ML20080T310
| ML20080T310 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 12/31/1994 |
| From: | Johannes R WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CO-95-0028, CO-95-28, NUDOCS 9503130259 | |
| Download: ML20080T310 (186) | |
Text
{{#Wiki_filter:, e , i W$LF CREEK
. NUCLEAD OPERATING CORNRATION i
Richard N. Johannes chef Administratwo offeer March 10, 1995 l CO 95-0028 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk ' Mail Station P1-137 Washington, D. C. 20555
Subject:
Docket No. 50-482: Wolf Creek Generating Station Annual I Safety Evaluation Report l 1 Gentlemen: Attached is. the Annual Safety Evaluation Report for Wolf Creek Generating Station which is being submitted pursuant to 10 CFR 50.59 (b) (2) . This report covers the period of January 1, 1994, to December 31, 1994. This report summarizes written safety evaluations approved during this period. If you have any questions concerning t.his matter, please contact me at (316) 364-8831 extension 4001 or Mr. Richard D. Flannigan at extension 4500. Very truly yours, IM ; Richard N. Joh es ; RNJ/jra
. Attachment i
!' cc: L. J. Callan (NRC), w/a , D. F. Kirsch (NRC) , w/a ! J. F. Ringwald (NRC) , w/a J. C. Stone (NRC) , w/a 130040 M l
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P.O. Box 411/ Burhngton, KS 66839 / Phone. (316) 364-8831 ' 9503130259 941231 An Equal %nunny Egoyer umCW PDR ADOCK 05000482 R PDR ; a >
Attachment to CO 95-002W Page i WOLF CREEK NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Docket No.: 50-482 Facility Operating License No.. NPF-42 ANNUAL SAFETY EVALUATION REPORT Report No.. 10 Reporting Period: January 1, 1994 through Decembe2- 41, 1794
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G; i i I JAttachment to CO 95 J0028 ; Page ii. , i r ;
SUMMARY
j , 1 [. This report provides m'brief description of changes, tests, and: experiments' il 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 wert. .] reviewed and found to be acceptable.by the Plant Safety' Review Comrrittee for .j the period beginning January l', 1994 and ending December 31,: 1994. 'This i report is submitted in accordance' with the - requirements of 10 CFR 50'.59 (b) (2) . j
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On the basis of these evaluations .of changes, the- following has 'been '! determined: ! e There is no increase in the probability of occurrence or '.ne . consequences I of an accident or malfunction of equipment important to safety previously .) evaluated in the Updated Safety Arnlysis Report ' (USAR) . .!*
- There is no possibility that an accident or malfunction .of equipment
.important to safety of a different type than any evaluated previously 'in -
the USAR may be created. [ i
- The margin of safety as defined in the basis for.-any- Technical Specification is not reduced. j Therefore, all items reported herein were determined not to involve an .l unreviewed safety question. l i
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[.! ' f l Attachment to CO 95-0028 Page 1 of 183 , Safety Evaluations- 59 92-0006 Revisions 0- [ l Limits for Ear.dling New and Irradiated Fuel This. modification supplement revises Section 9.1.4 of the Up dated' ; Safety Analysis Report (USAR) .to reflect'the load setting of the ! zefuel machine and clarifies USAR values. The hoist load limit switch f settings and Note 3 of Drawing M-716-00134-WO6
- Refuel Machine Gripper i 1 Hoist Set," are inconsistent with Technical Specification 3/4.9.6 and
'USAR 9.1.4. This revision changes the USAR, drawing M-716-00134-WO6, '
j i and drawing M-716-00183 "IM Refueling Machine," to be consistent with: , the allowable load. The load cell values have been'successfully used by WCNOC during all ! refueling outages and found to be satisfactory. Set points are adjusted prior to fuel movement and will trip equipment if exceeded. This modification will have no impact on accidents or malfunctions , evaluated as the licensing basis and there is no potential for the i creation of a new type 9pf unanalyzed event. There is no reduction in [ the margin of safety. j i 6 r 8 i l t i
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safety Evaluations. 59.92-0153 Revision 0' Revision of USAR to Provide Consistency Associated With Operator' ,[ L ' Actions { nThis revision to the Updated Safety Analysis Report (USAR) revises the' :l statement in'Section 10.4.9.2.3 which states that the Auxiliary l Feedwater - (AFW) flow can be. isolated to the affected loop within'10 i minutes and refers to Chapter 15 of the USAR for information regarding ; operator actions and time constraints. As described in Chapter:15 for- l the Main Steam Line Break (MSLB), indication is expected to be .l received within one minute, and AFW flow is expected to be terminated. i within 10 minutes of the event. However, other events described in t the USAR either require no AFW isolation or.have other time { constraints for-isolation. For the MSLB scenario, isolation is ; achievable within the specified 10 minute period, but this limitation . should not be construed to be applicable to all events. [ This statement is being revised because other events are postulated where it will be difficult to meet'this requirement of isolating AFW f I flow within 10 minutes. For other analyzed events, the time requirements are less stringent or isolation of AFW flow is not l required to show adequate protection and accident mitigation.
- Revision of this statement has no affect on plant operations,' safety,. ,
or existing analyses. , This revision does not modify any plant equipment, procedure, or operating / naintenance methods which can' affect the occurrence of an l accident. Time' requirements for secondary side breaks are specified. . elsewhere in.the USAR where specific. events are discussed. l Termination of AFW flow to the affected loop is a conservative action ! which will be accomplished, as necessary, ir accordance with operating '! procedures. This revision does not affect equipment important to safety. This revision will have no impact on accidents or- J 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. l l < i
}El 1 f I q Attachment'.to- 'CO.'95-0028 Page 3'_of 183 p Safety Evaluation: 59.92-0216 Revision 0
- Component Cooling Water Temperature Evaluation This modification allows operation of the component. Cooling Water (CCW) System at temperatures below 60 degrees Fahrenhe following a Safety Injection Signal (SIS).
The existing design basis minimum CCW temperature l's 60 degrees-
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Fahrenheit. This temperature is maintained by the CCW heat. exchanger (HX) and regulating valves (EGTV29/EGTV30) on bypass lines during. normal plant operation. During the cold season, the Essential Service Water (ESW) flowrate is reduced through the CCW HX by manually throttling valves on the ESW System. Upon initiation of a SIS all valves are positioned in post-Design' Basis Accident (DBA) configuration such that the minimum temperature cannot be maintained. This modification documents the engineerdng evaluation performed to demonstrate that the CCW System and the associated safety-related components are capable of performing their design basis safety-related function post-DBA, assuming a lake water temperature of 32 degrees Fahrenheit. This modification will have no impact on accidents and 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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Page 4 of 183 Safety Evaluation: 59 93-0148 Revision 0 ; service Water / Essential Service Water Radiation Monitor Removal j dais modification to' safety-related equipment permanently removes j radiation monitors (EART004A, EART004B, EFRT0035, and EFRT0036)'and l associated sampling valves and piping. The radiation monitors are ! being removed from service because of the redundant purpose of the l monitors, problems associated with keeping the monitors in service, i and the inability of the monitors to detect leakage.from'a potentially l radioactive source. Isolation valves EFHV0087 (ESW Train A Rad I
' Monitor Isolation Valve) and EFHV0088 (ESW Train B Radiation Monitor i Isolation Valve) are being removed because they are no longer needed. ,
t t The function of these radiation monitors is to detect inleakage from -! I' potentially radioactive systems and components served by the Service : Water (SW) and Essential Service Water (ESW) systems. Each fixed volume detector assembly monitors the return line of the associated l loop downstream of the return from the system. These radiation ! monitors serve no isolation function. There are no safety design bases associated with the SW or ESW radiation monitors. . I This modification reduces the number of safety-related components. f Modifications are in accordance with original design specAfications ; and a failure of the modified components will not prevent other safety-related equipment from performing safety-related functions. With the Li number of components reduced, the possibility to create a different ! type of malfunction of equipment important to safety is reduced. L{ There are no technical specifications for the SW or ESW radiation ! monitors and their removal _will not reduce the margin of safety or alter any technical specification. { t 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. ! I i t I v i f i i i t h I l l l
i Attachment to CO 95-0028 Page 5 of 183 Safety Evaluation 59 93-0157 Revision 0 Revision to Procedure Process and Numbering for Wolf Creek Generating Station Procedures This revision to the Updated Safety Analysis Report ('USAR) revises references to the Wolf Creek Generating Station Procedures Manual and the Wolf Creek Nuclear Operation Corporation Procedures Manual. These documents are renamed as Implementing Procedures. Program descriptions are added to the procedure hierarchy. The new procedure hierarchy is based on programs instead of organizations. This revision will reorganize procedures and programs making it easier to locate procedural requirements. This revision does not alter the content of existing procedures. This revision will have m impact on accidents or malfunctions evaluated as the licens' , basis and there is no potential for the creation of a new type of unanalyzed event. There is no reduction in the margin of satcty. l i l l
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Attachment to CO 95-0028 r Page 6 of 183 Safety Evaluation: 59 93-0162 Revision 0 Boron Injection System Modification # This modification to safety-related equipment permanently removes the . Boron Injection System (BIS). The BIS is part of the High Pressure Coolant Injection System and is no longer required for plant operation, nor is it required for safe shutdown of the plant during a Design Basis Accident. The equipment is not being used and has previously been removed from service. Analysis has shown that . adequate shutdown reactivity could be maintained with a 2000 ppm boron concentration in the Refueling Water Storage Tank (RWST) . As described in NUREG-0881, Supplement No. 5, " Safety Evaluation Report, WCGS" the NRC has concurred that there is no significant reduction in the margin of safety with the deletion of the BIS. As described above, the BIS no longer provides a safety-related function. However, portions of the BIS continue to maintain the existing High Pressure Coolant Injection System pressure boundaries. Therefore, unnecessary and costly maintenance activities and in- ; service inspections are still required on these portions of the BIS to ensure the High Pressure Coolant Injection System will perform the necessary safety-related functions. By permanently isolating the BIS i from the High Pressure Coolant Injection System these maintenance activities will no longer be required. Thia 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. l P h i 5 l i i i l i
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Safety Evaluation: 59 93-0190 . Revisions 0 Design and Configuration Change Process Revision
'The proposed revision to.the Updated Safety Analysis Report (USAR)- l f changes the' description of the Design / Configuration change process j into more generic terms. This will allow future enhancements to be j made.without affecting.the USAR Chapter 17 information. The new l definitions'for design change and' configuration change were based on work that is currently being performed by EPRI'on_ optimizing the [
engineering change process. The. review process by the. Plant Safety I Review Committee (PSRC) and the Nuclear Safety Review Committee (NSRC) I for design configuration changes that have Unreviewed Safety Question .;
-Determinations (USQDs) and Technical Specification changesLis ;
clarified by this revision. Chapter 13 is also being revised to be consistent with Chapter 17. ! i This revision.will have no impact on accidents or malfunctions j 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 'f the margin of safety. j I
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W 7 Attachment to- CO 95-0028 Page 8 of- 183
- Safety Evaluation .59 93-0196 Revision:0 o Battery Time Duration Modification I
This modification is being implemented to meet NUMARC 87-00,
" Guidelines and Technical Bases for NUMARC Initiatives. Addressing-Station Blackout at Light Water Reactors," guidance concerning battery duration for safety-related batteries NK11, NK12, NK13,'and NK14.
[: - Battery time duration.is being changed from 200 minutes.to 240 minutes. Calculations NK-E-001 "Clars 1E DC Voltage Drop,".and NK-E-002 " Class.1E Battery Sizing," were' issued to reflect this change. This modification revises the design drawings to reflect the extended duration. The design of the existing Wolf Creek Generating Station batteries exceed the requirements of NUMARC 87-00. The capability and capacity of the safety-related batteries are well within the requirements needed to meet the. coping assessments for Wolf Creek Generating Station. The increase of time duration t.o meet station blackout requirements.is within the scope of the' battery design requirements. 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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i Attachment to CO 95-0028 Page 9 of 183 Safety Evaluationt 59 93-0201 Revision 0 Updated Hydrogen Generation Analysis This modification is a revision to the post Loss of coolant Accident (LOCA) hydrogen generation analysis. This modification updates the , hydrogen generation analysis using amounts of aluminum, zinc, and zircaloy greater than the present amounts, and using the revised post LOCA containment temperature profile. .This is being done in order to quantify the margin between the present amounts of these materials and-amounts that would produce higher but still acceptable post LOCA hydrogen concentrations. This is desirable so that when modifications to the plant install these materials a complex engineering calculation will not need to be performed for each addition. This modification involves issuing new engineering calculations which supersede the original calculations, revising the Updated Safety Analysis Report (USAR), and drawing changes. As a result of the new analysis, valve GLV0148 " Containment Hydrogen Control System to Fuel Building Exhaust Isolation valve" must now be locked in a throttled position. This requires a one time test to determine the throttle position which will result in a hydrogen purge flow rate of 100 cfm with a 4 psig containment pressure. GLV0148 is a manual globe valve installed between the hydrogen purge line piping and the hydrogen purge line duct. The design purpose of valve GLV0148 is to control the purge flow rate so that the hydrogen purge line duct is not over pressurized. The results of the updated analysis indicate the increased hydrogen generation rate that could result from this change is within the capability of the Hydrogen Control System. The concentration of hydrogen in the post LOCA containment atmosphere would still remain below the required limits. This modification involves a slight increase in the calculated post LOCA containment atmosphere hydrogen concentration (from 2.1% to 2.8%). This concentration is of no physical consequence. Therefore, a hydrogen burn would not take place and containment integrity would still be maintained. 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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Safety Evaluatior. 59 93-0204 ' Revision 30 i Diaphragm Valve Replacement on'the Refueling Pool Drain Line' !
.t This modification to safety-related: equipment provides for replacement i of: valve-ECV7129_(Refueling-Pool. Drain to Reactor Coolant Drain Tank- -l Pump)" with a plug valve. In addition, a strainer is being added to !
the inlet of the refueling pool drain.line. The open port design of f
'the plug valve will provide higher' flow rates during drain down-of_the ;
refueling pool, which will minimize radioactive sludge. buildup in'the- l line. The addition of the drain inlet' strainer will prevent relatively large particles from entering and plugging the drain line.
-Valve ECV7129 is manually operated and has a safety-related passive function while in the closed position, q The replacement plug valve will be designed, purchased and installed [
to the requirements of the ASME Code Section III, Class 3; the same as I the existing diaphragm valve. The replacement valve is slightly ! heavier than the existing valve. However, the resulting stresses.in j the drain line 069-HCC-4" are within code allowables. The probability j of the replacement plug valve and the drain line 069-HCC-4" to fail is i no greater than the existing installation. The non safety-related: ; drain inlet strainer is constructed of 18 gauge stainless steel and will fit into the drain line with the flange end supported by the pool , linerLplate. The credible failure mode of the strainer would be' I blockage by debris. This is improbable because of the purity of the .; water in the refueling pool. If blockage should occur,-no safety. feature or function would be affected because the refueling pool- _, pressure boundary would remain intact-. Therefore,.this change will- :i, not increase the probability of occurrence'of an accident'previously , evaluated by the Updated Safety Analysis Report. ; This modification will have no. impact on accidents or malfunctions i evaluated as the licensing basis and there is no potential for the t creation of a new type of unanalyzed event. There is no reduction in the margin of safety. i i
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Attachment to CO 95-0028 Page 11 of 183 Safety Evaluation: 59 93-0205 Revision 0 Modification to Fire Protection Barriers for Auxiliary Feedwater Pump Supply valve This modification to safety-related equipment removes the three-hour fire barrier enclosure protection around valve ALHV0032 (Essential Service Water Supply Valve to the Motor-Driven Auxiliary Feedwater Pump). This modification provides for separation of cables, equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards. Fire detection and automatic firc suppression is also provided for this fire area. The installation of a wet pipe sprinkler system and 20 foot combustible-free separation zone in rooms 1206/1207, in conjunction with the detection provisions and at least a 1-hour fire wrap on the circuits associated with valve ALHV0032, will meet the requirements of 10 CFR 50 Appendix R for safe shutdown capability. The existing design is enhanced. 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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. Safety Evaluations :59 93-0209 Revision 0
- Installation of Radweste' Storage Building Power' Supply-This modification provides for installation of a_ power _ supply to the ;
new maintenance storage / shop structure adjoining the south face of.the-
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radwaste storage building. The scope of this design includes only the :
' installation of electrical service ~to the new structure. This d modification considers all electrical loads to be included in the
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building. ; This modification adds approximately 330 feet of electrical cable in j the Radwaste Building. Because some of this cable is'added.in an open. .{ ? top cable tray, a small increase in combustible loading is created by - this modification. However this increase is not significant and is j enveloped by the existing fire hazards analysis. 4 This modification will have no impact on accidents or malfunctions l 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 l the margin of safety. ! i , k i i,
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Page 13'of 183 :l
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Safety Evaluations 59 93-0214 Revision 0 -i' h,
' Drawing Update for the Water Treatment System This' modification updates Piping & Instrumentation Drawing.(P&ID) M-:
0'025, " Makeup Demineralizer. System" to reflect.the as-built , configuration'of the non safety-related' Water Treatment System. This' A modification assigns a component identification number for the level , instrumentation of the Cltarwell Tank. Six plug valves are added to , '} the P&ID to reflect the as-built condition. The component : identification number for.the Coagulant Aid Tank level indicator was- ' changed. The valve symbol for' valve IWM572 was changed from a globe t
.to a plug valve. These changes are administrative in nature. No components, systems, or' structures are affected by this modification. <
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Attachment to CO 95-0028 Page 14 of 183
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t Safety Evaluations 59 93-0217 Revision 0 Addition'of Main steam Tunnel supply Air Unit cooling Coil This modification adds a cooling coil to'the main steam. tunnel supply air unit (SGF01). This supply air unit is a non safety-related fan ' coil unit located on the 2047 level of the Auxiliary. Building. SGF01 was supplied with a coil space that allows addition of the cooling coil without external sheet metal modifications of the unit. The effect of this modification is a reduction in the peak l temperatures in the main steam tunnel. This modification will provide more stable temperatures in the main steam tunnel. . Implementation of-this modification will result in less adjustments to Main Steam - Isolation Valve accumulator pressures and the level of maintenance ~ , required will be. reduced. By lowering the summer time temperatures in the main steam tunnel,' equipment reliability will be improved. This ' modification will enhance equipment performance. l 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. a I f o 1 l
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r safAty Evaluations- 59 93-0218 Revision 0 e
' Replacement / Upgrade of Control Room Fire Protection Panel and Fire Detection system- j This modification upgrades the non safety-related WCGS Fire Detection l
"- System to minimize nuisance or spurious alarms,'to reduce maintenance.
. activities, and to enhance Control-Room human factors. This upgrade- ;
will accomplish these objectives by implementing the changes-identified below- - 1 1 Control Room Fire' Protection Panel (KC008) will be' modified to ; inccrporate a touch screen color graphics unit. This unit will ; provide pictorial layouts of monitored areas of'the site with identifying icons for each type of initiating device. A central i
, I control Network Display Unit will be installed which will' log events in a historic file and will display the current' network status on an LCD readout. The Network Display Unit would be available should the ,
color graphics screen fail. To accommodate these new installations, . much of the existing KC008 equipment will be removed. i Upgrades compatible with the above described modifications will be installed in the Security Building Monitoring Unit (KC102), and I&C Shop Monitoring Un it (KC101). Additionally, this modification will ;
. link the fire detection systems presently installed in the Dwight D. j Eisenhower Nuclear Training Center, Materials Management Building, On- [
the-Job Training Center, Waste Water Treatment Plant, and the' j Executive Office Building with the KC008 panel. ! The system impacted by this modification is non safety-related. The new system meets the requirements of existing equipment. This -l modification will have no impact on accidents or malfunctions -i evaluated as the licensing basis snd there is no potential for the ! creation of a new type of unanalyred event. There is no reduction in '! the margin of safety. ? r i
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{ LSafety Evaluation: 59 93-0219- Revision 1 '! Replacement of' Essential-Service Water System Containment Isolation' valves This modification to safety related. equipment includes the replacement j of the - carbon steel Essential Service Water (ESW) containment
- Isolation valves-(EF-HV-31, 32, 33, 34, 45, 46, 49, and 50) with stainless steel valves. 'This modification will replace the valve
- bodies only and will help reduce corrosion. Replacement of the valves ;
i ' in conjunction with removal of the bypass' lines discussed below allows removal of support EF03-H001. This modification also includes removal of bypass valves EF-HV-47 and
. 4 8, associated control switches, indicating lights, input to Balance ]
of Plant (BOP) computer, and associated piping and supports (lines EF-
- HBB-069/124-10" and EF-HBC-070/125-10" and supports 0-EF05-H003 and 0- ;
EF05-H004). The bypass lines are being removed because they are no ' longer needed following the post-Loss of Coolant Accident (LOCA) ESW [ flow reduction to the Containment Air Coolers implemented by a i previous modification and reported to the NRC as Unresolved safety Question Determination (USOD) 59 91-0176. Design documents revised by this modification (USQD 59 93-0219 Revision 1) reflect removal of the bypass lines only, and do not
-l change the design basis flows and heat loads. Current normal ~
operation utilizes the bypass valves in a throttled position to : control ESW flow. To support this, the modification reported as USQD E 59 91-0176 provide a safety injection signal (SIS) open signal to a valves EF-HV-47 and EF-HV-48. Therefore, removal of the bypass valves / lines requires moving the SIS open signal to valves EF-HV-49 and EF-HV-50. Removal of the bypass valves / lines requires installation of orifices to simulate the pressure drop that was j provided by the throttled bypass valves and bypass piping configuration. The orifices allow the remaining Containment isolation f valves (EF-HV-49 and EF-HV-50) to remain near full open position. ! Amendment 78 to Wolf Creek Generating Station Technical Specifications !
. approved the deletion of valves EF-HV-47 and EF-HV 48 from Table 3.6-1 l of the Technical Specifications. ,
This modification is designed to the original ESW system requirements j and satisfies all ESW system safety design bases requirements as j identified in USAR section 9.2.1.2 including Containment isolation,
. heat removal from components important to mitigating the consequences !
of a LOCA or main steam line break (MSLB). No new failure modes of l any components, systems or structures are being introduced by this , modification. This modification will not adversely affect the system function or operation, structural integrity, reliability, or : regulatory commitments. No new hazards are created, no previously ' evaluated hazards are affected, an no accident analyses assumptions or i t l l
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f' - Page 17 of: _183 parameters are altered. The proposed change will not' create the'
- possibility'of a new or different kind of. accident. No new
- malfunctions of equipment,=and no new failure modes.of any components,c systems or structures are' introduced. The ability of the Containment-m - isolation Valves to ensure that'the release of radioactive material-from the Containment atmosphere to.the environment will'be consistent -
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- margin of safety defined in any Technical Specification.
4 , bi ' l;g , p i . Attachment-to- CO 95-0028 Page 18 of. 183 isafety Evaluation: 159-93-0220- -Revision 0 Renumbering the High Pressure Coolant Injection Flow'Indicatiors-This modification revises drawings M-12EM01, "High Pressure Coolant' Injection System," M-13EM07, "High -Pressure Coolant Injection System :
~ Auxiliary Building," and J-14EM05,e" Instrument Isolation System Pump RCS Hot Leg 4 Test." This modification changes.the number of flow-indicator. number "EMFIO998" to "EMFIO928AA" to be more consist'nt e in Lcomponent numbering to relate flow indicator EMFIO928AA'toLit's function. This change is administrative in nature and does not affect
.any component, system,.or structure.in the plant.
