ML20136F920
| ML20136F920 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 12/31/1996 |
| From: | WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | |
| Shared Package | |
| ML20136F911 | List: |
| References | |
| NUDOCS 9703170112 | |
| Download: ML20136F920 (211) | |
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Attachment to ET 97-0017 Page i i
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WOLF CREEK NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station I
Docket No.: 50-482 Facility Operating License No.: NPF-42 l
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1 g ANNUAL SAFETY EVALUATION REPORT l
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. Report No.. 12 l
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' Reporting Period: January 1, 1996 through December 31, 1996 i
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l 9703170112 970311 PDR P ADOCK 05000482 PDR i
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-i Attachment to ET 97-0017 ;
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SUMMARY
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'This report provides' a brief ' description ' of ~ changes, tests, and experiments performed at Wolf Creek Generating Station pursuant to 10 CFR 50.59 (a) (1) .
. This report includes. summaries of.the. associated safety evaluations that were reviewed and found'to be acceptable by the Plant Safety = Review Committee for the - period beginning January 1,. 1996 and ending December 31, 1996. This '
- report -is submitted 'in accordance ~ with the requirements of 10 CFR 50.59 (b) (2) .
On..the basis of these evaluations of chat 49es , the following has been
-determined:
- . There'is no increase in.the probability of occurrence or the consequences-of an accident or-malfunction of equipment important to safety previously:
evaluated in the Updated Safety Analysis Report (USAR).
- There is no possibility that an accident or : malfunction ' of equipment.
important to. safety of a different type than any evaluated previously in 1
'the USAR may be created.
- The margin. of- safety as.-defined in .the basis for any Technical Specification is not reduced. -
, . Therefore, all items. reported herein' were determined not to involve ;an unreviewed safety: question.
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{ Attachment to ET 97-0017 I Page 1 of 209 1
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. Safety Evaluation: 59 93-0226 Revision:1 Positive Displacement Charging Pump Replacement j Revision 1 of this Unreviewed Safety Question Determination (USQD) 4 evaluates minor modifications to CVCS piping and pump foundations. ]
i l Also, the shutoff head of the Nornal Charging Pump (NCP), when _ l
! considering maximum suction head, is determined to exceed the existing
} discharge piping design pressure, which was not evaluated in the original USQD. The physical changes involved are consistent with
- carrent design criteria and original design codes, and are necessary to ensure proper fitup and installation of the vendor's equipment.
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~ The design function of the NCP and Chemical & Volume Control System I 4
(CVCS) is not being changed with this revision.
l Revision 0 of this USQD' evaluated replacement of the CVCS Positive y Displacement Charging Pump (PDP) PBG04 with a new centrifugal charging i
] pump for normal charging duties. Modifications to piping and pump j control and power supply were made to support the pump replacement. !
The pump is' passively safety related for maintaining reactor coolant pressure integrity, but is not required for the safe shutdown of the 4 plant, and is powered from a non-Class 1E electrical bus. Therefore, j it is not considered in accident mitigation analyses. Process instrumentation associated with the NCP performs passive safety functions of maintaining the system pressure boundary, but performs no active safety function with respect to control or indication. Power i
and controls are non-safety related.
l .Therefore, the modification was divided into two phases:
Phase 1 - Outage / Pre-outage tasks (PMR 04782)
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Phase 2 - Post outage tasks (PMR 04590) i The scope of this modification is to provide instructions and justification for all Phase 2 activities that may be implemented post-
. outage to support the PDP replacement. The following modifications are performed for the replacement of the PDP:
- 1) Replacement of the existing PDP with the new centrifugal pump and motor.
- 2) Replacement of existing hoist HKF19 with a larger capacity hoist to lift the new pump.
- 3) Replacement of existing PDP room fen coil unit SGLO7 with a higher capacity unit, due to increased room heat loads from the new pump motor.
- 4) Installation of a cross-tie line between the existing PDP recirculation line and the inle3 line to the Seal Water Heat Exchanger (SWHX), to allow cooling of recirculated water from the new pump.
Manual isolation valves are installed in the new cross-tie and existing recirculation piping.
I Attachment to ET 97-0017 Page 2 of 209
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balance line piping and supports, allowing room for installation of new in-line components.
- 6) Deletion of unncessary auxiliary supply lines and drain lines to/from the PDF, including component cooling water, demineralized water, instrument air, PDP gas purge eductor and tubing, and PDP oil level instrumentation.
- 7) Installation of new NCP flow control components and instrumentation, including a new flow control valve, and new flow ;
element and flow switch for NCP recirculation control.
- 8) Installation of new current transformers, ammeter and associated wiring changes in 4kV awitchgear PB03.
- 9) Reworking of the cable and conduit for the new room cooler. d Installation of a new thermal overload heater for the larger motor in the new room cooler.
- 10) Routing of the new electrical power and control cables to support ;
new pump motor and changes in control logic, which require new core bores and penetration seals.
The new pump will replace the PDP function of providing charging and RCP seal water flow during normal plant operations. The new pump is capable of continuously delivering all normal and maximum expected charging and seal water flowrates (87 to 132 gpm).
Neither accident related to CVCS malfunctions evaluated in the licensing bases Updated Safety Analysis Report (USAR) involves the !
existing PDP for initiating events, analysis input parameters, or accident mitigation measures. Therefore, the same will apply for the new NCP. The potential for a breach of the reactor coolant pressure boundary in the CVCS system has not increased as the modification I meets the applicable design codes. Therefore, this change will not increase the probability of occurrence of an accident previously evaluated in the USAR.
No change made by this modification involves components, piping or !
circuitry required for accident mitigation or safe shutdown, except !
for a change in the CCW (Component Cooling Water) pressure boundary. !
These changes will have the same effect on safety-related structures, l systems or components as the existing CVCS, CCW and other affected i systems. Therefore, this modification will not increase the !
consequences of an accident or a malfunction of equipment important to safety previously evaluated in the USAR.
This modification creates no new failure modes, and implementation )
would have the same effect on safety-related structures, systems or l components as the existing affected systems. Therefore, this change will not increase the probability of occurrence of a malfunction of ,
equipment important to safety previously evaluated in the USAR. l l
since no new credible failure modes are created by this change and failure of the NCP or other affected components and piping will not (
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l Attachment to ET 97-0017 Page 3 of 209 l prevent the CVCS and CCW systems from performing their safety i functions, the potential for the creation of a new type of unanalyzed I event is not created.
l No Technical Specifications are affected by this modification,
- herefore no acceptance limits are affected.
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Attachment to ET 97-0017 Page 4 of 209 l
l Safety Evaluation: 59 93-0228 Revision 0 Positive Displacement Charging Pump Replacement ,
This modification involves activities performed in support of the Positive Displacement Charging Pump (PDP) replacement with a l centrifugal type Normal Charging Pump (NCP) for delivering normal l charging flow to the Reactor Coolant System (RCS) and seal water to the Reactor Coolant Pumps (RCP).
The following activities are performed under the scope of this plant modification:
- 1) Replacement of existing hoist HKF19 with a larger capacity hoist to lift the new pump.
- 2) Replacement of existing PDP room fan coil unit SGLO7 with a higher j capacity unit, due to increased room heat loads from the new pump l motor. 1
- 3) Installation of a cross-tie line between the existing PDP l recirculation line and the inlet line to the Seal Water Heat Exchanger l (SWHX), to allow cooling of recirculated water from the new pump.
Manual isolation valves are installed in the new cross-tie and existing recirculation piping.
- 4) Relocation of CVCS PDP suction and discharge isolation valves and component cooling water isolation valves to allow system isolation for the PDP replacement.
- 5) Installation of new current transformers, ammeter and associated wiring changes in 4kV switchgear PB03.
- 6) Reworking of the cable and conduit for the new room cooler.
Installation of a new thermal overload heater for the larger motor in l the new room cooler.
- 7) Routing of the new electrical power and control cables to support new pump motor and changes in control logic, which require new core bores and penetration seals.
The modified CVCS piping and relocated valves are safety related, ASME Class 2, for maintaining the reactor coolant pressure boundary. The modified CCW piping and relocated valves are safety related, ASME Class 3, for maintaining the CCW system pressure boundry. The replacement room fan coil unit and hoist are both non-safety related.
Power supplies for the new NCP and room fan coil unit are non-Class 1E.
No components affected by this modification are required for safe shutdown of the plant, except for modified CCW piping, which must maintain pressure integrity. The CVCS piping and pressure-retaining components are safety related for maintaining the reactor coolant pressure boundary during normal plant operation. Piping modifications will not change any effect on safety related structures, systems, or components from the existing CVCS and CCW systems. The potential for a breach of the reactor coolant pressure boundary in the CVCS system
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has not increased. Therefore, this change will not increase the *
- j. probability of occurrence of an accident or malfunction previously I
evaluated in the Updated Safety Analysis Report (USAR).
! No change made by this modification involves components, piping, or j circuitry required for accident mitigation or safe shutdown, except ;
I for a change in the CCW pressure boundary. These changes will have l
{ the same effect on safety related structures, systems, or components '
as the existing CVCS, CCW and other affected systems. Therefore, this !
! modification will not increase the consequences of an accident or a l j malfunction previously evaluated in the USAR.
1, l This modification has no potential for the creation of a new type of unanalyzed event nor does it create the possibility of an accident of a different type than previously evaluated in the USAR. No Technical Specifications are affected by the modifications. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.
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Attachment to ET 97-0017 Page 6 of 209
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4 Safety Evaluation: 59 94-0175 Revision 1 Site Loop Gai-tronics Upgrade This modification revises U' pdated Safety Analysis Report (USAR) Figure I 9.5.2-2 to reflect as-built cable configuration. The number of l handset station additions on the site portion of the Plant Public Address System increased to the point where Centralized Power feed voltage dropped below the manufacturer's recommended minimum. The degraded voltage caused intermittent operation of a Plant Public Address System link to the offsite Education Center. In order to solve this voltage drop problem, a supplemental power cable was added to the site loop of the Plant Public Address System. {
This modification to power cable configuration affects only the non-safety related Public Address System. 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.
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Page 7 of 209 l
Safety Evaluation: 59 95-0106 Revision 3 0 s
Storage Location of SCBA Equipment for Fire Fighting
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This revision to the Updated Safety Analysis Report (USAR) Table 9.5A-1 revises the description of the location of self-contained breathing apparatus (SCBA) which is used for containment fire fighting and j damage control.
e USAR Table 9.5A-1 describes the method of compliance with the requirement as placement of SCBAs near the personnel airlock. In actuality the SCBA's are located in the Communications Corridor j stairwell at the Fire Brigade turn out area.
i Location of che SCBAs has no impact on any accidents or inputs and j assumptions used in the analysis of accidents described in USAR '
chapters 2, 3 or 15. Locating the SCBAs out of safety related areas removes the possibility of an SCBA becoming a missile hazard and affecting a safety related SSC. The location at the Fire Brigade turn l out area is considered an enhancement over their being located nearer
- to the containment entry point.
4 j This change will have no impact on accidents or malfunctions evaluated i
as the licensing basis and there is no potential for the creation of a new type of unanalyzed event.
l This is not a reduction in fire protection and the SCBA's are more readily available for Fire Brigade use for fire fighting activities.
Thus, there is no reduction in the margin of safety.
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Safety Evaluation: 59 95-0140 Revision 0 Replacement of Startup Strainers with Spacer Rings This modification involves changing out components by replacing startup strainers BLSS001 and BLSS002 (Reactor Make-Up Water Pump)
BGSS004 and BGSS005 (Chemical & Volume Control Chiller Pump) , and HBSS001 and HBSS002 (Reactor Coolant Drain Tank Pump) with spacer rings. The startup strainers are deemed as temporary equipment and shall be removed from the pump suction piping within their respective systems. These components were to be removed and replaced with spacer rings after initial flushing of the systens prior to startup operations. No safety related systems are affected by this change.
Replacement of startup strainers with spacer rings will not adversely affect the parameters of previously evaluated accidents in the U pdated Safety Analysis Report (USAR). Therefore the probability of occurrence of any of these accidents will not increase.
Replacement of startup strainers with spacer rings does not adversely affect the integrity of radiological barriers or the ability of the system to mitigate a radiological release. Therefore, the consequence of an accident previously evaluated in the USAR does not increase.
This modification will not degrade performance or seismic qualifications of safety systems designed to function in an accident.
This modification will not adversely affect equipment protection features, system redundancies or frequency of operation of the related safety systems. Equipment malfunction probabilities are determined using design specifications that assume spacer rings are installed.
Therefore the probability of an occurrence of equipment malfunctioning will not increase nor will the consequences of a malfunction previously evaluated in the USAR increase. Failure modes of safety related equipment will not change so there is no potential for the creation of a new type of unanalyzed event.
Replacement of startup strainers with spacer rings will not exceed or reduce the previously approved limits established in the Technical Specifications. Therefore, there is no reduction in the margin of safety.
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Page 9 of 209 Safety Evaluation: 59 95-0159 Revision 30 Effects of Temperature Above 150 Degrees Fahrenheit in the Upper Cavity Region This modification provides a single source document for the classification of the emergency lights as to whether they are Seismic Category II/I and require seismic mounting, are special scope and required for safe shutdown actions, or nen-safety related and used for personnel access or egress. This document includes the aiming requirements of each of the safety related lights that require
! aiming. Clarification of statements in the Updated Safety Analysis i Report (USAR) are also provided by this change. These changes should improve the ability of operations and Maintenance personnel to maintain and aim the emergency lights.
The emergency lights are not the initiators of accidents currently described in the USAR. No new failure modes are introduced by this modification. Therefore this change will not increase the probability cf occurrence or the consequences of an accident previously evaluated. The change does not increase the probability of occurrence l or the consequences of a malfunction of equipment important to safety than previously evaluated. No potential exists for the creation of a new type of unanalyzed event. No reduction in the margin of safety can result from this change.
Attachment to ET 97-0017 Page 10 of 209 Safety Evaluation: 59 95-0166 Revision 0 Delete Thermal Relief Valves From CCW System l
This modification removes Component Cooling Water (CCW) system thermal relief valves on the Centrifugal Charging Pump Oil Cooler Heat Exchanger (BGV0524, BGV0525), Safety Injection Pump Oil Cooler Heat Exchanger (EMV0188, EMV0189) , the Residual Heat Removal Pump Seal Cooler Heat Exchanger (EJV0156, EJV0157) and the Residual Heat Removal Heat Exchanger thermal relief valves (EJV084, EJV085) from the plant design and installs blind flange assemblies. The valves are not discussed in the Updated Safety Analysis Report (USAR). Since the l relief valves only provide a relief function when the component is out of service and the system is designed for the highest expected pressure it can be subjected to during operation, the valves may be removed. The only change to the USAR is a revision of system P& ids.
j The valves perform no design basis function during operating modes l except maintaining the pressure boundary.
This change 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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j Safety Evaluation: 59 95-0167 Revision 0 !
Revise' Main Dam Inspection Frequencies l j Specification C-403, Rev 5, " Periodic Surveillance of Non-safety i
Related Water Structures and Reservoir For the WCGS" has been revised
.for the following changes:
- f. 1) . Revised Table 5, " Frequency.of Monitoring" to decrease surveying of
- all monuments to once in 5 years, added column labeled forms to show-
! correct form to use for each type of surveillance, decreased the l underwater inspection of concrete structures to once in 5 years, decreased piezometer and inclinometer readings from a quarterly and
- semi-annual bases to an annual bases, added Sect 3.1.6 for the Service Spillway to make consistent with the text with an annual frequency, changed responsibility for Surveying from Maintenance to Engineering, and increased overall main dam embankment and saddle dam frequency to
! annual.
- 2) Added new paragraph to Section 3.1.2.5 to require monitoring the seepage near Monument 40.
- 3) Revised the Forms table in Section 8 to remove the column labeled
" Frequency" since this information is in Table 5 and was not always l consistent.
- 4) Added item 7 to Form B for documenting monitoring the seepage near Monument 40.
i The cooling lake, main dam and saddle dams are not considered back-up to the Ulimate Heat Sink (UHS) and are not considered safety related.
These surveillance changes do not affect any system, structure or component (SSC) nor change the performance of activities that are important to safe and reliable operation.
Section 2.5 of the Updated Safety Analysis Report (USAR) describes the inspection and performance monitoring program for the main dam, saddle dams,. baffle dikes and cooling lake. The revisions to the surveillance frequency will affect some of the descriptions and tables in this section that describe the inspection and monitoring during operation. Specifically.the change affects Table 2.5-90 and 91 which list the frequency for monitoring the performance instrumentation (pi'ezometers, settlement monuments and inclinometers) for the main dam, saddle dams and baffle dikes. Also, text changes were made on page 2.5-314, Rev 4 to clarify what frequency was initially monitored versus what has been monitored after reaching steady state and what frequency will be monitored from now on.
There are'no design basis accidents that could be impacted if any of the structures covered by this monitoring program were to fail. The main dam, saddle dams, baffle dikes and cooling lake are not safety related because they are not considered back-up to the UHS or any
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i Page 12 of 209 other SSC. Therefore, the probability of occurrence of an accident is not affected by this change nor are the consequences of an accident
- affected.
There are no credible accidents that this change could create, thus no.
accidents of a different type could be created. A failure of the main dam, saddle dam or baffle dike and the potential resulting loss of the
- cooling lake would prevent operation of the plant but would not i prevent the safe shutdown of the plant. Cooling water for safe i
shutdown is provided by the UHS. The UHS would not be affected by j failure of any of these structures. The Auxiliary Spillway located in the main dam is designed to pass the probable maximum flood, but failure of the dam and subsequent loss of the spillway would not adversely impact the flood analysis.
4 l The structures impacted by this change are not SSCs that are important
, to safety. There-is no credible malfunction of safety related
! equipment that can be attributed to a failure of any of the structures covered by this specification, i
The reductions in the surveillance frequency for these non-safety
- structures can not impact acceptance criteria established for fuel l cladding, RCS boundary, and containment. Thus no acceptance limits are identified that could be affected and the margin of safety is not i affected by this change, i
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Safety Evaluation: 59 95-0170 Revision:0 Filtration Upgrade l
This modification revises Updated Safety Analysis Report (USAR) Tables l 9.1-5 and 9.3-9 to reflect installation of .1 micron glass fiber
. filters in the Chemical and Volume Control System (CVCS) and the Spent Fuel Pool (SFP) cleanup system. The target size for all applications (RCS, Seal Return, Seal Injection, Boric Acid, SFP cleanup, SFP l Skimmer) is being reduced to .1 micron which is considered to be even more effective in removing iron oxides and cobalt particles than the currently approved .2 micron filter. This effort is part of the Wolf Creek Generating Station long range exposure reduction plan.
The CVCS and SFP filter cartridges are non-safety related and have no safety design. basis function. The CVCS filter housings however are safety related for pressure boundary considerations since integrity is assumed during accident mitigation, safety related cold shutdown and emergency boration. The spent fuel pool filter housings are non-safety related and perform no safety design basis function.
The use of glass fiber filter cartridges, even down to the .1 micron size, has no affect on accident analyses in any mode nor does it increase the probability or consequences of a malfunction of systems, structures or components important to safety.
It has been determined that the filters will not prevent the RCS and SFP chemistry from being controlled in ranges required by the Technical Specifications and plant procedures so margins of safety are not affected.
Accidents evaluated and referenced in Chapters 2, 3, and 15 of the USAR are not affected by the use of glass fiber filters in the CVCS and SFP cleanup system even down to .1 micron size. The use of this type of filter has no affect on the initiating events of these accidents. The use of this type of filter does not affect the response of plant systems to accidents analyzed in the USAR so the consequences of the accidents are not increased.
No new circumstances exist which could create a different type of accident. The .1 micron filter will remove smaller particles than the currently approved filters but the changeout criteria and radiation remain unchanged.
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Safety Evaluation: 59 95-0174 Revision 0 e i j Carbon Dioxide and Carbon Monoxide Monitor Abandonment This modification disables and abandons the carbon dioxide (CO2)and carbon monoxide (CO) monitors GKAIS0222 and GKAIS0223. These 3
instruments are designed to monitor CO and CO2 levels in the control room ventilation system. These monitors have s history of poor reliability and excessive drifting resulting 1:1 spurious alarms. They are not safety related and are not required for post accident
, monitoring.
These monitors are referenced in Updated Safety Analysis Report (USAR)
Sections 6.4 and 9.4 and depicted on USAR Figure 9.4-1. ]
There are no design basis accidents discussed or referenced in USAR l chapters 2, 3, or 15 that are impacted by this modification.
Disabling and abandoning these monitors will not create or affect existing accidents nor will it cause malfunctions to equipment important to safety. Disabling equipment not important to safety cannot create the possibility of a different type of unanalyzed event. There are no acceptance limits contained in the bases of any Technical Specification or other licensing basis documents that could be affected by this change.
