There will be no gaseous effluents as a result of this change. The Charging Pump flow rate is much lower than the RHR pumps. Therefore, the radiological consequences of a malfunction of a Charging Pump would be much less than that of the RHR.

The use of the Charging Pump in lieu of the RER pumps will not create the possibility of an accident of a different type than any previously evaluated in the USAR. This change will not create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the USAR.

The Technical Specifications do not have minimum flow limits for filling the refueling cavity. The Technical Specifications is concerned with maintaining RHR decay heat removal during refueling.

This change will assure continued uninterrupted decay heat removal throughout the process of filling the refueling cavity. Therefore, the proposed change will not reduce the margin of safety as defined in the basis for any Technical Specifications.

. Attachment to' ET 98-0014 page 211 of 238 Safety Evaluation: 59 97-0150 Revision: 0 Temporary Power Modification Temporary procedure TMP 97 036 provides instructions to transfer B-Train 120 VAC vital buses NN02 and NN04 from their respective normal Class IE power sources,' inverters NN12 and NN14, to a single alternate

. power. source, transformer XNN06. The procedure is limited for use only when the plant is in Operational Modes 5, 6'or E (defueled) and the A Train is operable. The effect will be that both Class 1E buses,.

NNO2 and NN04, will be supplied from a single transformer during execution of the procedure.

The purpose of this activity is.to keep NN02 and NN04 energized while the associated NB bus is de-energized for maintenance. Unreviewed Safety Question Determination (USQD) 97-0133 was initiated to allow the use of non-safety related power to energize XNN06 in Operational Mode 6-during fuel movement. Although it was known at that time that the temporary power arrangement involved connecting NN02 to NN04 by using two keys, it was not specifically. discussed in that USQD. This USOD addresses the use of two keys for that purpose. USAR section 8.3.1.1.5 discusses the key interlock which is normally used to ensure that only one of the busses is connected to the transformer at any time.

This change will not 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. This change does not create a possibility for an accident or malfunction of a different type than any. evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 212 of 238 Safety Evaluation: 59 97-0151 Revisions 0 Temporary Power Modification Temporary procedure TMP 97-045 provides instructions for temporary power to Distribution Panels NK42, NK44 and NK54 and batteries NK12 and NK14. The procedure is limited for use only when the plant is in Modes 5 or 6 and the opposite train is operable. The effect will be ,

that Train B equipment, that is not required to remain energized (but for which continued energization is preferred), can be temporarily powered from a welding receptacle. The USAR describes the electrical

. distribution system. That description will not be accurate when the temporary power is connected. This is acceptable because no credit is taken for the proper functioning of this equipment while the train is inoperable.

None of the design basis accidents discussed or referenced in the Updated Safety Analysis Report (USAR) Chapters 2,.3, 6, 9, or 15 are impacted by this temporary procedure. There are no credible accidents which can be created by the temporary procedure. The equipment is

.part of a non-functioning safety train. It is not required for safe shut down of the reactor, to maintain the reactor in a safe shutdown condition or to prevent or mitigate the consequences of an accident.

Technical specification requirements are not affected to any additional extent beyond those items which are normally identified for a safety related train outage.

This temporary procedure will not increase the probability of occurrence or the consequences of an accident or malfunction of i equipment important to safety previously evaluated in the safety ,

analysis report. This temporary procedure does not create'a  !

I possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of <

safety, as defined in technical specifications, is not reduced by this ;

temporary procedure. Therefore, this temporary procedure does not involve any unreviewed safety question. J l

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Attachment to ET.98-0014 Page 213 of 238 Safety Evaluation 59 97-0152 Revision:' O Snubber Replacement Snubber BG02-R008 was found frozen in place. This is one of the 37 anubbers chosen for functional testing in accordance with the Updated Safety Analysis Report-(USAR). The failed snubber, located on-Chemical and Volume Control System piping, was replaced with an identical new snubber. Thermal movement of piping at this location is less than one sixteenth of an inch in the snubber direction. Based on WRC Bulletin 300, NCIG-05, this small thermal movement will not induce any significant'secesses in the piping because of the frozen snubber.

The snubber passed the' acceleration test and would have behaved like a strut under seismic conditions. Since the snubber would have continued to function, even with the inoperable snubber, there is no safety significance.

There are no accidents identified in the USAR that will be impacted by this evaluation. Analyzed stresses at the affected locations do not significantly change and the thermal movements at the snubber location are insignificant. The snubber of equivalent replacement strut would have functioned under seismic and thermal modes as designed.

Therefore, the performance of the affected system and all other-systems, structures and components are unaffected and no new accidents can be created. No design basis accident were. identified as being i impacted by this change. Therefore, the probability of occurrence of I an accident ins not affected by this change. I The radiological consequences of an accident are unaffectd by this l

change. l This modification will not 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. This modification does not create a possibility for an accident or j malfunction of a different type than any evaluated previously in the  ;

safety analysis report. The margin of safety, as defined in technical I specifications, is not reduced by this modification. Therefore, this modification does'not involve any unreviewed safety question.

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Satety' Evaluation 59 97-0153 Revision 0

. Additional Changes Evaluated for security Plan Revision 29

.The following changes being made to the Security Plan Revision 29 are administrative changes. These changes are in addition to the changes reported by Unreviewed Safety Question Determination 59 97-0098.