- 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 CO 95-0028 Page 19 of 183 Safety Evaluation 59 93-0222 Revision 0 Replacement of the Core Exit Thermocouple Connectors This modification implsments a Combustion Engineering (CE) design to replace the core exit thermocouple connectors and conoseals on the reactor vessel. head. The Thermocouple / Core Cooling Monitor uses. fifty . core exit tharmocouples mounted in the reactor upper internals. The existing metal sheathed, mineral insulated thermocouple cables exit the reactor ressel head through the four thermocouple column assemblies and are connected to field routed cables by the use of. ! Westingb~;ae supplied LEMO connectors. Because each thermocouple utilizes a separate connector, fifty connectors must be disconnected and reconnected during each refueling outage to facilitate reactor vessel head removal. The LEMO connectors are mechanically weak and are prone to damage during disconnection and reconnection. The LEMO design includes a locking feature that becomes ineffective after repeated connection and disconnection. The conoseal assemblies provide the pressure boundary penetration for the thermocouple leads. Each of the existing four penetrations have two metal conoseal assemblies that must be disconnected during a refueling outage to facilitate head removal. The lower seal consists of a Marman clamp which requires a specialized hydraulic ram for securing. The upper seal uses an autoclave type joint secured by means of multiple sequentially torqued bolts. The repeated connecting and disconnecting causes wear and tear that may result in future leakage problems. To upgrade the thermocouple connectors, the CE design replaces the LEMO connectors with twin pin connectors. A short transition cable is then run to a multipin connector to group several thermocouple circuits into one connector, reducing the number of connectors from ' fifty to eight. These connectors provide a strong mechanical connection and reduce the radiation exposure time required during a refueling outage because only eight cennections are required to be disconnected and reconnected. The new CE conoseal design replaces the lower conoseal Marman clamp with a simplified quick-acting clamp. The upper seal, including the autoclave type joint and associated hardware, is completely replaced. The new seal consists of a graphite seal preassembled on a convenient carrier. The carrier is loaded on the thermocouple column assembly , with a simple tool. The major benefit of the CE design is that the lower seal does not need to be removed for reactor vessel head lifting This modification provides new cabling with multipin electrical connectors which are rugged quick acting assemblies. The connector
' Attachment to CO 95-0028 Page 20'of 183 design is such_that the contacts are fused'into a glass seal which firmly holds them in~pisce such that pin movement and wire breakage are eliminated. The cabling inside containment for core exit
- thermocouples is configured to reduce the number of electrical connectors to be disconnected and connected at a refueling outage from l 50 to 8. The Core Exit Thermocouple Nozzle Assembly and quick Acting clamp replaces the existing nozzle assemblies with components that allow the reactor vessel head to be removed and reinstalled.without disconnecting the new quick acting clamp on the lower conoseal seal.
1 t-This modification will not increase the probability of occurrence of an accident previously evaluated in the Updated Safety Analysis Report (USAR), nor will the consequences of and accident evaluated by the USAR be increased. Neither will the probability of occurrence of a malfunction of equipment important to safety be increased. This modification will not increase the consequences of a nelfunction of equipment important to safety previously evaluated by the USAR. This change will not create the possibility of an accident of a different type nor will it create the possibility of a different type of malfunction of equipment important to safety. This new equipment is part of the Accident Monitoring Instrumentation included in Technical Specification 3.3.3.6 (Instrument #17, Thermocouple / Core Cooling Detection System). The cables, connectors and nozzle assemblies are of the same quality requirements as the original equipment and their reliability is enhanced by numerous simplified design features. The ability of this instrumentation to monitor core temperature is not diminished and the margin of safety as described in the Technical Specification is not diminished. 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. l t I l l l J
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, Safety' Evaluation: 59 93-0233 Revision:0' 'I
. Replacement of. secondary Liquid Waste Interceptor Drain'Line This permanent. modification to the non safety-related Secondary Liquid
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Waste Interceptor.(FHF03)' replaces the existing one inch drain line P with-a'three inch line. This modification is being' implemented to-facilitate improved draining of the liquid waste tank. This drain , i line joins the Secondary Liquid Waste Transfer Pump Drain Line and drains to a Liquid Radwaste floor drain in the turbine building.
.The. existing one inch line is too small to allow draining the contents [
t, of the Secondary Liquid Waste Tank. Sludge buildup makes it necessary ! to remove the three inch by one inch outlet reducer and drain the tank- ,' through the three inch flanged outlet. t 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. [ i i , e 5 l l t i i l I l l i
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; 5afety Evaluation: 59 93-0235 Revision 0 i
Transfer of. Training Responsibilities-for Emergency Planning.
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This revision to the Updated Safety Analysis Report changes the z i responsibilities for ensuring the Emergency Response Organization , I receives training. commensurate with their respective Emergency Plan responsibilities from the Manager Training to'the'Vice President Nuclear Assurance. This change is administrative in nature and is a. B ' result of organizational. changes within WCNOC. ' Existing Emergency' Planning training programs are not altered by.this change. This revision 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. ; l L t i
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m J Attachment to CO 95-0028 l Page 23 of 183 i Safety Evaluation 59 93-0243 Revision 0 Emergency Diesel Fuel Oil System Modification This modification to the safety-related Emergency Diesel Fuel Oil System provides a logic change to run the Emergency Fuel Oil Transfer Pumps (PJE01A and PJE01B) continuously while the respective Emergency Diesel Generator (EDG) engine is running. This modification also provides for the addition of vent lines from the top of the Emergency Fuel Oil Day Tank standpipe to the existing Emergency Fuel Oil Day Tank vent line. In the current configuration, with the Emergency Diesel Generator running, the fuel oil transfer pump cycles on/off about every one or > two minutes. Most thermal cycling and stresses occur as a result of pump etarting torque. Therefore this modification will improve reliability of the fuel oil transfer pumps. During operation, excess fuel oil will be overflowed from the day tank standpipe back to the fuel oil storage tank. There are no hardware changes associated with the pump logic change except for changing the termination of cables and jumpers in the control circuit. Because there are no hardware changes and operational changes are within original design, there is no change in the probability of a malfunction of equipment important to safety. The new vent lines added by this modification assure that the fuel oil day tank would be adequately vented in the event of an inoperable storage tank vent and assures that the conclusions in the Updated Safety Analysis Report ('U SAR) Section 3.5.2.5.4 remain valid. A siphoning effect when recirculating fuel oil through the overflow line has been noted during fuel oil transfer pump surveillance tests, which causes erretic pump operation. Installation of the new vent will decrease this siphoning effect. Design of the new vent line is to the same criteria as that of the existing day tank vent line. Emergency Fuel Oil Transfer Pump capacity will not be changed by this modificatioa. Likewise, day tank control setpoints will not be changed. When the Emergency Diesel Generators are in standby, operation of the fuel oil transfer pumps to maintain day tank level will be unchanged. After the Emergency Diesel Generators have started, the modified logic will start the fuel oil transfer pump on low standpipe level rather than low day tank level. The day tank low level switch will be left in the circuit as a backup. This will be more conservative with respect to minimum day tank fuel inventory. This change will also cause the fuel oil transfer pump to start sooner after the Emergency Diesel Generator starts. The earlier start will have no adverse impact on load sequencing. With the fuel oil pumps ! running continuously when the Emergency Diesel Generator runs, fuel oil level will be maintained near the top of the standpipes. Therefore, the day tank inventory will be more conservative with the j l 1
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- 0. ~ ' Page 24'.of 183 .j Emergency Diesel' Generator running.
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. This modification enhances the ~ liability.of the' Emergency Diesel
% , ;j p Fuel Oil.Systemi This modification does not affect any fission- ';
product barriers'and will not degrade the Emergency Diesel Fuel Oil System. and the role it plays in supporting the EDGs in mitigating.the L;. consequences _.of an accident. . Credible. failure modas as previously evaluated are not changed and_ reliability of the Emergency Diesel '. j Generators is not degraded. 'There are no changes resulting from;this 3 modification which would initiate a previously unevaluated event. i Based on the above discussion, this modification will have no impact y on. accidents or malfunctions evaluated as the licensing basis and there is no potential-for the creation of a new type of unanalyzed j event. There is no reduction in.the margin.of safety. l. m
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. Safety Evaluations
- 59 93-0244 Revision 0. ;
' Piping & Instrument Diagram (P&ID) Changes.and. Additions to Reflect ;
the As-Built Condition of the' Instrument Air System. s
.This modification.to the safety-related Instrument Air System providesL ;
documentation for the as-built configuration of branch isolation
- valves to loads'off of the branch ieolations The portions of the y Instrument Air System affected by this modification are non safety- !
related. Currently the P& ids for the instrument air system do not ; indicate loads (end use components) off of branch isolation valves. The objective of this modification is to provide inproved information-and documentation of Instrument' Air System loads for plant operators. _; This modification does not' change any equipment in the plant. Ten new : drawing are being issued. These include'J-14DA80 through J-14KA89. These drawing are for the following areas: Condensate Storage Tank, Auxiliary Building, Reactor Building, Control Building, Communication ; Corridor, Diesel Building, Fuel Building, Radwaste Tunnel, Auxiliary , Boiler, and the Hot Machine Shop / Refueling Water Storage Tank Valve Pit. In addition,~ drawings M012KA03 "P&ID Instrument Air System," M- I 12KA04 "P&ID Instrument Air System," M-202A-04130-WOS "RIC-WIL-Conduit l Drawing," and M-1T7135 (Q) " Composite Radwaste Tunnel" are being ! revised to reflect the branch loads. 'This modification is. administrative in nature and does not affect any components, systems, ; or structures in the plant. 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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- 0 Safety Evaluation 59 93-0247 Revision 0 Low Pressure Feedwater Hase er Level Con.rols This modification to non safety-related equipment is being implemented to address excessive tube vibration in the low pressure feedwater heaters which have contributed to tube failures. It has been determined the excessive vibration is caused by insufficient liquid level at the drain cooler zone entrance to the heater. Maintaining the proper liquid level in the shell side of the low pressure feedwater heaters is imperative for proper operation of the heaters.
This modification improves the liquid level control system by
. relocating the instrumentation and installing new components which will allow the drains cooler approach tests to ha performed and control the liquid level at the optimum level in the heaters. This modification will improve the function of the Condensate and Feedwater systems by increasing the reliability of the low pressure feedwater heaters.
This modification does not increase the probability noc the consequences of feedwater system malfunctions that result in a decrease on feedwater temperature as described in the Updated Safety Analysis Report (USAR). The function of the Condensate and Feedwater System is not changed by this modification. There ara no technical specification requirements for the Condensate and Feedwater System. 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 'CO 95-O028 l f Page 27 of- -183 { e Safety Evaluation: 59 93-0251' Revision 0 Permanent Protective' Covering fr r Circulating Water Pumps , This Engineering Disposition provides for installation of.a permanent. -i
' structure to protect the Circulating. Water pumps,and piping. This . j structure-(Butler Building) will enhance personnel safety and provide j a more suitable area for maintenance and testing activities. The {
permanent structure will be independent from the circulating Water- , ' ~ Screen House structure and therefore is' excluded from the Plant .'[ Modification Request process- . . 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. I
Attachment to CO 95-0028 Page 28 of 183 Safety Evaluation: 59 93-0252 Revision 0 Hot Tool Room Remodeling This modification provides for the expansion and remodeling of the Hot Tool Room located on the 197%' level of the Auxiliary Building. This modification replaces the existing tool room cage with a modular i storage system to increase maintenance efficiency. Separation from safety-related components is maintained by this modification. The replacement of the existing tool room cage with a modular system [ 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. l 1 P
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Page 29 of 183 Safety Evaluation: 59 93-0254 Revision 0. f
-I Steam Seal. System Pneumatic Controller Replacement
'This modification to the non safety-related Steam Seal System .;
L pneumatic controllers provide for the replacement of Fisher 4160 ; Pneumatic Controllers (CAPC0013'and CAPC0014)with Foxboro Series 43 AP l Pneumatic Controllers. The'new controllers'are believed require less maintenance-than other models currently installed at Wolf Creek : Generating Station. The basic design functions of the original. [ controllers and the replacement controllers are identical. This modification will have no impact on accidents or malfunctions evaluated as the licensing basis and there is no potential for the -l creation of a new type of unanalyzed event. The 3 is no reduction in -! the margin of safety. , [ i
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g 1 i i t Attachment'to- CO 95-0028 ) Page 30 of- '183 Safety Evaluation: 59 93-0255 Revision 0 Clarification to Updated Safety Analysis Report Tables and System Description tor the Essential-Service Water System 1 l This revision clarifies Updated Safety Analysis Report (USAR) Tables 9.2-2, 9.2-3, and 9.2-4 and M-10EF, " System Description Essential' i
; Service Water System," by. adding .a note pertaining. to the screen wash, j prelube and strainer backwash. Adding a note to these documents will i help prevent incorrect'information being used for total Essential l Service Water Sr-tem flows. This revision is for clarification only. "_
This revision is administrative in nature. There are no components, systems, or structures affected by this revision. l i This revision will have no impact on accidents or malfunctions evaluated as the licensing basis and there is no potential fer the f creation of a new type of unanalyzed event. There is no reduction in i the margin of safety. f l e i r
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Safety Evaluation: 59 94-0001' Revision 0 . e Control' Room Air Conditioning Crankcase' Heater Replacement
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f This temporary modification to the Control Room Air Conditioning Unit l
~ (SGK04A) installs a non safety-related crankcase heater in place of a .!
safety-related crankcase heater in the unit compressor. The existing ,l crankcase heater failed and a safety-related replacement heater is not available. Therefore, it is necessary to install a temporary heater
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element. The replacement heater does not have the required documentation for installation as a safety-related heater. The existing heater is hard wired. The replacement heater is the proper ; replacement for the existing heater. However, the required j documentation for replacement is not available. !
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This modification will have no impact on accidents or malfunctions i 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 i the margin of safety. } f l i i I i i l r i I
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Safety Evaluation: 59 94-0002 . Revision 0 l Replacement of Cavity Cooling Fan With Ventilation Duct Section ' This temporary modification installs a ventilation duct section in place of the cavity Cooling Fan,which is being removed for repair. ' The zine and aluminum analysis was reviewed to determine the design , bases-limits. The added amount of the ventilation duct section is not .j outside design basis or the hydrogen analysis in the USAR. Other i aspects of this installation have been previously evaluated and : reported.by Unresolved. Safety Question Determination 59 93-0223 and 87 - SE-026. This revision will have no impact on accidents or malfunctions ! evaluated as the licensing basis and there is no potential for the ;
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creation of a new type of unanalyzed event. There is no reduction in the margin of safety. ; i
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' Attachment to. CO 95-0028 i Page 33 of 183 I Safety Evaluation: 59 94-0004 Revision 0 ,
Implementation of the Commnodity Discrepancy Report p This revision to-Section 17.2.15.3 of:the Updated Safety Analysis
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. Report adds provisions'for the Commodity Discrepancy Report as'another ;
means of reporting non-conforming materials, parts'and components ~ ; under: warehouse control. Commodity Discrepancy. Reports are used to' j
' document non-conforming materials identified at the time of~ receipt and after receipt inspection. This revision has no affect'on components,-systems, or structures in the plant. l This revision will'have no impact on accidents or malfunctions evaluated as the licensing basis and there is no potantial for the j creation of a new type of unanalyzed event. There is no' reduction in- 1!
the margin of safety. , f l
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.. Attachment to. :CO-95-0028 ,i Page 34-of. 183' Safety.Ryaluation: 59 94-0005 Revisics:0 Emergency Safeguard Features Actuation-System Panel Status Lights. j
. ;This temporary modification provides jumpers to eliminate the loss of' ;
power status panel' lights for the Emergency Safeguards Features j Actuation System on the Main Control Board. Valves EFHV87 and EFHV88, l
" Train A and B Essential Service Water Radiation Monitor Isolation Valves," are de-energized and isolated. Plant Modification Request i
-(PMR) 04701 will remove.these valves during a refueling outage. ,
Eliminating these lights will reduce the number.of' nuisance alarms on { the Main Control' Board, t The' safety function of valves EFHV87 and EFHV88 is to isolate.(close) on a Safety Injection Signal. Valves EFHV87 and EFHV88 are de- ! energized (fail-safe condition) and the upstream isolation valves are , closed. For this modification the status panel lights for EFHV87 and I EFHV88 are not required, the Essential Service Water line is isolated and remains isolated with EFHV87 and EFHV88 de-energized. This temporary modification will have no impact on accidents or. malfunctions evaluated as the licensing basis and there is no 1l potential for the creation of a new type of unanalyzed event. .There: .l is no! reduction in the margin of safety. l v F 9 f
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CO 95-0028 Page 35 of 183
' Safety: Evaluation: 59 94-0006 Revision 0 Radiological Controlled Area Access Control Water Heater Replacement
.This modification to non safety-related equipment removes the l- Radiological Controlled Area Access Control domestic water heater
'(TKD08) and replaces the instantaneous steam water heater (EKD01) with two 200 gallon electric water heaters. Additionally, this modification installs an expansion tank for thermal expansion of the water. No safety-related systems or components are affected by this modification.
This moditication 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. CO 95-0028' ,
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-Safety Evaluation .59.94-0007 Revision 0 Installation of Flex Gasket in Piping Systems JThis modification allows insulating flange gasket kits to be replaced with a single standard flex gasket in piping systems at the. transition .
from underground to above ground. Insulating' flanges were specified by Bechtel as part of the generic SNUPPS design to electrically , isolate power. block piping systems from underground piping. Th!.c was done to prevent the power block portion of a system from putting an electrical load on the Cathodic Protection System. In 1979 the site
- Architect Engineer, Sargent & Lundy (S&L), recommended that all
- insulating flanges on Bechtel piping be electrically. jumpered. This .
recommendation was based on S&L design philosophy for the Cathodic
' Protection System, which allowed for additional current to flow ,
through the power block systems and uses the Station Grounding System { as the negative return. Based on KG&E Engineering's concurrence Vith ; the S&L recommendation, Bechtel issued'IDCP PDL-501-QH, which added electrical jumpers across'each insulating flange, in 1984. Because of the installation of these jumpers, the insulating flanges have no , design function other than to limit leakage of fluids'from the piping. The Updated Safety Analysis Report is affected by this change > because a misleading note regarding the need for the insulating flanges on the Piping & Instrument Diagram for the Emergency Puel Oil System is being corrected. This modification affects only flanged joint design in piping systems and the new configuration is at least as good as the current design. The configuration allowed by this change is a standard proven design [ used elsewhere in the plant and throughout the industry. This change , does not affect any subject discussed in the basis for any Technical ; Specification or'in the Safety Evaluation Report. l This revision 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. i the margin of safety.
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' Attachment to CO 95-0028 Page 37 of 183 Safety Evaluation: 59 94-0008 Revision 0 Chlorine Detector System Deactivation this modification deactivates the chlorine detection instruments (GKAE2 /AT-0 /AI-2) and (GKAE-3/AT-3/AI-3). These instruments will be deactivated in place in a manner that the instruments can be made operable with a minimum of effort. These detectors provide protection from gaseous chlorine and automatically isolate and alarm the control Room in the event chlorine concentration reaches 5 parts per million ; or greater in the Control Room air intake system. The chlorine source (1 ton containers) were removed from Wolf Creek Generating Station on January 4, 1994, thereby eliminating the need for the chlorine detection devices. 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. , f
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l Safety Evaluation: 59'94-0009 Revision 0 : f
. , ' Reanovel'of Power to Essential Service Water System Flow Instruments j
.This modification,to safety-related equipment.provides for removing . j '
power from' Flow Indicator'EF-FI-3/4 and Flow Transmitter EF-FT-3/4 in: the Essential Service Water System. The.effect of-this modification ! is to spare'the.inctruments in: place.' These instruments are not'used: l to measure Essential Service Water System flow. A portable instrument i I is used for this purpose. i Failure of these instruments is not included in any accident. analysis l in the Updated Safety Analysis Report. These: instruments have no 'j active safety function. These instrument'are safety-related because- : of the pressure boundary associated with them. The pressure boundary' , j will remain intact. 'l This modification will.have no impact on accidents or malfunctions evaluated as the licensing basis and'there is no potential for.the
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'Page 39 of 183 !
Safety Evaluation ' 59 94-0010' Revision 0 i l Organizational Changes in the Engineering Department and Reporting . Relationship of Fire' Protection Group lt This proposed revision to the Updated Safety Analysis Report l consists" ; of. organizational _ changes including title changes, changes in ! reporting relationships;.and personnel changes in'the Engineering- f Department and the Operations Division. These changes improve the '
. operating philosophy of Wolf Creek. Generating: Station and do not j
' affect the safety and reliability or operation of the plant or i represent any decrease in the concern for the health and safety of the ,
public. 3 i This revision will have no impact on accidents or malfunctions [ evaluated as the licensing basis and there is no potential for the -i
. creation of a new type of unanalyzed event. There is no reduction in-the margin of safety. ] !
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Page 40 of 183. Safety Evaluation: 59'94-0011 -Revision:0 L < ,
-Engineering Department Organization Changes-
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This revision to procedure KPN-A-303, " Responsibilities'of the ~ f
, Engineering Department," Revision 3, supersedes KP-A200, " Statement of ,
Responsibilities Nuclear Engineering Division," Revision 12. LThe changes in this revision represent organizational changes,. including
. title. changes, changes in reporting relationships,'and personnel .
changes. .These' changes improve'the operating philosophy of the. Wolf '
- Creek generating station. This revision is administrative in nature.
There are no components,: systems, or structures affected by.this-i revision. l This revision 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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' Installation of Manual Isolation valve in the Safety. Injection Test-Return Line .j i This. modification to the Safety Injection System adds a manual .
isolation valve downstream oflthe "C" Accumulator test return valve 1 EPNV8877C. This modification is-intended to eliminate leakage through > EPNV8877C and'the subsequent reduced inventory in the Accumulator. -In' addition, by preventing leakage through EPHV8877C, pressurization of' i
' Residual Heat Removal-(RHR) discharge piping and cycling of RHR' 'l
' discharge valves will be reduced or eliminated. The new valve'will be installed on the non safety-related.section of the test return line which is isolated-from safety-related components. This' isolation ;
-valve is a passive component which meets the design requirements for- l the piping in which it will be installed. This installation will {
improve equipment reliability. . This valve installation will have no impact on accidents or malfunctions evaluated as the. licensing basis and there is no. i potential for the' creation of a new type of unanalyzed' event. There is no reduction in the margin of safety. :l t f d t i 1, t
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f (p . Safety Evaluations: 59'94-0013 - Revision 3 0 f Freese. Seal Installation'on Accumulator C Test Line
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~' 4 This temporary modification provides for the installation of a freeze. . seal on the' Safety Injection System Test Return line-to allow i installation of a new. valve (EM-V250) as specified by Plant. j Modification Request 04821.-Unresolved Safety Question Determination
'(USQD) 94-0012 was developed and approved for Plant Modification- {
-Request 04821. The freeze seal will'be applied in accordance with E approved procedures.
'This temporary modification will have no impact on accidents or malfunctions evaluated as the licensing basis and.there is no 1 4
potential for the' creation of.-a new type of unanalyzed event. .There. j
-is no reduction-in'the margin of safety.' .i I
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safety Evaluations .59.94-0014= . Revision 3 0 l Replacement'of Temperature. controllers '
- This non safety-related modification, replaces existing: servo-operated. j
. temperature' controllers;EBTIC-035, and EBTIC-036, FCTIC-055, and'FCTIC-
- 155, in the. Closed cooling Water System and the Auxiliary Turbine lj system, with microprocessor-based controllers. This modification _will .;
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' improve'the reliability of the controller and resolve the existing 3
- concern about spare parts. This change consists of replacing the old' !
- controllers with'a better designed control 1 system.
- This modification willLhave no impact on accidents or malfunctions evaluated as-the licensing basis and there is no potential _for'the j creation of a new type of unanalyzed event. There is no reduction in .. ;
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183 f L; Safety Evaluation:' 59 94-0015' Revision:D- f
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Verification and Maintenance.of WCNOC; Imaging System' ;
-This new procedure, AP 10-001, Revision 0, " Imaging System,"
. establishes the methods and responsibilities for the input, verification and maintenance of images in the WCNOC Imaging. System.