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Attachment to ET 97-0017 l Page 15 of 209 l Safety Evaluation: 59 95-0176 Revision 0 l
Radwaste Filter Upgrade This modification upgrades the filter media to polypropylene and glass fiber in the radwaste systems and increases the micron rating of radwaste filters from 50 to 70 microns for the systems that process waste water. Updated Safety Analysis Report (USAR) Tables 9.3-13, 10.4-15, 11.2-1, 11.3 and 11.4-5 list micron ratings of the filters that are being changed.
The new filter cartridges are non-safety related and have no safety design basis function. The increase in micron rating of non-safety l related glass fiber filter or polypropylene has no effect on accident analyses in any mode evaluated in the USAR. This is because the fine filtration to protect the equipment is no longer required. The I effluent is discharged to the lake rather than recycled back to the system.
The use of the equivalent or better filter cartridges will not increase the probability of occurrence or the consequences of an i accident previously evaluated, nor does it create the possibility of !
an accident different from any previously evaluated. The change does not increase the probability of occurrence or the consequences of malfunction of equipment important to safety than previously evaluated. Since no acceptance limits are identified, the margin of safety is not affected by this modification.
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Safety Evaluation: 59 95-0178 Revision 0 Emergency Lighting Classification and Aiming This modification provides a single source document for the classification of the emergency lights as to whether they are two over one and require seismic mounting, are special scope and required for safe shutdown actions, or non-safety related and used for personnel access or egress. This document includes the aiming requirements of each of the safety related lights that require aiming. Clarification of statements'in the U dated p Safety Analysis Report (USAR) are also provided by this change. These changes should improve the ability of Operations and Maintenance personnel to maintain and aim the emergency l lights.
The emergency lights are not the initiators of accidents currently described in the USAR. No new failure modes are introduced by this '
modification. Therefore this change will not increase the probability ,
of occurrence or the consequences of an accident previously l evaluated. The change does not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety than previously evaluated. No potential exists for the creation of a new type of unanalyzed event. No reduction in the margin of safety can result from this change.
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Safety Evaluation: 59 95-0180 Revision 0 Domestic Water System Cross Tie to Domineralized Water System This modification revises P& ids M012AN01 and M012KD01 to show the portions of the KD (Domestic Water) system inside the Radiological controlled Area (:RCA) being supplied by the demineralized water system (AN). The KD system is used for wash down inside the RCA. The KD water has a high concentration of anions which deplete the resins bed ;
of the Radwaste system and therefore increases the quantity of l radwaste to be shipped and buried. The KD system also provides water to emergency eye wash and showers. The shower and eye wash stations will be removed and portable eye wash stations with body spray will be located at the needed locations.
There are no accidents identified that will be impacted by this design change therefore, the probability of occurrence of an accident and the consequences of an accident previously evaluated are not affected.
This change does not increase the probability of occurrence or the i consequences of malfunctions of equipment important to safety since no i malfunctions were identified. No potential for the creation of a new type of unanalyzed event was created. The margin of safety was not ;
affected by this change'since no acceptance limits were identified. l 1
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Attachment to ET 97-0017 Page 18 of 209 Safety Evaluation: 59 95-0181 Revision 0 Installation and Removal of Hydro Plugs in the Safety Injection Accumulator System This temporary modification evaluates rework of check valves EP8956D, BB8948D, EP8818D, EP8956C, EP8956B, EP8818B, and EP8956A while the refueling pool is flooded up. Hydro plugs will be installed downstream of BB8948D, EP8956C, EP8956B, EP8956A, upstream of EP8818D, and upstream and downstream of EP8818B. These hydro plugs will be installed at mid-loop, with fuel removed from the reactor vessel (Mode E). The hydro plugs will be removed after the work on the check valves has been completed and the plant is again at mid-loop Mode E.
Before use, the plugs will have been tested for the expected pressures to be experienced and tied off in the unexpected case that the plug should break free. An emergency closure device will also be stationed so that if a leak did develop, the leak could be contained. If the check valve is not manned, a temporary cover will be installed to preclude any leakage that may occur while unattended.
No safety related equipment would be jeopardized if any of the plugs were to fail and the refueling pool were to partially drain out of the top of the check valve. The containment building is designed to contain the entire contents of the Reactor Coolant system (RCS) and the Refueling Water Storage Tank (RWST) . Any safety related equipment is either above flood 1svel, properly protected, or is in its fail safe position. In additava, there is no danger of uncovering fuel since the plant will be in Mode E and the refueling pool and fuel pool will be isolated from each other.
There are no design basis accidents identified. The installation, removal, and use of the hydro plugs does not adversely affect any system, structure, or component.
No credible accidents or malfunctions of equipment important to safety would be created by the installation, use, and removal of the hydro plugs. There is no fuel in the core. The refueling pool and fuel pool are isolated from each other, so a spent fuel pool accident could not be created by the failure of the plugs. Also, no acceptance limits contained in the bases for the Technical Specifications or in the licensing basis documents have been identified, so the margin of safety is not affected.
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Attachment to ET 97-0017 Page 19 of 209 Safety Evaluation: 59 95-0182 Revision 0 Removal of Feedwater Pump Startup Strainers The feed water pumps, PAE01A, PAE01B and PAE02 are designed to run without startup strainers AESS0006, AESS0007 and AESS0008 within the suction piping. The intended design function of the strainers was to remove significant debris that inadvertently entered the system during construction of the plant. After startup system flushing, temporary startup strainers were to be removed and replaced with spacer rings.
A source for significant debris no longer exists in the AE (Feedwater) system.
Replacement of startup strainers with spacer rings will not adversely affect the parameters of previously evaluated accident probabilities reviewed in Updated Safety Analysis Report (USAR) chapter 15.
Therefore, the probability of occurrence of any of these accidents will not increase. This modification does not adversely affect the integrity of radiological barriers or the ability of the system to mitigate a radiological release. Therefore the consequence of an accident previously evaluated in the USAR will not increase.
This modification will not degrade performance or seismic qualifications of safety systems designed to function in an accident.
The change will not adversely affect equipment protection features, system redundancies or frequency of operation of the related safety systems. Equipment malfunction probabilities are determined using design specifications that assume spacer rings are installed.
Therefore, the probability of an occurrence of equipment 1 malfunctioning will not increase nor will the consequence of a malfunction increase.
The replacement of startup strainers with spacer rings will not create i a condition that falls outside the approved limits of the design l criteria, therefore there is no potential for the creation of a new type of unanalyzed event. This change will not affect the margin of safety because there is no reduction in the approved limits established in the Technical Specifications.
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Attachment to ET 97-0017 Page 20 of 209 1
i Safety Evaluation: 59 95-0184 Revision: 0 i Security Building Modification.
This modification provides for A) The installation of a Credit Union Room to be located on the
) southwest corner of the first floor of the Security Building.
B) The addition of an annunciator sign (similar to that at the plant entrance) to be installed in the area above the entrance turnstiles.
The power for the unit will come from non-emergency sources. l C) Modifications to the Lunch Area in the basement of the Security )
Building. These modifications include the removal of two non-load bearing block walls, modifications to the lighting in the lunch room area, and modifications to the communications devices in the lunchroom area.
I D) Modifications to some Plant Communications equipment located in the Security Building Basement. This would include the addition of two Gai-tronics desk sets and the addition of two speakers.
This modification does not affect any Design Basis accident nor does it create any type of credible accidents. It affects only non-safety related equipment (Security Building and con" .s are not safety related). Therefore, this modification will .sve no impact on accidents or nalfunctions evaluated as the licensing basis and there is no potential for the creation of a new type of unanalyzed event There are no Technical Specification requirements dealing with any portion of the Security Building, therefore any modification to the building cannot have any impact on any requirement.
Attachment to ET 97-0017 Page 21 of 209 Safety Evaluation: 59 95-0185 Revision 0 l Revision to Organisation to Reflect Resignation of the Vice President Engineering This revision to the Updated Safety Analysis Report -(USAR) provides for organization changes, specifically filling the position of Vice President Engineering. Personnel filling the position of Vice President Engineering continue to meet the ANSI requirements for-Engineer in charge. This revision only affects the resume for the Vice. President Engineering in Chapter 13.1 of the USAR.
Since this is only a change to personnel, equipment, procedures, tests or experiments will not be affected. 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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Safety Evaluation: 59 95-0185 Revision: 1 Revision to Organization to Reflect Appointment of the Vice President Engineering This revision to the Unreviewed Safety Question Determination (USQD) 59 95-0185 Revision 0 is due to the announcement of a permanent personnel assignment to the position of Vice President Engineering.
R. A. Muench has been assigned to the position of Vice President Engineering replacing W. B. Norton. This is a personnel change only and has no effect on the organization since Mr. Muench meets the ANSI requirements for Engineer in charge. This revision only affects the resume for the Vice President Engineering in Chapter 13.1 of the Updated Safety Analysis Report (USAR).
Since this is only a change to personnel, equipment, procedures, tests or experiments will not be affected. 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 ET 97-0017 Page 23 of 209 Safety Evaluation: 59 96-0001 Revision:0 Drawing Modification to Show 011 Sampling valve This modification revises P&ID M-12CF01, " Lube Oil Storage, Transfer and Purification System," to show an oil sampling valve, CFV-1000, which is mounted on the Lube oil Conditioner vendor skid. It was installed and originally supplied by the vendor along with the skid per manual M-171-00008. This drawing change reflects the as-built configuration. No new valve is added.
This modification is a dacument change only with no effect on plant equipment or performance.
This modification will have no impact on accidents or malfunctions evaluated as the licensing basis and there is no potential for the creation of a new type of unanalyzed event. There is no reduction in the margin of safety.
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Attachment to ET 97-0017 Page 24 of 209 Safety Evaluation: 59 96-0003 Revision 0 l
Revision to Liquid Continuous Release Set Point Procedure AP 07B-003, Rev. O "Offsite Dose Calculation Manual" supersedes procedure ADM 04-019 Rev.5 "Offsite Dose Calculation Manual." Major revisions to the superseded procedure include, 1)
Convert the ODCM to the Procedure Writer's Guide format, 2) Allow I other comparison programs, 3)' Update Radiological Environmental Monitoring Program (REMP) sample locations, 4) Define the age group for Liquid and Gaseous Organ Dose Calculations, 5) Update the Liquid Continuous Release Points Alarm Setpoint Calculations, 6) Change the Safety Factor for Liquid Effluents from 1.0 to 0.5, 7) Change the Cumulative Dose Factor in the Projected Dose Calculation, 8) Remove the Particulate and Iodine sampler requirements from the Containment Purge.
l The alert alarm setpoints for effluent radiation monitors are not referenced in any of the design basis accidents previously evaluated in the Updated Safety Analysis Report (USAR). Therefore, there is no impa(, on accidents or malfunctions evaluated in the USAR.
The alert alarm function serves no automatic function and is used for indication only, the high alarm provides the automatic functions for radiation monitors. These changes do not affect the ability of the effluent radiation monitors to isolate the effluent stream, so a new type of unanalyzed event is not created nor is the margin of safety affected. f 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. Therefore, prior NRC approval is not required prior to implementation of this revision.
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Attachment to ET 97-0017 Page 25 of 209 1
1 Safety Evaluation: 59 96-0004 Revision 0 l Abandon Turbine Flow Meter I This modification abandons in place turbine flow meter HAFT 1094 and removes or spares RAFT 1094's associated electrical components. !
HAPT 1094 was used to monitor the amount of off-gas processed from the i
Volume control Tank (VCT) to a waste gas compressor skid. This turbine flowmeter has worked intermittently since start-up. Due to j the intermittent operation of the turbine flowmeter, procedures to off- l j gas the VCT without using the turbine flowmeter are used. Since the i MAFT1094 is infrequently used and the VCT purge can be performed without it, HAFT 1094 can be abandoned in place.
Instrumentation associated with RAFT 1094 are mentioned in Updated Safety Analysis Report (USAR) Section 11.3.6 and USAR Table 11.3-5 and
- are depicted on USAR Figure 11.3-1.
There are no design basis accidents impacted by this modification.
4 RAFT 1094 and associated electrical components are not initiators of
- accidents nor are they equipment important to safety. Therefore, abandoning and/or removing this equipment cannot create a credible
, accident or a new type of unanalyzed event, nor create malfunctions to i equipment important to safety. In addition, there are no acceptance y
limits contained in the bases of any Technical Specification or other licensing basis documents that could be affected by this change.
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-Safety Evaluation: 59 96-0005 Revision 0 Change to component Cooling Water Operation This change to Updated Safety Analysis Report (USAR) sections 9.1.3.2.3.1 and 9.2.2.2.3 clarifies operation of the Component Cooling Water (CCW) and Spent Fuel Pool (SFP).
Review of USAR section 9.1.3.2.3.1 shows that the USAR is different than the operating procedures. In addition, review of USAR sections 9.1.3.2.3.1, section 9.2.2.2.3 and 5.4.7.2.4 found that there are inconsistencies in the USAR descriptions concerning operation of the CCW during refueling.
This change has no effect on design basis accidents and will not create any new credible accidents. This change 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.
No Technical Specification sections are affected therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.
Attachment to ET 97-0017 Page 27 of 209 Safety Evaluation: 59 96-0006 Revision:0 Updating Domestic Water Supply System P&ID and 3/4" Valve Installation This modification approves installation of an additional 3/4" valve in j the Domestic Water Supply system in the Turbine Building (hot water supply line from TKD06) that does not appear on the system P&ID.
Installation of the valve requires a revision to the system P&ID (M-12KD02), which is Updated Safety Analysis Report (USAR) Figure 9.2-17 sheet 2.
The addition of the 3/4" valve in the Domestic Water system does not have any impact on accidents or malfunctions previously evaluated as the licensing basis and there is no potential for the creation of a new type of unanalyzed event, The Domestic Water system serves no safety function and has no safety design basis. The change does not reduce the margin of safety as defined in the basis for any Technical specifications.
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Safety Evaluationt 59 96-0007 Revision 0 1
Water Treatment Raw Water From Fire Protection Header Connection The temporary Procedure TMP 96-004, " Water Treatment Raw Water From 1 Fire Protection Header Connection," provides raw water from the l cooling lake through fire protection during times when normal raw water from John Redmond is not available, The procedural changes do not affect the performance of activities that are important to safe and reliable operation. There are no design basis accidents identified or evaluated for the non-safety related WM (Makeup Demineralized Water) and FP systems in the Updated Safety Analysis Report (USAR) chapters 2, 3 or 15.
Since the systems' functions are not changed, no crediible accidents that could be created are identified.
Since the temporary procedure would not affect the. systems' failure modes, controls on activity performance, the level of qualification or the effect on equipment important to safety, no credible malfunctions of equipment important to safety are identified.
No acceptance limits are identified that could be affected, so the margin of safety is not affected by this temporary procedure change.
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Safety Evaluation: 59 96-0008 Revision 0 Wolf Creek Steam Generator Chemical Cleaning Equipment Installation outside Containment
, This change implements Procedure IP-2408-1 " Wolf Creek SG Chemical Cleaning Equipment Installation Outside of Containment." This
- procedure details the tasks required to install, leak test, and verify the operation of the PN Services steam generator chemical cleaning equipment which will be used during the eighth refueling outage to chemically clean the steam generators. This equipment is temporary.
Flood analyses are described in U pdated Safety Analysis Report (USAR)
Section 2.4.2. The site is designed such that all water is shed away l from the plant or if the shed capacity is exceeded, runoff would occur l over the roadways, etc., to the lake prior to reaching a level that
, would endanger the safety related equipment on the site. Design i Engineering evaluated the positioning of the chemical cleaning equipment in regard to these Flood analyses and specified limitations for equipment location and configurations such that~the peak runoff 1 i with the equipment in place does not exceed that included in the USAR. Additionally, the positioning of the tanks needed for steam j- generator chemical cleaning was evaluated to ensure no unacceptable j loading of underground safety related structures, systems or
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There are no new credible accidents or malfunctions of a different I
! type that could be created that would affect any SSC provided the restrictions defined by Design Engineering are met. IP 2408-1 does not affect the acceptance limits as defined by the Technical Specifications. Therefore, there is no reduction in the margin of safety.
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Attachment to ET 97-0017 Page 30 of 209 l
1 Safety Evaluation: 59 96-0008 Revision 1 Wolf Creek Steam Generator Chemical Cleaning Equipment Installation
] Outside Containment This change implements a new procedure IP-2408-1
- Wolf Creek SG Chemical Cleaning Equipment Installation Outside of Containment".
This procedure details the tasks required to install, leak test, and verify the operation of the PN Services steam generator chemical cleaning equipment which will be used during the eighth refueling outage to chemically clean the steam generators. This equipment is temporary.
Procedure IP-2408-1 considers using a SG chemical cleaning booster pump located in the corridor of the Auxiliary Building elevation 2000', outside the hot machine shop. This area is not diked and may cause the chemicals used to leak into the floor drain system (FD).
The FD system is non-safety related and its failure will not impact any safety related systems or components. In addition, there are no safety related components in the corridor of Auxiliary Building that could be affected by the pump leakage. Revision 0 of Procedure IP-2408-1 was evaluated by Unreviewed Safety Question Determination (USQD) 96-0008 and is described below.
Flood analyses are described in Updated Safety Analysis Report (USAR)
Section 2.4.2. The site is designed such that all water is shed away from the plant or if the shed capacity is exceeded, runoff would occur over the roadways, etc., to the lake prior to reaching a level that would endanger the safety related equipment on the site. Design Engineering evaluated the positioning of the chemical cleaning equipment in regard to these Flood analyses and specified limitations for equipment location and configurations such that the peak runoff with the equipment in place does not exceed that included in the USAR. Additionally, the positioning of the tanks needed for steam generator chemical cleaning was evaluated to ensure no unacceptable loading of underground safety related structures, systems or components (SSCs) (Emergency DG Fuel Oil tank and ESW pipe and electrical duct bank) would occur.
There are no new credible accidents or malfunctions of a different type that could be created that would affect any SSC provided the restrictions defined by Design Engineering are met. IP 2408-1 does not affect the acceptance limits as defined by the Technical Specifications. Therefore, there is no reduction in the margin of safety
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Attachment to ET 97-0017 Page 31 of 209 i
8 Safety Evaluation: 59 96-0009 Revision 0 Supply Electrical Power to Essential Service Water Heaters This temporary modification allows the temporary connection of two non- l safety related 20KW electric heaters tc the NG (Low Voltage-480 V) '
l system safety related buses. The purpose of this modification is to
] remove and prevent the build up of ice on the ESW (Essential Service i
Water) traveling screens.
Since no components will be overloaded by this configuration, this condition will not increase the probability of occurrence of an accident. The circuit is protected by a safety related breaker and has been determined not to affect the voltage level of the safety related buses during an accident. Therefore, this condition will not 4
increase the probability of occurrence of a malfunction of equipment important to safety.
No increase in the consequences of an accident or a malfunction of equipment important to safety has been introduced by the addition of the non-safety related heater load as it has proper coordination and i is not a significant enough load to overload the bus or the diesel
- generator in an accident or during normal operation. There is no
, potential for the creation of a new type of unanalyzed event. The margin of safety as defined in the bases for any Technical Specification is not reduced by this temporary modification.
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- Safety Evaluation
- 59 96-0010 Revision 0
- Supply Power to Essential Service Water System to Prevent Freezing i This temporary modification allows the temporary connection of up to I
three non-safety related 20KW electric heaters to the NG system (Low i
- - voltage-480V) safety related buses (NG005EGF2). This is being done to j j remove and prevent the build up of ice on the ESW (Essential Service
{ Water) traveling screens.
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Since no components will be overloaded by this configuration, this condition will not increase the probability of occurrence of an
! accident. The circuit is protected by a safety related breaker and
} has been determined not to affect the voltage level of the safety
- related buses during an accident. Therefore, this condition will not increase the probability of occurrence of a malfunction of equipment
, important to safety.
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No increase in the consequences of an accident or a malfunction of 4
equipment important to safety has been introduced by the addition of the non-safety related heater load as it has proper coordination and is not a significant enough load to overload the bus or the diesel I generator in an accident or during normal operation. There is no potential for the creation of a new type of unanalyzed event. The margin of safety as defined in the bases for any Technical l Specification is not reduced by this temporary modification.
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Safety Evaluation: 59 96-0011 Revision 0 Use of Fire Protection Hose Station to Melt Ice at Essential Service Water Pumphouse pumphouse l This temporary modification allows the use of fire protection hose station KCHR141 in the B train ESW (Essential Service Water) Pump room to melt ice at the ESW Pumphouse with ESW Pump B providing the water.
A review of Updated Safety Analysis Report (USAR) Chapter 15 accidents I revealed the use of ESW hose station water to melt ice at the pumphouse will not affect the initators of any previously evaluated accidents. Therefore, the probability of accident occurrence is not increased. The use of hose station KCHR141 will not affect the barriers for confinement or radioactive materials and will not increase the consequences of any previously evaluated accident.
This temporary modification will not increase the probability of occurrence of a malfunction of equipment important to safety because the heat loads for post LOCA operation can still be removed even in the unlikely event of manual isolation valve failure. The Fire Protection System Header will not be impaired.