.In-Change 1, Chapter 13 was numbered and reflected in the table of contents. In Figure 5.1-4, shading for identification of vital sectors was reinserted for a portion of the fuel building which was inadvertently omitted in the Revision 28 submittal, some walls and' symbols were enhanced for clarity, and alarm point 73 was inserted inside its symbol.

In Change 2, the phrase, "and NUMARC 91-03," which was inadvertently deleted in Revision 28 was reinserted in 1.3.la. These changes are administrative'in. nature and do not change plan requirements.

In Change 3, the test frequency for soldom used hatches and doors is being extended for safety reasons and is shown to be acceptable based

.on. maintenance history for these hatches and-doors.

In Change 4, the number of security personnel required to man the search station at the Secondary Access Facility during peak personnel entrance is reduced. This is based on the low number of personnel who utilize the access facility.

This change does affect the ability to implement the Security Plan response for various events. These changes are administrative in nature. Therefore, this revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined'in technical specifications, is not reduced by this revision.

Therefore, this revision does not involve any unreviewed safety question. i

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Attachment to ET 98-0014 Page 215 of 238 Safety Evaluation: '59 97-0154 Revision: 0 Tesporary Procedure TMP 97-042 Temporary procedure TMP 97-042 provides instructions to transfer buses NN01 and NNO3 from their respective normal Class 1E power sources, inverters NN11 and NN13, to a single alternate power source, transformer XNN05. The procedure is limited to use only when the plant is in Modes 5 or 6 (or when fuel is off-loaded) and B-train is operable. The effect will be that both Class 1E buses, NN01 and NNO3, will be supplied from a single transformer during execution of the procedure. The purpose of this activity in'to keep NN01 and NNO3 energized while the associated 4160 volt NB bus ic de-energized for maintenance.

Updated Safety Analysis Report (USAR) Section 8.3.1 1.5 and Note 4 on USAR Figure 8.3-6-01, Rev. 10, show that the. breakers which connect buses NN01 and NNO3 to transformer XNN05 are key. interlocked to prevent both breakers from being closed at the same time. This is an administrative control which allows only one key to be used, during normal plant operation, to close a breaker. The breaker is designed so that the key cannot be removed while the breaker is closed.

Therefore, no key is available to close the second breaker.

l During implementation of procedure TMP 97-042, a second key will bo

.used to close the second breaker. This will allow both breakers to be closed at the same time, connecting both NN01 and NNO3 to transformer XNN05. Also during implementation of this procedure, other USAR descriptions for the Class 1E V3tal Instrument AC Power Systems, will temporarily be' inaccurate. Specifically, USAR sections 8.1.4.3 and 8.3.1.4 could otherwise be.affected by a permanent modification involving the same. scope. Since this a temporary procedure which will be implemented on a safety train, which is not regoired to be operable (No safety functions required from A Train equipment), the USAR will not be revised. A USAR change will not be drafted.

This temporary procedure will not 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. This temporary procedure does not create a possibility for an accident or malfunction of a different type than any. evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifier? ions, is not reduced by this i temporary procedure. Therefore, this temp.,rary procedure does not  !

involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 216 of 238 Safety Evaluation: 59 97-0155 Revision: 0 Temporary Modification 97-015-FP & Temporary Procedure 97-047 Temporary Modification (TMO) # 97-015-FP will install a non-safety related temporary Jockey fire pump at Fire Hydrant 1FP0138. The temporary Jockey fire pump will replace the non-safety related permanent Jockey pump (1FP02P) during the Service Water ano Auxiliary Power System (SL) Bus outages. During these outages, neither the permanent Jockey pump, nor the Service Water System that supplies its water, will be available. The Jockey fire pump provides the water necessary to keep the Fire Protection System' water solid without the need to activate the Main Fire Pumps. The Jockey pump may be lost without affecting system operability. During the SL-2 Bus outage, the main diesel fire pump will automatically start on a low system pressure signal if the Jockey fire pump fails to maintain system pressure at 125 psi.

The temporary Jockey pump will be powered by a diesel engine and will take suction directly from the cooling Lake. The pump is sized to supply sufficient capacity and pressure to maintain the static system pressure at 125 psi.

TMO 97-015-PP will not create a change to the USAR. However, for a short period of time, the USAR will not be correct as it will not depict the installation of the temporary Jockey fire pump at Fire Hydrant 1FP0138, as noted above.

Neither the temporary nor permanent Jockey fire pump is involved in any of the design basis accidents discussed or referenced in USAR ]

Chapters 2, 3, 6, 9, or 15. The alternate source of power for the i pump will be an independent diesel engine. It cannot have an affect  ;

on any accident scenarios based on the remote location of the  ;

temporary Jockey fire pump. The source of water for the temporary pump is ultimately the same water source as for the permanent pump, i.e. the cooling lake. Therefore, basic configuration is the same.

The only credible incident, not an accident scenario, that may occur as a result of the temporary Jockey pump is the same as that for the permanent Jockey pump. If either the permanent or temporary Jockey pump fails to maintain pressure in the system as required, then the l Diesel Fire Pump will be started and maintain the system pressure, as i required. The temporary Jockey pump will not alter this process in l any way.

i No credible malfunctions of equ.4pment important to safety may be directly or indirectly affected by the proposed activity. The non-safety related temporary Jockey pump will have no more of an affect on any equipment than the existing permanent Jockey pump. The loss of

f Attachment to ET 98-0014 Page 217 of 238 either jockey pump will just result in the start up of the Diesel Fire Pump, as noted above.