It also establishes the use of optical disks as a record storage ! medium-as described in Generic Letter 88-18, " Plant Record Storage on optical Disks." The useJof optical' disks as a storage medium for
-records does not decrease any other-record requirements.
i This new~ procedure will have no impact on accidents or malfunctionst
. evaluated ~as the licensing basis and there iis'no potential'for the ~!
creation of a new type of unanalyzed event. There-is-no reduction.in ! the margin of safety. l i n 3
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> Attachment to CO 95-0028 Page 45"of 183 Safety Evaluations. 59 94-0016 Revision 0 Disabling of Emergency Diesel. Generator Starting Air System Motor-
_ Operated Valves
'This" modification abandons in place the following. Emergency Diesel Generator System Starting Air valve KJHV0001, (Essential Service Water System "A" to Starting Air Compressor After' Coolers 5A/B Isolation);
KJHV0002, " Starting Air Compressors After Coolers SA/B Essential Service Water "A" Return Icolation Valve) , KJHV0101, (Essential Service Water "B" to' Starting Air. Compressor After Coolers'SC/D Isolation Valve), and KJHV0102 (Starting Air Compressor After Coolers SC/D ESW "B" Return Isolation Valve). All. valves affected by.this
. modification will be left in the closed position permanently. .These Emergency Diesel Generator System motor-operated valves are boundary valves for the Essential Service Water System that supplies cooling water to the starting air' compressor after coolers (EKJ05A, EKJ05B,.EKJ05C, and EKJ05D). The air compressors, dryers, and after -coolers of the Emergency Diesel Generator Starting Air System are non.
safety-related. The air reservoir tanks are the beginning of the. safety-related portion of the system. The tanks are sized large enough to start the Emergency Diesel Generator without.having to be recharged. The cooling water supply is non safety-related and is isolated on a Safety Injection Signal. The design of the air compressor allows the discharge to be adequately cooled before the compressed air exits the compressor unit. The accident position of the affected valves is closed and the U pdated , Safety Analysis Report specifically states in Section 9.5.6.2.2 that ! loss of cooling water does not damage the after cooler and the other ! components in the Emergency Diesel Generator Starting Air System. l' Because the valves will be disabled, no malfunction can occur and the consequences of an accident are not increased. This modification does , not affect the operability or function of the Emergency Diesel l Generators and does not create the possibility of an accident of a 3 i different type than previously evaluated in.the Updated Safety Analysis Report. Based on the above discussion, 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 j type of unanalyzed event, There is no reduction in the margin of ti safety. of safety. 1 l l I l 1
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~ Safety Evaluation 59'94-0017 Revision 0
-1 Reactor Coolant System and Fuel Pool Absolute Filtration Program' '
This modification provides-instructions for the. filter cartridge
' downsizing' program that will be implemented at Wolf Creek Generating- !
. Station. The intent of the filter cartridge downsizing program is to 'l l, reduce corrorion products in the Chemical and volume Control System ;
h: and Spent Fuel' Pool Cooling and Cleaning System. Downsizing'is a key - part of the Wolf Creek Generating Station long range radiation i exposure reduction plan. This modification will allow installation of f l alternate glass fiber filter cartridges in the Chemical-and Volume 1 Control System and Spent Fuel Pool' Cooling and Cleaning System in the submicron range (.2 microns) which is considered very effective in removing iron oxides and cobalt particles. . Specific filters involved include Seal Injection (FBG04A&4B), Seal Water Return'(FBG05), Reactor Coolant (FBG06), Boric Acid (FBG07), Puel' Pool Cleanup (FEC01A&lB), and Fuel Pool Skimmer (FEC02).
- Malfunctions described in Updated Safety Analysis Report Tables l9.1-6 (Spent Fuel Pool. Cooling and Cleaning System) and 9.3-10 (Chemical and Volume control System) are not affected by the use of submicron glass
-fiber filters.and the consequences of' malfunctions are not increased.
No new type of potential failures have been created by using glass' fiber submicron filters. The new filters are considered a systemi enhancement and will not affect systems, structures and components of the Chemical and Volume control System or and Spent Fuel Pool' Cooling and cleaning System in such a way as to create a different kind of malfunction. No margins of safety are affected by the use of submicron glass fiber filters. Use of these filters does not prevent plant chemistry from being controlled within ranges required by technical specifications. 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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Page 47.of. 183 ,! l h f Safety Evaluation: '59 94-0018 Revision 0 l
" Unit Vent Flow Probe Modification [
This modification to-non safety-related.equipmant provides for the use of available flow information compiled by the plant computer,. based on fan run contacts, to maintain isokinetic flow control and perform i , effluent calculations. This modifications: abandons in place the Unit ;; Vent velocity tubes (GTST0021BA and GTST0021BB) . which serve as process ; P radiation monitors. The Unit Vent-Velocity tubes provide a signal-that ;i is linearly proportional to stack flow and is used by the RM-80 j - computer to maintain isokinetic flow control during' sample taking and to perform effluent calculations. This modification allows for using i the substitute values available from the plant > computer. 'The method { provided by this modification will improve the reliability of process , radiation monitoring. The velocity tubes provides no isolation '[; function. Abandoning the velocity tubes can not cause a malfunction of equipment important to safety, i 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. e I a i [
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m-Attachment to CO 95-0028 Page 48 of 183 Safety Evaluation 59 94-0020 Revision 0 Revision to Inservice Valve Test Procedure This revision to procedure STS BN-06 " Borated Refueling Water Storage System Inservice Valve Test," Revision 2, provides for the performance of inservice valve testing to this system while in Mode 5 as well as Modes 6 and with no fuel in the reactor vessel. Previous revisions of this procedure have. required the plant to be in Mode 6 or with no fuel in the reactor vessel to perform this test. However, ASME/ ANSI OMA-1988, Part 10, requires that this test must also be capable of being performed in Mode 5. The revision to this procedure allows performance of the test with the plant in Mode 5. The only accident as a possible result of this procedure is a Boron Dilution Event. Administrative measures are in place in the procedure to preclude a Boron Dilution Event from occurring. Therefore this revision 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.
, =-- Attachment to CO 95-0028 Page 49 of 183 Safety Evaluation: 59 94-0021 Revision 0 Installation of Test Plugs in the Potable Water System This temporary modification to the non safety-related Potable Water System installs test plugs upstream and downstream of bypass valve
-lWM580. This temporary modification is being implemented to maintain availability of the potable water system to the plant facilities.
There are no design basis accidents evaluated in the Up dated Safety Analysis Report for the non safety-related Potable Water System. The test plugs do not affect the system's functions nor the failure , modes. There are no safety-related components, systems, or structures affected by this temporary modification. This temporary 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.
s Attachment to' CO.95-0028 Page-50 of 183 Safety. Evaluations. 59 94-0022- Revision 0 Nuclear Station Operator Training Program Procedure Change' This revision to the Updated Safety Analysis Report changes the, procedure'ADM 06-211 " Nuclear Station Operator (NSO) . Training Program" to AP 43-211'. There'are no other affects on descriptions.in the Updated. Safety, Analysis Report. This' revision is; editorial in nature.
. and no components,-systems, or structures important.to safety are affected-by this revision. 1 This revision sill have no. impact.on accidentsLor 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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p 5.'i n' I Attachment to CO 95-0028- ! L .. Page 51 of 183 l I,l Safety Evaluation: 59 94'-0023' Revision 0 , ll - Clarification of. cable Tray' Fill' Limits- -
' ;; .This revision to the Updated Safety Analysis Report:(USAR) references +
Wolf CreekiNuclea',r Operating Corporation documents which define. .; overfill limits.for; cable trays. The current description of' cable
-tray fill limits is. inaccurate. The current. description states that
. trays containing~ power' cables only are generally. limited to 30% fill.'
[ Fill limits for/ power cables only is determined by maintained space i Erequirements. *
.This change is a documentation change only and more explicitly defines 1 the limits for cable tray fill limits. This revision will have.no ,
E 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. q 7 i h h f o
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' Attachment.to .CO 95-0028 I Page'52 of 183 ,
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-Safety Evaluation: 59'94-0024 - Revision 0-Permanent Modification and Configuration Control Procedure Revision ;
' Revision 10 to procedure KGP 1131, " Permanent Modification and '
Configuration Control,"-implements changes ~in the areas of concepts i and definitions. In' addition, this revision expands the use of the l Configuration Control Package'and revises the modification' scope - criteria forlsome systeam and buildings. . The revisions'to'this .
-procedure continue to. ensure the design basca functions are maintained and that the Plant Safety Review Committee approves all Unresolved. ;
Safety Question Determinations before;implemenration in completed. [ This revision will have no impact on accidents or malfunctions- l 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^ i the margin of safety, L! T 3 4 e 6 f -I
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Page.53 of .183
"_ , safety Evaluation: 59 94-0025 Revision 0 [
Radiological Emergency. Response Plan Revision 39 f Revision 39 to the~ Radiological Emergency Response Plan incorporates',a grammatical change to the Emergency Action Levels (EALs)'in response ! to the Region.IV. Emergency Preparedness Inspector's comments. This , revision ensures the conditional statements in the EALs at the Alert , and Site Area Emergency' classifications applies to both the loss of the PK02 bus and a lossiof;75% of the control. room annunciators. .;
'This revision will have no impact on accidents or malfunctions $
' evaluated as the licensing basis ar.d there is no potential for the creation of a new type of unanalyzed event; There is no reduction ir '
i the margin of safety.
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. Page 54.of 183 '
l i Safety Evaluation: 59 94-0026 Revisica:O .. Feedwater Corresion Product Sampling Line This modification.to non safety-related equipment provides for the permanent installation of the existing sample panel which was j previously installed by Tempcrary Modification Order (TMO) 92-16-AE, TMO 92-16-AE was evaluated to the criteria of 10 CFR 50.59 and found to be acceptable.
- This modification provides permanent design for a sample panel to [
monitor Feedwater System corrosion products in the process water. l This modification differs from the existing temporary installation. ! This permanent design provides sample cooling from the Closed Cooling ., Water (EB) System and provides for the exiting water to the Feedwater' ! Reater Extraction Drains and Vents (AF) System. The sampling l connection remains at the same location, j This modification will have no impact on accidents or malfunctions l 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. l i i i i i I l l: I i t i
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Attachment to 'CO 95-0028 i t Page 55 of 183 , t Safety.Ryaluations' -59 94-0027 Revision 0 ! l Transfer of Recycle Holdup Tank (RNUT) "B" Contents'to REUT "A"
. Temporary procedure TP OP 305i " Transfer of 'B' RHUT Contents to 'A' RHUT," provides for transfer of the contents of RHUT "B" to RHUT "A" ,
by. equalization through the tank outlet isolation valves. Temporary; : procedure TP.OP 305 opens both of the RHUT outlet isolation valves at ! 9" , the same time to raise the tank level above the pump trip setpoint to: :l allow for processing the' effluent. Failures associated with the~ , Liquid Radwaste System have been analyzed in the Updated Safety . Analysis Report. The performance of temporary. procedure'TP OP_305 ; will.not affect the assumptions made in the failure analysis for the. l Liquid Radwaste System. This temporary procedure does not' affect the [ quantity, storage, production, or release of radioactive materials at 'i Wolf Creek Generating Station. i i This temporary procedure will have no impact on accidents or -f malfunctions evaluated:as the licensing-basis-and there is no. ll potential for the creation of a new type of unanalyzed event. .There . I is no reduction in the margin of safety. ! If
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Attachment toz CO 95-0028-
.Page 56 of 183 ,
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. Safety Evaluations. 59.94-0028- Revision 0 l Freese Seal to Replace. Auxiliary Feedwater System Valve.(ALNV0032) ;
This temporary modification installs a freeze seal on an eight inch' ! pipe an Essential Service Water System line~ to isolate valve ALHV0032 "ESW Supply to Turbine Driven Auxiliary Feedwater Pump." .This {
. temporary modification will temporarily isolate one train of the 1 Essential Service Water System supply to the' Turbine Driven Auxiliary Feedwater Pump.'This temporary modification'is being installed to !
facilitate the replacement of valve ALHV0032; The design function of .j the remaining ESW train.and the Turbine Driven Auxiliary Feedwater l- Pump is not affected by this temporary modification. Therefore, the ; equipment important to safety for these systems will continue to , perform their design function. l i This temporary modification will have no impact on accidents'or' .! malfunctions evaluated as the licensing basis and there is nd . potential for the creation of a new type of unanalyzed event. There is no reduction in the margin of safety. , I i j r 9
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Safety Ryaluations- 159 94-0030 . Revision 0 i Freeze Seal to' Replace Auxiliary Feedwater System Valve - , i , This temporary modification provides for the installation of a' freeze- . seal on an Essential-Service Water System line to isolate valve i- ALHV0030, (ESW to Motor Driven Auxiliary Feedwater Pump). This freeze _ seal will temporarily isolate. Motor Driven Auxiliary Feedwater Pump i "B." The "A" train of the Essential Service Water System will remain- [ operable. This freeze seal will allow the replacement of valve ! ALHV0030 without declaring the Easential Service Water System
' inoperable. !
I This temporary modification will;have no impact on accidents or : malfunctions evaluated as the licensing basis and there is no. ! i
. potential for the creation of a new type of'unanalyzed event. There is no reduction in the margin of safety.
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, Safety Evaluation: 59 94-0031 -Revision 0 -
Replacement of Fuel Handling Crane Auxiliary Roist This modification replaces the existing two-speed auxiliary hoist on > the Fuel' Handling Crane with a variable _ speed hoist. Both the ~ existing hoist and the replacement hoist are rated at 1.5 tons. The replacement hoist weighs approximately 555 pounds more than the. .! existing hoist. The Fuel Handling Crane will be structurally. j strengthened to maintain seismic II/I qualificationi: The Fuel- ; Handling Crane and the auxiliary hoist are non safety-related. The '
/' : Fuel Handling Crane is used only during refueling outages. :
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'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. T_here is no reduction in the margin of safety. ;
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;Page 59 of - 183 l l
Safety Evaluation ' 59 94-0035 Revision 0 !
'l Nuclear Sampling System Strainer Removal 'l
.. r This revision involves a total rewrite of procedure ADM 01-113 .;
" Scaffold Construction" and supersedes that procedure with AP 32-001'
" Scaffold construction." Scaffold constructed by this new procedure j!
b ' in safety related areas is. required to be evaluated for seismic ! qualification and will have no affect of the overall system
-performance. All equipment;important to safety.is protected from ,
degradation by this procedure. l t tais revision will have'no impact on accidents or malfunctions } evaluated as the licensing basis and there is.no potential for the l[ creation of a new type of unanalyzed event. There is no reduction in ; the margin of safety, f r o t h
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Attachment to CO 95-0028 Page 60 of 183 Safety Evaluation: 59 94-0036 Revision: 1 l Triaxici Seismic Switches Setpoint Changes This modification to non safety-related equipment provides for a change to the seismic Instrumentation System trigger setpoint from 0.01g to 0.03g. Additionally, the inputs for the system trigger and Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE) alarms are shifted to initiate from the free field instead of the containment base slab and/or operating floor. The Seismic Instrumentation System performs no automatic safety function, and it is not required for the safe shutdown of the plant. The Seismic Instrumentation System only provides information to the Control Room operators following a seismic event that could indicate that a shutdown is necessary. This modification does not alter any safety-related design functione or increase challenges to any system, structure, or component important to safety. 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. 9
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Attachment to lPage 61 of' 103 ! R safety. Evaluation: .l ~ 59-94-0038 Revision 0
.i scaffold Construction Procedute .
This revision involves a total rewrite of procedure ADM 01-113,
" Scaffold construction," and supersedes that procedure with AP 32-001, l
" scaffold Construction." Scaffold constructed by this new procedure ;
in safety-related areas is evaluated for seismic qualification and I will have no affect of the.overall system performance. 'All equipment- j
-important to safety is protected from degradation by-this procedure.
This revision will have no impact on accidents or nelfunctions l evaluated as the licensing basis'and there is no potential for the . .) e creation of a new type of unanalyzed event. There is no reduction.in i the margin of safety. {
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.Page 62 of:. .183 . .}
n , Safety Evaluations .59 94-0039 Revision 0 -- Primary to Secondary Steam Generator Leakage Procedure' j Procedure'AP 02-001 " Primary to Secondary S/G Leakage" is being. issued to address concerns identified by the Institute of Nuclear Power' ;
- Operations (INPO) Significant Operating Event Report (SOER) 93-1 :j
" Diagnosis and Mitigation of Reactor Coolant System Leakage' Including' i Steam' Generator Tube Ruptures." The purpose of this procedure is~to provide administrative guidance for the detection of:and response to a. ,
primary system to secondary system leak'. This procedure'provides ;
- guidance for operating the plant with primary system to secondary system leakage less than Technical Specification limits. This .
procedure instructs the control room operators to shutdown the plant when Technical Specification limits are' exceeded. This new procedure will'have no impact on accidents.or malfunctions evaluated as the licensing basis and there-is no potential for the ! creation of a nest type of unanalyced event. There is.no reduction in I the margin of safety. [
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j r l 1! Attachme'tn to CO 95-0028. { Page:63 of 183 safety Evaluation: 59 94-0040 Revision 0 ! Gas Decay Tank setpoint. Change' This modification changes to setpoints for the low range alarmsLfor HA '! PIS 1054;(Gas Decay Tank 1H Pressure Indicator) and HA PIS 1055 (Gas- ! ,1 Decay Tank 1D Pressure Indicator) in accordance' with design
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specifications for the pressure switch settings. . Design' drawings will l
~ also be changed -to reflect the ' correct setpoints. The high pressure' l setting of the Shutdown Gas Decay tanks ~is unchanged. The pressure ;
switch multiplier is set at 5.0. This causes the low range to alarm .;
- at 18 psig. when the high range is set at 90 psig. The change from i the existing 20 psig; set point for the low range alarm to 18 psig. is ,
in a conservative direction. ! i This modification will not affect any accident previously evaluated in j the Updated Safety Analysis Report nor will-it affect the consequences ! of any accident previously evaluated in the Updated safety Analysis ! Report. This modification will not increase the probability of ; occurrence of a malfunction of equipment important to safety nor will ! it increase the consequences of.a malfunction of equipment important , to safety. This modification will not create the possibility of an l accident of a different type than previously evaluated in the USAR nor l will it create a different type of malfunction of equipment important to safety than previously evaluated in the Updated Safety Analysis ;) Report This modification does not reduce the margin of safety as 'l- ' defined in technical specifications. I i i f f i l h 1 1 1
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i Attachment to CO 95-0028 l
'Page'64 of 183 .'
Safety Evaluations- 59-94-0041 -Revision 30 { Piping Replacement in Third Stage Extraction ^ to Moisture. Separator Reheaters ; This modification to non safety-related piping replaces existing. I carbon steel piping with Chrome Moly piping in the Condensate System,- t This modification is being implemented because of abnormal. pipe-wall .; thinning.in the third stage extraction of the Noisture' Separator. ) Reheaters. The replacement piping is more resiutant to wear than the j existing piping. , l i This modification which changes-the piping material from low carbon -; steel to low alloy steel will have no impact'on accidents or ! malfunctions evaluated as the licensing basis and_there is no f potential for.the creation of a new type of unanalyzed event. There 'I is no reduction in the margin of safety. i i
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safety Evaluation: 59 94-0042 Revision 0
-Process sampling Laboratory Air Conditioning Unit' Modification j This modification to a'non safety-related system replaces the evaporator and expansion valves (GEV0373 ' and GEV0739 respectively) for the Process Sampling Laboratory air conditioning. This modification ~ , t increases the heat removal capabilities of the air conditioning unit- [
in accordance with the design'apecification..There is no. Technical l SpecificationLassociated with the Process Sampling. Laboratory HVAC ! system. l r This modification will'have no: impact on accidents or malfunctions i evaluated as the licensing basis and there is no potential for the
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, Attachment to CO 95-0028 Page 66 of 183 Safety Evaluation 59 94-0043 Revision 0 Revision to Dose Calculation Methodology This revision to procedure EPP 01-7.2, " Computer Dose Calculations," incorporates the methodology of EPA 400 " Manual of Protection Action Guides and Protective Actions for Nuclear Incidents." The EPA 400 methodology is incorporated into the software used for dose assessment calculation. This revision changes the dose conversion factors and the emergency worker dose limits to be consistent with the more conservative guidelines of EPA 400. There will be no impact on accidents or malfunctions evaluated as the licensing basis and there is no potential for the creat1on of a new type of unanalyzed event. There is no reduction in the margin of safety. , h i i
i t ' Attachment to CO 95-0028 Page 67.of. 183 Safety Evaluation: 59 94-0044 Revision 0 _r
. Replacement of the Condensate System Drain Lines
'This modification provides for the replacement of existing carbon steel piping material with stainless steel in the: Condensate System
- drain, header. The drain header piping from the Main Feedwater Pump turbine stop valves to the Low Pressure Condenser. Abnormal pipe wall thinning has occurred in this drain header, and it will be replaced -
with stainless steel which is more resistant to-wear. This modification will have no impact on accidents or malfunctions n . 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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~Page 68 of 183
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. Safety Evaluation: 59 94-0045- Revision 0- ;
Replacement of Drain Lines.From Main Feedwater Pump High Pressure. ;
-Safety Valves :
This modification provides.for the replacement of existing' carbon- I steel piping material with low-alloy steell (2 1/4 Cr-1 Moly). in 1" &1 1/2". piping sections of the below-seat drain lines. These drain .I lines come from the. Steam Generator Feedwater Pump turbine high' j pressure safety valves. Abnormal pipe wall thinning has occurred in this piping, and it will be replaced with low-alloy steel which is more resistant to wear, , 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 j h the margin of safety. j
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1._ r ll f { Attachment to : CO 95-002Bn Page 69 of 183 , safety' Evaluations' 59'94-0047 LR evision 0
- Containment Todl Room L4 ,
This. modification provides for the installation of a permanent tool. j
, room in the containment building'at elevation 2051'. .This new tool :
J room has no active functions.and.is designed as seismic II/I. The .l 4 tool room is composed of modular cabinets.that are bolted together and- -l l 4 bolted to the supporting structural' steel. The' entire tool room is -l considered as a unit and has a minimum width of 15 feet and maximum- i height of 5 feet. '
. . I This modification will have no impact on accidents or malfunctions i evaluated as_the-licensing basia and there is no potential for the ;
creation of a'new type of unanalyzed event. There is no reduction in- , the margin of safety. I i t
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- Sofety Evaluations- 59,94-0048 Revision 0 -
Interface-Design for Radwaste Storage Building Fire Protection System- 'l This modification ~provides for connecting the Radwaste Storage. Building fore, protection equipment to the for the. Fire Protection' ; f System. This mcdification includes the design and-layout of-the
- interconnecting piping between isolation valve KCV0871,' located in the-Radwaste Storage Building, and the existing Fire Protection System; {
water supply header in the~ waste. drum storage area of the Radwaste ! Building. All circuits are wired to alarm locally and/or in the ~ control room, j The Radwaste Storage Building is classified as a non safety-related- ] ' structure. All the systems and components associated with this building are non safety-related. including the wet pipe sprinkler
. system. The Radwaste Storage Building will be used for storage of contaminated materials in sealed containers. Manual and automatic -i fire suppression features have been provided. including local and f central alarms. These provisions will ensure early identification and- [
suppression of a fire or a potential fire in the Radwaste Storage { Building. l This modification will have no impact on accidents or malfunctions l evaluated as the licensing basis and there is no potential.for the i creation of a new type of unanalyzed event. There is no reduction in 1
'the margin of safety. !
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! Safety Evaluation: 59-94-0049. Revision 0-Drawing Correction to P&ID for Secondary Liquid Waste System-p This modification corrects the location for hand switch HFHIS0095
[. (Secondary Liquid Waste to Waste Water Facility) on drawing M-12HF03, "P&ID' Secondary Liquid Waste System." Review of other. designf drawings: and field' inspection indicate the existing location in the field is~ correct. This modification involves no changes in the field, only a correction to drawing M-12HF03. 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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toJ CO 95-0028 e Page 72 of- 183 't i Safety Evaluations: 59L94-0050 ' Revision:0 { t Main Foodwater Pump Isolation _ Signal { This modification to the safety-related Main Feedwater System (AE)' -! adds a design feature to trip the Main Feedwater Pumps (MFPs) upon ; I ! receipt of a feedwater isolation signal (FWIS). ,Tais' design also includes hand switches that'will allow operators to block the new trip
' feature.'