Consequences of a malfunction of equipment important to safety will remain within the bounds of previous analysis because the operability of this equipment will not be affected by using water from the ESW hose station.
Evaluation of this temporary modification has concluded that it lacks the possibility of creating a different type of accident because operability of the Fire Protection System and the ESW System in removing heat loads post LOCA has not been adversely affected. There are no unique challenges to equipment important to safety created nor is any unique degradation to this equipment expected.
Technical Specification 3.7.4 bases and the USAR have been reviewed and no reduction to the margin of safety will result.
Attachment to ET 97-0017 Page 34 of 209 Safety Evaluation: 59 96-0012 Revision 0 Drawing Correction to Mastewater Treatment Facility This modification revises P&ID M-12WT03 in Up dated Safety Analysis Report (USAR) Fig. 9.2-25 to show a pipe cap on the overflow line (112-HBD-2") for the acid storage tank (TWT 01) in the Wastewater Treatment Facility. This is an administrative change only. The Wastewater Treatment System serves no safety related function. Failure of the system will not compromise any safety related systems nor prevent safe shutdown.
This change 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 Evaluation: 59 96-0013 Revision 0 l Temporary Modification of Essential Service Water Warming Line This temporary modification installs a 3/4" compressed air source at J EFV0318 and EFV319 in the Essential Service Water system (ESW). This l will allow sparging air to be directed into each (A and B train) ESW warming line. The purpose for this air flow is to assist in 1 mitigating ice formation at the trash racks by providing agitation and I mixing. This temp mod also installs a 2" hot water source to the A &
B train ESW warming lines. The spool piece in the chemical addition line is being removed and a hose from a vendor's truck mounted hot water heater is being connected to the 2" flanged connection on each , j train. The purpose for this hot water slow is also to assist in mitigating ice formation at the tracn racks by providing agitation and mixing. This temporary modification is an enhancement to warming system operation as described in Updated Safety Analysis Report (USAR)
Section 9.2.1.2, Essential Service Water System and USAR Figure 9.2-2.
Per USAR Table 3.2-1, " Classification of Structures, Components, and Systems," the Essential Service Water System and intake structure are ,
Seismic Category I, Nuclear Safety related. The new components have l
no direct safety related functions, except for pressure boundary, but '
may be important in maintaining the operability of the safety related ESW system during periods when ambient temperatures are extremely low. The new components are located such that their failure cannot affect any other SSCs important to safety.
J No systems or equipment important to safety, other than that discussed above, are germane to this evaluation. Many potential accidents evaluated in the USAR depend on the ESW system to remove decay heat and thus prevent or limit core damage. These modifications are enhancements to help ensure the ESW system is available and are not capable of adversely affecting the ESW system. Therefore, installing these modifications does not affect the consequences on any of the accidents evaluated in the USAR. A failure of the ESW system is not i an initiator of any accident or event previously evaluated in the USAR. Therefore, changes to them cannot affect the probability of occurrence on any of the accidents evaluated.
The ESW system will still function as required. No new failure modes or failure effects are being created. The new components are located such that their failure cannot affect any other systems, structures, or components important to safety. Therefore there is no creation of a new type of unanalyzed event. There is no margin of safety defined in the basis for any Technical Specification or Safety Evaluation Report.
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Attachment to ET 97-0017 l Page 36 of 209 safety Evaluation: 59 96-0017 Revision:O Installation of sparging Manifolds in Front of Essential Service Water Inlet structure Trash Racks This temporary modification removes grating above the ESW (Essential Service Water) Inlet Structure Bays so that air sparging manifolds can be installed in front of the ESW Inlet Structure trash racks in front of the bays to ESW pumps A and B. Hot water heJes will also be routed to these bays to melt the ice. The hot water .upply hoses and air sparging manifolds will be utilized to facilitate ice removal. These actions are being performed to supplement the 30 inch warming line to each bay which is designed to prevent ice accumulation on the trash j racks and traveling screens. l 1
Air sparging manifolds and hot water hoses in front of the ESW Inlet Structure trash racks are not described in the Updated Safety Analysis Report (USAR). Removal of the grating to provide access for these devices is not addressed in the USAR.
Because no design basis accidents were identified as being affected, the probability of occurrence of an accident is not increased nor are the consequences of a previously evaluated accident.
The ESW system single failure analysis is not affected by use of the sparging manifolds because they cannot limit flow or damage equipment downstream of the trash racks, therefore the probability of occurrence of a malfunction of equipment important to safety and its consequences are not increased.
No circumstances have been introduced which would create a different type of accident. No new types of potential single failures have been created, so no new types of malfunctions of equipment important to safety have been created.
l Since neither the operability or cooling capacity of the ESW system is affected, no margins of safety are affected.
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Attachment to ET 97-0017 Page 37 of 209 Safety Evaluation: 59 96-0018 Revision 0 Seal Water Heat Exchanger to Charging Pump Vent Valve Position This change consists of revising P&ID M-12BG03, " Chemical & Volume Control System," to show valve BGV0477 in the normal *1y closed position. BGV0477 is located on the return line (BG304HCB-3) from the seal water heat exchanger (EBG03) to the charging pump suction header. BGV4077 is used for filling and draining activities, but is normally closed and capped during plant operation. The subject components are part of the Seal Water Return Subsystem. The basic function of the Seal Water Return Subsystem is to take the coolant leaking past the No. 1 seal on the (4) RCPs, transfer this leakoff through the seal water return filter and the seal water heat exchanger to the suction side of the charging pumps, or by the alternate path to the Volume Control Tank, This is an administrative change that will have no impact on plant operations as described in the U pdated Safety Analysis Report (USAR).
In addition, the change does not have the potential to affect plant operation, prevent safe shutdown or maintain the plant in a safe shutdown condition.
This change 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 ma: gin of safety as defined in the basis for any Technical Specification.
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Attachment to ET 97-0017 Page 38 of 209 Safety Evaluation: 59 96-0019 Revision 0 Essential Service Water Warming Line Temporary Modification This temporary modification removes a spool piece from the chemical addition line. The chemical addition line is flanged off and taken
.out of service and a valve placed on the other end where the spool piece attached and closed. Another change removes the pipe cap from a vent valve and installs a check valve.
The Essential Service Water system is not an accident initiator, it removes heat from various components during an accident. This change will not increase the probability of occurrence or the consequences of an accident previously evaluated in the Updated Safety Analysis Report (USAR).
The ESW system will be able to perform its accident function and this modification will not change any single failure assumptions assumed during an accident nor increase the probability of occurrence or the consequences of a malfunction of equipment previously evaluated in the USAR, This modification does not create the possibility of a different type of unanalyzed event or affect tne margin of safety.
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Attachment to ET 97-0017 Page 39 of 209 Safety Evaluation: 59 96-0020 Revision 0 Secondary Sampling Cooling System Modification This modification to the secondary sample cooling system involves elimination of the failed chiller unit ERM01 and replacement with a l new heat exchanger that will be served by the central chilled water system chiller units SGB01A/B. The old RM chiller unit equipment will be removed. The cooling water supply to the condensate Demineralizer Sample Chiller EAK04 will be converted from closed cooling water to chilled water. The Central Chilled Water System (GB), the Condensate ;
Demineralizer System (AK) and the Secondary Sampling System (RM) are )
non-safety related. Failure of these systems will not affect safety '
related systems and will not prevent the safe shutdown of the plant.
The system P& ids (M-12GB01, M-12RM01, M-12EB01, M-12AK03) are changed by this modification.
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The GB, AK, and RM systems are not initiators for previously analyzed I accidents as documented in the Updated Safety Analysis Report (USAR). l They do not directly affect safety related systems and thus could not i increase the consequences of previously analyzed accidents or l malfunctions of equipment important to safety. The modification to {
these systems could not increase the probability of occurrence of )
malfunction of equipment important to safety. There is no potential for creation of a new type of unanalyzed event. There is no effect on l l
margin of safety. l l
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Safety Evaluation: 59 96-0022 Revision 0 Installation of Hydro Plugs to Allow Work On Reactor Coolant System check valves This activity reworks check valves EP8L56D, BB8948D, EP8818D, EP8956C, EP8956B, EP8818B, and EP8956A while the refueling pool is flooded up.
Hydro plugs will be installed downstream of BB8948D, EP8956C, EP8956B, EP8956A, upstream of EP8818D, and upstream and downstream of EP8818B at mid-loop, with fuel removed from the reactor vessel (Mode E). The Safety Injection Pumps will be tagged out and the Safety Injection Accumulators will be depressurized. The hydro plugs will be removed after the work on the check valves has been completed and the plant is j again at mid-loop Mode E. Before use, the plugs will have been tested '
for the expected pressures to be experienced and tied off in the unexpected case that the plug should break free. An emergency closure device will also be stationed so that if a leak did develop, the leak l could be contained. If the check valve is not manned, a temporary l cover will be installed to preclude any leakage that may occur while unattended.
No safety related equipment would be jeopardized if any of the plugs ;
were to fail and the refueling pool were to partially drain out of the top of the check valve. The containment building is designed to contain the entire contents of the Reactor Coolant System and the Refueling Water Storage Tank. Any safety related equipment is either above flood level, properly protected, or is in its fail safe position. In' addition, there is no danger of uncovering fuel since the plant will be in Mode E and the refueling pool and fuel pool will be isolated from each other.
There are no design basis accidents identified. The plant is in Mode E and the Clearance Order will preclude connection between the Refueling Pool and Fuel Pool. The installation, removal, and use of the hydro plugs does not adversely affect any system, structure, or component.
No credible accidents or malfunctions would be created by the installation, use, and removal of the hydro plugs. There is no fuel in the core. The refueling pool and fuel pool are isolated from each other, so a spent fuel pool accident could not be created by the failure of the plugs. Also no acceptance limits contained in the bases for the Technical Specifications or in the licensing basis documents have been identified so the margin of safety is not affected.
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l Safety Evaluation: 59 96-0023 Revision 0 .I control of Refueling Pool Concentration During Decontamination and Rinse Water Additions This procedure change to RPP 08-115, " Control of Refueling Pool Concentration During Decon and Rinse Water Additions," was developed to allow the use of demineralized water for deconning of tools and i
- equipment in the Refueling Pool. Administrative controls will be put i l in place to prevent a slug of unborated water from being swept into I the core region. These controls include 1) Decon the equipment at the farthest point possible from the Reactor Vessel, 2) Use the minimum amount of demineralized water required, and 3) Sample the Pool at most probable location for dilution to ensure minimum boron concentration is maintained.
t These changes affect only the procedural controls applicable for Mode 6 and Mode E (fuel removed from the reactor vessel). This revision i does not involve any hardware changes. There will be no change to
)] normal plant operating parameters or accident mitigation capabilities. Therefore, there will be no increase in the probability of any accident occurring due to these changes.
The Technical Specification limits on Mode 6 boron concentration will
- be met therefore, there will be no increase in the consequences of an ocuident or malfunction of equipment important to safety.
These changes do not involve any design change nor are there any 1
changes in the method by which any safety related plant system performs its safety function. Therefore, there will be no increase in the probability of occurrence of a malfunction of equipment important j to safety due to these changes.
Since Administrative Controls are established to minimize the likelihood of a rapid boron dilution caused by a slug of unborated water passing through the core, the possibility of an unanalyzed event
- is not created. Also there will be no impact on margin of safety as
- defined in the basis for any Technical Specification.
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Attachment to ET 97-0017 Page 42 of 209 A
Safety Rvaluation: 59 96-0025 Revision *0
- Flux Doubling Circuitry Disabled This change allows disabling the flux doubling circuitry by declaring both Source Range detectors inoperable if the Reactor Coolant System
-(RCS) boron concentration is greater than 2400 ppm in Mode 5 and boron dilution paths are isolated. This is done in order to comply with a A
Limiting condition for Operation (LCO) .
i This change will not increase the probability of occurrence of an accident or the consequences of an accident previously evaluated in the Updated Safety Analysis Report (USAR). The plant will be operated in accordance with a Technical Specification on the source range instrumentation which controls BDMS (Boron Dilution Mitigation System)
- instrumentation. No credible accidents could be created by this activity.
This change will not increase the probability of occurrence or the j consequences of a malfunction of equipment important to safety' previously evaluated in the USAR. No credible malfunctions of
- equipment important to safety will be affected by this activity. All requirements of the Technical Specification will be followed to prevent an inadvertent dilution and no negative reactivity would be
! seen by injecting Refueling Water Storage Tank contents to an RCS containing greater than 2400 ppm boron.
No acceptance limits which are contained in the bases for the i Technical Specifications are affected by this activity.
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a Attachment to ET 97-0017 Page 43 of 209 Safety Evaluation: 59 96-0027 Revision 0 Changes Within Fire Areas A-1 and A-19 This modification approves the addition of a consolidated dress out
] area in Fire Area A-1 room 1130 on the 1974 elevation of the Auxiliary i Building. Existing Health Physics (HP) dress out areas in room 1122 of Fire Area A-1 and room 1504 and 1506 on the 2047 elevation of the !
1 Auxiliary Building are being removed.
l Additionally, the Hot Tool Room in room 1122 has been enhanced to consist of closed metal cabinets, and the majority of stored combustibles in rooms 1128 and 1129 have been relocated to the j Radwaste Storage Building.
1 I Also, the Updated Safety Analysis Report (USAR) Fire Hazards Analysis
, (USAR Section 9.5B) is updated to reflect combustible fire loads for Fire Areas A-1 and A-19 as follows:
1
. a) The description of the anticipated transient combustible materials 2
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was revised from identifying a specific material, to identifying a i combustible material classification consistent with the fire
) protection administrative procedures. Descriptions of the fixed combustible fire loads were also revised to this format where appropriate, b) The fixed and transient combustible loads for rooms within the two Fire Areas were revised to be consistent with current plant i configuration and storage practices,
] The above changes represent a reduction in the fire loading for all 3 areas with the exception of Fire Area A-1, room 1130 where the consolidated HP dress out area has been added. Because the fire load has been reduced in other affected areas, no specific evaluation is required and the changes are bounded by previously performed fire hazards safe shutdown analysis. The increased fire load in room 1130 1 was evaluated for its impact on previously performed fire hazards safe
- shutdown analysis.
- The Electrical Fire Hazards Analysis and the applicable USAR 9.5B
, sections have been reviewed and this change in location and size of i combustible fire load does not affect any of the inputs, assumptions or results of the Fire Hazards Analysis for room 1130 or Fire Area A-1.
The fire load and presence of the dress out area does not impact the i ability to achieve safe shutdown in the event of a fire in the room or i
the Fire Area.
I USAR chapters 2, 3 and 15 have been reviewed and the changes do not
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Attachment to ET 97-0017 Page 44 of 209 I
impact the accident analysis or assumptions utilized for a design basis fire or the accidents described in these chapters.
- There are no acceptance limits in the USAR, Safety Evaluation Report,
( or Technical Specifications affected by this change and there is no ,
reduction in the fire protection for room 1130 or Fire Area A-1. l l
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Attachment to ET 97-0017 l Page 45 of 209 i Safety Evaluation: 59 96-0028 Revision 0 l l
Pressuriser Safety Valve Operability During Shutdown This change adds a footnote to the MODE 5 applicability to Limiting Condition for Operation (LCO) 16.4.1.1, which is relocated Technical '
Specification 3.4.2.1, Safety Valves. This LCO requires that a minimum of one pressurizer Code safety valve be OPERABLE, with a lift i setting of 2485 psig i 1%, in MODES 4 and 5. The footnote allows an i
exception to this requirement when the reactor coolant system (RCS) is vented to atmosphere by a 1 square inch or larger vent. ;
l As explained in Bases Section 3/4.4.2, these valves operate to prevent the RCS from being pressurized above the RCS safety limit of 2735 I I
psig, and the relief capacity of a single safety valve is adequate to )
relieve any overpressure condition which could occur during shutdown '
(typically referred to as cold overpressure protection). As explained l l
in Bases Section 3/4.4.9, it is acceptable to provide this cold I overpressure protection function by removing a pressurizer code safety l valve (as an alternative to opening an RCS vent) . I j
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The pressurizer Code safety valves are removed from their locations l each refueling outage for bench testing to verify operability and to I check and adjust setpoints. Because of the wording of Technical Specification 3.4.2.1, one valve has been left in place while the other two were removed for testing, then the third valve has been j removed after at least one of the first two are replaced and returned to operation. This practice is inconvenient, inefficient and time-consuming. Thus, the purpose of this change is to revise the wording in Technical Specification 3.4.2.1 to allow all three safety valves to )
be removed at the same time. l Updated Safety Analysis Report (USAR) sections 3.9(N).3.2, 5.2.2, 5.4.13, 5.4.7, 5.4.10, 5.4.13, 7.6.6 and 9.3.4 were reviewed for potential impact with respect to the low-temperature overpressure protection function of the pressurizer safety valves. Since this change affects only the shutdown Modes 5 and 6, then no design basis accidents, as discussed in USAR Chapter 2, 3 and 15 are affected by this change. )
1 Bases Section 3/4.4.9 states that it is acceptable to remove one of the Code safety valves instead of opening an RCS vent in order to provide RCS cold overpressure protection. Removing one Code safety valve provides a much larger vent path than opening an FIS vent.
Thus, the only effect of removing all three valves at one time would j be to enlarge the RCS vent path even more, which would not affect any l credible accident scenarios. Similarly, enlarging the RCS vent path
- j. by removing all three valves at one time would not adversely affect j any equipment important to safety and would not affect any credible
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malfunction of equipment scenarios. These valves are actuated by i
system pressure and do not directly interact with any other equipment important to safety.
j Therefore, this change will have no impact on accidents or
] malfunctions evaluated as the licensing basis and tha.re is no
- potential for the creation of a new type of unanalyzed event. There
- is no reduction in the margin of safety.
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Attachment to ET 97-0017 Page 47 of 209 Safety Evaluation: 59 96-0029 Revision 3 0 Addition of Check Valve to Chemical and Volume control system This modification adds Chemical and volume control- (CVCS) check valve BGV630 and. replaces the existing pump PBG04 suction side start-up strainer spacer ring and the associated flanges with a single welded pipe spool piece. The addition of the valve and the spool piece do not change any design parameters previously evaluated for the operation of plant / system.
No check valve exists to prevent backflow from the Reactor Coolant Pump Seal Water Return and Excess Letdown line 195-HCB-2", lines 501-
. HCB-2" and 161-HCB-2", when the Normal Charging Pump (NCP) is not running and valve BGV807 is left open. The flow would then turn the ,
I pump in a reverse direction causing additional unnecessary wear and tear to the rotating parts of the pump. Callaway has installed the check valve in their NCP modification for the same reason.
Also, a start-up strainer spacer ring and its associated flanges, existing on suction side of pump PBG04, are not required for the plant operation. The ring and the flanges have no design basis tunction.
The designation of valve BGV8394 has been revised to restore the original designation. Also, for line 161-HCB-2" previously designated as going to " Volume control Tank" has now been designated as " Seal Water Heat Exchanger" for consistency. These are administrative changes and do not impact the system function.
The above changes do not have any effect on the plant design bases nor any existing accident scenarios nor any acceptance limit.
The pipe stress run analysis performed in revision 2 of PMR 04590 was reperformed in revision 3 of the PMR. As a result of the new analysis, pipe stresses in high energy lines are altered. However, the safety evaluation previously performed for the pipe break analysis in USQD 59 93-0226 still remains valid, and therefore no new safety evaluation is required in this USQD. The change does not have any effect on the plant design bases nor any existing accident scenarios.
The margin of safety is not affected.
Attachment to ET 97-0017
. Page 48 of 209 i
Safety Evaluation: 59 96-0030 Revision 0 j
Guidelines for Replacement of Carbon Steel Piping That Has Been Degraded Due to Flow Accelerated Corrosion l The purpose of this modification is to provide guidelines for the replacement of existing carbon steel piping that has been degraded due to Flow Accelerated Corrosion (FAC), with a low alloy steel which improves piping resistance to FAC.
The pipe replacement does not change the cross sectional properties or geometric configuration of the piping system, or significantly affect j the mechanical propertiso, therefore there is no adverse effect on
- existing safety margins or structural integrity of the affected piping system. The change is an enhancement to the original design by I providing an increased resistance to FAC. All resultant stresses will
! remain acceptable within code allowables, therefore the probability of occurrence of an accident previously evaluated in the Updated Safety l Analysis Report (USAR) is not increased. The pipe replacement does not adversely affect any system, component or procedures required to mitigate the consequences of an accident previously evaluated in the USAR. All functions will continue to be performed, therefore the
! consequences of accidents previously evaluated in the USAR will not be
- increased.
Ductile fracture, corrosion, erosion / corrosion, loss of mechanical properties, excess strain, mechanical creep etc., are credible failure modes for which this change has been evaluated, and therefore this change does not introduce a new credible failure mode. Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR is not increased, nor is the possibility of a different type of malfunction important to safety created.