There are no Technical Specification or other licensing basis commitments pertaining to the operation of the permanent Jockey Pump.

Therefore, there are no Technical Specification requirements or other licensing basic commitments pertaining to the temporary Jockey pump affected.

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. Attachment to. ET 98-0014 Page 218 of 238 Safety Evaluationt 59 97-0157 Revision: O Changes to the USAR to Delete Unnessary References The Up dated Safety Analysis Report (USAR) is being revised.to delete references to. specific Technical Specification sections and to delete a reference to the radiologically controlled area (:RCA) fence in the discussion of outdoor tanks. .The deletion of specific references will have little or no physical effect'since the tie to the Technical Specifications.will be maintained in general terms. '

The deletion of the reference to the fence will have little or no effect since the' radioactivity limits allowed in the tanks are determined'in accordance' Chapter 16 of the USAR. The fence no longer

, plays a roll in exposure requirements associated with the tanks. The limits defined in Chapter 16 maintain the exposure within the regulatory limits.

These changes are administrative in nature and do not have any hardware or operational' modifications associated to them. These changes have no impact on equipment important to safety. These changes because of their administrative nature, will not affect-accident analysis, equipment malfunctions or margins of safety.

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Attachment to ET 98-0014 Page 219 of 238 Safety Evaluation: 59 97-0158 Revision: O Closed Cooling water system surge Tank Updated Safety Analysis Report (USAR) Section 9.2.8.2.3 describes the Closed Cooling Water System (ClCWS) surge tank as being located above the highest component in the system. This statement is inaccurate.

The exciter air cooler heat exchanger, which is cooled by Closed Cooling Water, is located at a slightly higher elevation. However, the surge tank is located at a high enough elevation to provide ample mc gin for pump suction head requirements. The design function of the system is otherwise not affected by this discrepancy. The closed Cooling Water System is Non-Nuclear Safety Related.

USAR Section 9.2.8.2.3 describes the ClCWS surge tank as being located above the highest component in the system. The exciter air cooler heat exchanger is actually located at a slightly higher elevation in the system. The functional description of the CICWS in Section 9.2.8 in the USAR is otherwise not affected. No other USAR descriptions or

'nclusions would change or be untrue due to this discrepancy.

There are no design basis accidents that are directly impacted by the CICWS. The Closed Cooling Water System is not directly tied to j automatic shutdown of components required to maintain feedwater flow or the turbine / generator on-line. However, loss of closed cooling water may lead to actions to manually trip the turbine to protect critical components such as the isophase conductors. A turbine trip is described in USAR Section 15.2, Decrease in Heat Removal by the Secondary Gystem. However, this USAR discrepancy does not affect l system function or reliability. Therefore the accident analysis in j Section 15.2 is not affected by this change. No other design basis )

accidents were identified in the USAR which could potentially be j impacted by the Closed Cooling Water System.

l Since there are no physical changes and the design basis function of the system is not affected by this change, no new types of accidents not previously analyzed could be created.  ;

Since the proposed change would not affect the system's failure modes, the systems design function, the level of qualification, or equipment important to safety, no credible malfunctions of equipment important to safety are identified. The closed cooling water system is not addressed in the Technical Specifications or USAR Section 16. i

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Attachment to ET 98-0014 Page. 220 of 238 Safety Evaluation: 59 97-0159 Revision: 0' Temporary Modification to Install Ultra Filtration Skid Temporary Modification (TMO) 97-14MB provides for_the installation of '

an Ultra Filtration (UF) System using skid mounted vendor equipment.

.The UF system will be used in series upstream of the vendor supplied Demineralizer Skid. The UF System will be used on a temporary basis to determine if it meets WCGS needs in a cost-effective manner.

It has been identified that certain riterials in the liquid radwaste system cannot be removed using the demineralizers, if they are not in a soluble rather than particulate form, and must be removed physically using filtration. Currently, filtration provided in the Liquid Radwaste System has the ability to remove material 0.45 micron and larger. The UF System is, designed to:

1. Remove all particulate larger than 0.06 microns in size,

2. Increase throughput and decontamination factors of downstream components by removing all organics and foulants with a molecular weight greater than 75,000,

3. Reduce radwaste generation by increasing throughput of downstream media vessels.

The UF Skid processes 10 gpm of waste feed, retaining a small percentage of the waste stream in a waste concentrating water recirculation loop. When the concentrated waste stream or slurry reaches certain limits, the slurry is drawn off by the vendor and transported to the vendor's waste processing facility. Slurries are dewatered (turned into a powder), packaged, and finally shipped to a permanent waste disposal facility.

The UF Skid equipment is designed to meet Regulatory Guide 1.143 requirements, i.e., Special Scope, D-Augmented / pressure boundary assured to reduce the chances of leakage. The equipment will be physically located on the 2031'-6" elevation of the Radwaste Building. Evaluations by Engineering have determined the acceptability of locating the UF Skid as proposed. Evaluations included both floor loading, including hypothesized shielding requirements, and electrical power requirements.

te UF Skid vendor will be responsible for,providing shielding of the

'n Skid and hoses. The shielding.will be subject to routine reviews by WCNOC Health Physics to determine the need for any additional shielding, as needed.