.I This' modification will trip the MFPs a few seconds earlier than would j '
normally occur when the Main Feedwater Isolation Valves (MFIV) close. The effect of this modification on the consequences of equipment l malfunctioning that can affect feedwater flow is minimal. On a ! feedwater line break, the MFPs trip is a normal and desirable ~ -t [' consequence. , The failure of this new deJign to perform the intended function will i result in a trip of the MFPs because of high discharge header . f pressure. This is the current design and thus has been previously- ll evaluated. A trip of the MFPs or a failure of the new trip, feature j does not result in a malfunction of equipment important to safety. j This modification will have no impact on accidents'or malfunctions l 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 d the margin of safety. l l t I
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p_ _ Attachment to CO 95-0028 Page 73 of 183 Safety Evaluation: 59 94-0051 Revision 0 Modification of Steam Generator Hydraulic Snubbers This modification to safety-related equipment eliminates all sixteen of the Steam Generator (SG) hydraulic snubbers from the four SGs. The two SG snubbers on the reactor vessel side (front side) of each SG are to be removed or rendered functionally inoperable. The associated hydraulic fluid tubing and reservoirs are to be removed, or abandoned in place. The two SG snubbers on the side of the SG that is away from the reactor vessel (back side) are to be modified with a compression collar device that will allow the snubber body to act rigidly. Additionally, the snubber will no longer utilize hydraulic fluid and , pressure sensitive valves to restrain dynamic loading events. The compression collar will be fitted w.4,th a gap so that the free thermal ; motion of the SG can occur, and the rigid bumper is available for dynamic loading. It has been demonstrated that with the front two snubbers deactivated > and the back two snubbers converte6 to compression struts, the loads and stresses in the loop piping, primary equipment supports, and main steam and feedwater piping remain within the applicable allowable limits. The modified snubbers, including the compression collars, are within the appropriate ASME Code limits. The function of the Reactor Coolant System (RCS) and attached auxiliary systems are maintained. The evaluation of affected systems after modification of the SG snubbers demonstrates that the loads and stresses are maintained within the appropriate allowable limits for the RCS equipment and supports. The elimination of the dynamic effects of postulated breaks from the design basis of the reactor coolant loop, the deactivation of the snubbers, and the addition of the compression bumpers do not affect the Emergency Core Cooling System, reactor containment functional design basis, equipment environmental qualification basis, or engineered safety feature systems response from those previously evaluated. Operation of the plant with the snubbers functionally inoperable will not affect any assumption previously made in evaluating the radiological consequences of an accident described in the Updated Safety Analysis Report. The operation of safety-related equipment is unaffected. 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. 1 I
y N ' Attachment to' 'CO 95-0028 Page'74 of. .183 Safety Evaluations 59 94-0052 Revision 0 Chemical and volume control System Filter Isolation. Valve Remote' operators This' modification to the-safety-related Chemical and volume' Control System filterfisolation valves removes the remote operators-(mechanical reach rods). 'This modification is being implemented-because of operability problems. The remote operators are being replaced with local manual operators on isolation valves (BG-V0101, BG- ! V0102, BG-V0105, and BG-V0106) for the' Seal Water Injection Filters (FBG04A&B). These valves are ASME Section III, Class 2, globe valves. The function of these valves is to isolate their respective h filters during change out. The safety function of these valves is to provide a pressure boundary. The replacement of the remote mechanical reach rods with local manual hand wheels only changes the manner in which the valves'are opened and closed. This modification does not affect their safety function. Therefore, 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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i .I" 'F Attachment to CO'95-0028 : Page'75 of. 183' ; b - LSafety Evaluation: 59 94-0053 Revision 3 0 " ^ De-energisation of Essential Service Water Valve Pit Meater 1 g .'. 3 I. Thisl permanent modification to non safety-related equipment allows the- [ ' space heaters from the Essential Service Water L(ESW) valve pits to be , removed from service. The heaters are no longer needed because the- ! equipment the heaters were designed to protect has.been removed by a
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i previous' modification. Calculation GD-M-001, Revision 0, was N' performed to show that the ESW valve pits will not' reach a temperature less than 32 degrees Fahrenheit. The space heaters will be left in. place and de-energized by their respective breakers. i
'This modification will have no impact on accidents or malfunctions [
evaluated as the licensing basis'and there is no potential for the j creation of a new type of unanalyzed event. There is no reduction in t the margin of safety. l i t
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'Page.76 of. 183 Safety Evaluation: 59 94-0054 Revision 0 .(
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o Separation of Fwedwater Relief' Valve Discharge Piping' I ip This modification to non safety-related equipment provides for the separation of a common discharge pipe from feedwater heater. relief- + valves (AEV0798 and AEV0979). This modification is being implemented j because of personnel safety considerations. Routing the two relief- ! valve lines to the floor drain separately will enable plant personnel-to work on one relief valve without the threat of the other relief 1 valve lifting and having steam come back up-the other line. , l l Modification of the relief valve discharge lines 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. f l i i 1 l I i 1 l
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p p Attachment to CO 95-0028 Page 77 of 183 Safety Evaluation: 59 94-0055 Revision 0 Room 3208 Combustible Load Evaluation This modification provides a reevaluation of the fixed combustible loading in room 3200 to address the addition of a proposed work station in the room. Room 3208 is in fire area C-6. Fire area C-6 includes the Access control araa for the Radiological Controlled Area above and below the suspended ceiling in rooms 3201 through 3211 and it is separated from other fire areas by three hour rated barriers. The additional combustible load created by the work station will be 22,000 Btu /Sq.Ft. Room 3208 has ionization type detectors installed above and below the ceiling. An automatic wet pipe sprinkler system is also installed above and below the ceiling. Alarms for both the detection and suppression systems alarm locally and in the Control
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Room. Manual suppression consists of a hose rack installed in the corridor within 20 feet of the doorway and a fire extinguisher which is located in room 3208. The addition of the new work station will result in a fixed combustible load of 22,000 Btu /Sq. Ft. in room 3208. This loading is well below the fire load required to challenge the suppression system or the three hour fire barriers of fire area C-
- 6. Additionally, transient combustibles and ignition sources are controlled administratively by the Fire Protection Program.
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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'I Attachment'to CO 95-0028 4
Page'78 of '183 [? f
- safety Ryaluations- 59 94-0056. ' Revision:0; .!
' Auxiliary stema Boiler stack Drain Line l
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- This modification-to the non safety-related drain line for the
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' Auxiliary Steam Boiler Stack replaces the two inch carbon steel' piping. t with two inch stainless steel piping. Abnormal corrosion is' evident. '
.in the' existing drain line from the Auxiliary Steam Boiler Stack
- because'this pipe drains potentially diluted sulfuric' acid created from.the mixing of water and ash in the stack.
Changing the type of material in the piping for the Auxiliary Steam j Boiler: Stack will have no impact on accidents or malfunctions t evaluated-as the licensing basis and there is no potential for the , creation of a new type of unanalyzed event. There is no reduction 11nL [ the margin of safety. -[
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7 L 1 Attachment to CC 96-0028 Page 79 of 183 Safety Evaluation: 59 94-0057 Revision 0 Reactor Coolant Pump Pin Replacement and Steam Generator Inlet Restraint Removal l This permanent modification provides for eight of the twelve safety-related Reactor Coolant Pump (RCP) tie rod pins (i.e. Reactor vessel side tie rods only) and the associated fastener hardware to be replaced with dimensionally modified pins in order to permit additional bracket pin slot clearance. In addition, this modification deactivates the Steam Generator (SG) inlet elbow whip restraints. The functionality of the SG whip restraint is unnecessary because of the elimination of primary pipe breaks resulting from the implementation of leak before break design. This modification will allow free thermal movement for the RCP tie rods through the entire heatup cycle to 100% power. The function of affected systems has been shown to be maintained; neither the replacement of the RCP tie rod pins, nor the deactivation of the inlet elbow whip restraints will affect any assumptions made in the radiological dose consequences. This conclusion also applies if the accident is the result of an equipment malfunction. Because the Emergency Core Cooling System (ECCS) analyses, containment analyses, and equipment environmental qualification are not affected by the elimination of pipe ruptures or deactivation of whip rest'raints, the operation of safety-related equipment and the ability to mitigate design basis accidents are also unaffected. The function of the affected systems and the. integrity of piping, supports, and equipment are maintained for the modified RCP tie rod pin design and whip restraint deactivation. Also, the elimination of pipe ruptures and deactivation of whip restraints in the reactor coolant loop piping would not require the function of any safety related equipment beyond that which is assumed in the current ECCS and containment analyses, which are u.affected by this change. Therefore no new failure modes are created. The margin of safety of the reactor coolant piping, components, and supports is defined by the structural criteria presented in Section III of the ASME Code. Considering the modified RCP tie rod pin design and deactivation of the SG inlet elbow whip restraint, it has been demonstrated that the appropriate criteria are met. The function of safety-related systems is maintained. 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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, tsafetyLRyaluations: - 59/94-0058- l Revision 0/ ,
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Closed cooling Water System = Relief Valve Relocation !
, This modification'to the non safety-related> heat exchanger. relief .j
' valves in'the closed-cooling Water system relocates the relief valves (from a side mounted' position'to the top side'of the heat exchanger. tj
... ~This will reduce'the amount of line. clogging.because'of. sedimentation 1 '
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'and thereby improve operability. Relocation of-these relief valves- d (EBV0100 and EBV0102) does not affect the function'of the' valves. The1
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l design function of these' valves:1s to provide thermal relief'to thel non safety-related heat exchangers EEB01A/B respectively. I
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.This' modification'will.have no impact on accidents-or malfunctions- f 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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i Lo-, - Attachment to. CO'95-0028: I 4 Page.81 of 183 '! r safety Evaluation: 59 94-0059 Revision 0 .
. Administrative Services Organisation Changes .;
This revision to the Updated Safety Analysis Report (US AR) : represents organizational. changes including title changes, changes in reporting. ;
-relationships and personnel changes. All functions continue to be i fulfilled. These changes improve the operating philosophy of the Wolf f Creek Generating Station'and do not affect the1 safety and'reliabilityJ
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, of pl' ant operation or represent any decrease.in.the. concern for the ;[
health and safety of the public. This change does'not affect any , system, component, structure, or procedure required to mitigate the i consequences of an accident previously evaluated in the USAR. i
~l This revision ~will have no impact on accidents or malfunctions evaluated as'the licensing basis and there is no potential for the j creation of a new type.of unanalyzed event. There is no. reduction in 1 the margin of safety.
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Page'82 of' '183 i i Safety Evaluation: 59 94-0060 Revisions 0 l
. Alternative Qualifications.for 10 CFR.50 Appendix 5 "
j m This revision to Chapter 17 of the Updated Safety Analysis Report. f F (USAR) ' allows the use of' nationally recognized atmadards which meet" { the intent of-10 CFR 50, Appendix B;.for qualifying cartain suppliers ! that no longer maintain Appendix B programs.. Nationally recognized , f f^ standards used to qualify.a supplier's Quality Assurante program will E .. be reviewed to ensure the supplier's program' meets the intent of 10 [ CFR 50 Ap;,andix B. i l This revision does not affect.any system component, structure, or j procedure required to mitigate the consequences of an accident l previously evaluated by the USAR. This revision has no affect on j equipment important to safety, because all safety-related procurements ; will be made from suppliers.with programs that have been evaluated and-audited to assure their program meets the applicable requirements of ; 10 CFR 50, Appendix B. Procured' safety-related items will be -[ controlled under programs which meet the intent of 10 CFR 50, Appendix j B. J This revision will have no impact on. accidents or malfunctions 't evaluated as the licensing basis and there is no potential for the j creation of a new type of unanalyzed event. There is no reduction in .i the margin of safety .i
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'Page 83 of- 183 i j
Safety Evaluation: 59 94-0061' Revision 0 !
- Main Turbine Drawing Discrepancies' This modification corrects a-discrepancy in Drawing M-12ACO2, "P&ID-Main Turbine." In. addition,._ Drawing M-766-02006, " Dimension Sheet i 445TS Motcr 727' Gen,",-is being superseded by Drawing M-766-01049,c ,
" Dimension Sheet 445TS Motor 727 Gen," because these drawings are duplicates. :
The change to Drawing M-12ACO2 is a document change only to reflectL j the a;-built configuration. The as-built configuration is in : , accordance with other design documents. Therefore the consequences of. , a malfunction remain unchanged. This modification will have no impact. _!
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on accidents or malfunctions evaluated as the licensing basis and- , there is no potential for the creation of a new type of unanalyzed I event. There is no reduction in the margin of safety. j i
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'Page 84 of 183 ,
Safety Evaluation: 59 94-0062 -Revision 0 > : t Shift Technica1' Advisor Training j 1 This procedurc revision involves the issuance of a new procedure, AP. ' 43-225, " Shift Technical' Advisor Training," which supersedes procedure f
, ADM 06-225, "STA-Duty Call-Technical Advisor Training." This revision f incorporates changes which create an on-shift position. This position I will be filled by an-individual who. meets the training and '[
qualification requirements of a shift Technical- Advisor (STA) . 'This l revision will provide additional engineering expertise on-shift. c [ This procedure revision.will have no impact'on. accidents or .j malfunctions evaluated as the licensing basis and there is no- i potential for the creation of a new type of unanalyzed event. .There' j is no reduction in.the margin of safety. ;
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g _ - _ . . i LAttachment to CO 95-0028-Page 85 of '183 i l Safety Evaluation
- 59 94-00'63 Revision 0 Piping & Instrumentation Update to Delete Valve Status From the '!
Drawings _ i This modification removes the locked status of valves from the P&ID j drawings. 'The status of all locked valves is controlled-by procedure' ; ADM 02-102, " Control of Locked Component Status." This; modification eliminates the possibility for discrepancies between documents as they- , relate to the locked status of valves. A note will be added to the P& ids to reference procedure ADM 02-102 for locked valve status. This is a administrative change only and involves no changes in the field. .j This modification does not affect any components, systems, or j structures. This modification will have no impact on accidents or malfunctions -t evaluated as the licensing basis and there is no potential for the l creation of a new type of unanalyzed event. There is no reduction in ' the margin of safety. ! e s l l 5 e f v r i I
Es t [-- Attachment to- CO 95-0028 Page'86 of 183' Safety Evaluation: 59 94-0064- Revision 3 0 Modification to Door Locking Mechanisms-This modification changes the locking mechanisms of watertight pressure resistant, hollow metal and woven wire doors such that the s doors can be left unlocked. Also, the manufacturer of door hardware is being changed from Corbin or Russwin to Corbin Russwin.- This: modification effects the locksets of selected doors throughout the Power. Block and Security Building, and allows the use of equiva',ent door hardware manufactured by Corbin Russwin. The modification effects the security. function of the doors only and is being made to
; allow easier access to plant areas.
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.
Attachment to CO 95-0028 L Page 87 of 183 Safety Evaluation: 59 94-0065 Revision 0 Depressurization of the Boron Injection Tank This revision to procedure STS EM-205, " Safety Injection System Inservice Valve Test," adds a step to the procedure to depressurize the Boron Injection Tank (BIT) by aligning it to the Boron Recycle Holdup Tank (BRHT). The alignment is established in order to prevent flow through the BIT outlet check valves, which could occur when the BIT has a higher pressure than the Reactor Coolant System (RCS). As a result of this procedure revision, Reportability Evaluation Request (RER) 94-018 was initiated. The concern of RER 94-018 is that this configuration could cause a potential flow diversion should a safety Injection (SI) actuation occur following a primary or secondary system pipe break. Based on review of plant systems, it was a determined that this flow path could leak as much as 20 gallons per minute to the BRHT. The 20 gallon limitation is established by a flow orifice in the flow path to the BRHT. The following is an evaluation of this concern. The creation of a path for 20 gallons per minute of high pressure water from the Centrifugal Charging Pump does not increast the probability of previously analyzed accidents. Twenty gallons per minute leakage is too small an amount to qualify as a break, and is within the margin for the SI flow to be delivered during a Loss of Coolant Accident or a Steam Generator Tube Rupture. An operator will be stationed to isolate this path during the performance of the procedure. The additional flow diversion during the execution of this procedure does not affect the consequences of failure of any equipment important to safety. No accident different from those evaluated in the Updated Safety Analysis Report are created by this procedure revision. The 20 gallons per minute of diversion flow does not approach the pump flow limits set by the Technical Specifications. This revision 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. I
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.Page 88 of '183' I l
Safety Evaluation: 59 94-0066 Revision 0 Drawing Change to Correct Component Number in Fire Protection System This modification to the special scope fire protection drawings is I [_ h editorial in nature and requires no equipment modifications or replacement. Basket strainer (KCBS0010) will be renumbered to, -l 6 (KCYS0010) .to conform with~the as installed strainer type. Drawing M- ; 12KCO2,."P&ID Fire Protection System," will be revised to show the new-; L component number, to show' correct symbology designated for a Y . [ strainer, and to show flush valve KCV0710 at the correct location on { strainer KCYS0010. . Drawing M-650-00170, " Auxiliary Feedwater Pump t Turbine," will be revised to show a 1" strainer flush valve.in lieu of -i a 3/4" strainer flush valve. Drawing M-13KC17, "PP ISO Fire b Protection Auxiliary. Building," will also be revised to show the ;
- ' correct continuation, drawing number (M-650-00170), at the ,
L owner / vendor boundary. t Because this modification is administrative in nature there will be no : ) impact on accidents or malfunctions evaluated as the licensing basis. i and there is no potential for the creation of a new type of unanalyzed ; event. There is no reduction in the margin of safety. [. r
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l l Attachment to CO.95-0028 , Page 89 of 183. i
.f Safety Evaluation 59 94-0067' Revisions 0
-Nuclear Sample Panel Tubing Modification
.This modification provides for an increase in tubing sizes and tubing !
accessory sizes inside the Nuclear Sample Panel- (SJ-144) . -This modification will diminish the sample line. flow restriction. 'This ; modification is for applications where flow control is not critical,. ,
.where small amounts of sample line drainage is acceptable, and where i viscous' samples or samples with solids / sludge are common. ;
Implementation of this modification will lessen the possibility'of' .
. block. age of sample lines in the Nuclear Sample Panel. j The tubing and tubing accessories specified in this' modification are ,
subject to the original design specification requirements. No change f to sampling techniques,'results, or test accuracy will.be incurred. l Sampling will continue to be performed within the same operating range. ,
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This modification will have no imp'act on accidents or.malfunctione f evaluated as the licensing Ptsis and'there is no potential for the' ! creation of a new type of unanalyzed event. 'There is no reduction in ! the margin of safety. , i t l l f I i
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' Attachment.to.
CO 95-0028 , Page 90 of: 183 f I Safety Evaluation:' 59 94-0068- 1 Revision O. l Source and Intermediate' Range Detector Replacement- ] This modification to safety-related equipment in the Ex-core Neutron' { Monitoring System will upgrade the intermediate and source range l detectors with a replacement unit. The replacement unit'has better { noise reduction capabilities. Therefore,;the replacement should
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reduce the probability of-equipment failure. ;This is a-like for like _ i change out of an improved model from the original supplier and has !
- been field proven at other' plants for over three years. El Based on the above discussion, this modification will.have'no-impact Ji on accidents or malfunctions evaluated as the licensing basis and i r there'is no potential for the creation of a new type of unanalyzed ;
event. There is no reduction in the margin of safety. ! I
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1 Attachment to CO 95-0028 Page 91 of 183 3 1 Safety Evaluation: 59 94-0069 Revision:0 Emergency Diesel Generator Engine Injection Cooling Line Removal This modification to the safety-related Standby Emergency Diesel Generator Engine removes the fuel oil injection nozzle cooling water lines. This change has been demonstrated by the vendor (Colt-Pielstick) to be safe and acceptable. This change has no adverse ; impact on the safety-related design basis function of the Emergency Diesel Generator. The removal of the cooling water line from the injection nozzles will decrease the potential for a failure of the Emergency Diesel Generator, when called on to operate, by eliminating the potential for fouling of the cooling water tank or. failure of the cooling water pump. Based on discussions with Colt-Pielstick, this change will reduce the operating life of the Emergency Diesel Generator from approximately 120,000 hours of operation to approximately 9,000 hours of operation. The Emergency Diesel Generator is not expected to operate more than 6,000 hours for the life of the Wolf Creek Generating Station. The elimination of corrosion and scale concerns will improve the reliability, availability, and operability of the Emergency Diesel Generator. 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.
l Attachment to CO 95-0028 p Page 92 of 183 Safety Evaluation: 59 94-0070 Revision 3 0 Nuclear Sampling System Liquid Over-Pressure Return Line This modification to drawings for the Nuclear Sampling System revises drawings M-12SJ01, "P&ID Nuclear Sampling System," M-12SJ04, "P&ID Nuclear Sampling System," and J-352-00006, "P&ID Sample Panel," to reflect the as-built configuration and provide references to other drawings and notes to clarify system operating characteristics. This modification will show a liquid over-pressure return line connecting to the weste holdup tank header. These changes to the above mentioned drawings are editorial in nature and have no effect on installed equipment. 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.
t F ' Attachment to CO 95-0028 Page 93 of 183 i Safety Evaluation: 59 94-0071 Revision 0 Operations Department Organizational Changes This revision to the Updated Safety Analysis Report (USAR) reflects organizational changes in the Operations Department. Two Shift Supervisors have been reassigned. Three new Shift Supervisors have been appointed. The position of Administrative Supervisor reporting to the Manager Plant Support has been removed. The changes made in this revision represent organization changes, changes in reporting relationships, and personnel changes. These changes improve the operating philosophy of the Wolf Creek Generating Station. These changes do not affect the safety and reliability or operation of the plant nor represent any decrease in the concern for the health and safety of the public. This revision does not affect any system, component, or procedures required to mitigate the , consequences of an accident previously evaluated in the USAR. This revision has no effect on eg'tipment important to safety. The newly appointed Shift Supervisors meet the requirements established by the NRC and Wolf Creek Nuclear Operating Corporation. Based on the above discussion, this revision 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. I I 1 i i I
C .C si ?Z , _ _, i ut H _. . LAttachmentito .CO 95-0028 < ; Page.94'of~ 183 i i f L . . i' l'i Safety Evaluations- 59 94-0072 Revision 0 - E . . - - New Telephone Systeen , This revision to the Updated SafetyLAnalysis Report'provides for the replacement of the. plant-to-offsite communication touch-tone telephone ,;
- system with a new model and brand. The new' system will have the same- l i functional characteristics as'the telephone system. The new telephone :!.
switch will replace all existing telephone ~ switches which are ! currently located in'the' Administration Building,' construction
-Administration Building, and the Learning Center with a new main ;
. telephone switch located on the second floor.of the Administration '!
p Building and a satellite hub at the Learning' Center. } l This revision 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- i
.the margin of safety.