The piping change is a design enhancement which does not adversely affect equipment that is important to safety. Therefore, the replacement does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the USAR.
The piping change is an enhancement to the original design by providing an increased resistance to FAC. All resultant stresses will remain acceptable within code allowables. Therefore, the possibility of an accident of a different type than any previously evaluated in the USAR will not be created.
The margin of safety will not be reduced.
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Safety Evaluation: 59 96-0031 Revision:0 Clarification of 60 Degree Minimum Temperature From a review of the Mechanical / Nuclear Design Criteria document (:M-000) Section 2.5.3, it has been concluded that the 60*F min, normal temperature is based on operator comfort and not equipment operability. Heating systems are designed to be capable of maintaining this minimum temperature under normal design conditions. ;
Design calculations and Specifications reference this 60*F value as l
the minimum normal temperature.
However, at times the temperature of these areas have been lower. The primary concern of a lower ambient air temperature is thickening viscosity of oils and lubricants and hardening of elastomeric subcomponents such as gaskets.
Based on a review of equipment important to safety in the Electric Penetration Rooms (1409/1410) and the Charging Pump Rooms (1107/1114),
these areas may be as low as 45'F without challenging operability.
The design basis documents will continue to reflect the nominal minimum design value of 60*F.
There are no accidents in the USAR that are initiated by low temperature in the electrical penetration rooms or the charging pump rooms. Thus, this change could not increase the probability of accidents previously analyzed. !
A lower initial operating temperature (45*F to 60*F) for the affected rooms might actually improve equipment function by lessening the chance of overheating. No analyzed accidents create consequences of low room temperature in the affected rooms. Thus, the consequences of accidents previously analyzed cannot be increased by the lower room temperature.
For the affected rooms, there are no malfunctions of equipment important to safety previously analyzed, that were initated by a low room temperature. Thus, no increase in probability of occurrence is possible.
l No malfunctioning equipment is postulated resulting from low temperatures in the affected rooms. Thus, no increase in consequences of malfunctioning equipment is possible. No potential for the creation of a new type of unanalyzed event is created. There are no Technical Specification bases that deal with a low room temperature {
limit for the affected rooms. 1 l
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t Safety Evaluation: 59 96-0032 Revision 0
, Acceptability of operation With Polypropylene Filter Membrane Material i in the Spent Fuel Pool
, This modification is required to determine the acceptability of operating with polypropylene filter fragments in the Spent Fuel Pool j (SFP) and/or the Reactor Coolant System (RCS). Once the SFP and the refueling cavity are connected to perform refueling activities, a potential for the foreign material entering the RCS will exist. Since
.4 the presence of foreign material in the SFP and RCS creates the potential for damage to system components, evaluation is necessary to justify operation of Wolf Creek Generating Station (WCGS) . The Updated Safety Analysis Report (USAR) does no,t specifically allow f operation of WCGS with foceign materials within the RCS or the SFP.
1 The probability of occurrence of an accident previously evaluated in j the USAR will not increase due to foreign material within the SFP or the RCS. Any' accidents assumed to occur will not occur at any greater frequency due to the presence of foreign material. The basis for this
, is the continued applicability of the safety analysis design parameters. Foreign material will not affect the initiators of an I event, and thus, the probability of occurrence is not increased.
a The consequences of an accident previously evaluated in the USAR will 1
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! not increase. The effect of foreign material has been evaluated and I i has been found to have no significant effect since the safety analysis design parameters remain valid, i-The probability of a malfunction of equipment important to safety as previously evaluated will not be increased. Foreign material will not degrade the performance of any safety system assumed to function in the accident analyses. The consequences of a malfunction of equipment will not be increased. The impact of the changes on the safety analyses has been evaluated and the results indicate that safcty limits continue to be met.
+
Foreign objects have been evaluated and it has been shown that the possibility for the creation of a new type of unanalyzed event is not
{ created by their presence in the SFP or the RCS. )
Evaluation of foreign material'has shown that safety analysis I conclusions remain valid and that the margin of safety is not reduced.
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Safety Evaluation: 59 96-0033 Revision: 0 Temporary Radwaste Storage Tanks
{ This temporary modification provides temporary storage tank (s)and pump skid with hose connections to the Liquid Radwaste System. The storage tank (s) will store the liquid radwaste demineralizer effluents during the refueling outage and condenser modifications. The temporary pump j
will transfer the demineralizer effluent back to the radwaste system ;
- for further processing. The temporary skid will be located outside l the radwaste_ building on a concrete pad with a dike to contain any possible leakage from the tanks, piping or equipment. The overflow or ;
vents from the temporary tanks are routed to the radwaste building I floor drain system to ensure no gas or overflow effluent is released to the atmosphere or environment. All hoses exterior to the building l and diked area will be sleeved, i Since no Updated Safety Analysis Report (USAR) accidents are impacted, l there is no effect on the probability of occurrence or the l consequences of an accident previously evaluated in the USAR.
Since no malfunctions are identified, the probability of occurrence and the consequences of a malfunction of equipment important to safety
^
previously evaluated in the USAR have not been impacted. Also, no malfunctions of a different type could be created.
1 Since no credible accidents that could be created are identified, the
, possibility of an accident of a different type than any previously evaluated in the USAR cannot be created.
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No acceptance limits are identified, so the margin of safety is not affected by this temporary change.
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_w Attachment to ET 97-0017 i- Page 52 of 209 4 Safety Evaluation: 59 96-0034 Revision:O Delete Reporting Requirements for Meteorological Tower Instrumentation Out of Service.
j This change deletes the reporting requirements in chapter 16.3.1.2 of the Updated Safety Analysis Report (USAR) for meteorological instrumentation out of service. This change will eliminate the need to report to the NRC whenever met tower instrumentation is out of service for longer than the specified time. This change will not affect any design bases accidents described in chapters 2, 6, 9 or 15
- of the USAR. This change only deletes the reporting requirement to I
the NRC, which will not affect any credible accidents in the USAR.
This change does not affect any design acceptance limits for the met tower instrumentation, or any equipment important to safety. All
).
equipment is expected to function as designed.
1
, Because this change does not affect any design bases accidents, it i
does not increase the probability of occurrence of those accidents nor does it change the consequences of those accidents discussed in the
- USAR. Also no new accidents of a different type are created.
Since this change does not affect equipment, it does not affect the j probability of occurrence of malfunctions or the radiological
- consequences of equipment important to safety. Also there are no new malfunctions of a different type created.
1 There is no impact on the margin of safety.
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l Attachment to ET 97-0017 Page 53 of 209 Safety Evaluation: 59 96-0035 Revision 0 Revision 47 of the Radiological Emergency Response Plan This revision of Section 2.0, " Emergency Classifications," of d'9 Radiological Emergency Response Plan indicates which operating modes apply at each branch of the emergency action level flow charts. OFN BB-031, " Shutdown Emergency" was added to decision block 3-LRCB1 I since, depending on the plant mode, it could be used for Reactor coolant System (RCS) leak isolation and the wording in decision block 6-LEP/AC1 was modified to say, *
...which lasts or is predicted to last greater than 15 minutes." This was added to allow the flexibility of answering yes to this box without having to wait fifteen minutes, if it vere clear that the event would exceed that time.
These changes do not alter the operation of the plant equipment. The intent is to improve the logic tree used to develop emergency classifications. Therefore, there is 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 ET 97-0017 Page 54 of 209 Safety Evaluation: 59 96-0037 Revision 0 Temporary Use of Fire Protection System For General Electric (GE)
Induction Heater This temporary modification provides cooling water to the GE induction heaters from the Fire Protection System. As part of the Generator disassembly process, GE will be using a water cooled induction heater to heat and remove the generator retaining rings. It was anticipated ;
that a connection to Service Water would be available for cooling flow to the heater, however, during the anticipated duration of work, the Service Water system will not be available.
The tie-in to the Fire Protection system will be made at the 3/4" vent valve KCV335 associated with the standpipe and hose rack KCHR076. The existing pipe cap on the vent valve will be removed and connection made using Chicago fittings and 1" red rubber air hose. The cooling water demand is estimated to be 30 to 40 gpm. The Fire Protection system pressure will be 150 psi with a head loss of approximately 25 psi due to elevation. The theoretical flow available through the 3/4" line at 125 psi is several times the estimated demand and enough that I friction losses in any of the connection hose is considered insignificant. The discharge from the induction heater will be routed to a Turbine building floor drain.
The design basis accidents in U dated p Safety Analysis Report (USAR) chapters 2, 3 and 15 have been reviewed and there is no effect on the accidents or any assumptions used in the evaluation of the subject accidents.
No new hazards are created and the temporary modification will be installed in a non-safety related portion of the power block and will be utilized on non plant equipment.
No new credible accidents are created by this temporary modification.
Since no acceptance limits are identified, the margin of safety is not 1
- affected, j
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Attachment to ET 97-0017 Page 55 of 209 Safety Evaluation: 59 96-0038 Revision 0 Use of Safety Injection Pump for Boration in Mode 6 This modification revises procedures STS BG-001, " Boron Injection Flow Path Verification," and OFN BG-009, " Emergency Boration," and Updated j Safety Analysis Report (USAR) Section 16.1.2 to allow use of a Safety i Injection (SI) pump injecting into the Reactor Coolant System (RCS) as an alternate boration flow path during Mode 6 with the reactor vessel head removed (Refueling Mode), when water level above the top of the reactor vessel flange is greater than or equal to 23 feet (Technical Specification 3.9.10.1). This will allow refueling operation to proceed when both Centrifugal Charging Pumps are INOPERABL2.
This alternate means of boration ensures that negative reactivity control is available during refueling.
This change affects only the description of the USAR and the procedural controls with regard to boration requirements applicable for Mode 6 with the vessel head removed. This change does not involve any hardware changes nor are there any changes in the method by which any other safety related plant system performs its safety function.
There will be no change to normal plant operating parameters or accident mitigation capabilities. As stated in USAR Section 15.4.4.2, an inadvertent boron dilution is precluded by administrative controls during Mode 6, i.e. closing and locking dilution source manual valves.
Therefore, there will be no increase in the probability of any accident or malfunction of equipment important to safety occurring due to these changes.
The boron concentrations of the RCS, the refueling pool, and the refueling cavity during refueling will be maintained in compliance with the limit specified in the CORE OPERATING LIMITS REPORT (COLR).
The boron concentration limit ensures that a core Keff of 0.95 is maintained during fuel handling operations. Given the above, there will be no increase in the consequences of any accident or malfunction of equipment important to safety due to these changes. Also, since no hardware changes are involved the possibility of a different type of unanalyzed event is not created.
The procedural controls and USAR change are sufficient to maintain the boron concentration of the RCS, the refueling pool, and the refueling cavity during refueling above the limit specified in the COLR. There will be no effect on the manner in which safety limits or limiting safety system settings are determined, nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions. Therefore, there will be no impact on margin of safety as defined in the basis for any Technical Specifications.
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Attachment to ET 97-0017 Page 56 of 209 Safety Evaluation: 59 96-0038 Revision 1 Use of Safety Injection Pump for Boration in Mode 6 i
Revision 1 of this Unreviewed Safety Question Determination (USQD) is made to clarify that maintaining 23 feet of water above the vessel flange is not a requirement for using the alternate boration flow path. The reference to 23 feet in Revision 0 was only a description of the condition.
This change was originally described in Unreviewed Safety Question l Determination (USQD) 59 96-0038 Revision 0 which revised procedures l STS BG-001, " Boron Injection Flow Path Verification", and OFN BG-009, l
" Emergency Boration", and Updated Safety Analysis Report (USAR) .
Section 16.1.2 to allow use of a Safety Injection (SI) pump injecting l into the Peactor Coolant System (RCS) as an alternate boration flow l path during Mode 6 with the reactor vessel head removed (Refueling Mode). This allows the refueling operation to proceed when both Centrifugal Charging Pumps are INOPERABLE.
This alternate means of boration ensures that negative reactivity control is available during refueling.
These changes affect only the description of the USAR and the procedural controls with regard to boration requirements applicable for Mode 6 with the vessel head removed. They do not involve any hardware changes nor are there any changes in the method by which any other safety related plant system performs its safety function. There will be no change to nornal plant operating parameters or accident mitigation capabilities. As stated in the USAR Section 15.4.4.2, an inadvertent boron dilution is precluded by administrative controls during Mode 6, i.e. closing and locking dilution source manual valves.
Therefore, there will be no increase in the probability of any accident or malfunction of equipment important to safety occurring due to these changes.
The boron concentrations of the RCS, the refueling pool, and the refueling cavity during refueling will be maintained in compliance with the limit specified in the CORE OPERATING LIMITS REPORT (COLR).
The boron concentration limit ensures that a core Keff of 0.95 is maintained during fuel handling operations. Given the above, there will be no increase in the consequences of any accident or malfunction of equipment important to safety due to these changes. Also, since no hardware changes are involved the possibility of a different type of unanalyzed event is not created.
The procedural controls and USAR change are sufficient to maintain the boron concentration of the RCS, the refueling pool, and the refueling cavity during refueling above the limit specified in the COLR. There
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Attachment _to ET 97-0017 l Page 57'of 209 will be no effect on the manner.in which safety limits or limiting safety system settings are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions. Therefore, there will be no-impact on margin of safety as defined in the basis for any Technical Specifications, i
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j, Safety Evaluation: 59 96-0039 Revision 0 l Temporary Mitrogen Supply to Volume control Tank This temporary modification provides a temporary nitrogen supply to VCT (Volume Control Tank) while the KH (Service Gas) system low l pressure nitrogen is out of service for installation of nitrogen sparging to the condensate storage tank. This evaluation also applies
, when the VCT is supplying charging pump suction with a nitrogen cover blanket during plant modes 5 and 6.
J No new types of accidents are created by this temporary modification.
The use of a temporary nitrogen bottle as described to supply nitrogen will not cause any systems, structures or components important to safety to malfunction, or to malfunction in a way not previously analyzed. No Technical Specifications bases are affected.
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Attachment to ET 97-0017 Page 59 of 209 Safety Evaluation: 59 96-0040 Revision 0 Procedure for Breathing Air System Operations Procedure SYS-KB-200, Rev. O, " Breathing Air System Operations,"
l provides temporary cooling to the aftercoolers by means of Service Water or Fire Protection.
There are no design basis accidents identified or evaluated for the non-safety related KB (Breathing Air), KC (Fire Protection) and EA (Service Water) systems in the Updated Safety Analysis Report (USAR) chapters 2, 3 or 15.
Since the systems' functions are not changed, no credible accidents that could be created are identified.
1 Since the procedure change would not affect the systems' failure
- modes, controle on activity performance, the level of qualification,
, or the effect on equipment important to safety, no credible malfunctions of equipment important to safety are identified.
Since the KB, EA and KC systems are not included in the bases of the Technical Specifications, no acceptance limits are identified that could be affected.
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Safety Evaluation: 59 96-0041 Revision 0 l Installation of a Pressure Gauge In Residual Heat Removal System (RNR) l This temporary modification will install a pressure gauge downstream of valves EJV0056 and EJV0063 (Residual Heat Removal System) in crder j to monitor system pressure downstream of EJHV8840. The added pressure ,
gauge has no direct safety function. It will be installed at the test i connection and will be valved out of service at all times except when taking brief pressure readings.
Adding a pressure gauge to the test connectior. will not increase the probability of occurrence of any previously evaluated accidents since the gauge is rated higher than design pressure of the associated RHR piping and will only be valved in for a short time with a dedicated individual present. i l
The potential radiological consequences of a postulated failure that
. results from a release are bounded by the calculated radiological l consequences of failures outside the containment of small lines l connected to the primary coolant pressure boundary as described in l Updated Safety Analysis Report (USAR) section 15.6.2. There is no
- potential for the creation of a new type of unanalyzed event, i
4 Installation of a pressure gauge that will be valved in for only a 4
short period of time while a dedicated individual is present will not -l increase the probability of occurrence of a malfunction of equipment. i The RHR system is designed to single failure criteria. The consequences of a breach is bounded by the small line failure occurring outside the containment and does not increase the consequences of this malfunction.
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] The only Technical Specification that could be affected by this change j is operational leakage T.S. 3.4.6.2. The basis for this Technical !
Specification is early detection of RCS pressure boundary leakage.
Any leakage caused by the gauge installation would not be a pressure boundary leak and would be observed by the dedicated individual and isolated immediately.
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J Attachment to ET 97-0017 Page 61 of 209 Safety Evaluation: 59 96-0042 Revision 0 1
Drawing Correction cf the Steam Generator Slowdown System This modification revises P&ID drawing M-12BM02, " Steam Generator l Blowdown System" to reflect the as-built condition.
l Engineering reviewed all the revisions of the above P&ID drawing. It was discovered that during Startup, valve BM-VO72 was removed from the field and incorporated in the drawing M-13BM14 Rev 01. However, it was inadvertently not incorporated in M-12BM02. Drawings M-12BM02 and M-13BM14 are now consistent.
This is an administrative change. The drawing has been corrected to restore the originally approved configuration. There is no technical change and there is no change to any design basis.
Since this is only an administrative change and the drawing has been corrected, there is no impact on accidents and malfunctions evaluated I as the licensing basis. There is no potential for the creation of a new type of unanalyzed event, and there is no impact on margin of safety.
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l Attachment to ET 97-0017
' i Page 62 of 209 Safety Evaluation: 59 96-0043 Revision 0 New Fuel Elevator Hard Stop Modification This modification installs a mechanical hardstop to the New Fuel Elevator (NFE). The purpose of this modification is to provide a physical stop to the spent fuel basket if the electrical interlock mechanism and the power disconnects were to fail. This assures a water shielding distance of at least 8 feet when an irradiated fuel assembly is being repaired.
This modification will not increase the probability of occurrence of an accident previously evaluated in the Updated Safety Analysis Report (USAR) because the only accident evaluated which could involve the NFE is a postulated Fuel Handling Accident (FHA).
A supplemental radiological evaluation has concluded that a FHA with a water shielding distance of 8 feet above the assembly remains bounded by the current FRA analysis. Additional conservatism is supplied in this evaluation since the USAR assumes subsequent damage to a spent fuel assembly. As only one assembly can be inserted into the NFE, margin is inherently provided. Therefore, this modification cannot increase the consequences of an accident previously evaluated in the USAR.
This modification will not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR because the modified NFE assembly is expected to remain structurally sound during all plant conditions.
This modification will not increase the consequences of a malfunction of equipment important to safety previously evaluated in the USAR because a hypothetical catastrophic failure of NFE during a fuel repair will not result in off-site doses exceeding those currently calculated in the USAR for a postulated FHA.
There is no potential for the creation of a new type of unanalyzed event because the NFE will continue to perform its intended function following this modification. No failure mode or accident condition has been identified that would result in a malfunction of equipment different than the design basis FHA currently analyzed in the USAR.
This modification does not reduce the margin of safety as defined in the basis for any Technical Specification. This modification does not violate criticality limits for fuel assemblies in the spent fuel pool. Technical Specification section.3/4.9 is not affected by this modification. Minimum water shielding distances above fuel assemblies in the spent fuel racks are not affected.
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Attachment to ET 97-0017 Page 63 of 209 l
Safety Evaluation: 59 96-0045 Revision 0 Spent Fuel Pool Cooling System Design and Analyses This revision to the Updated Safety Analysis Report (USAR) revises sections 9.1.3, 9.1A.4 and 9.1A.5 relating to design and analyses of the spent fuel pool cooling system. In addition, Tables 9.2-9 and 9.2-10 will be updated. These revisions clarify confusing and ambiguous language in the description of the spent fuel pool cooling system and s correct inconsistencies in the USAR text sections and associated tables.
The USAR changes do not involve any hardware changes. There will be l no change to normal plant operating parameters or accident mitigation capabilities. Also, the processes of refueling operations and the methods for removing decay heat produced from the irradiated fuel l stored in the spent fuel pool are not changed. Therefore, there will I be no increase in the probability of any accident occurring due to !
these changes.
l The changes will correct inconsistencies in the USAR. They will not
]
prevent, change, or degrade actions assumed in an accident, and will not have an effect on the radiological consequences of any accident described in the USAR.
i The changes do not involve any design change nor are there any changes in the method by which the safety related plant system performs its safety function. The refueling sequence associated with a full-core offload is not changed. Therefore, there will be no increase in the probability of occurrence of a malfunction of equipment important to q safety due to these changes.
There will be no increase in the consequences of a malfunction of
- equipment important to safety previously evaluated in the USAR, since no hardware changes are involved. Since there are no physical modifications to the facility or changes in methods of operation, there is no potential for the creation of a new type of unanalyzed event.
There will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any
, effect on those plant systems necessary to assure the accomplishment of protection functions. Therefore, the change will have no. impact on the margin of safety as defined in the basis for any Technical Specification.