The remaining services / utilities such as air, water, and electricity shall be provided, but are all non-safety related.

6 Attachment to ET 98-0014 Page 221 of 238 Hoses to and from the skid'are being routed through existing pipe penetrations to intercept the flow of fluid that would have been going to the Demineralizer Skid. Once filtered, the fluid will then continue on to the Demineralizer Skid for processing there. After appropriate UF and Demineralizer processing, the ' clean' water that results may either be released to the WCGS Lake or reused internally at WCGS.

Any spills from the hoses or the UF Skid will be naturally directed to the existing floor. drains for collection and processing. Therefore, any liquid spills will be contained within the Radwaste Building that has been designed to contain such possible events. All of the tanks that will feed the UF Skid are vented to the Radwaste Building HVAC, resulting in any gases having already been extracted or vented prior to reaching this skid. Therefore, any gaseous releases will be less than any previously evaluated tank ruptures analyzed in the USAR (boron recycle holdup tank and primary evaporators bottoms tank, reference USAR Section 15.7.2; refueling water storage tank, reference' USAR Section 2.4.13).

During the time the UF Skid is installed and operating, the USAR, rpecifically section 11.2, will not reflect the existence of the UF equipment. The USAR will also not reflect the improvements in effluents expected as a result of the use of the equipment.

The flows to the UF Skid will be less than or equal to the flows currently going to the Demineralizer Skid. Therefore, the.only significant incident would be the possibility of a spill as discussed above. Any spill that might occur as a result of the use of the UF Skid is bounded by the analysis of the rupture of the worst case liquid radwaste storage tank as noted in chapters 2 & 15. This is due to processing only one tank at a time through the UF Skid.

Therefore, a spill would be no worse than the rupture of these tanks already analyzed in the USAR.

The installation of the UF Skid will not create any credible accidents. The only credible incident associated with this skid is ('

the possibility of a spill. However, as also noted, the spill would be naturally contained by the existing design of the Radwaste Building and is bounded by existing analyses.

No equipment important to safety will be directly or indirectly affected by the proposed activity. All of the equipment affected is either Special Scope D-Augmented or non-safety related. The equipment will be located in the Radwaste building in an area that physically

• separates it from possible interaction with any safety related equipment.

None of the equipment is safety related nor is it required for the safe start-up, operation, or shutdown of the WCGS. Therefore, there (L

Attachment to ET 98-0314 Page 222 of . 238 are no acceptance limits contained in the bases for the technical specifications or other licensing basis documents that could be negatively affected by the proposed activity. The use of the UF Skid can, however, have a positive affect on the plants radiological and >

effluent' releases, and the associated release permit commitments. By removing more of the various particulates, foulants, and organic compounds from plant effluents WCNOC is meeting or exceeding the regulatory requirements and thus improving the quality of that effluent.

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Safety Evaluation: 59 97-0160 Revision 0 i

Title Change for the Superindent Operations Support l This Unreviewed Safety Question Determination evaluates a change to an Operations Organization title. The title of Supervisor Operations Support is changed to Superintendent Operations Support. This will l

require changing references to that position in section 13.1 of the ,

Updated Safety Analysis Report (USAR). '

This USAR change does not change any administrative controls which would reduce the level of qualification of personnel, nor does it affect any system structure, or component. This change does not ]

change the performance of activities t,'at are important to the safe  !

and reliable operation of the plant.

l This revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety  !

analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 224 of 238 -

l Safety Evaluation: 59 97-0161 Revision 0  !

Essential Service Water Pipe Space Basement Door Breach Authorization During certain maintenance activities, it is desirable to breach 1974' elevation, Essential Service Water (ESW) pipe space basement door

1. 31011. This door breach authorization is limited to Modes 5 and 6,  ;

or with the core off loaded.. i l

Whenever a Control Room Ventilation Isolation Signal (CRVIS) is  !

received, the Control Room Emergency Ventilation System (CREVS) I operates to maintain the Control Room at a positive pressure three and one fourth inch water column. The CREVS is not designed to provide pressurization air to the ESW pipe space, 1974' elevation. Only the-normal HVAC systems supply these areas, and these systems shut down when a CRVIS is initiated. The normal ducts located in this area are isolated from the Control Room pressure envelcpe by safety related dampers when a CRVIS is initiated.

Door #31011 is part of the RCA boundary and is not to be used as an access into the ESW pipe space (RCA) area without specific controls ,

from the Health Physics (HF) department and Security. When this door ]

is open, the door 11011 accessing the Auxiliary Building and electric chase doors 31051 and 31061 cannot be left open.

For the determination of dose to the control Room personnel, the worst case was analyzed to be the failure of a Control Room filtration j system. A failure of the filtration system would allow a potential l leak pathway directly into the suction side (negative pressure side)  !

of the Control Room A/C recirculation unit. This would allow air from  ;

the control building to enter the control room without first passing through the filtration unit. For this accident scenario (ref. USAR 15A.3), operator action must be taken within 8.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to isolate the system. A single failure of any one of the SR dampers isolating the 1974 elevation would not open a path to the suction side of the Control Room A/C unit.