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Attachment to- CO 95-0028- I! Page 95 of' 183 *
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. Safety' Evaluation: 59 94-0073 . Revision 0 l t
Removal of Internals Fram Instrument Air Check Valve ! This temporary' modification provides.for the removal of internals from-check valve >(KA-V772) to allow-system air pressure to flow back to the
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i . . Instrument and Service Air. Compressor,' internal air ~ discharge check l valve. This temporary modification.will allow the air outlet' pressure
! sensor (CKA01A) to accurately measure system pressure and.thus enable i the compressor'to properly respond to system pressure changes. The 'I safety-related portions of'the Instrument and' Service: Air System'are l completely independent of the check' valve being modified. .i H
This modification will have no impact on accidents or malfunctions evaluated as the licensing basis and there is no potential-for:the l creation of a new type of unanalyzed event. There is no reduction in' ~ the margin of safety. t i I l 1 f
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i Attachment to CO 95-0028 Page 96 of 183 Safety Evaluation 59 94-0074 Revision 0 , Modification to Cycle 8 Fuel Assembly Design This modification to Region 10 (Cycle 8) fuel assembly design adds a protective bottom grid, extends the length of end plugs, relocates grids 1 and 2, incorporates a variable pitch plenum spring, rotates the mid-grid in combination with intermediate flow mixers,-and adds a high burnup bottom grid. The mechanical changes will not increase the , probability of occurrence of an accident previously evaluated in the Updated Safety Analysis Report (USAR). The clad integrity is maintained and the structural integrity of the fuel rods, fuel assemblies, and core is not affected. The mechanical features have no impact on fuel rod performance or dimensional stability nor will they cause the core to operate in excess of design basis operating limits. Flow parameters within the reactor vessel are unaffected. Changes to the fuel assembly do not result in any change to plant equipment. The probability of all loss of coolant accidents (LOCA) in the USAR has not been increased. Adherence to applicable standards and criteria precludes new challenges to components and systems that could adversely affect the availability of existing components and systems to mitigate the consequences of any accident or adversely affect the integrity of the i fuel rod cladding as a fission product barrier. Furthermore, adherence to applicable standards and criteria ensures that these fission product barriers maintain the design margin of safety. The radiological consequences of all accidents previously evaluated in the USAR are not increased. No new performance requirements are being imposed on any system or component such that any design criteria will be exceeded nor will they cause the core to operate in excess of design basis operating limits. ; No new failure modes or limiting single failures have been created. The functions of safety-related systems are unaffected. Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR has not increased. The evaluation of the affects of the use of these fuel features show that the acceptance criteria for all accidents described in USAR Chapters 6 and 15 as well as the criteria contained in 10 CFR 50.46 continue to be met. Therefore, the consequences of a malfunction of I equipment important to safety have not been increased. j Based on the above discussion, the possibility of an accident of a j different type than any previously evaluated in the USAR has noc been l created. Also, the possibility of the malfunction of safety ~7uipment l different from that evaluated in the USAR will not be created. Plant operation, with the modified grid elevations, does not adversely i j l l
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+ i+ _ Page 97 of - 183' affect the fuel assembly structural response. The' margin of, safety as. i
' defined in the Technical' Specification Bases is not reduced in'any. i Chapter 6.and'15 non-LOCA' accidents. The non-LOCA safety analysis' ?
acceptance criteria continue'to be met. Further, the LOCA evaluation , demonstrates that all 10 CFR 50.46 criteria continue'to be met. i
^" This modification will have no. impact.on accidents or malfunctions; ;
- evaluated as_.the licensing basis and there is no potential for the. j creation'of.a new type of unanalyzed event. There is no reduction in }
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U. Attachment to- 00-95-0028'
-Page_98 of-l 183
. Safety Evaluation: 5994-00h5 . Revisions'O Installation of Secondary Personncl Access Facility to Protected Area This modification provides-a Seconda*;y Personnel Access Facility which y 'will enhance' ingress to and egress from.the Protected Area Boundary.
The Secondary Personnel Access Facility will be located on the north
' fence line of' the Protected. Area Boundary, just east of , the Construction Administration Building. This facility will contain equipment to provide the required access screening prior to granting access to personnel entering the protected area. In addition, intrusion detection hardware will be modified to provide the required detection of unauthorized access to.the protected area. 'The Secondary Personnel Access Facility does not connec+. to, nor does it affect any safety-related system, component, or structure.
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 co 95-002B. ! cr ; Page 99 of .183 ;
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a ; l' i !, ' Safety Evaluation 59 94-0078. Revision:0 : L: l , Removal of Rod Control Disable Function on Loose Parts Monitor I i :~ I
. This modification removes the disable function'of-the Loose Parts :
. Monitoring System'on control rod motion demand. The disable function ;
on the loose parts monitor provides a momentary override for any loose ~! parts alarm. The Loose. Parts Monitoring System remains operable with l
.this modification. This modification will remove the disable j function,' which will increase the possibility for loose parts alarms. -l
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'This modification will have no: impact on accidents.or malfunctions- j 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 j the margin of safety. ,t i i 1 i
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p Attachment to CO 95-0028 Page 100 of 183 Safvaty Evaluation: 59 94-0079 Revision 0 Plant Public Address System Upgrade This modification to the Plant Public Address System involves the instaA:stion of supplemental wires to ensure design voltage is maintainti. In addition, selected speaker amplifiers are being modified to inprove their alerting capability for Evacuation Alarm broadcasts. This modification involves improvements and enhancements to the Plan
- public Address System. No safety-related components, systems, or structures are affected by this modification.
This revision :*ill 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 j the margin of safety. I i l l l
f,, , Attachment to CO 95-0028 Page 101 of 183 Safety Evaluation: 59 94-0080 Revision: 0
-Thermal Expansion Monitoring Program Noise' Monitoring Equipment Installation
~This modification provides for the installation of noise monitoring
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. 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 perforni no design function. 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 ROS. No new failure modes are created by this installation. The capabilities of safety-related plant equipment to perform their required safety function is not affected by this modification. This modification does 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- -CO'95-0028 f n f
%. Page 102'of 183 i
Safety Evaluations. 59 94-0081 Revision 0 Updated safety Analysis Report Revision to Correct Pipe Connections
'This modification revises Up dated Safety Analysis Report, Figure 9.4- l 4, Sheet 2, to be consistent with revised drawing M-12GE02, "P&ID ,
" Turbine Building HVAC System." The revised figure reflects the correct !
V arrangement of two pipe. connections to the Turbine Buildng HVAC System ;l ventilation' duct (GE- 07 9-5NL-14 ) .. Unrelated drawing errors were
' discovered during close-out of Plant Modification Request 03811. l-This modification will have.no impact on accidents or malfunctions- l 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. l l ( E t
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5 Attachment to CO 95-0028 Page 103 of 183 Safety Evaluation: 59 94-0082 Revision 0 Radwaste Storage in New Radwaste Storage Building This modification provides for the storage of radioactive waste in the. new Radwaste Storage Building. A previous evaluation provided justification for the storage of any fixed or internal contamination item such as contaminated materials, equipment, tools, scaffolding, portable units, and protective clothing. This justification is based on the engineering judgment that any fixed or internal contamination which would not become loose during transit or maintenance, is not expected to create an unacceptable radiological condition in the Radwaste Storage Building. This modification expands the use of this building to cover any item with loose or potentially loose contamination. Because the Radwaste Storage Building is not built to the criteria of power block structures, it is necessary to assume failure of the structure as a result of natural phenoinena such as a tornado, earthquake or fire. The administrative limit on the total curries of radioactive materials permitted in the building will limit potential doses to a small fraction of the 10 CFR Part 100 exposure limits. Because appropriate administrative controls and procedures are in place, Wolf Creek Nuclear Operating Corporation meets the intent of regulatory requirements for preventing, mitigating, or controlling inadvertent or accidental release of radioactivity to the environment. The storage of waste in the Radwaste Storage Building 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.
J i P . t Attachment to CO 95-0028 l Page 104 of :183 1! Safety Evaluation: 59 94-0083 Revision 0 ' Component Identification correction for Condensate Domineraliser System This modification corrects valve number duplication'in drawings ~for ! the Condensate Demineralizer System and assigns a unique number to the t [ valves. Component number AK0967 appears.twice on the Piping and F Instrument Diagrams.for the. Condensate Demineralizer System. Drawing [ M-12AK01, "P&ID Condensate Demineralizer System," identifies AKV0967 [ as a 1/2" ball 1 valve which. serves as the bypass isolation for the cervice inlet. valve (HV022A) to.the condensate demineralizer t'ank (TAK01C). Drawings M-12AK02, "P&ID Condensate Demineralizer System," t
.and M-13HF09,." Piping Isometric Drawing," describes valve AK0967 as a !
l 1/2" plug valve' located in the flush line~that branches off of the 1" ! line downstream of isolation valve AKV0010. A new component identification number (AKV1010) has been assigned for the 1/2" plug valve located in the flush line. This modification revises. drawings M-12AK02 and M013HF09 to reflect the new valve identification number. In addition, this r.cdification revises drawing , M-12AK02 to show the reducing insert on the 1/2" line upstream of AKV1010.that was inadvertently left off of the-drawing because of a drafting error. i This modification is an administrative change and will have no impact ; on accidents or malfunctions evaluated as the licensing basis and i there is no potential for the creation of a new type'of unanalyzed event. There is no reduction in the margin of safety. t
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i l 1 Attachment to CO 93-0028 Page 105 of 183 i Safety Evaluation 59 94-0086 Revision 0 Drawing Correction for the Makeup Domineralizar System This modification () corrects discrepancies between the drawings and the as-built configuration. Drawing M-0025, Sheet 4, "P&ID Makeup Demineralizer System" does not show the exact lineup of the connection points for branch piping lines. In addition, a valve is installed in the system on line IWM077C-3/4 that is not indicated on the Piping & Instrument Diagram. A system walkdown verified the installed valve is a 3/4," stainless steel globe valve that is manufactured by Dragon Valves, Inc. Component identification number 1WM1285 has been assigned to the 3/4" valve. The function ~of this valve is to provide isolation of line IWM077C-3/4 which supplies the 1/4" lines which supply the sample sink and silica analyzer racks. This modification revises the drawing M-0025 to reflect the as-built configuration. 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' 00 95-0028 .
- Page'106 of' 183 l
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Safety Evaluations- .59 94-0085 Revision 0 '!
' Revision:t'o General Mmployee Training Procedure '
B - -This. revision tolthe' procedure 7DM 06-200 " General Employee' Training," I L changes the procedure number to AP-30A-001. This~ revision changes , E . training course' definitions to be-more consistent with INPO. guidance ; L' " contained in'ACAD'93-009, " Guidelines for General Employee' Training." ij In' addition, this revision adds a provision.to allow unfettered access 1 h to the protected area for' selected personnel. , The changes in this revision are administrative in nature. This , -- revision will have no impact on accidents or malfunctions evaluated as ! the licensing basis and there is no rotential for~the creation of a. ,; L new type of unanalyzed event. There cs no reduction in the margin of l safety. , j
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. . Attachment to CO 95-0028 Page 107 of 183 k- '. -
l Safety Evaluation: 59 94-0086- Revision 0 Drawing Correction to the Service Gas System Piping and Instrument , Diagram ' 1This modification revises drawing M-12KH01, "P&ID Service Gas System," ;
'to clarify that valve KHV0309, (Condenser. Nitrogen Blanket Relief
-Valve,) serves.as a relief valve rather than a check valve. System walkdown has shown that the installed'KHV0309:is a swagelok, poppet check valve. A notation will be added to drawing M-12KH01 to clarify. ;
the' application as a vacuum relief vent check vaAve. This .l modification is administrative in nature and does not affect any components, systems, or-structures.
'ntis modification will have no impact on accider.ts or malfunctions '
evaluated as the licensing basis and there As no potential for the creation of a new type of unanalyzed event. There is'no reduction in . the margin of safety. ' 4
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"' 8 , ..s Atrachment to- CO.95-0028 g Page 108 of: 183' j'
Safety Evaluation: 59 94-0087 Revision 0 $ Addition of-Pressure. Gauge in the Residual Heat Removal' System O - Thisfmodification provides for the: installation of a pressure gauge.. l using.the. existing process connection (high pressure side) of the flow ! transmitter .EJFT0619 ~ (Residual Heat. Removal Heat Exchanger "1B" to , Accumulator Injection Line). This modification was' initiated as .j ~ corrective'actionLfor relief valve bellows failures'of Residual Heat
. Removal Heat Exchanger "1A" and ~"1B" Outlet Check Valves' (EJ8730A/B);
- This modification provides hardware to monitor the pressure on the.
.i bellows. The' relief valves.have been installed to protect the system
' t from over pressure because of thermal expansion of fluid in the pipe. i
.In addition,' leakage'through check valves from the Reactor Coolant; j System'could pressurize the system. Therefore, it is necessary to monitor the system pressure at the relief valve onLa daily basis. If j the pressure exceeds a preset value, the' pressure can be relieved. zl without lifting the relief valve.
This. modification installs the' safety-related pressure gauge which '~ j will be used to monitor pressure on the bellows as discussed'abovei The design, materials, and pressure gauge are all qualified for safety- 't related application. .The design is stress analyzed and the calculated i stress is within the allowable limit. l 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. . i P
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' Attachment to' CO 95-0028' Page 109 of 183.
_ Safety Evaluation L 59'94-0088 ' Revision:0
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' Modification to Provide for Main Condenser Tube Cleaning This modification provides a " hot tap" connection in the circulating Water System piping to allow for on-line mechanical condenser tube ,
cleaning. The " hot tap" connection will be provided in:the discharge
- piping from the "C". Circulating Water Pump and will consist of'a four. 1 inchiweldolet, a' weld neck flange, a flanged full port gate valve, and a blind flange. .This assembly is welded to the outside of the l pressurized Circulating Water System piping.
Precautions are provided to assure that extensive tube plugging will- .l' not'take place and to avoid making the condenser unavailable for a p nteam dump and causing a-turbine. trip. .The Circulating Water System ,; serves no safety-related function and-has no safety design basis. ,The installation'and operation of the condenser tube cleaning system will have no effect on any safety-related system structure on component. This modification will have no impact on accidents or malfunctions
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evaluated as the licensing basis and there is no potential for the l creation of a new type of unanalyzed event. There is no reduction in 1 the margin of safety. ! 3 t I l I i l 4 1 l l I i
i l Attachment to CO 95-0028 Page 110 of 183 Safety Evaluation: 59 94-0090 Revision: 0 Modification to Canopy Seal Clang Assamblies This modification provides the necessary documentation to install
- canopy seal clamp assemblies on any spare Core Exit Thermocouple (CET) or active penetration in order to encapsulate the canopy seal weld.
Wolf Creek Generating Station (WCGS) has previously experienced leakage at canopy seal welds on spare Control Rod Drive Mechanism (CRbM) penetrations #24, #25, and #29 on the reactor sessel head. The leaking canopy seals were repaired by installing canopy seal clamp assemblies. Plant Modification Request 04230 installed CSCAs on penetrations #24 and #25. Plant Modification Request 04506 installed a canopy seal clamp assembly on penetration #29. The reactor vessel closure head contains 78 penetrations. Each head adapter flange has an identical stainless steel flange welded on the top of the penetration. The stainless steel flange has male ACME threads (to mate with an attachment) and a canopy lip. The attachments determine the type of penetration. The attachments are as follows: 13 head adapter plugs, 4 female flanges, 8 part length CRDMs, and 53 full length CRDMs. Each of the attachments has a female ACME thread and a canopy lip. The head adapter is designed such that when the attachment is threaded onto the stainless steel flange (at original construction) , the two canopy lips come together and are seal welded. This seal weld is required because ASME Section III states that threaded joints in which threads provide the only seal shall not be used. Therefore, the canopy seal weld was provided to seal the ACME threaded pressure boundary connection. It is important to note that the ACME threads of the threaded joint are the load caring part of the joint design. The canopy seal weld is designed to retain the pressure of the reactor coolant while the strength of the joint is maintained by the ACME threads. The canopy seal clamp assembly consists of a top plate, housing, tensioning studs and a Grafoil seel carrier. The top plate is positioned over the top part of the attachment and the head adapter. The seal carrier is positiened over taa seal weld area of the canopy seal and the housing surrounds the seal cc.rrier. The tensioning studs arc installed between the top plate and the housing, and are hydraulically compressed so that the pressure o7 the Grafoil seal is greater than the internal Reactor Coolant Systed fluid pressure to prevent / terminate leakage in the seal welded area. The Grafoil seal carrier, which contacts the outside of the weld, ir rot degraded by the effects of radiation. The canopy seal clamp assemblies are designed and fabricated in
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I Attachment to CO 95-0028 Page ill of 183 accordance with the requirements of ASME Section III, subsection NB, 1986 Edition. The vendor has provided a code reconciliation for the t canopy seal clamp assemblies back to the original construction code year of 1971 Edition through Winter 1972 Addenda. The canopy seal c1 cmp assemblies have been qualified for a 40 year life under all , normal and worst case accident environmental conditions. All stresses on the reactor vessel head and CRDM penetrations created because of the addition of the canopy seal clamp assemblies have been analyzed ; under all postulated conditions, including a Loss of Coolant Accident and seismic conditions. All stresses remain within code allowable limits. 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. i e
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' Attachment.to- CO.95-0028 'l Page 112 of 183
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f Safety Evaluationt 59 94-0091' -Revision 0 + t Rod. control slave Decoder Card Diode Repositioning { This modification'to the non~ safety-related rod control' slave decoder- 3 card diode positions is being performed in accordance with. ; Westinghouse Bulletin NSD-TB-94-05-RO " Rod Control -.CRDM Timing l Change" and NRC Generic Letter 93-04 " Rod Control System Failure and i
-Withdrawal of Rod Cluster Assemblies." 'The modification involves .!
repositioning of diodes on Rod Control Slave Decoder Cards to obtain .; improved current timing. 'This modification is being performed to preclude possible asymmetric r'od withdrawal effects when_ failures occur as observed at the Salem plant. This modification also provides1 jl for one-time acceptance testing in accordance with the outline in l Westinghouse Bulletin NSD-TB-94-05-RO to demonstrate this modification does not affect normal Rod Control System operation, j The timing changes involved with this modification are designed only ; to reduce the probability of asymmetric rod withdrawal in the event of ; certain specific timing circuit failures. Repositioning of the diodes does not adversely _ affect any aspects of the Rod Control System that j are accident related. This change does not degrade the. function or- 1 design of the decoder cards. i This modification will have no impact on accidents or malfunctions I evaluated as the licensing basis and there is no potential for the l creation of a new type of unanalyzed event. There is no reduction in the margin of safety. l I Y f 6 4 b
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' Attachment to CO 95-0028 i i
Page 113"of 183 j L- ! Revision 30
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Safety Evaluation ' 59 94-0092 ! Revision 41 to Radiological Emergency Response Plan This revision to the Radiological' Emergency Response Plan provides'for
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I the combining of the Technical- Support Center (TSC) and the Operations' l Support Center '(OSC) into one location in the TSC. This change. allows l better communications between the TSC and OSC staff. Also, the' F getc Emergency Operations Facility Engineering Team has been combined'with v the TSC Engineering Team and will report to the TSC. This change !*
' allows all of the Engineering Team to be located in one area, which
.will produce an_ environment conducive to enhanced teamwork and improved information flow. In addition, the Wichita Phone Team is-being relocated to the Learning Center because the Wichita office has ,
been moved to the site. This revision will have no impact on accidents or malfunctions j 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 l the margin of safety, ! t
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4 4 . I Attachment to' CO 95-0028 i-, i Page 114.of 183 , t Safety Evaluation 59 94-0093- -Revision:0 l Drawing Correction ~on the Piping and Instrument Diagram-(PEID) for the. j
. Containment Cooling system This modification revises drawing M-12GN02 "P&ID Containment Cooling i system" to reflect the as-built configuration of the dampers for Cavity Cooling Fan "2B" Backdraft Damper (GND0081) and Cavity Cooling Fan "2A" Backdraft Damper (GND0082). These dampers were transposed at installation. This modification is a drawing change only and will not ;
affect equipment, components, or structures. This modification will have no impact on accidents or malfunct!.ons ; evaluated as the licensing basis and there is no potential for the .j creation of a new type of unanalyzed event. There is no reduction in the margin of safety, i I h r b i e t i t l i e I l 1 l
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.l Page 115'of 183 j i
t Safety Evaluations 59-94-0094 Revision:D-Drawing' Revisions to Reflect the Condensate System As-Built, f . Configuration' This modification incorporates the as-built configuration of the
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i condensate System into the appropriate design drawings. These drawing j are'M-12AD03, "P&ID Condensate System," M-13AD07, " Piping Isometric j Turbine Building," and M-15AD07,." Hanger Location Drawing." The "[ drawing revisions involve pipe-spool and' hanger configurations. -This ! modification is administrative in nature and does not affect any j equipment, components, or structures in'the plant. ; i This modification will.have no impact on accidents or malfunctions i 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, j l l i
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I 1 Attachment to' CO'95-0028-Page 116-'of 183- l
, Safety Evaluations 59 94-0097 Revision 0 : -
Cheadcal and volume Control' system' Alignment for Normal Operations ,
- This ~ revision to the Updated Safety Analysis ' Report . (USAR) provides l clarification for specific operator actions'taken during an inadvertent safety injection actuation incident. 'The current wording. '
of this section does not. clearly indicate. alignment of the chemical ! Jand Volume control System' (CVCS) for normal charging. "The revision j states-that the operator stops the'high head Safety Injection Pumps i l-
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and aligns the charging system for normal operation. The existing. j plant procedures include the necessary steps to align the CVCS for ; normal charging following an inadvertent safety injection actuation, j This revision will provide consistency with existing plant procedures. !
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This revision will have no impact on accidents or malfunctions evaluated as the licensing basis and there.is no potential.for the j creation of a new type of unanalyzed event. There is no reduction in'- j the margin of safety. .:
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Page 117 of 183 Safety Evaluation: 59 94-0098 Revisions O' Migh' Pressure Injection System Isolation Valve
. Revision 3 to Plant Modification Request .(PMR) 04145 includes newi
, actuator thrust settings for long term use for motor operated valve EM.
HV8924 " Residual Heat Removal Heat Exchanger "A"/ Chemical Volume control System to Safety Injection. Pump A Upstream Isolation Valve). In additioni this revision clarifies the breaker status for EMHV8924
'in Updated Safety Analysis Report Figure 6.3-1. . Valve EMHV8924 is.
administratively controlled to remain in the safeguards position (open) when the Emergency Core Cooling System function is required.
- Previous' revisions to this modification have been previously evaluated Unresolved Safety Question Evaluation 59 92-0002.
I This revision 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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Page 118 of- 183 [,,.. Safety' Evaluation: 59 94-0101' ' Revision 1
' Temporary Procedure for Turbine Generator Torsional' Response' This temporary procedure, TMP EN-169 " Turbine Generator' Torsional Response Test," provides for.the performance of the turbine generator ~
torsional response testing to determine the natural frequencies of torsional vibrations. Turbine generator torsional response testing is ! being performed to measure frequencies at which the turbiae generator -t is subject to resonance.~ Frequencies near 120 hz are of concern. ; A torsional stimulus will be applied to the: unit by introducing a i negative sequence current within design limits. This will be done by-shorting one of the phases downstream of the generator switchyard ; breaker and applying a very low excitation current. Data will also be - collected during turning gear engagement and during normal synchronization to the grid. 1 The accident analyses'in Chapter 15 of the Updated Safety Analysis { Report'(USAR) have been' reviewed and it has been determined that the performance of this test'will not increase the probability of . i accidents evaluated.by the USAR. f Revision 1 to the Safety Evaluation, summarized and reported in USQD 94-0101, Revision 0, provides a more detailed analysis of the l probability and consequences of a turbine / generator overspeed event. l The results of this analysis provided no changes in the conclusions '{ described in Revision O. 1 This revision will have no impact on accidents or malfunctions l 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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;p , Attachment to CO 95-0028 Page 119 of 183 5
! 'l, safety Evaluation: 59 94-01'03 Revision 3 0' !
-i Drawing Revision to Correct Chemical and Volume Control System Valve. .;
status i This modification corrects the valve status of Chemical-and Volume. {
' Control System valves on drawing M-12BG05, "P&ID Chemical and Volume' [
. Control System."- Sampling system isolation valves (BG-V0330 and BG- .j V0329)'and return isolation valves.(BG-V0349 and BG-V0348) are '
incorrectly shown on drawing M-12BG05 as being.normally closed. This e modification will revise M-12BG05 to show these valves as normally i open. The resolution of this drawing-discrepancy is an administrative -! change. - l L 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 l the margin of safety. l t I t t f i i i l i 3 l
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'i Page 120 of' 183 -;
i (, Safety Evaluations- 59'94-0104 Revision 0. ! N i
~ Personnel Title' Changes i
;This revision.to the Updated Safety Analysis Report - (USAR) changes the titles of Wolf Creek Nuclear. Operating Corporation personnel. Thel -{
. positions of Manager Chemistry, Manager Security, and Supervisor !