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Attachment to ET 97-0017 Page 64 of 209 ;
1 Safety Evaluation: 59 96-0046 Revision 0 l
Draining of the Fuel Transfer Canal and Cask Pit '
Procedure SYS EC-401, " Draining the Fuel Transfer Canal from the cask Loading Pit" is being revised to fill the empty Fuel Transfer Canal (FTC) which is isolated from the Spent Fuel Pool (SFP) and Refueling Pool (RFP) by using a portable submersible pump. This pump will be placed in the Cask Loading Pit (CLP), unisolated from the SFP and pump water out of the CLP/SFP and over into the FTC by way of a hose '
concurrent with makeup to the CLP/SFP from the Refueling Water Storage Tank (RWST). This will allow for filling the FTC with CLP/SFP water while making up to the CLP/SFP from the RWST. The waters being ;
transferred in this process are all borated to a concentration at or l above the concentration of the CLP/SFP and RFP.
The Spenc Fuel Pool Cooling and Cleanup (SFPCC) System is described in
- l Updated Safety Analysis Report (USAR) Section 9.1 and shown on Figure 9.1-3 Sheets 1 and 2. Pumping CLP/SFP water by utilizing the j submersible pump and hose in the CLP and over to the FTC is not j described in the USAR nor is this process reflected on USAR Figure 9.1- !
3 Sheet 1. l The Fuel Handling accident in the Fuel Building was identified but not affected by the procedure actions. No other design basis accidents were identified. Therefore, the probability of occurrence and I consequences of all previously evaluated accidents remain as evaluated. l This procedure change does not affect the SFPCC System's failure !
modes, controls on activity performance, the level of qualification, or the effect on equipment important to safety. No credible malfunctions of equipment important to safety are identified. l Therefore, the probability of occurrence and consequences of all previously evaluated malfunctions of important equipment remain as evaluated. Since no malfunctions were identified, malfunctions of a different type than previously evaluated in the USAR could not be created.
Drainage of the CLP/SFP has been evaluated and the procedure changes will not compromise this evaluation. No other credible accidents besides drainage could be created or identified. Therefore, the potential for a unique accident has not been created.
The margin of safety provided by Technical Specification 5.6.2 is maintained and not reduced by the procedure actions.
Attachment to ET 97-0017 Page 65 of 209 Safety Evaluation: 59 96-0047 Revision 0 Use of Fire Protection System as Source of Cooling Water for Steam Generator Cleaning Process This Unreviewed Safety Question Determination (USQD) evaluates the acceptability of the use of the Fire Protection - (FP) system as a source of cooling water for the Steam Generator. (SG) cleaning process. New Procedure OP-2408-2, " Operating Procedure for Chemical Cleaning of Wolf Creek Steam Generators AED" uses the FP system as a supply of cooling water for the SG chemical cleaning process.
The use of the FP system is not a reduction in PP. Also, a contingency plan has been prepared to provide guidance for securing all alternate cooling load demands on the FP system in the event of an actual fire.
The design basis accidents in Updated Safety Analysis Report (USAR) chapters 2, 3 and 15 have been reviewed and this use of the FP system for alternate plant cooling has no effect on the accidents or assumptions used in the evaluation of the subject accidents.
l No new hazards or failure modes are created by this change and associated connections will installed in an area outside of any safety related buildings and used on non-plant equipment. No new credible accidents are created by this change, and the condition is enveloped by the design basis fire evaluation which has been performed and is documented in USAR section 9.5B.
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4 Attachment to ET 97-0017
! Page 66 of 209 i 1 Safety Evaluation: 59 96-0048 Revision 0 Fuel Transfer Cart With One Mechanical Latch j This. modification allows movement of the fuel transfer car with only one functional mechanical latch device. The function of the mechanical latch is to serve as a backup interlock to prevent the movement of the transfer car when the fuel container is not in its fully horizontal position. This evaluation assumes that administrative controls are in place which replace the function of the latch device.
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I This modification replaces the function of the mechanical latch with
! administrative controls. The function of preventing the movement of the transfer car remains redundant and can still withstand a single J
~
failure. Therefore, no design basis accidents described in the Updated Safety Analysis Report (USAR), including the fuel handling l accident, are impacted by this proposed change.
Since all the safety functions associated with movement of the 4
transfer car when the fuel container is not in its fully horizontal position remain redundant and can still withstand a single failure, there are no credible accidents or malfunctions of equipment important
- to safety identified that this proposed change could create.
There are no acceptance limits identified that could be affected by this proposed change.
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4 Safety Evaluation: 59 96-0049 Revision 0 Operating Procedure for Chemical Cleaning of Steam Generators "A" and "D"
This Unreviewed Safety Question Determination (USQD) evaluates Procedure OP-2408-2, " Operating Procedure for Chemical Cleaning of Wolf Creek St0am Generators A & D," Revision 1. This is a new j procedure that details the steps required to chemically clean Steam l Generators A and D by PN Services. Removal of deposits including corrosion products and impurities from the secondary side of the steam j 3 generators by chemical cleaning is being performed to reduce the l potential for secondary side corrosion of tubes and to possibly '
improve thermal hydraulic performance.
The implementation of this procedure will not create any new credible accidents that could affect any system, structure or component. I
- Chemical cleaning of the steam generators will be conducted while the i plant is in shutdown condition and the process will be conducted in accordance with approved procedures.
I j No credible malfunctions of equipment important to safety will be l developed with the implementation of this procedure. The plant will j
, be in a shutdown condition and the appropriate equipment necessary to '
! keep the plant in a safe shutdown condition will be maintained
! J operable as defined in the Technical Specifications. l The use of this procedure will have no effect on any acceptance limits that are contained in the bases for the Technical Specifications or the licensing basis.
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Attachment to ET 97-0017 Page 68 of 209 Safety Evaluation 59 96-0050 Revision O Power Operated Relief Valve Temperature Computer Point This modification will activate an existing spare computer point to monitor the discharge pipe temperature of the PORV's (Power Operated Relief Valves). Presently, the temperature is monitored by temperature indicator TI-463 and is annunciated on the RK (Plant Annunciator ) system.
There are no credible accidents or malfunctions of equipment important to safety which this modification would create. There are no acceptance limits in the Technical Specifications or the licensing basis _ documents which would be affected by this modification.
I Attachment to ET 97-0017 Page 69 of 209 Safety Evaluation: 59 96-0052 Revision 0 Installation of Temporary Fixtures for Monitoring Formation of Frazil Ice on Essential Service Water Pump Suction Bays Outside of Trash Racks This temporary modification allows the placing of a ~30 foot chain on l the grating of each ESW (Essential Service Water) inlet structure bay. The chain will be allowed to hang through the grating and be submerged in the lake water. Because the section of trash racks submerged in the lake cannot be easily inspected from the grating, the chain provides an alternative method for checking if ice is being formed on the trash racks. During the periods when the environmental conditions are conducive to frazil ice formation, the chain shall be placed in-service and periodically pulled out to inspect for any ice formation.
The weight of the fixture with the chain is less than 100 lbs. The grating is capable of withstanding this load. The worst failure mode of this fixture is that the chain may fail and drop in the intake bay. Since the suction velocity in this region is not high and the chain is made of steel, it will fall to the bottom of the intake bay and not be drawn to the pump suction.
There are no types of credible accidents or malfunctions of equipment important to safety that this modification could create.
This change does not reduce any margin of safety as no operating parameters such as temperature, flow or pressure are impacted.
Attachment to ET 97-0017 Page 70 of 209 Safety Evaluation: 59 96-0053 RevisionsO Replacement of Carbon Steel Elbow with Stainless Steel in Essential Service water A Strainer Backwash Piping This modification will replace a carbon steel elbow with a stainless steel elbow in ESW (Essential Service Water) A Strainer backwash piping to restore base metal thickness that has been reduced below minimum wall due to corrosion.
Since there is no adverse impact on the structural integrity of the piping, no new accidents are initiated.
This modification does not have any adverse impact on the performance of the strainer.
There is no impact to any acceptance limits contained in the bases for -
the Technical Specifications, i
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Attachment to ET 97-0017 Page 71 of 209 f Safety Rvaluation: 59 96-0054 Revision 0 Installation of Blind Flange Valve in Accumulator Safety Injection
, system This modification installs a blind flange spool piece in 92-FBD-1" (Accumulator Safety Injection System) line to prevent damage of the j Bourdon tube by isolating this line from high pressure from the steam
- generator.
The Bourdon tube has been deformed several times over the years. The deformation is a result of leaking from steam generator double 1 isolation valves, through valve EPV-0121 and up to the Bourdon tube in
! controller EP PIC-0001. This line was designed for supplying nitrogen I
to steam generators during layup.
y This is a Special Scope line and is not a part of the accident
. analysis. However, piping stress calculations and pipe supports EP08-H504/C502/C501/R510 and R511 are reviewed for the additional loads and 1
stresses. There is adequate margin to accommodate the additional loads and stresses without exceeding the code allowables. Therefore, the probability of occurrence and the consequences of an accident previously evaluated in the Updated Safety Analysis Report have not j been increased. The probability of occurrence of a malfunction of
- equipment or the radiological consequences of a malfunction of
- equipment important to safety are not increased. The possibility of an unanalyzed event is not created. There is no reduction in the
- margin of safety as defined in the basis for any Technical
, Specifications.
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3 Attachment to ET 97-0017 Page 72 of 209 Safety Evaluation: 59 96-0056 Revision 0 i
Cycle 9 Reload Design I The safety evaluation presented in the Reload Safety Evaluation (RSE) and the Cycle 9 Core Operating Limits Report (COLR) demonstrate that the probability of an accident previously evaluated in the Updated Safety Analysis Report (USAR) is not increased. Operation of Wolf Creek Generating Station (WCGS) in Cycle 9 with the introduction of Regions 11A and 11B reload fuel has been analyzed in accordance with l methodologies reviewed and approved by the NRC. Demonstrated l adherence to applicable standards and acceptance criteria precludes any new challenges to components and systems that could increase the i
probability of any previously evaluated malfunction of equipment '
important to safety. The Cycle 9 Reload Design does not violate any safety limits and all cycle 9 design criteria are met. Therefore, the i probability of occurrence of a malfunction of equipment important to l safety previously evaluated in the USAR has not increased. l The safety evaluation presented in the RSE and the Cycle 9 COLR l demonstrate that the consequences of an accident previously evaluated '
in the USAR are not increased. The demonstrated adherence to I applicable standards and design criteria precludes new challenges to '
components and systems that could a) adversely affect the ability of existing components and systems to mitigate the consequences of any accident, and/or b) adversely affect the integrity of the fuel rod cladding as a fission product barrier. 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 accomodated within the conservatism of the existing analyses.
Therefore, the consequences of an accident and malfunction of equipment important to safety previously evaluated in the USAR are not increased.
The safety evu2uation demonstrates that the Cycle 9 Reload Design does not create the possibility of malfunction of equipment important to safety than any other previously evaluated in the USAR. All design and performance criteria continue to be met and no new failure modes have been introduced. The implementation of the design and adoption of the COLR 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 evaluation demonstrates that the Cycle 9 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
Attachment to ET 97-0017 Page 73 of 209 components and systems that could introduce a new type of accident not previously described in the USAR. All design and performance criteria continue to be met and no new failure modes have been introduced.
Operation of WCGS in Cycle 9 has no effect on either the LOCA or non-LOCA accidents considered in the USAR. The Cycle 9 design does not create a condition outside of the 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.
The safety evaluation demonstrates that the Cycle 9 Reload Design does not reduce the margin of safety as defined in the Bases to any Technical Specification. The Cycle 9 Reload Design and COLR establish that all design and safety analysis limits continue to be met and that these limits are supported by the applicable Technical Specifications. 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 USAR Chapter 6 or Chapter 15 accidents. Therefore, the margin of safety as defined in the Bases of the Technical Specification has not been reduced.
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Attachment to ET 97-0017 Page 74 of 209 Safety Evaluation: 59 96-0057 Revision 3 0 Fill and Vent the Circulating Water System with Service Water Secured This temporary procedure provides instructions to fill and vent the Circulating Water System (CW) with Service Water secured. The procedure uses a temporary pump to supply water to the CW system through the condenser tube cleaning connection isolation valve for filling the CW system.
Since the Circulating Water System is not in service, this procedure would not affect the performance of activities that are important to safe and reliable operation.
Since the CW system is not in service and the system's functions are not changed, no credible accidents that could be created are identified.
There are no design basis accidents identified or evaluated for the non-safety related CW system in Updated Safety Analysis Report chapters 2, 3 or 15.
No credible malfunctions of equipment important to safety are i identified and no acceptance limits are affected.
Attachment to ET 97-0017 Page 75 of 209 Safety Evaluation: 59 96-0058 Revision:O Low Pressurizer Pressure Trip Setpoint Increase This modification revises the low pressurizer pressure reactor trip setpoint from 1915 psig to 1940 psig. The increase in the low pressurizer pressure reactor trip setpoint ensures core thermal limit protection will be maintained with the assumed 3.5% reduction in the thermal design flowrate.
Current Technical Specifications state that the low pressurizer pressure reactor trip setpoint must be greater than or equal to 1915 psig. Since the trip setpoint of 1940 psig is greater than the Technical Specification limit, the revision of the low pressurizer pressure reactor trip setpoint may be implemented under the current licensing basis.
The low pressurizer pressure reactor trip setpoint assumed in the Updated Safety Analysis Report (USAR) Chapter 15 accidents is assumed at a conservatively low value. Therefore, raising the low pressurizer pressure reactor trip setpoint does not increase the probability of occurrence or radiological consequences of an accident previously evaluated in the USAR.
This change does not impact the single failure assumptions assumed in the USAR Chapter 15 accident analyses. The low pressurizer pressure reactor trip bistables are designed to be recalibrated on a periodic basis, therefore recalibration of the bistables will not increase the probability of occurrence or radiological consequences of a malfunction of equipment important to safety previously evaluated in the USAR.
Raising the low pressurizer pressure reactor trip setpoint does not result in changing or modifying any plant equipment, only recalibration of the trip bistables. Therefore, this change will not create the possibility of a different type of malfunction than any previously evaluated in the USAR. Since raising the setpoint assumed in the analyses can only result in less severe analysis results, this change will not create the possibility of an accident of a different type than any previously evaluated in the USAR.
The current safety analyses bound the proposed value for the low pressurizer pressure reactor trip setpoint. This, in combination with the fact that current Technical Specifications allow for increasing the setpoint above the Technical Specification limit, demonstrates that the margin cf safety is not reduced by the change.
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Attachment to ET 97-0017 Page 76 of 209 Safety Evaluation: 59 96-0059 Revision:0 Post Maintenance Essential Service Water Cross Tie Isolation Valve Leak Test The purpose of this temporary procedure is to allow post-maintenance testing of the Essential Service Water (ESW) Cross tie isolation valves by quantifying the leak rate through these valves. The two cross tie isolation valves will be closed. A we.ter source from the Fire Protection system at approximately 100 psi will be fed.into the piping drain line located between the two isolation valves filling this piping. The combined leakage of the two isolation valves will then be established by measuring flow through the fire protection ;
connection using a calibrated flow meter and normalizing the flow rates based on the measured pressure values upstream and downstream.
The plant is in Mode 6 or Mode E (fuel removed from the reactor
- vessel) during the performance of this test, and no new failure modes are created by this procedure. The accidents of Updated Safety Analysis Report (USAR) chapters 2, 3 and 15 have been reviewed and are j
not affected by this test. 'Use of the Fire Protection system as'a ,
source of water for this test is not discussed in the'USAR. There are' ;
I no Technical Speci.fications applicable to the operability of the ESW system in plant' Mode 6 and E.
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1 Attachment to ET 97-0017 .
Page 77 of 209 1
i Safety Evaluation: 59 96-0060 Revision 0 1
. Elimination Of One Overload and Underload on the Refueling Machine This change will eliminate one of the automatic overloads on the
,~
refueling machine to allow overcoming any frictional forces between the fuel assembly bottom nozzle and lower core alignment pins during the raising of a fuel assembly. This change will also eliminate the automatic underload to overcome frictional forces between the fuel 4
assembly bottom nozzle and lower core alignment pins during lowering of a fuel assembly. The automatic bypass of the one overload and the underload at the upender and the RCCA (Rod Cluster Control Assembly) change fixture, to produce a slack cable indication to allow latching and unlatching of a fuel assembly, will also be allowed by this change.
, An automatic overload is still active at a value less than or equal to 250 pounds above the indicated weight of the suspended fuel assembly
- during lifting. The automatic underload is not needed when the fuel assembly is less than or equal to 2 inches above full down in the core, upender, or RCCA change fixture because there is no longer any
- possibility of snagging adjacent grid straps at this level.
Therefore, this change does not increase the probability of occurrence of an aceitant or an accident of a different type previously evaluated in the Updated Safety Analysis Report (USAR).
The most severe accident would be dropping a spent fuel assembly as
, analyzed in the USAR Section 15.7.4. This change does not increase j the consequences of an accident or malfunction of equipment important ,
l to safety because it does not change the load bearing components of 1 the refueling machine. Therefore, the consequences of an accident J
described in section 15.7.4 remain the same, and the possibility of a
- different type of malfunction of equipment important to safety is not
- created.
)
Elimination of one automatic overload and the underload when the l bottom nozzle of the fuel assembly is less than or equal to 2 inches I above the full down position in the core, upender, and RCCA change l
'1 fixture has no effect on the equipment used. Therefore, this change cannot affect the probability of occurrence of a malfunction of 4
equipment important to safety.
The Technical Specifications do not address the automatic overloads and underloads for the refueling machine, therefore the margin of safety cannot be reduced.
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Attachment to ET 97-0017 Page 78 of 209 Safety Evaluation: 59 96-0061 Revision 0 Scaffold Storage Area This modification revises the Updated Safety Analysis Report (USAR)
Figure 1.2-13 to delete the protective clothing storage area from Room 1504 on the Aux Bldg 2047' level. A portion of this area will instead be used for scaffold material storage. A II/I hazards evaluation performed for this area indicates that no safety related components exist in the area that would be affected by seismic failure of the scaffold storage enclosure.
Also, a fire hazards analysis update is performed for Fire Area A-19 as additional fixed combustible materials will be stored there. The effects of a design basis fire are analyzed where it is assumed that all equipment in the area is destroyed by fire. Even with this condition, USAR Section A.19.7.2 concludes that a fire in this area will not prevent safe shutdown of the plant. Although the additional fixed combustibles represent an increase in fire loading, and USAR Appendix 9.5B will be revised to reflect this, the fire load in Fire Area A-19 will be maintained below the fire barrier ratings for this area. Therefore, this change does not impact assumptions used in the design basis fire and no reduction in fire protection of Fire Area A-19 will result.
It is concluded that use of this area for scaffold material storage will not introduce any seismic II/I hazards or fire hazards in the area. Accordingly, use of this area for scaffold material storage will not impair the function of any safety related plant equipment / components or systems, and will not adversely affect the j ability to achieve and maintain safe shutdown in the event of a fire. '
II There are no design basis accidents identified because this change has I no effect on the inputs, assumptions or components involved in the l accidents evaluated in the USAR. l 1
Since no seismic II/I hazards are introduced by this change and since the area's fire barrier ratings are not being challenged, no credible accidents that could be created are identified.
Since no new seismic or fire hazards, which could impair the function of safety related equipment, are created by the change, no credible malfunctions of equipment important to safety are identified.
Since the change does not affect any safety related plant equipment / components or systems in any.way, no acceptance limits are identified that could be affected.
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l Safety Evaluation: 59 96-0062 Revision 0 l
Radiological Baergency Response Plan Revision 49 This revision to the Radiological Emergency Response Plan (RERP) ;
changes wording of the Administrative Emergency Action Levels (EAL) to 1 correct an error and give clear direction on which path to use when i determining proper actions to take. This change allows high offsite I doses to result in the declaration of a General Emergency.
Previously, it was possible to have high offsite doses at an emergency classification of an Alert.
i These changes do not impact design basis accidents or create any new credible accidents. This change increases flexibility within the EALs l to address additional equipment failures which could cause less I protection than is required to protect the public. l 1
Malfunctions of equipment important to safety are not created by these changes. )
Acceptance limits of Technical Specifications are not affected.
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Attachment to ET 97-0017 i Page 80 of 209 1
i Safety Evaluation: 59 96-0064 Revision 0 Transient Combustible Update l This modification revises the Updated Safety Analysis Report (USAR) in l the following ways:
, combustibles in terms of general combustible material classifications instead of specific materials.
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- 2) Transfers approved storage area combustible materials now l
, identified as a transient combustible into the fixed combustible load l calculations. This change adds to the fixed combustible load )
calculations the maximum amount of material which was evaluated and l approved for the areas.
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- 3) Adds an additional statement to the USAR FRA providing guidance on !
how the transient combustibles are controlled via an administrative )
procedure. j l
- 4) Adds a statement to FHA that some class A combustibles located in closed metal containers may not be considered as contributing to the fire load in the fire hazard area. This addresses the numerous gang boxes and metal lockers located in plant corridors which have minor amounts of class A type combustibles and are used by Operations, Plant Support, and Health Physics during the conduct of day to day plant activities.