Door #31011 being open during normal operating conditions would have negligible impact on the rest of the ventilation systems. The ventilation duct in the basement area is already open to the oute:.

pipe chase area. Thus opening the door would have a minimal effect on the' pressure conditions in that area; therefore, the air flowing into and out of this area would not be expected to change. The air flow i througLout the other areas of the building would be maintained.

Door 31011 is seismically qualified with the ability to remain functional following a seismic event This function is not diminished if the door was in its open position.

Attachment to ET 98-0014 Page 225 of 238 Breaching door #31011, with the contingency in place to close this door when a tornado or extreme wind conditions are threatening or following an event that could challenge the control room habitability, will not affect the adequacy to prevent accidents or mitigate the consequences of an accident.

USAR Section 15A.3, Control Room Radiological Consequences calculational Models, identifies a LOCA as the limiting case for radiation doses to a control room operator. In determining the dose to control room personnel, a worst case was ascertained to be the failure of the filtration fan in one of the control Room filtration system trains while in CRVIS lineup. A failure of the filtration system would allow a potential .'.eak pathway directly into the suction side (negative pressure side) of the Control Room A/C recirculation unit. This would allow air from the control building to enter the control room without first passing through the filtration unit. The effects of the loss of air from this potential leak path would be no more limiting than a failure of a pressurization system which would also allow a potential leak path from the control building; therefore, this condition would be less limiting than the previously analyzed failed Control Room filtration system.

Breaching door #31011 will not create a credible accident. This door is a pressure door designed to withstand the forces generated during a tornado or extreme wind conditions. The door will not be open when these conditions occur. When this door is open all breach permit requirements will be enforced. There are no other credible accidents that could occur when this door is open.

If a failure occurs to any one of the safety related dampers isolating the 1974 elevation from the Control Room pressure envelope, an unfiltered flow path to the suction side of the Control Room A/C unit would not be created. The effects of the loss of air from this potential leak path would i s to more limiting than a failure of a pressurization system, why.. would cVso allow a potential leak path from the control building. Therefore, this condi*alon would be less limiting than the previously analyzed failed Control Room filtration system. All locations in the Control Room pressure envelope where these dampers are located are pressurized areas; therefore, the flow of air would be out of the control building.

Tests have been conducted which show the ability to maintain the Control Room pressurization requirements are not affected. Opening door #31011 will not violate the assumptions used in the accident analysis for those accidents that could occur during these modes of operation.

The most limiting accident that would affect dose to control room personnel would be a LOCA. This change is to be implemented only in Modes 5 and 6 or with the core off loaded. Breaching door #31011 in l

Attachment to ET 98-0014 Page 226 of 238 i

these modes will not increase the probability of occurrence of an accident previously evaluated in the USAR since breaching this door will not create a LOCA.

Breaching of door #31011 in Modes 5 and 6 or with the core off loaded t

will not exceed the limits-for the conditions evaluated in Section 15A.3 of the USAR.

Opening the door to this area docs not increase the risk to any safety actuation systems. This door provides protection from the pressures generated during a tornado or extreme winds. The door will be closed during these weather conditions. There is no increase'to.the probability of occurrence of a malfune' ion of equipment important to safety.

Opening the door will not introduce a condition which will change the habitability conditions for the control room. Breaching door #31011 when not in a CRVIS lineup is essentially the same as the plant design since there is an existing duct opening in this same wall that will isolate only during a CRVIS. There is no increase in the radio *:gical consequences of a malfunction of equipment important to safety.

It has been shown the ability to maintain the control rooms required positive pressure has not been affected when the door is opened. This change does not reduce the margin of safety as defined in the Technical Specifications.

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Attachment to ET 98-0014 Page 227 of 238 Safetv Evaluation: 59 97-0162 Revision: 0 USAR Correction to Polar Crane Auxiliary Hoist Use Over Reactor Vessel This revision of the' Updated Safety Analysis Report (USAR) provides for a change to the values indicated in Section 9.1.4.2.2. The polar crane main and auxiliary hooks are administratively controlled by procedures to prevent travel of potentially damaging loads over the reactor vessel when the upper internals have been removed and fuel is in the reactor vessel except for required reactor vessel servicing activities. Section 9.1.4.2.2 of the USAR states that the loads carried over the vessel.are light (less than 300 lb) during Reactor Vessel servicing activities such as irradiation sample removal. The sample and sample tool weigh about 300 pounds. In addition, plant Procedure FHP 02-014 requires a chain fall, load scale and sling to be attached to the auxiliary hoist block / hook. The additional weight of' these items (approximately 175 pounds) make the 300 pound value in the USAR inaccurate. It is necessary to change this value to be make the plant procedure consistent'with the USAR.

The fact that the auxiliary hoist load block / hook is carried over the reactor vessel when the upper internals is removed and fuel is in the reactor and the main hook is within close proximity of the vessel (the two b^ tet are four feet and six inches apart on the trolley) for irradiated sample retrieval was specifically discussed in WCNOC's response to Generic Letter 81-07 and the FSAR prior to final Licensing. These statements of operation were reviewed by the NRC and documented NUREG-0881, Supplement 5.

The probability of an irradiated sample retrieval / fuel accident is very low because of the safety features, administrative controls, and design characteristics of the polar crane as required by NUREG-0612.

This revision will not 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. This j

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revision does not create a yossibility for an accident or malftnction j of a different type than .ny evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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ttachment to ET 98-0014 Page'228 of 238 Safety Evaluation: 59 97-0163 Revision: 0 Missle shield Evaluation for Fire Area A-23 l

This revision to the Updated Safety Analysis Report (USAR) is being  !