Operations Support are being changed to reflect the title of
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Superintendent:in their respective areas'. In addition,' the titles of l Superintendent of Operations and Plant Manager'in Section'1B of the. 'i USAR are Deing updated to reflect current titles. { The changes made in this'revisi6n represent' organizational changes. l
-only. Reporting relationships are unchanged by this. revision. All' ;
functionsEcontinue to be fulfilled. This revision'does not affect any ; system, component, structure or procedure' required to mitigate the '
.. consequences of an accident previously evaluated in:the USAR. This :
, revision is administrative in nature'.
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This revision will have no impact on accidents or malfunctions. j 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 : 'I' the margin of safety. t i a e i 4 h
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'Page 121 of. '183 .
j 2 N Safety Evaluation: 59 94-0105 Revision:0- .;
-Fire Rxtinguisher Relocation'in the Radweste Building l This modification provides.for the relocation of a fire extinguisher. ,
from Room 7225 in the Radwaste Building to another location in the ;
-proximity to provide more' accessibility. This fire extinguisher will be available to Fire Brigade personnel and Station personnel accessing-
' Room 7225. This modification is not a reduction in. fire protection- ;
for Room 7225 or'for the Radwaste Building. ! This modification will have no impact on accidents or malfunctions .. L evaluated as the licensing basis and there is no potential for the { L creation of a new type of unanalyzed event. There is. no redt : tion in ! the margin of safety. ' ;
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LAttachment'to' fCO 95-0028 i - Page 122 of 183 Safety Evaluations' 59 94-0106 Revision 0' Temporary Modification to Provide Temporary Power to Security Systems This temporary modification provides temporary power to Security Systems while maintenance activities are being performed on inverter XSK01A. 1This temporary power will be supplied by a portable power-pack connected to a welding receptacle in the Auxiliary Building. This temporary power will allow security to maintain redundant systems-capabilities. . This_ change does not' affect any accident analysis evaluated in the Updated Safety Analysis Report. This revision 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 'CO 95-0028 l-Page'123 ofJ . 183- t f
Safety Evaluation 59-94-0107 Revision 0 I t
= Standby Emergency' Diesel Generator' Drawing Correction ;
This modification revises drawing M-12KJ02(Q), ." Standby Diesel '!
' Generator "A" Intake, . Exhaust, & Starting Air System," to correct' l duplicate valve numbers and to correct a drawing discrepancy .)
concerning pipe class. Errors in the drawing occurred at the last: l revision during Computer Aided Drafting restoration and'are.not i
. associated with any design change. These changes are administrative j
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' This modification will have no impact on accidents or' malfunctions
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; i evaluated as the licensing basis and there is no potential for the l i
creation of a new type of unanalyzed event. There is no reduction in the margin of safety, f; I
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. Safety Evaluations: 59 94-0109 Revision:0 L!
i F Turbine Generator Bearing Fire Protection Modification l t This modification to non safety-related equipment provides for i changing the turbine generator bearing fire protection system from a .
. manual water spray' system.to a fully automatic pre-action system. The.
pre-action system'shall be located in a separate fire area and can be .
. actuated without requiring operator action if the' area is filled with _i
. fire or smoke. This modification is being implemented to comply with {
p an.American Nuclear Insurers recommendation. ; i j This modification enhances the fire protection program. There are no previously evaluated accidents which would be' impacted or initiated by .r the proposed modification or by a failure of the automatic pre-action- l
, system. The automatic pre-action system does not serve any safety- j related function. The pipe and components installed by this. )
modification are not in an area which would effect any equipment ~ important to safety.
. This modification will have no impact on accidents or malfunctions j
evaluated.as the licensing basis and there is no potential for the .lt creation of a new type of unanalyzed event. There is no reduction-in the margin of safety.' i t l
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~ Attachment to 03 95-0028 Page 125 of 183 ;
i [ Safety Evaluation's 59 94-0110 Revision 0 - .{ i
" Jockey Fire Pump Discharge Orifice Resizing a%d Pipe Material
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, Replacement ; I
.t E This modification to the non safety-related jockey fire pump discharge i backpressure flow orifice reduces the opening size of the orifice. [
The static system pressure will be increased from 80 pounds per square inch (psi) to 125 pai. The orifice opening is being reduced to i compensate for the increase in pressure drop across the orifice. In ; addition, this modification will improve the cavitation which has been .{ experienced in the spool piece and elbow immediately downstream of the- , orifice. This spool piece and elbow is carbon steel and is being' s replaced with stainless steel which is more wear and erosion i resistant. This change affects only the non safety-related portion of f the fire protection system. ; This modification will have no impact on accidents,or malfunctions j evaluated as the licensing basis and there is no potential for.the { creation of a newl type of unanalyzed event. There is no reduction in ! the margin of safety. . t I i i i t r [ f I
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. Page 126.of; 183 j
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Safety Evaluation: '59 94-0111 Revision 3 0 ! Revision.24 to the Security Plan. ; a
- This revision to the Security Plan represents administrative changes' .. i and changes to security requirements. This revision does not affect !
the safety, reliability or operation of the plant. This revision does -l not' affect-any system, component,.or procedures required to mitigate
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- theLconsequences of an accident previously evaluated or to operate the ,j plant. This. revision does not affect any equipment'important'to :l safety. Details of Revision 24 to the Security Plan are Safeguards l Information and are available for review at Wolf Creek. Generating [
station. 'l This revision will have no impact on accidents or malfunctions evaluated as the licensing basis and there is no potential for the' l creation of a new type of unanalyzed event. There is no reduction in i the margin of safety. l
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- Attachment to CO 95-0028 ,
i Page 127 of 183 S-Safety Evaluation 59 94-0114 RevisionsO. Chemical Degassing of the Reactor Coolant System Using Rydrogen. [ Peroxide t This modification provides for the chemical degassing of the Reactor j Coolant System (RCS) using hydrogen peroxide. The addition of hydrogen peroxide into the RCS will have no impact on operability or ! The. addition of hydrogen peroxide into the RCS '
' integrity of the RCS.
will have no impact on the RCS components. 'The chemical degassing guidelines do not conflict with the plant shutdown procedures nor do they conflict with the shutdown chemistry operations. :When the. - correct dosage of hydrogen peroxide is added to the coolant, there will not be any excess oxygen accumulated in the gas spaces in the j RCS. However, if an excess amount of hydrogen peroxide.is added, the precautionary measures in procedure ADM 04-021 " Chemistry Surveillance During Refueling Shutdown" provide a means of preventing a flammable ; or explosive mixture of hydrogen and oxygen. Therefore, the . probability of an accident is not increased by this modification.
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This modification does not affect the integrity of the fuel assemblies ; or the reactor internals. In addition, chemical degassing does not1 affect any fission product barrier. The modification does not change, degrade or prevent the response of safety-related mitigation systems to accident scenarios as described in Updated Safety Analysis' Report , (USAR). There is no affect on the mitigation of the radiological ; consequences of an accident described in the USAR. Therefore, the consequences of an accident previously in the USAR will not be , increased. The chemical degassing process is consistent.with the plant shutdown procedures and the shutdown chemistry operations. Chemical degassing does not cause the initiation of any accident nor create any new failure mechanisms. This modification does not create the possibility of an accident different than any evaluated in the USAR. , This modification does not result in an increased probability of ,
. scenarios previously deemed improbable nor does it create any new l failure modes for safety related equipment. LThe chemical degassing of !
the RCS will not increase.the probability of a malfunction of , equipment important to safety previously evaluated in the USAR. This modification does not result in a different response of safety-related systems and components to accident scenarios than those ; postulated in the USAR. No new equipment malfunctions have been ~l 6 introduced that will affect fission product barrier integrity. This modification introduces no new or additional equipment. Chemical ! degassing does not create any new failure modes that could adversely impact equipment important to safety. This modifs tion will not ; I increase the consequences of a malfunction of equipment important to i I
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'Page 128 of 183
' safety previously' evaluated in the USAR.- The possibility of a
- i. malfunction-of equipment important to safety of a different type.than' -
., ;previously evaluated'in the USAR is not created.
Chemical' degassing'will have no affect on the availability, operability, or performance'of safety-related systems.and components. Therefore, the proposed modification will not reduce the margin of' safety as described in the bases.to any Technical Specification. m l.; l' t
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~Page 129 of . 183 Safety Evaluation: 59 94-0116 RevisionsO
-Installation of Rosin Sample Valve _
1l: This modification to non safety-related equipment provides for_ thel j _ installation of a resinLsample valve in'the resin transfer line. .. This j will' allow samples _to be taken for evaluation of resin condition and- i effectiveness of the regeneration process. This modification will, j involve the replacement of a 90 degree elbow with a tee and sample i valve, j This modification will have no impact on accidents or malfunctions ; evaluated as the licensing basis and there is no potential for the l' creation of~a new type of unanalyzed event. There is no reduction in the margin of safety. j r i f I
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,. Attachment'to CO 95-0028 Page 130 of 183 safety Evaluations' 59 94-0117 Revision:O
. clarification of Hydrostatic Testing Requirements This modification revises the Updated Safety Analysis Report (USAR) for clarification regarding hydrostatic testing requirements of Quality Group "D" (Augmented) piping systems ~and for consistency with ANSI B31.1. Table 3.2-5 of the USAR is being changed to reflect the
. alternative approach to hydrostatic testing allowed by ANSI B31.1. In addition, the Wolf Creek Nuclear Operating Corporation position to Regulatory Guide 1.143 " Design Guidance for. Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-cooled Nuclear Power Plants" is being clarified.
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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V t . . l [ , Attachment-to CO 95-0028 l t Page 131'of 183 i i Safety Evaluations. 59 94-0118 Revision 0 ~ ! Temperature' Indicating Controller Setpoint' Change. , I This modification changes the setpoint of temperature indicating i- controller BMTIC0023'(Steam Generator Blowdown Non-Regenerative Heat. Exchanger Discharge Temperature) to control the non-regenerative heat , exchanger blowdown fluid discharge temperature from 135 degrees ni Fahrenheit to 120 degrees Fahrenheit. The original setpoint of 135 . p . degrees Fahrenheit was based on prolonging domineralizer resin life j and to preclude the possibility of eluting the radioactivity absorbed
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- by the resin. This modification revises the setpoint of BMTIC0023 to j the most limiting component in the Steam Generator-Blowdown System !
which is radiation monitor, BMRE0025 (Steam Generator Blowdown Non- l Regenerative Heat Exchanger' Discharge Radiation). BMRE0025 has a j maximum design sustained sample temperature of 122 degrees ! Fahrenheit. Therefore, the new setpoint will be 120 degrees [ Fahrenheit to prevent BMRE0025 crystal failure and/or loss of I calibration. BMTIC0023 is a non safety-related component. Changing ! the setpoint of this component has no affect on any safety-related- l equipment. ! This modification will have no impact on accidents or malfunctions l evaluated as the licensing basis and there is no. potential for the l creation of a new type of unanalyzed event. There is no reduction in ! the margin of safety. j
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1 4 f. ?: Attachment to- CO 95-0028 Page.132 of 183 Safety Evaluation: 59.94-0120 Revision O' Essential Service Water Drawing Discrepancies
, This modification provides clarification'to Note 6 of drawing M h[3; 12EF02, "PEID Essential Service Water System." 'The clarification
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provides instruction concerning use of the annubar test connection that was provided to set.or verify containment cooler flow rates. The
, Essential Service Water System has been re-balanced and flow. rate verification procedures'have been revised to utilize other
, instrumentation for measuring flow rates. This modification is-
' edministrative only and- will not affect any equipment, components,;or-structures.
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. 4
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- i. - Attachment to CO 95-0028 Page 133 of 183 Safety Evaluation: 59 94-0121 Revision 0 Seismic Loading for Evaluation of Temporary Conditions This revision to the Updated Safety Analysis Report adds evaluation criteria for evaluating temporary conditions in the plant.
Calculation AN-94-037 " Seismic Loading.for Evaluation of Temporary conditions" provides reduced seismic accelerations for evaluation of plant temporary conditions such as temporary rigging, lead radiation shielding, scaffolding, freeze plugging and temporary alterations of supports or boundary conditions.
- r. The minimum seismic design input to be util.lzed for these temporary conditions shall be based on an earthquake 'nagnitude which has no greater probability of being exceeded over the duration of the condition than the probabilit? of exceeding the design basis earthquake for permanent strr.ctures during a year. The appropriate floor input for the locarica of interest may be scaled down by the ratio of the ground response zero period acceleration for the reduced earthquake magnitude for the duration of the temporary condition to the Wolf Creek Generating Station site ground response zero period acceleration. Loads calculated from these reduced seismic inputs may be utilized in place of Safe Shutdown Earthquake or Operating Basis Earthquake loads when assessing compliance with the design requirements.
This revision 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- CO 95-0028- i Page 134 of 183 l Safety Evaluation: 59 94-0122 ' Revision:0'- !
Installation of Fuller Earth Filter Drain valve This modification to non' safety-related equipment adds a drain ~ valve: .i' to the Electro-Hydraulic Control System Puller Earth Filter. 'The Electro-Hydraulic Control. System provides over speed protection for-
'the main turbine. The drain valve'will be installed on filter housing 1 (FCH03A)'just above the filter. outlet connection, This modification ,
will reduce the probability of.an oil spill.and will improve'the filter change out' process by reducing theunmount of cleanup required.
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.- l This modification will have no impact ~on accident'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 E the' margin of safety. , j I
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a., Attachment to CO 95-0028 Page 135 of 183 Safety Evaluation: 59 94-0123 Revision:0 Changes to Cycle 8 Fuel Design , This modification addresses the design changes to Cycle 8 fuel assemblies which were made to enhance fuel reliability. This modification is Revision 1 to Configuration Change Package 05120 Revision 0, which was reported by Unresolved Safety Question Determination 59 94-0074. Revision 1 addressed concerns regarding the fuel thermal performance degradation because of implementation of intermediate flow mixers (IFM) and rotated mid-grids. An engineering review (ER-93-33) was conducted after the flow testing was completed on the 17x17 V5H fuel assemblies. The data collected during flow testing included test assemblies with and without alternately rotated mid-gride. The excessive vibrations present in the standard configuration was virtually eliminated when the alternate vaned grids were rotated by 90 degrees about the axis of the fuel assembly. However, tests performed by Columbia University indicated that rotation of alternating grids would reduce the thermal performance in fuel which also included IFMs in the assembly design. The results of the safety evaluation performed to address this concern confirm the acceptability of plant operation with V5H fuel with IFMs and rotated mid-grids incorporated into the fuel design. This > justification is based on the departure from nuclear boiling (DN3) margin gained from replacing the thimble plugs in the core (decreasing bypass flow), the recovery of margin allocated in the maneuvering analysis, and the removal of conservatisms in the uncertainties accounted for in the Statistical Core Design. This Unresolved Safety Question Determinaion confirms that there is no increase calculated in the transient specific fuel rod failure. Evaluations and analysis support the conclusion that all safety analysis acceptance criteria continue to be met. The probability of occurrence and the consequences of an accident evaluated previously in the Updated Safety Analysis Report (USAR) will not be increased because of the use of V5h fuel with IFMs and rotated mid-grids. Analyses, evaluations, and minimum DNB calculations confirm that the USAR conclusions remain valid for this modification. The use of V5H fuel with IFMs and rotated mid-grids does not increase the probability of occurrence on a malfunction of equipment important to safety or increase the consequences of a malfunction of equipment- ; evaluated in the USAR. The use of the specified fuel design does not ] ' affect equipment operation, or assumptions regarding the malfunction of equipment important to safety. The use of V5H fuel with IFMs and rotated mid-grids does not create the possibility of a new or different kind of accident from any accident previously evaluated. The use of the specified fuel does not I
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, f Page 136'of. :183 -
. change the plant configuration in-a way that introduces'a.new ,
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p potential hazard to the plant. For this reason,;the possibility'of a'
.new accident which is different-fromLany already. evaluated in-the USARL is not created. The'use of V5H. fuel with'IFMs'and RMGs does not involve a'significant reduction in the' margin of safety. : The analyses :
and evaluation discussed in1this safety evaluation demons': rate that : all applicable safety. analysis acceptance criteria contisae to be met i' l; for this modification. .Therefore, it is concluded'that the margin of safety as described'in the bases to any Technical specification is not;- : reduced. r
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'Page'137.of 183'-
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. Safety Evaluation: 59'94-0124 Revision 0 ;
Re-analysis of Main Steam Line Break This revision to the Up'ated. d Safety. Analysis Report.(USAR) is a. result -j of changed methodologies in calculating mass and energy release during; j a: postulated main steam line break, and the change in calculating main; j steam-line tunnel pressurization. The original pressure analysis for -; the main steam line tunnel utilized blowdown.information which did not. ~! consider water entrainment. The revised methodology take into { consideration water entrainment released through a break. The new-methodology demonstrates that the peak pressure is below the peak ; pressure previously calculated. This revision does not change plant- ; configuration, design, or processes. Ji i This revision 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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3- i r Attachment to CO 95-0028 , Page 138 of 183 { 59 94-0125' Safety Evaluation: Revision:0 I
. Updated safety Analysis Report Revision to Improve Consistency With Emergency Operating Procedures !
This revision to the Updated Safety Analysis Report. (USAR), Section ! 7.3, " Engineered Safety Features Systems," deletes the last two' j sentences of paragraph six which states, "The likelihood of this ! situation occurring.(LOCA, SIS reset, LOSP) is further reduced by l
. administrative action'which would delay the time of reset of safety injection to just prior to the firs'c manual operator. action associateu ;
with changeover from che safety mode to the recirculation mode. r following a LOCA. This modificatisn to the Emergency Operating
- Procedures' ensures that the minimum time ofter the pipe rupture that i the safety injection signal is reset is greater than 10 minutes." !
These sentences are being replaced with a sentence.which states
" Guidance for the Operator Actions is provided in the Emergency Operating Procedures." This revision is being implemented to achieve consistency between the USAR and the Emergency Operating Procedures ;
(EMGo). [ The EMGs give. adequate cautions that inform the operators of the consequences of a Safety Injection Signal (SIS) reset. The reset of i the SIS provides operator flexibility during the accident to more ' effectively mitigate the accident. The consequences of previously ; evaluated accidents in the USAR will not increase.
- l This revision 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. Ther's is no reduction in l the margin of safety. l I i I I i l t i
- Attachment to' CO S5-002B Page 139 of 183 f
' Safety' Evaluations' ~59 94-0126 Revision:0 Emergency Diesel Generator Main Lube Oil Strainer Drain Piping Modification This modification to the safety-related Emergency Diesel Generator.
Lube oil system provides a-change in the piping configuration which will assure that the lube oil drain piping cannot be cross connected. With the current configuration this cross connection is possible. This modification will change the configuration of drain line piping and prevent debris from getting into the bearing supply header. This modification will have no impact on accidents or malfunctions l '. 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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m_ 1 Attachment to CO 95-0028 Page 140 of 183 Safety Evaluation: 59 94-0127 Revision 0 Testing of Alternate Valve Parts in Main Steam Dump Valves This temporary procedure TMP-EN-172, Revision 0, " Testing of Alternate Valve Parts in Main Steam Dump Valves" provides for testing of safety- , related valve parts designed and produced by a third party vendor and being substituted for similar parts in the non safety-related main steam dump valves. This test is not described in the Updated Safety Analysis Report (USAR). The third party parts were originally intended for the safety-related Steam Generator Atmospheric Relief valves. Mowever, the design is sufficiently different to raise some question concerning the stability of these parts during operation. No testing has previously been accomplished to determine their suitability. The main steam dump valves have a body and bonnet l identical to the steam generator atmospheric relief valves. In an effort to prove this design and provide assurance that installation in the ARVs is acceptable, it is necessary to perform acceptance testing of this design. Thus, it is appropriate that this testing be performed in the less safety-sensitive main steam dump valves. This procedure accomplishes this acceptance testing. Adequate protection against loss of heat removal capability is maintained by the Steam Generator Safety Valves (primary) and the ARVs , (secondary). During the performance of the tests on the steam dump valves, the cooldown rate is not expected to exceed 100 degrees Fahrenheit in one hour as specified in Technical Specifications. The step change will be limited by Main Steam Isolation Valve closure if the steam dump fails after a thirty degree decrease while performing. The performance of this temporary procedure 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. > O
w , - i l Attachment to CO 95-0028 Page 141 of- 183 o i Safety Evaluation: 59 94-0128 Revision 0 I Auxiliary Feedwater Pump Turbine Tachometer Relay Modification j This modification provides for the replacement of tachometer relays.
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used in the controls for the Auxiliary Feedwater Pump turbine f r (KFCO2). The Airpax tachometer relays are no longer available and ! will be replaced with a syncro-Start speed switch. ')'; The present configuration utilizes.two separate tachometer relays, '[ The tachometer relays provide high and low speed alarms'and overspeed ' trip for the Auxiliary Feedwater Pump turbine. This modification f (f. replaces the two tachometer relays with a single device that perforne ! ,[ ! ..the same functions as the two devices. The failure modes and effects i lp is the same for both the original Airpax tachometer relays and the .I replacement Synchro-start speed switch. Tachometer relay functions, performance, and failure modes of plant equipment are unchanged by ; this modification. ! i This modification will have no impact on accidents or malfunctions { evaluated as the licensing basis and there is no potential for the j creation of a new type of unanalyzed event. There is no reduction in ; the margin of safety. .i f i 3 I I i t t 1 L i I
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n - l Attachment to- CO 95-0028- j Page 142 of 183 I
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' Safety Evaluation: 59 94-0129 Revision 0 !
r Main steam Line' Drain Valve stroke Time change This revision to the Updated Safety Analysis Report (USAR) revises [
- Figure 6.2.4-1 (Sheets 1 through 4) to change the stroke time limit of S the. main steam line drain valves (ABLV-07, ABLV-08, ABLV-09, and ABLV-10)- from five to ten seconds. The main' steam line drain valves are automatically isolated upon receipt of a steam line isolation signal. ;
The main steam drain line is located between the steam generator l safety valves and the Main Steam Isolation Valves (MS1V)' . The drain ; line, which connects the main steam line to the condenser, consists of 7 a drain drip leg, a level control valve (main steam line drain' valve), a number of manual valves and level control' instrumentation. The Main .i Steam Line Drain Valve is a two inch globe valve connected to one inch :I drain pipe. This drain valve is normally closed during the plant 'i operation. However, it opens on a high level signal from the level f control transmitter, upstream of the valve, to release the condensate- - to the condenser and maintain condensed fluid in the reservoir at a e desired level. During an abnormal condition and upon receipt of a steam line isolation signal, the drain valve will automatically j I isolate to maintain the secondary side pressure boundary and to provide protection to mitigate the effects of an accident. During inservice testing, it was discovered that ABLV-07 and ABLV-09 failed to meet.the stroke time requirement. These valves have a { different design configuration than valves ABLV-08 and ABLV-10, and j this design configuration requires more time to stroke the valves. The main seam line drain valves are automatically isolated upon receipt of a steam line isolation signal which is generated from a low steam line pressure signal, a high-high containment pressure signal, a I manual actuation, a Solid State Protection System signal or a stream j line pressure-negative rate-high signal. Review of the loss of , coolant accident analysis and the main line break analysis will have l no impact on events already analyzed in the USAR. Because the main steam line drain valves are located outside containment, a five second l stroke time relaxation will not cause additional mass and energy to l release to the containment during a loss of coolant accident inside .! containment and previously evaluated accident conditions remain valid.- ! This revision will have no impact on accidents or malfunctions i evaluated as the licensing basis and there is no pot..ntial for the -! creation of a new type of unanalyzed event. There is no reduction in j the margin of safety. j r l l 1 l l
ti l Attachment.to' 'CO'95-0028 f t Page 143 of 183 .