The intent of these changes is to provide a consistent approach in I identifying combustible materials in the USAR and the procedure AP 10-102, " Control of Transient Combustible Materials." This change will also provide a better separation between transient combustible materials taken into the plant versus those which have been evaluated by Engineering in the FHA for permanent location in the plant.
These changes do not add or approve any additional combustible materials beyond those already previously evaluated. These changes do not approve any different types of materials for use in the plant than had already been previously evaluated.
These changes do not affect the design basis accidents discussed in USAR Chapters 2, 3, 6, 9 and 15. There are no new failure modes initiated as the amounts, types and locations of the materials have not been changed.
The changes are bounded by the design basis fire for all affected fire areas where it has already been assumed that all equipment in the fire area not protected is consumed by fire.
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Attachment to ET 97-0017 Page 81 of 209 There are no acceptance limits relative to the above changes. Fire Protection has been removed from Technical Specifications with the exception of the QA audit requirements. Therefore, the margin of safety is not reduced.
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1 Attachment to ET 97-0017 Page 82 of 209 Safety Evaluation: 59 96-0065 Revision 0 Radiological Emergency Response Plan Revision 48 This change revises Radiological Emergency Response Plan (RERP) to update EAL (Emergency Action Levels) flow charts to cover different possible equipment troubles which could cause less protection than is required to protect the public. This change was made due to the ESW )
(Essential Service Water) icing event.
This change does not impact design basis accidents nor does it create ,
any new credible accidents. '
Credible malfunctions of equipment important to safety are not i increased.
This change does not affect plant equipment or operational setpoints, '
but rather it better defines conditions for declaration of emergency conditions. Therefore, the margin of safety is not reduced.
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l Safety Evaluation: 59 96-0066 Revision:0 I Essential Service Water System Repair on Discharge and Warming l Crosstie Lines j The excavation efforts described in this modification are necessary l for the purpose of facilitating the modification of the underground portion of the Essential Service Water System (ESWS) piping so that l the ESWS discharge system pressure can be upgraded from 10 psig to 75 psig. This modification consists of reinforcing by welding four (4) existing 30" dia, blanking reinforcement plates on the Unit 1/2 ESWS Discharge and Warming crosstie lines. Construction shoring will prevent any adverse effects from construction of the pits on the piping of either train, and also on the existing adjacent structures.
Upon completion of these excavation activities, the ESWS, Security and other non-safety related systems will be restored to their original configuration.
This excavation has been evaluated with respect to hazard analyses.
ESWS lines are moderate energy lines, therefore, a high energy pipe break hazard does not exist. No new flooding source is created by this excavation.
Safe shutdown capabilities of the plant will not be impacted as administrative controls are in place and separation is maintained between the trains.
This excavation does not affect the ability of any safety related system, component or structure to perform its safety related function.
The construction of the excavation pits will not increase the probability of occurrence of aa accident previously evaluated in the Updated Safety Analysis Report (USAR) because the operating and accident conditions that jeopardize the integrity of the piping have been evaluated and taken into account by the design of the excavation pits and the administrative controls imposed.
The safety design bases relating to structural integrity, function and operability of the ESWS are unaffected by the excavation. Thus, the excavation will not increase the consequences of any accident previously evaluated in the USAR.
The excavation has been evaluated with respect to missile, flooding, seismic and heavy loads hazards. The integrity of piping and components will not be affected. Thus, the excavation will not increase the probability of occurrence.or consequences of a malfunction of equipment important to safsty previously evaluated.
The excavation has been evaluated for all relevant hazards to the
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Thus, the excavation will not create the possibility of a different l type of accident or malfunction of equipment important to safety than
- any previously evaluated in the USAR.
The bases of any Technical Specification are unaffected by the excavation. Therefore, there is no reduction in the margin of safety.
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Safety'Ryaluation: 59 96-0066 Revision 1 I i
Essential Service Water System Repair on Discharge and. Warming Crosatie Lines j Revision 1 of this Unreviewed Safety Question Determination (USQD) 59 96-0066 makes the following changes to Revision 0 for clarification: l
- 1) adds the words "with strong backs" to the statement, "The )
modification consists of reinforcing the four (4) existing 30" dia. I blanking plates on the Unit 1/2 Essential Service Water System (ESWS) )
Dischamge and Warming crosstie linas" in section 1, page 1. j
- 2) replaces the words "their original configuration, function,
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capability availability and reliability" with the words "an i engineering approved equivalent configuration, function, capability, '
l availability and reliability" in section 1, page 1.
- 3) adds a period after the sentence ending with " worked" in section 1, page 2.
- 4) spells out 4'-6" to read four (4) feet six (6) inches in section 1, page 3. l
- 5) replaces the word "will" to "may" in section 1, page 4.
Revision 0 is described as follows:
The excavation efforts described in this modification are necessary for the purpose of facilitating the modification of the underground portion of the ESWS piping so that the ESWS discharge system pressure can be upgraded from 10 psig to 75 psig. This modification consists of reinforcing by welding four (4) existing 30" dia, blanking reinforcement plates on the Un it 1/2 ESWS Discharge and Warming crosstie lines. Construction shoring will prevent any adverse effects from construction of the pits on the piping of either train, and also on the existing adjacent structures. Upon completion of these ;
excavation activities, the ESWS, Security and other non-safety related I systems will be restored to their original configuration i This excavation has been evaluated with respect to hazard analyses. l ESWS lines are moderate energy lines, therefore, a high energy pipe j break hazard does not exist. No new flooding source is created by I this excavation.
Safe shutdown capabilities of the plant will not be impacted as administrative controls are in place and separation is maintained between the trains.
This excavation does not affect the ability of any safety related system, component or structure to perform its safety related function. l The construction of the excavation pits will not increase the 1
1 Attachment to ET 97-0017 Page 86 of 209 probability of occurrence of an accident previously evaluated in the
~
Updated Safety Analysis Report (USAR) because the operating and
[ accident conditions that jeopardize the integrity of the piping have been evaluated and taken into account by the design of the excavation pits and the administrative controls imposed.
l
- The safety design bases relating to structural integrity, function and i operability of the ESWS are unaffected by the excavation. Thus, the
- excavation will not increase the consequences of any accident i previously evaluated in the USAR.
j The excavation has been evaluated with respect to missile, flooding,
- seismic and heavy loads hazards. The integrity of piping and components will not be affected. Thus, the excavation will not increase the prosability of occurrence or consequences of a malfunction of equipment important to safety previously evaluated.
The excavation has been evaluated for all relevant hazards to the affected ESWS piping including flooding, fire and II/I concerns.
Thus, the excavation will not create the possibility of a different type of accident or malfunction of equipment important to safety than any previously evaluated in the USAR.
The bases of any Technical Specification are unaffected by the excavation. Therefore, there is no reduction in the margin of safety.
O O
1 Attachment to ET 97-0017 Page 87 of 209 i
Safety Evaluation: 59 96-0066 Revision 2 Essential Service Water System Repair on Discharge and Warming Crosstie Lines Revision 2 of this Unreviewed Safety Question Determination (USQD) 59 96-0066 makes a change to Revision 1 in section 1, page 3. This change removes the provision for using a temporary cover at the end of each workday over the excavation site and establishes a communication 4 link with the Control Room so that excavation or welding activities I can be immediately halted in the event of a National Weather Service tornado warning posting. Administrative controls will be put in place to restore the soil cover to the 4'-6" level or greater, in the event of a missile threat, such as a tornado warning. This shall be accomplished by 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> coverage of loaded dump trucks and earth moving equipment stationed at the work site. Also, in the event the train opposite to the one on which the excavation / welding is being i performed becomes inoperable for an unplanned failure which could last for more than a shift, the train on which excavation / welding is being performed shall have at least 4'-6" of sand dumped on it.
Revision 1 of this Unreviewed Safety Question Determination (USQD) 59 l 96-0066 makes the following changes to Revision 0 for clarification: l
- 1) adds the words "with strong backs" to the statement, "The modification consists of reinforcing the four (4). existing 30" dia. ;
blanking plates on the Unit 1/2 ESWS Discharge and Warming crosstie I lines" in section 1, page 1.
- 2) replaces the words "their original configuration, function, capability availability and reliability" with the words "an engineering approved equivalent configuration, function, capability, availability and reliability" in section 1, page 1.
- 3) adds a period after the sentence ending with " worked" in section 1, page 2
- 4) spells out 4'-6" to read four (4) feet six (6) inches in section 1, page 3
- 5) replaces the word "will" to "may" in section 1, page 4.
Revision 0 is described as follows:
The excavation efforts described in this modification are necessary for the purpose of facilitating the modification of the underground portion of the Essential Service Water System (ESWS) piping so that the ESWS discharge system pressure can be upgrade 6 from 10 psig to 75 psig. This modification consists of reinforcing by welding four (4) existing 30" dia. blanking reinforcement plates on the Unit 1/2 ESWS Discharge and Warming crosstie lines. Construction shoring will prevent any adverse effects from construction of the pits on the piping of either train, and also on the existing adjacent structures.
_- - , ~ . . -- . . ~ . . - _- - . . . - - - . ~ . . - . - . . ..
I 1
+
, Attachment to ET 97-0017 Page 88 of 209 i
i Upon completion of these excavation activities, the ESWS, Security and
, other non-safety related systems will be restored to their original 3
configuration.
This excavation has been evaluated with respect to hazard analyses. )
4
- )
ESWS lines are moderate energy lines, therefore, a high energy pipe !
break hazard does not exist. No new flooding source is created by
, this excavation.
4 6
, Safe shutdown capabilities of the plant will not be impacted as l
administrative controls are in place and separation is maintained '
between the trains.
I
( This excavation doe 9 not affect the ability of any safety related
] system, component or structure to perform its safety related function.
The construction of the excavation pits will not increase the ,
probability of occurrence an an accident previously evaluated in the
- Updated Safety Analysis Report (USAR) because the operating and accident conditions that jeopardize the integrity of the piping have l been evaluated and taken into account by the design of the excavation pits and the administrative controls imposed, i 1
The safety design bases relating to structural integrity, function and
- operability of the ESWS are unaffected by the excavation. Thus, the i excavation will not increase the consequences of any accident previously evaluated in the USAR.
i-j The excavation has been evaluated with respect to missile, flooding,
, seismic and heavy loads hazards. The integrity of piping and components will not be affected. Thus, the excavation will not increase the probability of occurrence or consequences of a l malfunction of equipment important to safety previously evaluated. '
1 i The excavation has been evaluated for all relevant hazards to the affected ESWS piping including flooding, fire and II/I concerns.
~
Thus, the excavation will not create the possibility of a different I
type of accident or malfunction of equipment important to safety than any previously evaluated in the USAR.
The bases of any Technical Specification are unaffected by the j excavation. Therefore, there is no reduction in the margin of safety.
J
- e w i
Attachment to ET 97-0017 Page 89 of 209 Safety Evaluation: 59 96-0066 Revision: 3 Essential Service Water System Repair on Discharge and Warming Crosstie Lines Revision 3 of this Unreviewed Safety Question Determination (USQD) 59 96-0066 makes editorial changes to Revision 2 in order to clean up the language without making technical changes or changing the intent.
With this Revision 3 of the USQD the entire modification is issued for the first time.
Revision 2 of this Unreviewed Safety Question Determination (USQDl 59 96-0066 removes the provision for using a temporary cover at the end of each workday over the excavation site and establishes a communication link with the Control Room so that excavation or welding ,
activities can be immediately halted in the event of a National Weather Service tornado warning posting. Administrative controls will be put in place to restore the soil cover to the 4'-6" level or greater, in the event of a missile threat, such as a tornado warning.
This shall be accomplished by 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> coverage of loaded dump trucks and earth moving equipment stationed at the work site. Also, in the event the train opposite to the one on which the excavation / welding is being performed becomes inoperable for an unplanned failure which could last for more than a shift, the train on which excavation / welding is being performed shall have at least 4'-6" of sand dumped on it.
Revision 1 of this USQD makes the following changes to Revision 0 for clarification:
- 1) adds the words "with strong backs" to the statement, "The modification consists of reinforcing the four (4) existing 30" dia.
blanking plates on the Unit 1/2 ESW Discharge and Warming crosstie lines" in section 1, page 1.
- 2) replaces the words "their original configuration, function, capability availability and reliability" with the words "an engineering approved equivalent configuration, function, capability, availability and reliability" in section 1, page 1.
- 3) adds a period after the sentence ending with " worked" in section 1, page 2.
- 4) spells out 4'-6" to read four (4) feet six (6) inches in section 1, page 3.
- 5) replaces the word "will" to "may" in section 1, page 4.
Revision 0 is described as follows:
The excavation efforts described in this modification are necessary for the purpose of facilitating the modification of the underground portion of the Essential Service Water System (ESWS) piping so that
Attachment to ET 97-0017 Page 90 of 209 the ESWS discharge system pressure can be upgraded from 10 psig to 75 psig. This modification consists of reinforcing by welding four (4) existing 30" dia blanking reinforcement plates on the Unit 1/2 ESWS Discharge and Warming crosstie lines. Construction shoring will prevent any adverse effects from construction of the pits on the piping of either train, and also on the existing adjacent structures.
Upon completion of these excavation activities, the ESWS, Security and other non-safety related systems will be restored to their original ,
configuration. '
This excavation has been evaluated with respect to hazard analyses.
ESWS lines are moderate energy lines, therefore, a high energy pipe ;
break hazard does not exist. No new flooding source is created by I this excavation. l l
Safe shutdown capabilities of the plant will not be impacted as l administrative controls are in place and separation is maintained I between the trains. I l
This excavation does not affect the ability of any safety related system, component or structure to perform its safety related function. l
, 1 The construction of the excavation pits will not increase the I probability of occurrence of an accident previously evaluated in the l Updated Safety Analysis Report (USAR) because the operating and l accident conditions that jeopardize the integrity of the piping have been evaluated and taken into account by the design of the excavation pits and the administrative controls imposed.
i The safety design bases relating to structural integrity, function and operability of the ESWS are unaffected by the excavation. Thus, the excavation will not increase the consequences of any accident previously evaluated in the USAR.
l The excavation has been evaluated with respect to missile, flooding, seismic and heavy loads hazards. The integrity of piping and components will not be affected. Thus, the excavation will not increase the probability of occurrence or consequences of a malfunction of equipment important to safety previously evaluated.
The excavation has been evaluated for all relevant hazards to the affected ESWS piping including flooding, fire and II/I concerns.
Thus, the excavation will not create the possibility of a different type of accident or malfunction of equipment important to safety than any previously evaluated in the USAR.
The bases of any Technical Specificatien are unaffected by the excavation. Therefore, there is no reduction in the margin of safety.
l l
! Attachment to ET 97-0017 l l
Page 91 of 209 l l
Safety Evaluation: 59 96-0067 Revision: 1 1 1
! Revision to Main Turbine Overspeed Protection Valve Testing Frequency Revision 1 of this Unreviewed Safety Question Determination (USQD) 59 96-0067 will revise Surveillance Requirement 16.3.2.1.1b, items a, b and c, of the Updated Safety Analysis Report (USAR) Section 16.3.2 (3/4.3.4), " TURBINE OVERSPEED PROTECTION." This change will change the frequency of testing the four high pressure turbine stop vsives, six low pressure turbine reheat stop valves, and six low pressure turbine reheat intercept valves from once per 7 days to once per 92 days in 16.3.2.1b, item a. In addition, the current 31 day test
- requirement for the four high pressure main turbine governor valves in 16.3.2.1.lb, item b, will be moved to item a (which will change their surveillance from 31 days to 92 days), the requirement for direct observation of the movement of the valves will be moved from item c to item a (which will change this surveillance from 31 days to 92 days) and items b and c will be deleted. Items d and e will be " renumbered" to b and c. Revision 0 of this USQD is superceded by Revision 1.
There are no design basis accidents that this change would affect.
Turbine overspeed is discussed in USAR Sections 10.2.3 and 3.5.1.3.
The only change is to reduce the frequency of the current surveillance. The surveillance procedure, the equipment or components l being tested, and the methods used for testing are not being changed, ;
nor is system design or operation being changed. Therefore, this l change will not increase the probability of occurrence of an accident l previously evaluated.
l The consequences of a turbine-generated missile, as described in the USAR, would not be affected by the change. The only change is to reduce the frequency of the current surveillance. Therefore, this change will not increase the radiological consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
Changing the weekly test to quarterly for these valves would result in an increase in the probability of missile generation from turbine overspeed. However, based on review of calculations performed at other plants with GE turbines, which the NRC has approved this change for, this increase in probability of missile generation would not exceed the current NRC acceptance criteria of 1 X 10(-5) events per year. Thus, this change would not increase the probability of occurrence of a malfunction of equipment important to safety or affect the acceptance limits of a postulated turbine-generated missile.
Implementing this change would not create the possibility of a new type of unanalyzed event since the only change is to reduce the frequency of the current surveillance.
_ _ _ _ _. , _ . . . _ _ _ . . - _ _ _ _ _ _ . _. --.m _.____ _ _ ._m.._ _ _ . - _ . m_____._ _
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Attachment to ET 97-0017 1 4
Page 92 of 209-
)
A I l
1 Safety Evaluation: 59 96-0068 Revision 0 s supplier Evalution Process
\
'This revision to the Supplier Quality Supplier Evaluation procedure '
evaluates incoming information "real-time" as defined in AP 24C-002.
This change will actually allow evaluation of suppliers on a more frequent basis, but at least on an annual basis. This involves j removing the reference co the annual evaluation in the Updated Safety j Analysis Report (USAR) Chapter 17.2.18.6. 1 This change only applies to the administrative control of a supplier by supplier Quality and therefore, will not impact accidents and malfunctions evaluated as.the licensing basis.
There is no potential for the creation of a new type of unanalyzed I event.
There is no impact on margin of safety.
1 1
I
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,- n- ~ . - . - . . . . ... -. ~ . . -- -~ - . ~ - _ , ~ - . . - . - - - . - _ _ . , . -
l
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. Attachment to ET 97-0017 '
Page.93 of 209 l
l- . Safety Evaluations .59 96-0069 Revision: 0 -
l-l Minor Thermal Restriction at steam Generator Upper Lateral Restraints-j This modification evaluates the Reactor Coolant Loop (RCL) system for l- a minor horizontal thermal restric*. ion at the Steam Generator (SG) l: upper lateral restraints (ULS) for all 4 loops. A visual inspection documented scratching, galling and gouging of some contact surfaces at ,
the SG-ULS on loops B, C and D. Shims on Loop D were re-worked to !
- . provide more clearance. Review of heat up thermal monitoring was lJ
- planned to determine if further work was indicated on the other loops. ;
k 4 It has been demonstrated that with a 1/4" thermal restriction at the.
1: steam generator upper lateral support location, the loads and stresses .
- in the loop piping, primary equipment supports.and main steam and
- . feedwater piping remain within the applicable allowable limits. The f j application of leak-before-break (LBB) technology as a basis for the
- - elimination of the dynamic effects resulting from postulated ruptures f of the RCL piping is shown to remain valid. The functions of.the 7 j reactor coolant system and attached auxiliary systems are maintained.
l Therefore this evaluation will not increase the probability of j occurrence of an accident previously evaluated in the Updated Safety
] Analysis Report (USAR).
l ,
4 The functions of the affected systems, components and supports have l been shown to be maintained. The restrained thermal expansion does j not affect the engineered safety feature systems response from that
!' previously evaluated. Operations of the plant with the upper steam generator support thermal restriction will not affect any assumptions previously made in evaluating the radiological consequences of an j accident or malfunction of equipment described in the USAR.
1 l The evaluation of the affected systems including the thermal restraint 8
at the steam generator ULS demonstrates that the loads and stresses f
are maintained within the appropriate allowable limits for the RCS j equipment and supports. The operation of safety related equipment is
- . unaffected, therefore the probability of occurrence of a malfunction 1 of equipment important to safety previously evaluated in the USAR is
- not increased.
i
- j. The loads and stresses within the primary equipment components and j supports, and main steam and feedwater piping, are maintained within
- appropriate limits. The normal and accident condition functions of ,
, the affected systems are not affected by the ULS thermal restriction.
j The application of LBB and the elimination of arbitrary intermediate j breaks (AIB) have considered all of the design loadings and have been j shown to remain valid. Thus, the thermal restriction will not create the possibility of a new accident scenario.
i i
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, . . . . . . . . _ . . . .. _ - - . ~ . . . _ ~ .
Attachment to ET 97-0017 f Page 94 of 209 i
j It has been demonstrated that the function of the affected systems, and the integrity of the piping, supports, and equipment are
,- maintained for operation with the thermal restriction at the SG ULS location. Such. operation would not require the function of any safety ,
related equipment beyond that which is currently assumed. Therefore, l no new failure modes are created. i The margin of safety of the RCS piping, components and supports, and ,
main steam and feedwater piping, is defined by the structural criteria presented in Section III of the ASME Code. Considering the thermal l restriction at the SG ULS location, it has been demonstrated that the i appropriate criteria are met. The function of safety related systems is maintained.