. initiated to change the description of the A-23 and T-2 fire areas in USAR Section 9.5B to include a discussion of the non-rated missile shields. ;In the Fire Hazards Analysis, USAR Section 9.5B, the I description of Fire Area A-23 states that " Fire Area A-23 is separated l from all adjoining areas and buildings by 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated fire  !

barriers". Contrary to this statement, at the 2065', elevation there are missile shields installed separating-fire area T.2 from-fire area A-23 which do not meet the requirements of a fire testad, 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> raten fire barrier. Fire delineation drawing A-1807 shows tnat the missile shields are not part of the rated fire barrier. l An evaluation has been conducted and documented in the Unreviewed Safety Question Determination (USQD) for the non-rated missile shields following the guidelines of NRC Generic Letter (GL) 86-10 to establish that the non rated missile shields are commensurate with the fire hazards in the areas'and that the missile shields will not affect the j ability to achieve and maintain fire safe shut down. This USOD ]

contains a GL 86-10 evaluation for a non-rated fire protection feature. ;

i Based on evaluation of this change, the missile shield separating fire areas A-23 and T-2 will provide an equivalent level of protection from the largest expectc3 fire event initiating in fire area T-2 and l

protect safety related equipment in fire area A-23. .The missile )

shield in the current configuration does not affect the ability of the plant to achieve and maintain safe shutdown. This change does not affect the missile protection capability of the missile shield. This ]

change does not create any new failure modes or accidents. There are no acceptance limits applicable to fire barrier functions of the missile shields.

This revision.will not increase the probability of occurrence or the consequences of aa accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 229 of 238 Safety Evaluation: 59 97-0164 Revision 0 j

' Swing Charger Technical Specification Bases Change This revision to Technical Specification (TS) Bases 3/4.8.1, 3/4.8.2, )

and 3/4.8.3 addrssses the issuance of Amendment'104 and the acceptable use of the swing charger to meet the LCO requirements. Specifically, the following is being added to the Bases:

"A D.C. electrical source consists of a battery bank, associated full capacity charger and the D.C. bus. The associated full capacity charger is the charger that provides the normal supply to the bus and battery bank (NK21 for NK11, NK23 for NK13, NK22 for NK12 and NK 24 for NK14). The spare charger can supply either battery bank and bus on its respective train (NK25 for NK11 or NK13.and NK26 for NK12 or NK14). LCO 3.8.2.1 and 3.8.2.2 are met by restoring the inoperable or required full capacity charger to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by either 1) providing the normal supply to its associated battery bank and bus or 2) by supplying the battery bank and bus using the spare l charger (powered from its safety-related bus) in that train. The j spare charger is normal.'.y maintaiaed energized when not in service to allow immediate (< 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) alignment to the battery bank and bus when the associated full capacity charger is inoperable. Surveillance Requirement 4.8.2.la is only performed on those chargers connected to a battery bank and bus."

The proposed change to the TS Bases will incorporate the appropriate language associated with Amendment No. 104. Amendment.104 revises the 125-volt D.C. Sources (3.8.2.1 end 3.8.2.2) TS to include provisions for installed spare chargers. TS 3.8.2.1 and 3.8.2.2 were revised to indicate that spare charger NK25 may be connected in place of charger NK21 or NK23 and spare charger NK26 may be connected in place of charger NK22 or NK24.

This revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 230 of 238 Safety Evaluation: 59 97-0165 Revision 0 Temporary Procedure to Provide for Flushing of Seal Injection Line Installation of design modification DCP 07118 will add a short drain line to the seal injection line for reactor coolant pump PBB01B.

Temporary procedure TMP 97-041 will be used to flush the section of pipe line that was cut and re-welded. This procedure will use Reactor Makeup Water (BL) system for the flush water supply. This work will be done with the plant at half pipe water level. A temporary blind flange will be installed at the flange connection between the seal injection line and the seal housino inlet line for RCP'B'. This will pre ent any BL water or foreign material from entering the RCS. When flushing is complete, the section of line used to flush will be drained of BL water. No BL water or debris will enter the Reactor Coolant System (RCS) due to this temporary flushing procedure.

USAR Section 15.4.6.2 " Dilution During Refueling" states: "An uncontrolled boron dilution transient cannot occur during this mode of j operation. Inadvertent dilution is prevented by administrative i controls which isolate the RCS from the potential source of unborated water. Valves BGV-0178 and BGV-0601 (or 602) in the Chemical and Volume Control Syste, will be locked closed during refueling op' rations. These valves block all flow paths that could allow ur orated makeup water to reach the RCS. Any makeup which is required during refueling will be borated water supplied from the Reactor Water Storage Tank (RWST) by the Residual Heat Removal (RHR) pumps."

USAR Table 9.2-23, Summary of Reactor Makeup Water Requirements, does not list a one time use temporary flush for this seal injection line.

The required flow for this flush is 25 gpm maximum. The pressurizer relief tank is out of service during this flush which requires 150 gpm minimum. Therefore, adequate margin exists with the Steam Generator Blowdown System to do this flush. Lines BB-157-BCB-2" and BB-158-BCA-2" are approximately 190 feet long. Volume of pipe to fill initially is approximately 23 gallons.