-Safety Evaluation: 59 94-0131 Revision:0 l
Steam Generator Pressure Pulse Cleaning [ W Westinghouse procedure, SSS 2.2.2 SAP-5, " Steam Generator' Pulse ! Cleaning for Wolf Creek Unit-1," provides for the pressure pulse l cleaning of steam' generator tube support plates and tubes. The : objectives of this process are to remove suspended and dissolved- ~j solids through pulsing and recirculation and to loosen and move the. ! large sludge particles down to the tubesheet where'they may be ;
' subsequently removed from the' generator by sludge lancing and blowdown.
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An evaluation to determine the effects of pressure pulse cleaning on l the WCNOC steam generators has been performed. The. evaluation ! considers the effects of pressure pulse cleaning on all relevant steam ! generator components. ! Bared on the evaluation, performance of pressure pulse' cleaning 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.
i unanalyzed event. There is no reduction.in the margin of safety. . f I i I a i z i
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r C . Attachment.to CO 95-0028 l i o Page 144.of- 183 : l
' Safety Evaluation: 59'94-0132 Revisions 0 ;
i Drawing Correction for the Process Sampling System j cThis modification corrects editorial errors in drawing M-12RM02,~"P&ID [ Process Sampling System," to associate Note 4 with Main Steam Line "A" 'l Sample Flow Control Valve'(RMV0612) and Low Pressure Feedwater Header- l Outlet Sample Flow Control Valve (RMV0617). In addition, RMV0617 will l be correctly shown as a globe valve instead of a needle valve. These ; changes ~nre. editorial"in nature and do not affect any components in. !
.the plant. :
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. l I e r i 1 5 I l l i h 4 t' I
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Attachment to CO 95-0028 Page 145 of 183 Safety Evaluation: 59 94-0133 Revision:0 Cycle 8 Core Reload Design This safety evaluation reviews the Cycle 8 core reload design and changes to the Core Operating Limits Report for Cycle 8. This evaluation demonstrates that the core reload will not adversely affect : the safety of the plant. This evaluation was accomplished utilizing the methodology described in NSAG-007, " Reload Safety Evaluation Methodology for the Wolf Creek Generating Station." Wolf Creek Generating Station operated in Cycle 7 with both Westinghouse 17X17 Vantage 5H (V5H) and 17X17 Vantage Er,with Intermediate Flow Mixers (IFM) fuel assemblies. It is planned to reload the core with Westinghouse 17X17 Vantage SH fuel assemblies which feature the IFM grid straps. A safety evaluation was performed for the phased transition from Westinghouse Standard and Vantage 5H with IFM in a transition core and for operation with a full Vantage 5H with IFMs core. This report is consistent with the evaluation and analyses given in the Vantage 5H with IFM implementation report and the Cycle 7 Technical Specification change which contains re-analyses of applicable Updated Safety Analysis Report Chapter 15 accidents. 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. I
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.i t Attachment to CO 95-0028 f Page 146 of 183 l i
Safet; Rvaluation: 59 94-0134 Revision 0 l Revision of Setpoints for Safety Related Radiation Monitors'
.This modification recalculates and changes the Engineered Safety Feature Actuation System (ESEAS) setpoints of the safety-related area !
radiation monitors in the Control Room, fuel Building and in containment.. This modification-is based on the identification of the correct controlling isotope for accident conditions. Setpoints being ! changed include Control Room gaseous monitors (GK-RE-0004 and GK-RE- l 0005), Fuel Building gaseous monitors (GG-RE-0027 and.GG-RE-0028), and. I Containment area radiation monitors (GT-RE-0031 and GT-RE-0032). l l The new setpoints are' based'on the exposure limits as specified in' ! Technical Specification 3.3-6 and using Xe-133 as the' controlling. i isotope. The assumption of Kr-85 as the controlling isotope in , calculating the existing setpoints is erroneous.. Nuclear Engineering ' has determined that the controlling isotope should be Xe-133, which is l a large contributor to radioactivity'in these areas. ! These new setpoints calculated in this modification will continue to maintain the radiation exposure dose limits below 2mR/hr for the' 'j Control room, 4mR/hr for the Fuel Building, and 9mR/hr in ; containment. These safety-related radiation monitors, upon reaching j the calculated setpoint will perform the assigned area isolation I function in accordance with the design criteria. The design function of these radia* ion monitors is not changed. i l This mcoification will have no impact on accidents or malfunctions ; evaluaced 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 nnargin of safety. { l l i
y e Attachment to CO 95-0028 Page 147 of 183 Safety Evaluation: 59 94-0136 Revision 0 Revision to the Auxiliary Steam Chemical Addition Design Drawings This modification revises design documents to reflect the as-built conditions of the non safety-related Auxiliary Steam Chemical Addition System. The following design documents are being revised; drawing M-12FE01, "P&ID Auxiliary Steam Chemical Addition System," drawing M-13FE01, "Small Pipe Isometric Turbine Building," drawing M-AN10, "Small Pipe Isometric Demineralized Water Storage and Transfer System," Instruction Manual M-116A-00001, " Auxiliary Steam Chemical Add Unit 30 Gallon Tanks," and Instruction Manual M-116-00016,
" Supplement to Instruction Manual Chemical Addition Pumps and Tanks."
The Auxiliary Steam Chemical Addition System is skid mounted with a locally operated control station and is made up of two mixing tanks, two metering pumps and the necessary network of piping and valves. The system is designed to inject hydrazine and ammonia solutions into the feedwater of the Auxiliary Boiler and the Auxiliary Steam Reboiler. Injection of the chemicals occurs downstream of the Auxiliary Steam Feedwater Pumps directly into the upper drum of the Auxiliary Boiler. The Auxiliary Steam Chemical Addition System in conjunction with the Demineralized Water Makeup Storage and Transfer System will provide the required chemistry for all modes of operation. This modification does not affect any other equipment components or structures. 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.
l Attachment to CO 95-0028 Page 148 of 183 Safety Evaluation: 59 94-0137 Revision 0 Removal of Power from High Pressure Safety Injection System Isolation valve This modification removes power from the motor operated valve EMHV8924 (Residual Heat Removal Heat Exchanger "A"/ Chemical Volume Control System to Safety Injection Pump "A" Upstream Isolation Valve) . This modification is being performed in accordance with Revision 4 of Plant Modification Request 04145. Previous revisions of this modification were evaluated by Unresolved Safety Question Determination 59 92-006 and Unresolved Safety Question Determination 59 94-0098. The circuit breaker for EMHV8924 will be locked open at all times and the valve will be operated manually when operation is required. The probability of a Loss of Coolant Accident occurrence, has not been changed by modifing EMHV8924 to a manually operated valve which is administratively controlled. This modification has no affect on the frequency of accident occurrence in Modes 1, 2, and 3 because it is in the open safeguards position during these times. The status of this valve in Modes 4, 5, and 6 does not affect the probability of { previously evaluated accidents because the Emergency Core Cooling System (ECCS) recirculation is not required. Maintaining this valve as a manually operated valve which can only be closed in Modes 4, 5, and 6 has not increased the malfunction probability of the ECCS i equipment in Modes 1, 2, and 3 when the equipment is essential. No new or different variables have been introduced or changed by this modification that affects the valve 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. t i 1
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' Safety Ryaluation: 59:94-0138 Revision 0 .
HDrawing Corrections , A LThis' modification revises drawing M-12AF02 "P&ID Feedwater Heater- Li 4 Extraction Drains" and MS-01.and MS-02." Mechanical Standards" to be-consistent with as-built configurations; This revision is editorial = in nature and corrects the piping material and line number of the drawing. There are no changes to plant components, equipment, ,
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systems, or structures as a result of this' modification. l
,l This modification will have no impact on accidents or malfunctions :l evaluated as the licensing basis and there is no potential for the_- j creation of a new type of unanalyzed event. There is no reduction in
- the margin of safety. '
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L m f I L- Attachment to CO 95-0028' ! h ; Page 150 of 183 ! i t Safety Evaluations 59 94-0139 Revision 0 r Main Condenser Chemical Cleaning This temporary procedure provides controls for the actions.necessary' ; h
- to clean calcium carbonate from the Main Condensers using Betz 860. .
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The chemical is to be injected through a fabricated flange;on the. water box drain standpipe. . Service water will be used to: fill the' .; condensers through the condenser drain piping. The solution will then ! be' recirculated through each tube bundle through'the' Condenser Drain l Pump. A. fabricated recirculation line attached to the shell of check- l
.val ve DA-Vill and routed to a series of fabricated injection'manways ;
on water boxes will serve ~as the recirculation path. The water box i venting system will be utilized to vent off any CO2 created by the . j chemical reaction with the scale. Once the cleaning is. complete, the ; spent solution will be discharged to the lake through the service j Water System. j All previously evaluated accidents which could occur as a result of- f this procedure are bound by.the Turbine Building Flooding Evaluation. l and by the toxic chemicals discussion in Section 2.2.3.1.3 of the' q Updated Safety Analysis Report. No safety-related equipment is- l associated with this procedure. ! This temporary procedure will have no impact on accidents or I malfunctions evaluated as the licensing basis and there is no- , potential for the creation of
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e Attachment to CO 95-0028 I' Page 151 of 183 Befety Evaluations. 59 94-0140 Revision 30 Revision to Local. Leak Rate Testing Procedure This revision to procedure CKL PE-024, "LLRT Valve Lineup," changes ! : the status of_ valve BG-V227 (Seal Water Return Test Connection-Upstream of Penetration P-24 Isolation) and valve BG-V456 -(RCP Seal Water Return Line Drain /Te: Connection Upstream P-24) from locked closed to closed. Both valsL. are vent and drain isolation valves
- associated with containment penetration 24. Both valves are outside the containment isolation valves associated with containment
. penetration 24. There is no equipment important to safety that will'
- be adversely affected by changing the status of these valves from locked closed to closed.
Changing the status of these valves 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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3, e' , Attachment to CO 95-0028 Page'152 of 183 > 1 Safety. Evaluations 59 94-0141 Revision 0-4 Fire Load in Containment Tool' Room
- This' modification limits the permanent containment tool room to 1000 p lb. of flammable materials.- These materials shall be solids only. In E addition,?this modification limits zine coated surfaces to-less than
- 600 square feet for zine coated tools or other zine coated surfaces.
The limit of less than 600 square feet of zine coated surfaces lis-
- considered insignificant.and does not require revision of1 design basis calculations. Based'on engineering ~ review thisfchange would.have a negligible impact on the available design margins.
l The additional combustible loadingfof the containment tool. room will
. create a total. fixed' combustible loading of 16,613 Btu / square foot in-
- Fire Zone-RB-5. This loading is well below the fire' loading of 85,000 Btu / square; foot' required to challenge the capabilities of suppression systems or fire barriers for Fire Zone RB-5.
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. r
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; = a-Attachment to' CO 95-0028 Page 153 of- 183 Safety Evaluation: 59.94-0144 Revision 0
; Defeat of' Interlocks on Spent Fuel Pool Bridge Crane For Fuel' Reconstitution.
' Revision 2 to procedure MGE KE-003'" Defeat Interlocks on SFP Bridge Crane for. Fuel Reconstitution" adds additional bypasses that are required to perform fuelfreconstitution. When'this procedure is performed the' south bridge limit for the Spent Fuel Pool Bridge. Crane
. (HIE 04): and the five foot exclusion boundary' between' the crane and the
-New- Fuel Elevator (HKE10) will be bypassed.
, This revision defeats the interlock between the Spent Fuel Pool Bridge. Crane and the New Fuel Elevator to allow the bridge crane to travel over the cask loading pit area-when the elevator is not in.the full down position. This will provide support to the fuel reconstitution activity. This procedure will.be performed for fuel reconstitution. During fuel reconstitution both devices-are used simultaneously. .This activity is controlled by an administrative' procedure. Defeating the~
. interlocks does not alter the crane design basis functions nor'does'it:
affect the cranes' structural integrity. The New' Fuel' Elevator is non safety-related equipment. This revision will have no impact on accidents or malfunctions' evaluated as the licensing basis and there is no potential for the 4 creation of a new type of unanalyzed event. There is no reduction in the margin of safety. I i 1, n i
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l-i Attachment to CO 95-0028 Page 154 of. 183 I- Safety Evaluation: 59 94-0146 Revision:O Fuel Transfer System Operating Instructions Revision The revision to procedure FHP 03-006, Revision 9, " Fuel Transfer System Operating Instructions," provides for defeating the electrical position limit switch interlock in transfer car system (HKE09A/B) and using a camera to verify the position in place of the limit switch. In addition to the electrical position limit switch interlock, there is a backup interlock that is a mechanical latch device on the lifting arm that is opened-when the car is moving into position. Therefore, two methods of verifying. position is still available and the probability of occurrence of an accident previously evaluated in the Updated Safety Analysis Report (USAR) does not increase The visual verification in place of the position limit switch interlock does not alter the transfer car design basis functions. Thus, the proposed change does not increase the consequences of a malfunction of any equipment important to safety as evaluated in the USAR. A fuel handling accident is analyzed in Section 15.7.4 of the USAR. The proposed change does not create the possibility of an accident of a different type than any previously evaluated in the USAR. Defeating the interloch does not effect the design function of the transfer car. Thus, the proposed change does not increase the possibility of malfunction of equipment important to safety as evaluated in the USAR This revision 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.
r l' Attachment to- CO 95-0028. Page 155 of- 183 Safety Evaluation: 59 94-0147 Revision 3 0 Removal of Procedure Titles From the Updated Safety Analysis Report This revision to the' Updated Safety Analysis Report-('USAR)' replaces specific procedure. numbers'and titles in USAR Section 13.2 and Table p 5.2-5 with a statement;which indicates that the associated activity or control is controlled by an administrative procedure. A paragraph was also added to Section 17.2.5.1 to clarify that'a search should be conducted based on the description of the activity or. control rather than on the procedure' number. Th4a revision provides USAR' descriptions.of activities and controls consistent with the intent of 10 CFR 50.59. This revision is administrative in nature and does not affect any plant equipment, components, systems, or structures. There are no USAR design basis accidents associated with this revision. This revision 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. i h i
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i Attachment.to CO 95-0028-Page 156 of 183 ; 1,- I Safety Evaluation: 59 94-0148' ' Revision 0 [ Temporary Procedure to Provide for Testing of the Static Exciter , Voltage Regulator ! This temporary: procedure, TMP MT-032 Revision 0,'" Testing of the f Static Exciter Voltage' Regulator on NE01 and NE02," provides a means { to test the operability of the Static Exciter Voltage Regulator (SEVR) :j without running the Emergency Diesel Generator. The SEVR is normally { powered by a Potential Power Transformer which is fed at the generator ! output. Implementing this procedure provides'an alternate, non safety- [ related source feed to the SEVR. This will allow verification of the -i correct function of the SEVR. . The transformer is not. energized and is- ! not supplying the SEVR during implementation of this procedure. f Because the system will be inoperable at the time this procedure is- , ! performed, there is no increase in the probability of' occurrence of an j accident previously evaluated in the Updated Safety Analysis Report. , All systems will be returned to their designed configuration at the completion of.this test. There is no permanent design change as a j result of this procedure. j t 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. ; I f i
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- safety Evaluation: 59 94-0149 Revision 0 Unit Parallel Relay for Emergency Diesel Generator A EThis temporary modification, TMo 94-09-NE, Revision 0, provides for the installation of jumpers and the lif ting ~ of leads on the contacts i of the Unit Parallel Relay for the "A" Emergency Diesel Generator. [
This modification is being implemented because it is. suspected that
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this relay is being effected by electromagnetic fields in the. ' vicinity. This will place this safety-related relay in the. safeguards state. This temporary modification effectively removes a safety related relay from the circuit. The function of this relay is to become de-energized. This temporary modification will jumper across j the normally closed contacts and remove the leads of the normally open ; contacts. .Therefore, all of the contacts will be in the de-energized ; state. '! i Because this relay will continuously be in it's safeguards state, this j modification will.have no impact on accidents or malfunctions j 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. g I ( l i i
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5 i Attachment to- CO 95-0028 i Page 158 of; 183 1 Safety Evaluation: 59 94-0152 Revision 0 , i Redesign of Cycle 8 Core' Loading "This modification to the original Cycle 8 Core reload design provides ; for the operation of Wolf Creek Generating Station with the introduction of Regions 10A and 10B reloaded fuel. This modification ; has been analyzed in accordance with methodologies reviewed and. approved by the Nuclear Regulatory Commission. Demonstrated adherence , to applicable standards and acceptance criteria precludes any new ] challenges to components and systems that could increase the. l probability of any previously evaluated malfunction of equipment important to safety. The Cycle 8 reload design does not violate any ! safety limits and all Cycle 8 design criteria are met. Therefore, the ; probability of occurrence of a malfunction of equipment important to ; safety previously evaluated in the Updated Safety Analysis Report i (USAR) has not increased. ,
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The safety evaluations presented in the Reload Safety Evaluation and the Cycle 8 Core Operating Limits Report demonstrate that the .I consequences of an accident previously evaluated in the USAR is not { increased. The demonstrated adherence to applicable standards and , design criteria precludes new challenges to components and systems- j that could adversely affect the ability of existing components and l systems to mitigate the consequences of any accident or adversely , affect the integrity of the fuel rod cladding as a fission product barrier. Furthermore, adherence to applicable standards and design ; criteria ensures that these fission product barriers maintain design ! margin to safety. The Cycle 8 reload design and Core Operating Limits Report were developed and verified using NRC approved codes and- f methods. The mechanical design of each fuel region that will be ! resident in Cycle 8 have been shown to meet all design criteria. The nuclear design confirms that all reactivity and kinetics parameters , are with design limits. The effect of the reload on the design basis -j accidents considered in the USAR have been examined and in all cases, j it was found that the effects of the reload were accommodated within i the conservatism of the existing analyses. Therefore, the { consequences of an accident previously evaluated in the USAR are not ! increased. I The safety evaluations presented in the Reload Safety Evaluation and the Cycle O Core Operating Limits Report demonstrate that the probability of a malfunction of equipment important to safety previously evaluated in the USAR is not increased. Demonstrated adherence to applicable standards and acceptance criteria precludes any new challenges to components and system that could increase the probability of previously evaluated malfunctions of equipment important to safety. The Cycle 8 reload design and Core Operating Limits Report were developed and verified using NRC approved methodologies. No new performance requiretnents are imposed on any I i I
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Page 159 of- 183 system or component such that any design criteria will be' exceeded nor will the Cycle 8 reload design cause the core to operate such that operating limits are challenged. No new failure modes or limiting single failures have been introduced with the Cycle 8 reload design. There is no known mechanical or electrical impact on equipment
.important to safety for LOCA or non-LOCA events due to the Cycle 8 reload design which would increase the probability of a malfunction of equipment important to safety. The functions of all safety-related systems are unaffected by the Cycle B reload design. Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR has not increased.
The safety evaluations presented in the Reload Safety Evaluation and the Cycle 8 Core Operating Limits Report demonstrate that the consequences of a malfunction of equipment important to safety previously evaluated in the USAR is not increased. The demonstrated , adherence to applicable standards and design criteria precludes new challenges to components and systems that could adversely affect the ability of existing components and systems to mitigate the . I consequences of any accident or adversely affect the integrity of the , fuel rod cladding as a fission product barrier. Furthermore, adherence to applicable standards and criteria insures that the fission product barriers maintain design margin of safety. The Cycle 8 reload design and Core Operating Limits Report have been developed and verified utilizing NRC approved codes and methodologies. The ' effect of the reload on the design basis accidents considered in the USAR have been examined and in all cases, it was found that the effects of the reload were accommodated within the conservatism of the - existing analyses. Therefore, there is no increase in the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. l The safety evaluations presented in the Reload Safety Evaluation and the Cycle O Core Operating Limits Report demonstrate that the Cycle 8 , reload design does not create the possibility of an accident of a different type than any other previously evaluated in the USAR. The demonstrated adherence to applicable standards and design criteria precludes new challenges to components and systems which could introduce a new type of accident not previously described in the . I USAR. The Cycle 8 reload design and Core Operating Limits Report were developed and verified utilizing NRC approved codes and i methodologies. All design and performance criteria continue to be met and no new single failure mechanisms have been introduced nor will the core operate in a manner such that the operating limits are ' challenged. Operation of Wolf Creek Generating Station in Cycle 8 has no effect on either the LOCA or non-LOCA accidents considered in the USAR. The Cycle 8 design does not create a condition outside of the i design basis accident criteria. Therefore, the possibility of an . accident of a different type than any previously evaluated in the USAR has not been created. { 1 I
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Attachment to CO 95-0028 Page 160 of 183 .The safety evaluations presented in the Reload Safety Evaluation and the Cycle 8 Core Operating Limits Report demonstrate that the Cycle 8 reload design does not create the possibility of malfunction of equipment important to safety than any other previously evaluated in ' the USAR. The Cycle 8 reload design and Core Operating Limits Report provide demonstrated adherence to applicable standards and criteria which precludes new challenges to components and systems which could introduce a new type of malfunction of equipment important to safety not previously considered in the USAR. All design and performance criteria continue to be met and no new failure modes have been introduced for any system, component, or piece of equipment as a result of the Cycle 8 reload design. The implementation of the Cycle 8 reload design and adoption of the Core Operating Limits Report will not create the possibility of a new equipment malfunction since these changes do not impact normal operation of the plant. Therefore, the , possibility of the malfunction of safety-related equipment different , from that previously evaluated in the USAR is not created. { The safety evaluations presented in the Reload Safety Evaluation and the Cycle B Core Operating Limits Report demonstrate that the cycle 8 reload design does not reduce the margin of safety as defined in the Bases to any Technical Specification. The Cycle 8 reload design and Core Operating Limits Report establish that all design and safety analysis limits continue to be met and that these limits are supported by the applicable Technical Specifications. Evaluation of the Cycle 8 reload design accounts for both normal operation and postulated accident conditions for the Wolf Creek Generating Station. The LOCA l evaluation demonstrates that all 10CFR50.46 criteria are met. The non-LOCA safety analysis acceptance criteria remain unchanged and continue to be met. The core design parameters.and assumptions incorporated in the safety analysis remain bounding and thus, the conclusions in the USAR remain valid. The margin of safety as defined in the bases is not reduced for any Chapter 6 or Chapter 15 accidents. Therefore, the ! margin of safety as defined in the Bases of the Technical Specification has not been reduced. l f I l i I l
i p I Attachment to CO 95-0028 Page 161 of 183 Safety Evaluation: 59 94-0153 Revision 0 Boron Dilution Transient Roanalysis This revision to the Updated Safety Analysis Report (USAR) Section 15.4.6, " Chemical and Volume Control System Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant" reflects changes of key parameters associated with cycle 8 specific I analyses for the boron dilution transients. The inadvertent boron dilution event (BDE) presented in the USAR is postulated to be a initiated by operator error or Chemical and Volume control System (CVCS) or Reactor Makeup Water System (RMWS) malfunction. This revision to the USAR does not affect the frequency of the possible , error or malfunction because the changes involved are only analytical revisions to the safety analysis bases as a result of cycle 8 design. Revising a safety analysis parameter such as the initial boron concentration does not increase the probability of these operator i errors or mechanical / electrical failures. This revision affects only the time required for an operator or the automatic flux doubling Boron Dilution Mitigation System (BDMS) to mitigate an inadvertent BDE. The results calculation AN-94-043 " Cycle 8 Boron Dilution Event Analysis" indicates there will be sufficient time for either an operator or BDMS intervention to terminate the dilution event before complete loss of shutdown margin should an unplanned BDE occur during Modes 1 through 5 of reactor operation. Because the reactor will be maintained sufficiently suberitical during an inadvertent BDE, the consequences of an accident previously evaluated in the USAR will not be increased. This revision affects j only the time required for the BDMS or the operator to mitigate the inadvertent BDE and does not involve any changes to equipment important to safety. This revision does not change any components, equipment, systems, or structures. All associated equipment continue to meet the design specifications and there is no increase in the consequences of a malfunction of equipment important to safety. The normal operation of the plant is not affected by this revision. No new transient scenario, event precursors, failure mechanisms or i limiting single failures are introduced as a result of the revision. Therefore, the possibility of a new or different kind of accidents not created. The results of Cycle 8 specific BDE analysis, based on the boron I requirements for meeting the technical specification shutdown margin limit of 1.3% Delta k/k, indicates that there will be sufficient time i for either the automatic flux doubling BDMS or operator intervention to prevent a complete loss of shutdown margin should an unplanned BDE occur during all operating modes. Because shutdown margin is not
Attachment to CO 95-0028 Page 162 of 183 . lost, the minimum Departure From Nuclear Boiling Ratio remains well above the safety analysis limit values, no over pressurization would occur and therefore, there are no fuel failures. Based on the above discussion the margin of safety as defined in the technical ! specifications will not be reduced as a result of this revision. I This revision 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. 2 8 I f I
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' Attachment to- 'CO 95-0028 Page 163 of= 183 .}
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~ Safety Evaluation: 59 94-0154 Revision O L
Revision'to Rod Drop Time Measurement Procedure STS RE-007 w .