I I
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Attachment to ET 97-0017 Page 95 of 209 I
- Safety Evaluation
- 59 96-0071 Revision 0 '
i Essential Service Water Warming Line Flow Increase j This modification considers the effects of adding two new orifices EF- I FO-07 and EF-FO-08 at the Essential Service Water (ESW) discharge f
outlet and increasing the ESW return piping design pressure to 75 i psig. These orifices were sized such that a part of the pressure drop l across the existing orifices EF-FE-03/04 was transferred to new orifices EF-FO-07/08. The overall pressure drop in the ESW discharge header is not changed, with warming lines isolated. The resulting increased pressure in the ESW discharge piping between EF-FE-03/04 and
, EF-FO-07/08 ensured that the warming lines would be at a positive pressure relative to atmospheric pressure with a warming flow rate of 5000 gpm and a Ultimate Heat Sink (UHS) water level of 1068' (1060' is
- the minimum pump submergence requirement).
l The effect of these changes was to enhance the existing design to allow the required warming flow rate to be achieved. The ESW flow balance to critical components in the power block was unaffected. The l only Updated Safety Analysis Report (USAR) changes were the P&ID
- change showing the addition of EF-FO-07/08 (Figure 9.2-2-03) and a
- change to the stated ESW return piping design pressure from 50 to 75 psig (Section 9.2.1.2.2.2).
No design basis accidents are impacted by this modification, therefore the probability of occurrence of an accident previously evaluated is )
not affected by the change.
4 The change enhances the availability of ESW pumps in cold weather. ,
The ability to safely shutdown the plant is improved. Thus, there is I no impact on the consequences of the accidents evaluated in the USAR.
Since the orifices are passive components and designed, fabricated and installed to the same quality level as the ESW system, they cannot increase the malfunction of any equipment identified in the ESW system single failure analysis described in USAR Table 9.2-6. Since the malfunction of equipment is not impacted, the radiological consequences are not increased.
The commercial, industrial or military events, hazards involving missiles, pipe whip and natural phenomena, and condition I through IV faults are not impacted by these orifices. Therefore, no new type of accident is created. No new types of single failures have been created, therefore, no new types of malfunctions are possible. The operability of_the ESW system, its capacity to cool as defined in Technical Specification 3/4.7.4 are not adversely impacted by this change. No margins of safety are affected.
Attachment to ET 97-0017 Page 96 of 209 Safety Evaluation: 59 96-0071 Revision 1 Essential Service Water Warming Line Flow Increase Revision 1 of this Unreviewed Safety Question Determination (USQD) 59 96-0071 considers the effects of adding twelve new orifices at each warming line outlet nozzle in front of the trash racks at the Essential Service Water (ESW) pumphouse and of adding ESW warming line vent valves EFV406/407.
The effect of these changes was to ennance the existing design to allow a warming flow rate of ~5,000 gpm to be achieved without throttling the warming valves (EFV0264/0265) . The ESW balance to provide flow to critical components in the power boek was verified to be unaffected. The only U pdated Safety Analysis Report (USAR) changes were additional changes to the P&ID (Figure 9.2-2-03) showing the addition of EF-FO-09 through 20 and EFV406/407.
Revision 0 of this USQD 59 96-0071 considers the effects of adding two new orifices at the ESW discharge outlet and increasing the ESW return piping design pressure to 75 psig.
No design basis accidents are impacted by this modification, therefore the probability of occurrence of an accident previously evaluated is not affected by the change.
The change enhances the availability of ESW pumps in cold weather.
The ability to safely shutdown the plant is improved. Thus, there is no impact on the consequences of the accidents evaluated in the USAR.
Since the orifices and the vent valves are passive components and designed, fabricated and installed to the same quality level as the ESW system, they cannot increase the malfunction of any equipment identified in the ESW system single failure analysis described in USAR Table 9.2-6. Since the malfunction of equipment is not impacted, the radiological consequences are not increased.
The commercial, industrial or military events, hazards involving missiles, pipe whip and natural phenomena, and Condition I through IV faults are not impacted by these orifices and valves. Therefore, no new type of accident is created. No new types of single failures have been created, therefore, no new types of malfunctions are possible.
The operability of the ESW system, its capacity to cool as defined in Technical Specification 3/4.7.4 are not adversely impacted by this change. No margins of safety are affected.
l
._ _ _ . - . _ _ _ ____-___ .m._ . _ _ _ _ . _ . _ _ . _ . _ _ _ ,_. _ __.
A 4
Attachment to ET 97-0017 Page 97 of 209
?
I Safety Evaluation: 59 96-0071 Revision: 2 Essential Service Water Warming Line Flow Increase d
Revision 2 of this Unreviewed Safety Question Determination (USQD) considers the combined effects of all prior changes which, 1) installs l new orifices EF-FO-07 and EF-FO-08 located at each Essential Service Water (ESW) discharge outlet, 2) increases the ESW return piping
- design pressure to 75 psig, 3) installs twelve new orifices EF-FO-009 4
through EF-FO-020 (six per train), 4) installs new vent valves EFV-406/407 on the ESW warming lines, 5) provides operational requirements that ensure ESW warming flow minimum temperature will always be sufficient, 6) provides a detailed discussion of the design basis for 9
warming line delta T and frazil ice prevention, 7) provides j surveillance requirements and performance monitoring recommendations t
for warming flow and pressure. Revision 1 of this USQD considered the effects of adding twelve new orifices at each warming line outlet nozzle and of adding warming line vent valves EFV406/407. Revision 0 of this USQD considered the effects of adding two new orifices at the j ESW discharge outlet and increasing the ESW return piping design I pressure to 75 psig.
i In summary, earlier revisions modified the ESW system to provide a nominal warming water flow rate of approximately 5,000 gpm, thus l bringing it into compliance with the existing design as described in i
'the Updated Safety Analysis Report (USAR), Because of the icing event last January, and ensuing investigations and evaluations, it is now recognized that it is necessary for there to be adequate heat load on the ESW system as well as adequate flow to prevent the blockage of the ESW trash racks with frazil ice when lake temperature is less than 32*F and an ESW pump is running.
No design basis accidents are impacted by this change. The probability of occurrence of an accident is not affected.
4
- The change enhances the availability of ESW pumps in cold weather.
The ability to safely shutdown the plant is improved. Thus, there is no impact on the consequences of the accidents evaluated in the USAR, as the performance of the ESW system has not been compromised.
Since the orifices are passive components and designed, fabricated and installed to the same quality level as the ESW system, they do not increase the probability of malfunction of any equipment previously evaluated in the USAR. Since the malfunction of equipment is not impacted and a single failure cannot prevent an adequate heat load from being present, the radiological consequences are not increased.
The commercial, industrial or military events, hazards involving missiles, pipe whip and natural phenomena, and condition I through IV
- - - .- _. . . - . . - - . - . . . . . - . - - - . - . - . . . . - , - - . . . - . _~..
1 l
i Attachment to ET 97-0017 Page 98 of 209 l
faults are not impacted by these orifices and valves. The functions -I failure modes and their effects of the systems have not changed.
Therefore, ru) new type of accident is created. j Leakage past the seals installed in EF FO-7and 8 could, if it got significantly worse, eventually result in a reduction of warming j flow. The leakage at present is insignificant in terms of system j functional performance, and is not expected to get significantly '
worse. To ensure that the leakage continues to be insignificant in I terms of system functional performance, warming line flow and pressure- )
will be re-verified once more on each train prior to winter, and at least once per year thereafter. Therefore, seal degradation to the point where warming line flow is compromised is not credible. Since it is not credible, it cannot be a failure of a different type. ;
The operability of the ESW system, its capacity to cool as defined in l Technical Specification 3/4.7.4 are not adversely impacted by this l change so no margins of safety are affected. I l
l l
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.. . .- .~ . - - - . . . _ . - . - - _ . .~ -. _ .- . . .~ .-.
i
} Attachment to ET 97-0017
. Page 99 of 209 1
i Safety Evaluation: 59 96-0071 Revision:3 Essential Service Water Warning Line Flow Increase Revision 3 of this Unreviewed Safety Question Determination (USQD) 4 provides design drawings and evaluations justifying installation of vent lines on the Essential Service Water (ESW) warming lines. If air l were allowed to accumulate in the ESW warming line high points, i warming line flow could be reduced to a less than acceptable value.
Calculations show that it may be possible under some conditions for air to come out of solution in the low pressure portions of the ESW
, return piping. Adding the vent lines will preclude this previously 4
existing condition. This modification will provide a continuous flow path from the existing safety-related 3/4" vent valves to a floor drain.
Revision 2 of this Unreviewed Safety Question Determination (USQD) l considered the combineu effects of all prior changes which, 1) installed new orifices EF-FO-07 and EF-FO-08 located at each Essential
] Service Water (ESW) discharge outlet, 2) increased the ESW return j piping design pressure to 75 psig, 3) installed twelve new orifices EF-FO-009 through EF-FO-020 (six per train), 4) installed new vent valves
! EFV-406/407 on the ESW warming lines, 5) provided operational requirements that ensure ESW warming flow minimum temperature will
- always be sufficient, 6) provided a detailed discussion of the design
< basis for warming line delta T and frazil ice prevention, 7) provided
- surveillance requirements and performance monitoring recommendations for warming flow and pressure. Revision 1 of this USQD considered the effects of adding twelve new orifices at each warming line outlet i nozzle and of adding warming line vent valves EFV406/407. Revision 0 j of this USQD considered the effects of adding two new orifices at the ESW discharge outlet and increasing the ESW return piping design I pressure to 75 psig.
In summary, earlier revisions modified the ESW system to provide a
, nominal warming water flow rate of approximately 5,000 gpm, thus bringing it into compliance with the existing design as described in the Updated Safety Analysis Report (USAR). Because of the icing event 5 last January, and ensuing investigations and evaluations, it is now recognized that it is necessary for there to be adequate heat load on j the ESW system as well as adequate flow to prevent the blockage of the ESW trash racks with frazil ice when lake temperature is less than
, 32*F and an ESW pump is running.
4 No design basis accidents are impacted by this modification. In fact, the overall effect of this change is to make the ESW system function regarding warming flow conform to that described in the Updated Safety Analysis Report (USAR). Therefore,the probability of occurrence of an accident is not affected by this change.
f
_ . _ . .._ _- . - =_ -_ -. . _ .
Attachment to ET 97-0017 Page 100 of 209 l
This change enhances the avai'.a' a llity of ESW pumps in cold weather, j The ability to safely shutdown the plant is improved. Thus, there is )
4 no impact on the consequences of the accidents evaluated in the USAR. I l
Since the orifices are passive components and designed, fabricated and installed to the same quality level as the ESW system, they do not increase the probability of malfunction of any equipment previously evaluated in the USAR. Since the malfunction of equipment is not impacted and a single failure cannot prevent an adequate heat load from being present, the radiological consequences are not increased. l l
The commercial, industrial or military events, hazards involving i missiles, pipe whip and natural phenomena, and Condition I through IV 2
faults are not impacted by these orifices and valves. Therefore, no ,
new type of accident is created. I l
Leakage past the seals installed in EF FO-7 and 8 could, if it got significantly worse, eventually result in a reduction of warming flow. The leakage at present is insignificant in terms of system a functional performance, and is not expected to get significantly worse. To ensure that the leakage continues to be insignificant in
- terms of system functional performance, warming line flow and pressure will be re-verified once more on each train prior to winter, and at .
l least once per year thereafter. Therefore, seal degradation to.the point where warming line flow is compromised is not credible. Since it is not credible, it cannot be a failure of a different type.
j The operability of the ESW system, its capacity to cool as defined in Technical Specification 3/4.7.4 are not adversely impacted by this
- change. No margins of safety are affected.
l I
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Attachment to ET 97-0017 Page 101 of 209 Safety Evaluation: 59 96-0072 Revision 0 Containment Recirculation Sump Screens This change to Updated Safety Analysis Report (USAR) Table 6.2.2-1 identifies the Safety Injection (SI) pump and Component cooling. Water pump (CCP) downstream throttle valves as having the smallest restrictions found in systems served by the Containment Recirculation sump. Implementation of this change will revise the USAR to reflect the existence of a smaller restriction.
The containment recirculation sump screens are nominally 0.125" square holes. The diagonal dimension of the screens is nominally 0.176".
Therefore, debris up to 0.176" may potentially pass through the screens into the Emergency Core Cooling System (ECCS) during a postulated post-LOCA (Loss of Coolant Accident) recirculation scenario. A review of the four 3/4" charging pump throttle valves, EMV0107, EMV0108, EMV0109, EMV0110 and the eight safety injection pump throttle valves, EMV0095, EMV0096, EMV0097, EMV0098 and EMV0089, EMV0090, EMV0091, EMV0092 finds the valves are throttled with gaps less than 0.176". It is shown that the debris effect on these valves is not a concern. The change identifies a minimal restriction less than that already identified in the USAR. This will have no effect on the safety or operation of the plant. Thus, there will be no increase in the probability of occurrence of an accident previously evaluated in the USAR.
The capability to shutdown and maintain core coverage remains unaffected. Therefore, the radiological consequences of an accident will not be affected.
The material expected to be drawn into the containment Recirculation system has already been evaluated to be equal to or less dense than water allowing it to float. It is also expected to be small and then further fragmented during passage through the CCP or SI pumps.
Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR will not be increased.
As the equipment, CCP and SI pumps and their respective throttling valves, remain unaffected by this finding, there will be no increase in radiological consequences as a result of a malfunction of equipment.
There is no potential for the creation of a new type of unanalyzed event as all equipment remains unaf fected. All margins of safety remain unchanged from that defined in the bases.for Technical Specifications.
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j Attachment to ET 97-0017 Page 102 of 209 l i l Safety Evaluation: 59 96-0073 Revision 0 Safety Injection Seal Drip Pocket Drain l
This modification provides a separate flow path to collect the drip l 4
from the mechanical seal from the ceal pocket drains for each Safety l Injection pump to a floor drain. The routing of the seal pocket drain j line separately to the Dirty Rad Waste (DRW) system will allow the oil on the skid to be intercepted before it enters the drain system.
The drain lines are added to the P&ID (USAR Fig 6.3-1).
I Since this change would not affect the system's functions or failure
, modes or the effect on equipment important to safety, there is ne j impact on accidents and malfunctions evaluated as the licensing basis. Since no credible accidents or malfunctions are identified,
- there is no potential for the creation of a new type of unanalyzed event. Nor is there any effect on margin of safety as defined in the basis for any Technical Specification.
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Attachment to ET 97-0017 l Page 103 of 209 Safety Evaluation: 59 96-0075 Revision 0
- Deletion of Special Reporting Requirement from Relocated Technical Specificatiens This change to the Updated Safety Analysis Report (USAR) Chapter 16, j Relocated Technical Specifications, deletes the requirement to submit a Special Report to the NRC currently in Operational Requirements 16.3.1.4, " Accident Monitoring Instrumentation," 16.3.1.5, " Loose-Part Detection System," and 16.7.4.1, " Area Temperature Monitoring."
Specifically, this change will eliminate the requirement in these relocated technical specifications to submit a Special Report to the NRC when selected Accident Monitoring Instrumentation is out of service for more than 30 days, when the Unit Vent High Range Noble Gas Monitor is out of service for more than 7 days, when one or more Loose l Part Detection System channels is inoperable for more than 30 days, and when the room temperature in certain monitored plant areas exceeds
, prescribed limits for more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or 30*F.
This change will not affect any design basis accidents described in Chapters 2, 6, 9, or 15 of the USAR, and will not affect any system or component operational requirements, design, surveillance requirements or safety limits. This change only deletes the reporting requirement i a to the NRC, which is not discussed elsewhere in the USAR. Therefore, l
there is no impact on accidents and malfunctions evaluated as the I
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component, there is no potential for the creation of a new type of l unanalyzed event. Nor is there any impact on margin of safety.
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- Safety Evaluation 59 96-0076 Revision 0 .
1 Fuel Building Emergency Filter Adsorber Transmitter This revision to Updated Safety Analysis Report (USAR) Table 3.11(B)-3 adds component GGPDT0019, Fuel Building Emergency Filter Adsorber i 2
transmitter, safety related equipment in a mild environment. This l component is added in order to be consistent with the intent of the I table, to list all safety-related equipment. The table is also being l revised to change the spec number for components ACPT0505 and ACPT0506 '
(Turbine Impulse Pressure Transmitter) from Westinghouse specification to M-771, i
These.are administrative changes only, and do not make any physical changes in the plant or changes to calculations. Therefore, there is I no impact on accidents or malfuctions evaluated as the licensing basis, no potential for the creation of a new type of unanalyzed event and no it. pact on margin of safety.
Attachment to ET 97-0017 Page 105 of 209 Safety Evaluation: 59 96-0079 Revision:0 Remove Control Building Condenser Relief Valves From Service This modification removes the Control Building and Class 1E Equipment air conditioning unit condenser relief valves (GKV076 9, GKV0770, GKV0771, GKV0772) from service. The valves are only required when the heat exchanger is out of service and therefore, are not required by American Society of Mechanical Engineers (ASME)Section III Code. The function of the valves are easily compensated for by the recommended opening of a vent or drain when the heat exchanger is isolated. Any overpressurization, should it occur when the heat exchanger is out of service, would pose no nuclear safety concern. Since the relief valves only provide a relief function when the component is out of service and the system is designed for the highest expected pressure it can be subjected to during operation, the valves may be removed.
The valves will be replaced with blind flanges, using ASME material, which will provide an equivalent pressure boundary function. The only change to the Updated Safety Analysis Report (USAR) is a revision of system P& ids.
The valves do not perform any relief function during operation. Their purpose is to protect the heat exchanger when it is out of service. 1 Thus, the function of these valves are not part of any design basis l accident scenario. Therefore, the probability of occurrence of a previously evaluated accident would be unaffected.
Since the valves perform no design basis function during operating modes except maintaining the pressure boundary, their absence cannot increase the consequences of any accident or malfunction of equipment important to safety. Also, their absence cannot increase the probability of occurrence of any malfunction important to safety.
There is no potential for the creation of a new type of unanalyzed event and, since the valves are not discussed in the Technical Specifications and do not form the basis of any Technical Specification, there is no impact on margin of safety.
Attachment to ET 97-0017 Page 106 of 209
- , Safety Evaluation
- 59 96-0081 Revision 0 o
Operations Organizational Change Chapter 13, conduct of Operations, of the Updated Safety Analysis Report (USAR) is being changed to reflect the organizational change that will occur in Operations. K. K. Davison is being promoted to l Supervisor Operations Support to replace R. L. Sims. There are no j effects due to this organizational change. The changes are administrative in nature. The candidate for the promotion is fully qualified and meets the minimum qualifications for the position he is being promoted to. 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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l Safety Evaluation: 59 96-0085 Revision 0 Revision 27 to the Wolf Creek Physical Security Plan This Unreviewed Safety Question Determination (USQD) evaluates changes l to Revision 27 of the Physical Security Plan, Safeguards Contingency Plan and Security Training and Qualification Plan. Major changes include: 1) Realign the Security organizational structure 2) Revise l 1 the access authorization commitments to incorporate NUMARC 91-03' l (Nuclear Power Plant Personnel Access Authorization Data Exchange Guidelines) 3) Eliminate the requirement to perform a metal detector ,
search on Security personnel who leave the protected area to perform l
. official duties 4) Improve consistency of text regarding authorization l
- for increasing the visitor-escort ratio 5) Allow cargo to be exengt I from search at the protected area entrance 6) Allow cargo to be exempt
- from search under specified conditions 7) Revise drawings to
- incorporate changes to site structures 8) Incorporate the addition of a vehicle Barrier System 9) Remove the requirement to inventory ACADs i every six months 10) Remove the old process for making ACAD badges 11)
Remove the " Licensee Designated Vehicle" requirement for vehicles i going to and from the Essential Service Water Pumphouse 12) Specify that Gate-10 will be controlled by a Security Trained Individual whenever the gate is open to allow cargo to enter the Warehouse from
- the Cargo Search Area 13) Remove the requirement to post the l l Containment Personnel Hatch during periods of mejor maintenance and refueling and clarifies other text relating to access controls within protected and vital areas 14) Revise text associated with barrier
} descriptions 15) Revise the Physical Exercise Performance Test 16)
J Correct typographical errors, revise procedure numbers and other minor grammatical and administrative changes.
There is no impact on accidents and malfunctions evaluated as the licensing basis. There is no potential for the creation of a new type of unanalyzed event. The margin is safety as defined in the basis for l I
any Technical Specification has not been reduced.
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- g. Safety Evaluation: 59 96-0086 Revision 0 Downgrade of Chemical and volume system component This modification revises the Updated Safety' Analysis Report (USAR)
Figure 9.3-8-01 to identify BGV133, root isolation valve to pressure gauge BGPIO184 (RCP No. 1 Seal Back Pressure),- as a normally closed valve. USAR Table 3.11(B)-3 is revised to delete component BGPIO184 as it has been reclassified as non-safety related and therefore shall not be listed-in the safety related components listing.