1 This temporary procedure will not 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. This temporary procedure does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this temporary procedure. Therefore, this temporary procedure does not involve any unreviewed safety question.

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Page 231 of 238

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Safety Evaluation: 59 97-0166 Revision 0 Radiological Emergency Response Plan Revision Revision 58 to the Radiological Emergency Response Plan (RERP) provides for a revision to the Emergency Action Level (EAL) table for Loss of a Reactor Coolant Boundary to ensure a Notification of Unasual Event (NUE) is declared for certain tube leakage which would constitute a Loss of Coolant Accident (LOCA) or if safety injection is actuated due to pressure loss. This change will ensure more conservative emergency action levels are declared.

This change will not affect'any descriptions in the Updated Safety Analysis Report (USAR). No design bases accidents are impacted by this change because this is a change to che emergency plan and the emergency plan does not have any interaction with plant equipment or I accident analysis. This is a change to the emergency declaration  !

process to plant evolutions already described in the USAR.

l EAL activities do not impact plant equipment. Therefore, there are no credible malfunctions of equipment important to safety affected or

. created. i There are no acceptance limits associated to the EALs documented in the USAR or the Safety Evaluation Reports.

This revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

Attachment to ET 98-0014 Page 232 of 238 Safety Evaluation: 59 97-0167 Revision: O l Startup Operation - Blowdown System This revision to Procedure GEN 00-003 allows Operations the flexibility to start-up with or without heat from the blowdown system l on the steam generator blowdown regenerative heat exchanger. Starting- l up without heat on the blowdown regenerative heat exchanger will delay the requirement of pumping the heater drain pumps forward until later in power ascension. This will allow the heater drains to be routed to the condenser where it has the capability of passing through the Condensate Demineralizer System thus improving cycle chemistry.

Starting up with heat on the regenerative heat exchanger will maximize l blowdown flowrates which may or may not out-weigh the benefits of routing the heater drains to the condenser. Therefore, this change permits Operations to have the option of placing the regenerative heat exchanger in service or not during plant start-ups. USAR Section 10.4.7.2.3, System Operation of the Condensate and Feedwater System states: " . .the condensate is directed to the steam generator blowdown regenerative heat exchanger where it is heated by the liquid discharge from the steam generator blowdown flash tanks. The heated fluid is then pumped..." l 4

J The change to GEN 00-003 allows Operations to place blowdown inservice with or without the regenerative heat exchange - in service.

Condensate will still be routed through the retenerative heat exchanger. The only change is the reference to being heated from the liquid discharge off the flash tank.

Preheating of feedwater via the blowdown system per the regenerative heat exchanger is relatively minor compared to the preheating performed by main steam via the #6 and #7 feedwater heaters. The technical requirements of maintaining feedwater temperature within 250 degrees of the steam generators to minimize thermal stresses on the feedwater piping can be accomplished with or without heat from the regenerative heat exchanger. This 250 degree requirement is procedurally controlled and is not affected by this change to GEN 00-003.

This revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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Attachment t.o ET 98-0014 Page 233 of 238

.Bafety Evaluation: 59 97-0168 Revision: 0 Deletion of Duplicate Information in USAR Table During the final review of Updated Safety Analysis Report Change

-Request (USARCR)97-110, which was evaluated by Unreviewed Safety Questions Determination 59 97-0051, by a discipline group, it was identified that there still existed duplicate information within Table 0.3-4 for several circuit breakers as well as some other equipment.

In addition, the information contained on sheet 23 for NB0117 and NB0217 applied to Callaway (Union Electric) and not to Wolf Creek Generation Station (WCGS). WCGS uses these two circuit breaker cubicles as a storage space for test equiprent. Correcting this table will bring it in line with another drawing in the Updated Safety

. Analysis Report (USAR) and with an approved design drawing.

This is an administrative change to the USAR. This. revision will'not 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. This revision does not create a possibility for an accident or malfunction of a.different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is r.ot reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

Attachment to' ET 98-0014 Page 234 of 238 Safety Evaluation: 59 97-0169 Revision 0 Revision to Radiological Posting Practices Revision 3 of AP 25A-001, " Radiation Protection Manual," will allow personnel outside of the Health Physics Department to establish and remove radiological postings. Revision 3 allows Supplier Materials Quality and Quality control radiography personnel who are appropriately trained, to make radiological postings, which are i reqaired by I-SMQ-008, " Material conformation (X-Met)" or AP 25B-200, j

" Radiography Guidelines". These procedures control activities l

performed with the alloy analyzer and other radiological equipment. j The effect of'the procedure change to allow properly trained individuals to make radiological posting in accordance with approved procedures is to increase the efficiency of Health Physics (HP),

supplier Materials Quality (MS0) and Quality control (OC). Work by SMQ or QC will not be dependent on the availability of a HP technician to make the postings and the HP technician will not be pulled from ,

other plant duties.

This revision will not increase the probability of occurrence or the {

consequences of an accident or malfunction of equipment important to l safety previously evaluated in the safety analysis report. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical l specifications is not reduced by this revision. Therefore, this I revision does not involve any unreviewed safety question.