'This-revision to procedure STS RE-007, " Rod Drop Time Measurementi" -l
,provides for the use of test equipment to test dropping:an entire bank '[
of control rods at one time verses one rod at a' time. The dropping of' an entire control-rod bank will be accomplished by opening reactor , trip. breakers, which leads to a feedwater isolation signal with Tavg , below 564 degrees' Fahrenheit. In order to suppross'the'feedwater , isolation signal, the reactor trip input.into the feedwater isolation circuitry will be defeated for.approximately eight hours during rod j drop testing. Only the feedwater isolation' signal on reactor trip will be disabled during-testing. 'All other'feedwater isolation signals will remcin active and function normally. . This safety evaluation is valid only with the Reactor Coolant System i pressure above 1970 psig. Evaluation of this change assumes that'the ; rod drop test is performed in Mode 3, with the reactor subcritical, and greater than 350 degrees Fahrenheit. The scope encompasses j events which rely on feedwater isolation in conjunction with-a reactor- j trip to meet the applicable acceptance criteria. These parameters' correspond to the accident analyses evaluated.for the zero power j! condition and the overcooling events. Accident: analyses considered by , .! this evaluation include Feedwater Malfunction (decrease in feedwater "' temperature), Feedwater Malfunction (increase in feedwater flow), Excessive Increase in Secondary Steam Flow, Inadvertent Opening of a i Steam Generator Safety Valve, Main Steam Line Break,-Rod Withdrawal :I From Subcritical, and a. Rod Ejection Accident. Accident analyses not I listed here have been shown to be limiting assuming.the reactor.is-critical (i.e. power level greater than zero percent) or because the event is initiated above 564 degrees Fahrenheit and results-in and- 'j Reactor Coolant System heatup. Therefore, only the analyses listed above may be affected by defeating feedwater isolation on a. reactor , trip signal based on the plant condition during the rod drop. i
-The evaluation confirms that current analyses remain bounding or are not affected by the change in rod drop testing. .The evaluation supports the conclusion that all safety analysis acceptance criteria i will continue to be met. Because this evaluation shows that there is no impact to the current U dated p Safety Analysis Report (USAR) '!
analyses, the probability of occurrence and the consequences _of an accident evaluated previously in the USAR are not' increased and the proposed test change does not affect any of the mechanisms postulated 'l in the USAR to cause a Loss of Coolant Accident (LOCA) or non-LOCA
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l , design basis event. The evaluation confirms that the USAR conclusions remain valid for this change because the current analyses have been 3 shown to remain bounding. On these bases it is concluded that the- ; probability and consequences of the accidents previously evaluated in the USAR are not increased. Therefore, it is also concluded that i i i _--_____=!
. Attachment to CO 95-0028-Page 164 of 183 '
-there is no reduction in the margin of-safety. This revision 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. , t S t I f k f 3
t i LAttachment to CO 95-0028' ; Page 165 of - ' 183 - , safety Evaluations. '59 94-0156 - Revision 0 - Revision of Essential 1 Service Water Low Flow Rate
.This mo'ification d revises the minimum Essential Service Water (ESW).
flow rate required to remove the heat' loads in the Auxiliary Feedwater' Pump Rooms after a Loss of Coolant Accident (LOCA). Room coolers-
- (SGF02A and SGF02B) are capable of removing.the required heat. loads with the reduced ESW flow rates of 100 gallons per minute each. The ;
flow rate of 100 gallons per minute provides~a flow velocity through .; the tubes above three feet per second. This flow rate;is the' minimum l flow rate needed to control microbiological induced corrosion. The ;
; design basis safety-related function of the coolers is unchanged. ,
- There are no hardware changes involved. -
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 i the margin of safety, j
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.Page 166 of' -183 I Safety Evaluation: 59 94-0157 Revision:0- ,
Component cooling water surge Tank- Level' Equalisation ' ! This temporary procedure TMP 94-OPS-00B-0, " Component-Cooling Water-
-Surge Tank Level Equalization," provides direction to connect the two '
Component Cooling Water System surge tank drain lines'to equalize the ; tank levels. This procedure does not change the design basis or ! safety function of the tanks or the Component Cooling Water System. l This temporary procedure will not increase the consequences of an : accident previously evaluated in the Updated Safety Analysis Report. l This temporary procedure does not result in any malfunction of any { safety-related equipment. The temporary hose which will;be used to ; connect the two drain lines will be non safety-related. Because the [ work will be done under constant supervision and a monitor will be ll posted at the drain valves, this procedure will not create the { possibility of an accident of a different type. The potential for ; cross connection of the Component Cooling Water trains'does not j constitute a different type of accident than previously evaluated in 1 the USAR. There is no possibility for a different type of malfunction !
-than previously evaluated. The cooling capability of the Component !
Cooling Water System is not reduced by the performance of this f temporary procedure. { This temporary procedure 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, f
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-n Attachment to CO 95-0028 Page 167 of 183 Safety Evaluation: 59 94-0160 Revision 0 Modification to Improve Technical Support Center This modification provides for the reconfiguration of the Technical Support Center (TSC) to allow the Emergency Response Organization to work more effectively. This modification includes the installation of a new internal door, rearranging of computer locations, repositioning of telephones, and new friskers and step-off pads. In addition, a shower stall with an eye wash station will be installed in place of the existing shower head and eye wash station.
The changes in the TSC cannot affect the probability of occurrence of an accident because the facility is completely separate from the power block. Decontamination of emergency response personnel in the TSC will create an unaccounted for release of radiation to the sewaga treatment plant and then to the cooling lake. The water from the shower would be very low in activity and would not increase the consequences of an accident previously evaluated in the Updated Safety Analysis Report (USAR). This modification will have no impact on accidents or malfunctions evaluated as the licensing basis and there is no potential for the l creation of a new type of v. analyzed event. There is no reduction in the margin of safety.
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Safety Evaluations 59 94-0162 Revision 0 Drawing Revision to Reflect a Capped Fitting in the Reactor Makeup water system The purpose of this modification is to correct the applicable design documents to reflect the as-built conditions as they pertain to a fitting installed downstream of Reactor Makeup Water to Refueling Pool Clean-Up (BLV0078) isolation valve and. Reactor Makeup Water to Reactor Vessel Head Decontamination and Refuel Pool Clean- Up (BLV0079) isolation valve. Design drawings M-12B101, "P&ID Reactor Makeup
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Water System," and MS-02, " Pipe Fitting Class Sheets" will be revised to reflect a capped fitt2ng downstream of the above identified components. This modification will have no impact on accidents or malfunctions ( evaluated as the licens.'.ng 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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-1 Attachment to CO 95-0028 l Page 169 of .183 Safety Evaluation: 59 94-0163- Revision 0 i
Replacement of Feedwater Heater Relief Valve [
- t This modification to non safety-related equipment replaces and !
upgrades Feedwater Heater Relief Valves (7A and ?B). The replacement . valves have hard faced trim and slip type flanges. The relief valves l outlet lines will be rerouted to the vent. stack for Blowdown Flash l Tank Relief Valve (BMV052) to improve the valves ability to'rescat j properly. {
.i' This design change will not increase the probability of a loss of normal feedwater flow nor any other accident identified in the Updated ,
Safety Analysis Report. The design change will minimize the loss of l feedwater through the relief valves. Hard facing on the valve seat 1 and the valve disk will provide additional durability and allow the j
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valve to rescat repeatedly without loosing the seating surface. l l The feedwater heater thermal relief valve provides protection from l excessive pressure caused by the thermal expansion of the water when j the flow is isolated. The relief valve provides'a leak path for [ feedwater to the atmosphere The relief valve capacity is very small l compared to the.feedwater system capacity. The failure of the relief f to reseat after relieving would be-slight' loss of feedwater flow. The small flow path that is the result of the relief opening will not' j increase the probability of a lass of normal feedwater flow. I r This modification does not affect the failure modes of safety-related J components, structures or equipment. This modification will have no j impact on accidents or nalfunctions evaluated as the licensing basis. I and there is no potential for the creation of a new type of unanalyzed ! event. There is no reduction in the margin of safety. ; i t t
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.. l Attachment to'. 'CO 95-0028.
- Page 170 of ,183 Safety Evaluation: 59 94-0164 Revision:0 !
'l Organizational Changes I i
This revision to the Updated Safety Analysis Report ('U SAR) reflects { changes to the Wolf Creek Generating Station Organization. This ! revision includes a change in reporting relationship for the Fire. ! Protection organization, changes in reporting relationship for the [' Inservice Inspection and Inservice Testing programs, and a change.in. title and reporting relationship'for the Environmental Management { organization. j i The organization changes made in this revision do not' affect the -! safety and reliability of the plant or the operation of the plant. ! There is no affect on equipment important to safety. These changes do not represent any decrease in the concern for the health and safety of j the public. There is no increase in the probability of occurrence of [ a malfunction of equipment important to safety previously evaluated in i the USAR. t j This revision'does not affect any system, component, or procedures ~i required to mitigate the consequences of an accident previously f evaluated in the USAR. All functions continue to be performed. , Therefore, the consequences of accidents previously evaluated in the : USAR will not be increased. There is no increase in the consequences -I of a malfunction of equipment important to safety previously evaluated in the USAR. There is no change in the possibility of an accident of 3 a different type or a malfunction of equipment the,a previously [ evaluated in the USAR. This revision does not change the overall i operating philosophy of capability of Wolf Creek Nuclear Operating' , Corporation. There is no reduction in the margin of safety as defined- l in the basis for any technical specification. . 1 ; e t i I t i j i
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i a s E si Attachment to CO 95-0028 - r - Page 171 of 183- l Safety Evaluations- 59 94-0165 Revision 0 l
,5 Revision to'Ouidelines for WCGS Staff Working Mours- f Procedure ADM 01-023 " Guidelines for WCGS Staff Working Hours" is being replaced by AP 13-001 " Guidelines for WCGS Staff Working l Hours." Changes incorporated by this revision include the use of a !
- form for documenting the reasons'for an individual exceeding working hour limits. It'will no longer be required that the documentation be recorded in the Station Log (Shif t Supervisors Log) as specified in' Updated Safety Analysis Report, Section 18.1.3.2. The revision to-
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't This revision will have no impact on accidents or malfunctions !
evaluated as the licensing basis and there is no potential for the :I creation of a new type of unanalyzed event. There is no reduction in the margin of safety. I
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Safety Evaluation: 59'94-0166- Revision:0 I i Clarification of Service Water System Design Flow Rates This revision to the Updated Safety Analysis Report (USAR) is'to clarify that the Service Water flow through the various Service Water I System and Essential Service Water System components _may be lower than } the nominal design flow rates given in Table 9.2-1 and 9.2-2 or the l USAR during normal power operation. When the lake temperature is ; below the design temperature of 90 degrees Fahrenheit less flow is' required to remove the heat load from the heat exchangers. In a addition, a clarification is provided for minimum flow required to l control Microbiological Induced Corrosion (MIC) . ! This revision does not change USAR information concerning Essential- ! Service Water Loss of Coolant Accident flow rates or Essential Service f Water System flow rates required _during any other accident previously i evaluated. Therefore, the consequences of any accident previously i evaluated in the USAR is not increased. The Essential Service Water .[ System flow rates required to cool safety-related equipment are not l affected by this revision. Therefore, this revision will not increase , the probability'of occurrence of a malfunction to components. Because ; the Essential Service Water System flow rates are not affected by this
- revision, the consequence of any malfunction of this equipment which l has been previously evaluated will not be affected and no new type of [
unanalyzed event will be created. Based on the above discussion, i there is no possibility of a different type of malfunction of safety l related equipment cooled by the Essential Service Water System and the [ margin of safety as defined in technical specifications is not reduced. ! t 1 t i [ i I i J i
7 l- ! l Attachment to CO 95-0028 Page 173 of 183 Safety Evaluation: 59 94-0168 Revisiont0 Change In Residual Heat Removal Boron Concentration Revision 24 to procedure SYS EJ-120, "Startup of a Residual Heat Removal Train" provides for a change in boron concentration level in the Residual Heat Removal (RHR) System. This revision ensures the ; Reactor Coolant System (RCS) boron concentration is maintained above ! the minimum shutdown requirements. The revision minimizes a system line up which has the potential to challenge the RHR pumps if a drain down event were to occur. This revision ensures that plant parameters are consistent with the assumptions in existing analyses. It also guards against the unlikely scenario of localized dilution by maintaining RHR boron concentration at an acceptable level prior to the recirculation mode. No equipment important to safety is adversely affected by this revision. This revision ensures the margins of safety in reactivity control systems are maintained. This revision 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, i r 1 4 P f l
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' Attachment to :CO 95-0028 .l Page 374 of 183' '(
Safety Evaluation: 59 94-0170 . Revision 0 j i Winter Operation of the Chilled Water System l This revision to the Up dated Safety Analysis Report (USAR)r Section j 9, 4 .10. 2'. 3, provides for deletion of a statement which refers to '; winter shutdown of the Chilled Water System. 'This statement is ! erroneous because many of the loads on the Chilled Water System are not seasonal, but instead are related to plant mode and power. Operation of the Chilled Water System'in winter months will not ; increase the probability of occurrence of an accident previously j evaluated in the USAR. j This revision 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- t the margin of safety, i i
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;Page 175 of 183 ;j i i t
' Safety'Byaluation: -59 94-0171 Revision 0 !
Correction to Plant Cooldown Sequence This revision to the Updated Safety Analysis Report ('SAR), U Section ;
.5.4A.3.2, corrects minor. errors innthe description of the cooldown and .;
depressurization from' Hot Standby to Cold Shutdown conditions. In . [ addition, Sections.5.4.7.2, 7.6.2, and 10.4.9 of the USAR are' revised to reflect the correct pressure at which the Residual Heat Removal, , (RHR) System is placed in' service during plant cooldown. The changes associated with this revision will not change.the l l probability of occurrence of previously evaluated accidents. This ; revision clarifies the pressures at which the Emergency Core Cooling '! System accumulators are isolated and the RHR System is placed in. l service. However, this revision has no affect on operation of'these { systems. Therefore,-the' consequences of previously evaluated i accidents will not be changed. Because this change is for ; clarification only, there is no increase in the probability of ; occurrence of malfunctions of equipment important to safety. . Clarification of the USAR descriptions will not increase the .! consequences of any equipment malfunctions. This revision cannot. create the possibility of a different type of malfunction of equipment important to safety than previously evaluated in the USAR and the 3 margin of safety has not been reduced.
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safety Evaluations. 59 94-0172 Revision:O t- ,
' Revision to Auxiliary Building KFAC Drawing, J This' modification provides for the' revision of drawingHM-12GLO2, "P&ID 1 Auxiliary. Building HVAC," to add a previously unidentified valve to 'I the drawing. 1This valve.was discovered during a walkdown of the !
Auxiliary / Fuel Building. Normal Exhaust Filter Absorber Unit. The j valve is located on the_ instrumentation-low pressure valve. header. _ In '
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addition, this revision will address the symbology of - the pressure
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differential instrument individual isolation valves (GLV0712,7 GLV0713, j GLV0715, GLV0716, GLV0747 and GLV0748) in the Auxiliary Building HVAC. : System. 'This revision is administrative in nature and has no affect , on any component, _ system, or structure in the plant. i
~This modification =will have no impact on accidents or malfunctions evaluated as the licensing basis and.there is no potential for the 'I creation of a new type.of unanalyzed event. There is no reduction in the margin of-safety. :
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7--- m-h- I ' Attachment to- CO 95-0028-Page 177 of 1183 Safety Evaluation: 59'94-0179 Revision 0 Maintenance Organization Changes This revision to the Updated Safety Analysis Report (USAR) changes the l' title of Manager Maintenance and Modifications to Manager Maintenance i and identifies the positions reporting directly to the Manager i Maintenance. Those positions include Superintendents Mechanical, l Electrical, I&C, and Maintenance Support and Planning. 7tn addition,'a- ! - staff position of Assistant' Manager Maintenance has been added. This
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USAR Change Request also changes the reporting relationship'of the Superintendent Modifications from the Manager Maintenance to'the Vice-President Engineering. This change is administrative in nature and does not affect any components, equipment, systems or structures. Only position titles and reporting relationships have been changed. This revision 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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. Page 178 of 183' >
Safety Rvaluation: 59 94-0180 Revision 0 i
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Change to Seismic Peak Recording Accelerometer Range _ This revision to Table 16.3-1 of the Updated Safety. Analysis Report (USAR) provides the correct measurement range of all the Peak
- Recording Accelerographs (Peak Acceleration Recorders) listed in the .
USAR. Amendment 75 to Wolf Creek Generating Station Technical l Specifications has been previously approved by the NRC. -Amendment 75 l removes the Seismic Instrumentation Section from Technical .) Specifications to the USAR Section 16. Peak Acceleration Recorders provide no automatic safety function and , are not required for safe shutdown of the plant. Therefore, changing their measurement range cannot increase the probability of occurrence [ r of an accident previously evaluated in the USAR. Peak Acceleration l Recorders only design function is to provide seismic information I through scratch plates to operators and other personnel following a l seismic event. This revision does not change their design function. [ Therefore, the consequences of an accident are not increased. ; I Peak Acceleration Recorders are not associated with any safety-related i equipment. .Therefore, this revision cannot increase the probability [ of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR. There are no safety-related design functions or increase in challenges to and system, structure or component important to safety as-a result of this revision. , Therefore, the consequences of a malfunction of equipment important to i safety is not increased. This revision cannot create an accident of a . different type than any previously evaluated in the USAR, Changing [ the Peak Acceleration Recorder ranges will not introduce any new I equipment failure modes, nor does it create the possibility'of a single failure affecting multiple trains. This change does not create . the possibility of an equipment malfunction of a different type than 1 previously evaluated in the USAR. Because Amendment No. 75 to i operating license No NPF-42 of Wolf Creek Generation Station relocates ; the seismic Instrumentation requirements from the Technical l Specifications to the USAR, this revision cannot reduce the margin of [ safety as defined in the basis for any Technical Specification. i 5 i
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Page 179 of 183' ! i i 1 Safety Evaluation: 59 94-0181 Revision:0' h
. Replace Metrological Tower Temperature Transmitters This' modification'to non safety-related equipment replaces the NY-CAL
. temperature transmitters, which have been discontinued by the
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manufacturer, with Transmission Inc. instrumentation. This- , modification is being performed because of the unavailability of the j HY-CAL instruments. The transmitters affected are located on the : Meteorological Tower and. transmit temperatures to the Nuclear Plant Information System.
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Replacement of ' the obsolete HY-CAL transmitters will not increase the , probability of an accident nor increase the consequences of an- l accident previously evaluated in the Updated Safety Analysis Report' ('USAR) . This modification will not increase the consequences of malfunction of equipment'important to safety,previously evaluated-in' '! the USAR. { 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. , i i e b I f k a f P i
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-t Page 180 of- 183 '!
Safety Evaluation: 59 94-0182 Revision O L' Revision to Comatitment to Regulatory Guide.1.64 to Reflect Design verification Activities f This revision-to the Updated Safety Analysis Report'(USAR) changes,.in
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part, the commitment to comply with. Regulatory Guide 1.64 " Quality ! Assurance Requirements'for the Design of Nuclear Power Plants." This revision complies with the recommendations of Regulatory Guide- 1.64 with the following exception to Section C.2: " Design ve.l!.ication, as ; stated in Section 17.2.3.6 of the USAR, is performed by qualified i verifiers who are noc directly responsible for the design or the. , design change. In unusual' eases, the designer's supervisor may-perform the verification ifs he is the only technically qualified ; individual, the need for him to perform the review is approved and I documented in advance by the supervisor's management, and Quality Assurance audits monitor the frequency of the supervisor's review to guard against abuse." , i This revision is in accordance with ANSI N45.2.11-1974 and is an j administrative requirement only. .This revision does not affect the j design or operation of equipment assumed to perform to prevent accidents. This revision does not increase the probability of and
- accident previously evaluated in the USAR. Systems, structures and components are not affected by this revision.
-This revision 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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~i Attachment to CO 95-0028 Page 181 of- 183- j
. h safety Evaluation: 59 94-0183 Revision 3 0
' Technical Specification Clarification for Containment spray and safety Injection Room Coolers' ,
This clarification to Technical Specifications. identifies a single room cooler.is acceptable for supplying the Safety Injection and Containment Pump Rooms. Inoperability of che Containment Spray Pump , Room Cooler.is not cascaded to the related pump, provided that the ;{ Safety Injection Pump Room Cooler on the same train is' operable with no tubes plugged. The intent of this clarification is'to allow the cooler to be removed from service only for actual outage time to work j r on the cooler or if cooler leakage exceeds design allowed outage time to work on the cooler or if cooler leakage exceeds the allowed leakage rates. Operational expectation is to minimize the amount of time the cooler is out of service. , Using the Safety Injection Pump Room Cooler to cool the Safety Injection and Containment Spray Pump Rooms will not increase the probability of occurrence of an accident previously evaluated in the ., Updated Safety Analysis Report (USAR). Using the Safety Injection l Pump Room Cooler can keep a-suitable environment for both pumps. l Using the Safety Injection Pump Room Cooler to cool the Safety ! Injection and Containment Spray Pump Rooms will not increase ~the { consequences of an accident previously evaluated in the USAR because , the Safety Injection Pump Room Cooler can provide a suitable i environment for both pumps. Loss of the Safety-Injection Pump Room Cooler will cause the. loss of the Safety Injection and containment Spray Pumps. This. failure is bounded by the. loss of a single safety 'l train. Therefore,-there is no increase in the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. This clarification 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. e l
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[Q ,- p Attachment to' CO 95-0028 Page 182 of 183 (~ Safety Evaluation: 59 94-0185 _ Revision 0 Installation of Freeze Seal on Closed Cooling Water Lines This temporary modification provides for - freeze ~ seals on the Closed - Cooling Water System lines at the discharge of each Generator Isophase Bus Duct' Cooler. These seals must be used to isolate Closed Cooling. Water to allow for replacement of' flow switches (EB FS-041 and'EB FS-042) because the switches are located down stream of the Bus Duct g ~ Cooler isolation valves. The freeze seals are not applied simultaneously, allowing one Generator Isophase Bus Duct Cooler to be in service at all times during the switch replacement process. The freeze seals are applied to non safety-related closed cooling Water piping in the Turbine Building and the potential leakage of the system inventory, because of leakage of the freeze seal,- has no affect on equipment important to safety. There is no equipment important to safety that would be affected by the spraying of water from the closed cooling Water System caused by a failure of the freeze seal while the flow switch is removed. Failure of a freeze seal would not increase the probability or consequences of a malfunction of any equipment important to safety. This temporary 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. I' P
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.Page 183aof 183' i l
' Safety.Ryaluations- 59 94-0187 Revision 0 ,
Temporary Jumper on Secondary Spent Rosin Storage Tank Level t Transmitter This temporary modification to non safety-related instruments-installs ; jumpers in the ultrasonic transmitter HCLITS0034 to allow transmitter . HCLT0033 to function properly. .These instruments are connected in ' series in the current configuration. Both of these-level transmitters - monitor the level in the Secondary Spent Resin Storage Tank. This . modification only changes the electrical wiring on the backup level l transmitter for the Secondary Spent Resin Storage Tank. This modification will not affect resin sluicing activity. 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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