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l Attachment to ET 98-0014 j Page 235 of 238 Safety Evaluation: 59 97-0171 Revision: O Personnel Assignments for the Positions of Manager Resource Protection and Manager Maintenance ,

4 New personnel assignments have been made for the Manager Maintenance i and the Manager Resource Protection. The resumes' for the individuals filling these positions are included in the Updated Safety Analysis Report (USAR). These individuals are experienced individuals and conform to the recommendations of Regulatory Guide 1.8 and Chapter 13.1 of the USAR. Equipment, procedures, tests or experiments are not  ;

affected by these changes since the changes are assigned to experienced personnel.

This is a personnel change only and will not affect any accidents discussed in the USAR, nor will this personnel change create any new accidents. These personnel assignments will not affect any malfunctions of equipment important to safety. Since these personnel meet the recommendations of Regulatory Guide 1.8, there are no changes to acceptance limits.

This revision will not 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. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.

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Attachment to ET 98-0014 Page 236lof 238 Safety Evaluation: 59 97-0172 Revision 0 USAR. Change to Refueling Water Storage Tank (RMST) Information Updated Safety. Analysis Report (USAR) Section 7.4.1 defines the Wolf.

. Creek Generating Station (WCGS) licensing basis for safe shutdown as

" hot standby". This section includes a listing of system / controls and-monitoring indicators required to maintain a safe hot standby during non-accident condition. The USAR is incorrect, as the RWST level indication is not required for the safe shutdown of the plant during a non-accident condition. This USAR change will delete reference to RWST level in Section 7.4.1 Item b.5. (a) on USAR Page 7.4-4.

This change is consistent with other sections of the USAR. USAR Table 7.5-2 lists the display information'provided for operator use during hot and cold shutdown operations. This table does not list the Refueling Water Storage Tank (RWST) level indicators. USAR Table 3.11(B)-3 which provides listing of all safety related equipment and components required for hot and cold shutdown does not identify that RWST level transmitters'or indicators are required for hot or cold shutdown. The Auxiliary Shutdown Panel, which is used in the event the Control Room is evacuated, also does not have any RWST level indication. The RWST level is not used by the operators in any.

procedure during non-accident conditions for safe shutdown. The deletion of the reference to RWST level does not imply that the borated water in the RWST is not available.

This is an administrative change to USAR only. The RWST level instrumentation is not impacted. No operating procedures are being revised. The design basis functicn of the RWST level is unchanged. {

Therefore, this change has no impact on accidents and malfunctions previously evaluated in the USAR. This change will not create'new l I

type of unanalyzed event and has no impact on any margin of safety.

There is no unreviewed safety question.

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Attachment to ET 98-0014 Page 237 of 238 4

Safety Evaluation:

I 59 97-0173 Revision 0 USAR Change to Waste. Gas Holdup System and Analyzer j This change to the Updated Safety Analysis Report (USAR) incorporates a definition into the USAR that is based on Technical Specification Clarification (TSC) 029-85. This definition addresses what constitutes operation of the Waste Gas Holdup System and when the Waste Gas Analyzers must be in operation in accordance with USAR Section 16.0.2. The definition will read, "The Waste Gas Holdup system is in operation when there is any gas input to the system or when a recombiner is in operation." This includes " degassing operation" which is defined as the operation of switching from a H2 to a N2 blanket (or vice versa) in the VCT while in Modes 3, 4 and 5.

If the Waste Gas Holdup System has no inputs and the recombiners are not running, then the system is not in operation. The operation of the Waste Gas Holdup System on recirculation only, then the system is not considered to be in operation. If the Waste Gas Holdup System is not in operation, per this definition, then the Waste Gas Analyzers do not have to be in operation."

Evaluation of this change has concluded that this revision will not 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. This revision does not create a possibility for an accident or malfunction of a different ,

type than any evaluated previously in the safety analysis report. The margin of safety as defined in technical specifications is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question. l l

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I Attachment to ET 98-0014 Page'238 of 238 Safety Evaluation: 59 97-0177 Revision: 0 Revision t9 Corrective Lense Requirements While Using Self Contained Breathing Apparatus Procedures associated with the wearing Self-Contained Breathing Apparatus (SCBA), and negative pressure respirators are being revised to allow the use of contact' lenses, except the non-gas permeable hard contact lenses, while donning respirators. This change will have no affect on Operators, the fire brigade or health physics re.sponse.

This change is administrative in nature and will implement the current OHSA and ANSI codes. Health Physics Position Paper 162 establishes the bases for acceptability for this change because it provides a reasonable argument and justification for the use of contact lenses with respirators.

This revision will indirectly affect the Up dated Safety Analysis Report (USAR) where Regulatory Guide 8.15 is referenced. Regulatory Guide 8.15 takes " advice" from NUREG 0041, which does not recommend the use of contact lenses with respirators. However, Regulatory Guide 8.15 and NUREG-0041 were issued in the 1970s when contact lenses were not as medically advanced as they are today. This change will have no impact on plant equipment, tests or experiments since there will be no detriment to plant response activities with the use of respirators.

Design Basis Accidents will not'be affected by this change since there will be no affect on operator, fire brigade, or health physics personnel actions. Therefore, there will be no new accident created nor will there be any credible malfunctions of equipment associated with this revision.

This change will not 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. This change does not create a possibility for an accident or malfunction of ,

a different type than any evaluated previously in the safety analysis i report. The margin of safety as defined in technical specifications is not reduced by this change. Therefore, this change does not involve any unreviewed safety question, i

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