BGPIO184 provides indication only of RCP #1 seal leakoff and provides no control or mitigating function in the event of an accident and therefore isolation of the gauge and classification as non-safety related does not increase the frequency of' occurrence or radiological consequences of an accident previously evaluated in the USAR.
Since BGPIO184 provides indication only of RCP #1 seal leakoff and neither provides, nor affects other systems providing accident mitigation, this change does not increase the frequency of occurrence or-the radiological consequances of a malfunction of equipment important to safety.
There are no existing or new accidents the change could affect or create. The possibility of a different type of malfunction of equipment important to safety is not created. There are ru) acceptance limits contained in the bases of any technical specification or other i licensing basis documents that are affected by this change.
Attachment to ET 97-0017 Page 109 of 209 Safety Evaluation: 59 96-0087 Revision:0 Identification of Supervisor Chemistry as the ANSI /ANS 3.1-1978 Chemistry and Radiochemistry Position During a previous organization change, a temporary assignment was made until such time the Superintendent chemistry qualified for that position, fulfilling the requirements of the ANSI /ANS 3.1-1978 chemistry position. This change only affects the Resumes in the Updated Safety Analysis Report (USAR) and has no effect on the organization since the Superintendent Chemistry now meets the ANSI requirements for the chemistry position. No other changes have been made to the organization. This change will not affect equipment, procedures nor test or experiments.
l This change does not affect nor create accidents as are or could be described in the USAR. Malfunctions of equipment important to safety i and acceptance limits for the Technical Specifications are not affected.
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Safety Evaluationt 59 96-0088 Revision 3 0 l
steam Generator Blowdown Regenerative Heat Exchanger Vent and Drain '
Addition This modification removes the one inch upper shell drain line from the
- Steam Generator Blowdown Regenerative Heat Exchanger and installs valves, elbows, caps and pipe as required to provide an independent drain and vent for the shells. The Steam Generator Blowdown Regenerative Heat Exchanger is double shell design with the shells arranged in a vertical orientation. A one inch pipe ties the two shells together to allow for draining the upper shell through the lower shell. The difference in thermal expansion between the shells has induced stresses into the drain pipe and cracked the upper shell drain line attachment weld. This change will eliminate the tie between the shells thus preventing reoccurrence of the problem.
The installation of the independent drain and vent will not affect the design bases accidents discussed in Updated Safety Analysis Report (USAR) chapters 2, 3 or 15.
Since the system functions are not changed, no credible accidents that could be created are 3dentified.
Since the change would not affect the system's failure modes or affect equipment important to safety, no credible malfunctions of equipment important to safety are identified.
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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l Safety Evaluation 8 59 96-0009 Revision 0 Circulating Water screen House Enclosure This modification will add a three-sided addition to the circulating water screen house. The roof will have removable hatches to allow the crane to remove equipment as needed. A new walk way will be added to l the east and south sides of the existing structure. Heating and lighting will be installed for the new addition. The existing exterior wall that will be enclosed by the new addition will have the siding and insulation removed. Updated Safety Analysis Report (USAR) figures 9.2-1-03 r/9, 10.4-04 r/9 and 10.4-1-05 r/9 will be revised to reflect the new addition.
There are no design basis accidents associated with the circulating water (CW) screen house (structure Z019), service water system (WS),
traveling water screens (SW), acid feed (AX), chlorine (CL), water
- gland (WG), vent system (VH) at ZO19, service air (SZ) , heat tracing (HT) or fire protection (FP) systems.
The circulating water screen house, AX, CL, CW, FP, HT, SW, SZ, WG, VH and WS systems are not evaluated for a design basis accident initiator in Chapter 15 of the USAR. These systems are non-safety related.
No credible malfunctions of equipment important to safety will be affected by the modification due to these systems, AX, CL, CW, FP , HT, SW, SZ, WG, VH and WS being non-safety related located in a building that is non-safety related.
This modification will nava 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 as the AX, CL, CW, FP, HT, SW, SZ, WG, VH and WS systems are not listed in any Technical Specifications.
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Attachment to ET 97-0017 Page 112 of 209 Safety Evaluation: 59 94-0090 Revision 0 l
- Elbow Replacement Modification Downstream of AEFV0001B (Feedwater Flow Valve)
This modification replaces an elbow immediately,down stream of AEFV0001B (Feedwater Flow valve) which is severely eroded at its {
extrados. There is a high differential pressure across the valve. I Previously, the subject elbow had been replaced by a thicker schedule elbow and from carbon steel to stainless steel. This modification replaces the subject elbow along with its attached elbowlet and cap to a flanged tee. There is a cap welded to the blind flange as a i secondary safety measure. To prevent any leakage through the flanges, )
Plexitallic, style Compressed Gasket Inner-ring (CGI) spiral wound I gasket is approved.
This is a non-safety related system. The modification will have no impact on accidents or malfunctions evaluated as the licensing basis and there is no potential for the creation of a new type of unanalyzed ,
event. There is no reduction in the margin of safety. )
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Attachment to ET 97-0017 Page 113 of209 Safety Evaluation: 59 96-0091 Revision 0 Organisational Changes In operations, Engineering and the Adaministrative organisation This organizational change includes the realignment of the Operations Organization to include System Engineering and the Quality function.
The change to the organization also includes changes to Engineering l (to include Information Services) and the Administrative organization (to change the structure of the communications functions) . Titles and personnel are changed and typographical errors are corrected.
There are no accidents discussed in the Updated Safety Analysis Report (USAR) that rely on organization to mitigate. A change to the organization will not create new accidents since'all functions continue to be performed and the personnel performing the functions meet the ANSI qualifications. For the same reasons that no accidents 4 are impacted or created, no malfunctions of equipment important to safety are affected. There are no acceptance limits affected by this change since no functions have been deleted.
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Safety Evaluation: 59 96-0091 Revision 1 Organisational Changes In Operations, Engineering and the l Administrative Organisation )
Revision 1 of this Unreviewed Safety Question Determination (USQD) addresses a Updated Safety Analysis Report (USAR) commitment that the l Independent Safety Evaluation Group (ISEG) reports to a corporate I official who holds a high-level, technically oriented position that is not in the management chain for power production. Currently, the I
Performance Improvement and Assessment group reports to the Chief Operating Officer (COO). The Performance Improvement and Assessment Group also maintains direct reporting lines to the Chief Executive Officer (CEO) for matters where independence is of importance. Due to this independence, the ISEG/ Performance Improvement and Assessment reporting chain to Otto Maynard will not invalidate the USAR commitment. l Revision 0 of the USQD is described as follows.
This organizational change includes the realignment of the Operations Organization to include System Engineering and the Quality function.
The change to the organization also includes changes to Engineering (to include Information Services) and the Administrative organization (to change the structure of the communications functions). Titles and personnel are changed and typographical errors are corrected.
There are no accidents discussed in the USAR that rely on organization to mitigate. A change to the organization will not create new accidents since all functions continue to be performed and the personnel performing the functions meet the ANSI qualifications. For the same reasons that no accidents are impacted or created, no malfunctions of equipment important to safety are affected. There are no acceptance limits affected by this change since no functions have been deleted.
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Attachment to ET 97-0017 Page 115 of 209 1 Safety Evaluation: 59 96-0092 Revision 30 Reflection of the As-Built Configuration of Potable Water System This modification revises the Updated Safety Analysis Report (USAR)
Table 1.7-3 and Figure 9.2-5A to reflect the as-built configuration of the underground potable water supply piping and isolation valves.
$ This configuration has existed since site construction and was changed out due to construction supply piping corrosion leakage and site improvements not shown previously. This change is outside the power i block structures and has no effect on safety relnted structures, systems and components.
There are no design basis accidents affected by these changes.
l There are no types of credible accidents created. <
There are no credible malfunctions of equipment important to safety affected as a result of this change.
s I The potable water system has no effect on acceptance limits found in any Technical Specification or licensing basis document.
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Attachment to ET 97-0017 Page 116 of 209 Safety Evaluation: 59 96-0093 Revision:0 Radiological Emergency Response Plan Revision 50 This modification revises the Radiological Emergency Response Plan, to reflect changes to the Public Information Organization, including: 1) restructuring of the Public Information Organization staff and responsibilities, 2) the addition of an Information Clearinghouse (IC) and Media Center (MC) at the Wolf Creek Learning Center, 3) moving the backup IC/MC to Topeka, and 4) potential activation of the Wolf Creek Public Information Organization at an NUE (Notification of Unusual Event) rather than at an ALERT. Further, this revision changes EPP 03-1,1, " Emergency Planning Program" references to those of the new j procedure, AP 17C-024, " Emergency Planning Responsibilities." l The proposed changes do not impact design basis accidents. As stated above, this change impacts organizational structure and has been made to improve the dissemination of information to the public.
This change does not create any new credible accidents.
This change does not increase the credible malfunctions of equipment l important to safety. '
This change does not affect the acceptance limits of Technical Specifications.
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Attachment to ET 97-0017 Page 117 of 209 1 l Safety Evaluation 8 59 96-0094 Revision 0 l Freeze Seals on Closed Cooling Water Lines This temporary modification installs freeze seals on the 3" closed cooling water lines to and from the condensate polishing and secondary sample chillers to allow cutting and capping the lines which go to the condensate polishing sample chillers. These freeze seals support implementation of Design Change Package 06256 which provides chilled water cooling water supply to the sample chillers and removes the closed cooling water supply. Cooling water to the sample chillers will be isolated, which will not affect plant operation. The freeze i seals will be installed to ensure qualification and testing at higher l than design pressures and that emergency contingency measures are in place.
The freeze seals will provide isolation for the subject lines which -
I are not indicated on Updated Safety Analysis Report (USAR) Figure 9.2- )
- 14. l This temporary modification does not create the possibility of credible accidents nor affect equipment, systems or components 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.
No acceptance limits are identified that are affected by this temporary modification.
Attachment to ET 97-0017 Page 118 of 209 Safety Evaluation 59 96-0096 Revision 0 Root Cause Investigation of Cycle 8 Rod Insertion Anomaly This Unreviewed Safety Determination (USQD) evaluates all phases of an investigation in the root cause analysis of the Cycle 8 rod insertion anomaly. Field examinations of irradiated fuel assemblies will be performed to assist in the root cause analysis. As part of this investigation, assemblies H38 and H50 will have all fuel rods removed and placed in unirradiated skeletons. The irradiated skeletons will then be sectioned and transported to Westinghouse for hot cell examination.
This investigation will not increase the probability of occurrence of an accident previously evaluated in the Updated Safety Analysis Report (USAR). The only accident scenario applicable to the scheduled activities is the fuel handling accident covered in Section 15.7 of the USAR. The fuel handling accident is postulated to occur after the fuel assembly is moved through the transfer gate and prior to being lowered into the designated spent fuel pool area. This section of the USAR describes the dropping of a fuel assembly which results in 100%
cladding failure. During the examination phase, a designated number of fuel assemblies will be transferred to alternate Spent Fuel Pool locations under controlled procedures. The breakage of a fuel rod during reconstitution activities is bounded by existing safety analyses and the reconstitution activities are performed using controlled procedures. Inadvertent criticality based on fuel rods positioned in the unirradiated skeleton in a more reactive configuration has been analyzed as well as reactivity effects from moderation effects while removing fuel rods.
This investigation will not increase the radiological consequences of an accident previously evaluated in the USAR. The fuel handling accident analyses contained in Section 15.7 of the USAR utilizes conservative assumptions. In the unlikely event of a fuel handling accident, fission product inventories would be sufficiently below accident analysis assumptions. The fuel handling accident is postulated to occur 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown while the fuel assemblies have actually been in the spent fuel pool several months. Fission products have had adequate time to decay, especially noble gas inventory assumed for offsite release. Fission product gas release to the public is minimized by the emergency exhaust system. One train is required to be running during all fuel movements.
The probability of occurrence of a malfunction of equipment important to safety will not increase due to the. investigation. The change in fuel assembly configuration has no impact to plant equipment. Fuel building emergency exhaust is unaffected by the work being performed in the fuel building. Spent fuel pool cooling is unaffected by this
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4- Attachment to ET 97-0017 Page 119 of 209 change. Hoists and cranes used during fuel or cask movement will be
, limited to the maximum load. Weekly crane and hoist checks will be q
performed per procedure.
The only equipment important to safety which would increase the l radiological consequences if it malfunctioned during a postulated accident is the emergency filtration. Because the work being i
performed does not affect this system, the radiological consequences of a postulated fuel handling accident would not increase.
l The investigation will not create the possibility of an accident of a different type than previously evaluated. The activities planned for reconstitution and examination take place in the Fuel Building. None ,
of the activities impact plant resources outside of the fuel building '
with the exception of service air and electrical demand. Electrical
, out?.ets are supplied by non-vital buses in the fuel building. ~
Similarly, service air is non-vital and ultimate failure would not I j affect plant equipment to cause an accident of a different type. All I postulated fuel handling accidents have been previously evaluated. j Movement of irradiated skeleton segments will not create the l possibility of an accident of a different type. The dropping and !
l potential failure of a single rod is bounded by analyses assuming 2
failure of 100% of the cladding of 264 fuel rods.
4 i The investigation will not create the possibility of a different type of malfunction of equipment important to safety than any previously 4 evaluated in the USAR. All plant equipment will be used in accordance with plant procedures. Westinghouse equipment will have appropriate procedures controlling their usage. All Westinghouse procedures will
, be reviewed and approved. Equipment, parts or tools dropped in the i cask loading pool will not affect operation of the spent fuel pool i cooling system. Items in the bottom of the cask loading pool have no
! means to enter the-spent fuel pool. Foreign Material Exclusion j procedures will be adhered to throughout the examination.
- The margin of safety as defined in the basis for any Technical Specification is not reduced by this investigation. Rebuilt fuel !
4 assemblios will have the same burnup as they had prior to relocating ,
fuel rods- . An evaluation performed by Westinghouse concluded that '
4 there would be no impact to reactivity by relocating fuel rods in alternate locations of an assembly. Reactivity effects of removing a
] fuel rod from an assembly from moderation will not occur if soluble j boron concentrations are kept above 1000 ppm. This is below the required boron level for the spent fuel pool. Fuel building emergency ;
exhaust will not be affected by the work that will occur in the fuel
- building. Boration of the spent fuel pool will not be changed during the work activities. Heat loads in the spent fuel pool will not be j increased.
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Attachment to ET 97-0017 Page 120 of 209 Safety Evaluation 8 59 96-0096 Revision: 1 Root Cause Investigation of Cycle 8 Rod Insertion Anomaly Revision 1 of this Unreviewed Safety Determination (USQD) allows a modification to position a failed rod storage rack on an 18" spacer in the spent fuel pool during fuel rod transfers to the failed rod storage rack. The spacer will elevate the rack to allow greater access and visibility. The reduction in the minimum water height above the top of fuel rods will not impact the current safety analyses which require 23 feet of water above the top of an irradiated assembly.
This investigation, as previously considered in Revision 0 of USQD 96-0096 is shown below.
This USQD evaluates all phases of an investigation in the root cause analysis of the Cycle 8 rod insertion anomaly. Field examinations of irradiated fuel assemblies will be performed to assist in the root cause analysis. As part of this investigation, assemblies H38 and H50 will have all fuel rods removed and placed in unirradiated skeletons.
The irradiated skeletons will then be sectioned and transported to Westinghouse for hot cell examination.
This investigation will not increase the probability of occurrence of an accident previously evaluated in the Updated Safety Analysis Report (USAR). The only accident scenario applicable to the scheduled activities is the fuel handling accident covered in Section 15.7 of the USAR. The fuel handling accident is postulated to occur after the fuel assembly is moved through the transfer gate and prior to being lowered into the designated spent fuel pool area. This section of the USAR describes the dropping of a fuel assembly which results in 100%
cladding failure. During the examination phase, a designated number of fuel assemblies will be transferred to alternate Spent Fuel Pool locations under controlled procedures. The breakage of a fuel rod during reconstitution activities is bounded by existing safety analyses and the reconstitution activities are performed using controlled procedures. Inadvertent criticality based on fuel rods positioned in the unirradiated skeleton in a more reactive configuration has been analyzed as well as reactivity effects from moderation effects while removing fuel rods.
This investigation will not increase the radiological consequences of an accident previously evaluated in the USAR. The fuel handling accident analyses contained in Section 15.7 of the USAR utilizes conservative assumptions. In the unlikely event of a fuel handling accident, fission product inventories would be sufficiently below accident analysis assumptions. The fuel handling accident is postulated to occur 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown while the fuel assemblies
l Attachment to ET 97-0017 l
Page 121 of 209 have actually been in the spent fuel pool several months. Fission products have had adequate time to decay, especially noble gas l inventory assumed for offsite release. Fission product gas release to j the public is minimized by the emergency exhaust system. One train is l required to be running during all fuel movements. i l
The investigation will not increase the probability of occurrence of a j malfunction of equipment important to safety. The change in fuel '
assembly configuration has no impact to plant equipment. Fuel I building emergency exhaust is unaffected by the work being performed l in the fuel building. Spent fuel pool cooling is unaffected by this l change. Hoists and cranes,used during fuel or cask movement will be l limited to the maximum load. Weekly crane and hoist checks will be performed per procedure.
The only equipment important to safety which would increase the radiological consequences if it malfunctioned during a postulated accident is the emergency ff.ltration. Because the work being performed does not affect '.his system, the radiological consequences of a postulated fuel handl..ng accident would not increase.
The investigation will not create the possibility of an accident of a different type than pre *Aously evaluated. The activities planned for reconstitution and ext.nination take place in the Fuel Building. None of the activities impact plant resources outside of the fuel building I with the exception of service air and electrical demand. Electrical i outlets are supplied by non-vital buses in the fuel building.
Similarly, service air is non-vital and ultimate failure would not affect plant equipment to cause an accident of a different type. All postulated fuel handling accidents have been previously evaluated.
Movement of irradiated skeleton segments will not create the possibility of an accident of a different type. The dropping and potential failure of a single rod is bounded by analyses assuming failure of 100% of the cladding of 264 fuel rods.
The investigation will not create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the USAR. All plant equipment will be used in accordance with plant procedures. Westinghouse equipment will have appropriate procedures controlling their usage. All Westinghouse procedures will be reviewed and approved. Equipment, parts or tools dropped in the cask loading pool will not affect operation of the spent fuel pool cooling system. Items in the bottom of the cask loading pool have no means to enter the spent fuel pool. Foreign Material Exclusion procedures will be adhered to throughout the examination.
This investigation does not reduce the~ margin of safety as defined in the basis for any technical specification. Rebuilt fuel assemblies will have the same burnup as they had prior to relocating fuel rods.
An evaluation performed by Westinghouse concluded that there would be
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Attachment to ET 97-0017 Page 122 of 209 no impact to reactivity by relocating fuel rods in alternate locations of an assembly. Reactivity effects of removing a fuel rod from an assembly from moderation will not occur if soluble boron concentrations are kept above 1000 ppm. This is below the required boron level for the spent fuel pool. Fuel building emergency exhaust will not be affected by the work that will occur in the fuel building. Boration of the spent fuel pool will not be changed during the work activities. Heat loads in the spent fuel pool will not be increased.
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1 Page 123 of 209 3afety Evaluation: 59 96-0096 Revision: 2 Root Cause Investigation of Cycle 8 Rod Insertion Anomaly Revision 2 of this Unreviewed Safety Determination (USQD) includes additional controls to be followed during shipping cask movement.
While the shipping cask is being lifted by the cask handling crane, interlocks and mechanical stops will be installed to prevent shipping cask movement over spent fuel assemblies or fuel rods in the spent fuel pool. Interlocks are removed when transferring inspection equipment into and out of the spent fuel pool. However, these are not considered heavy loads because their weight is less than a fuel assembly and handling tool. Also, the shipping cask will be moved in safe load positions in the fuel building only.
The cask handling accident analyzed in the Updated Safety Analysis Report (USAR) assumes a spent fuel assembly shipping cask with -
irradiated fuel assemblies inside the cask. The shipping cask used for this investigation will not contain spent fuel assemblies.
This investigation has been previously considered in USQD 96-0096 Revision 0 and Revision 1 and is shown below.
Revision 1 of this USQD installs an 18" spacer upon which to position a failed rod storage rock in the spent fuel pool during fuel rod transfers to the failed rod storage rack. The spacer will elevate the rack to allow greater access and visibility. The reduction in the minimum water height above the top of fuel rods will not impact the current safety analyses which require 23 feet of water above the top of an irradiated assembly. The 18" loss in water shielding will not impact current safety analyses because the rack is capable of only holding up to 52 fuel rods while an assembly holds 264.
Revision 0 of this USQD evaluates all phases of an investigation in the root cause analysis of the Cycle 8 rod insertion anomaly. Field examinations of irradiated fuel assemblies will be performed to assist in the root cause analysis. As part of this investigation, assemblies H38 and H50 will have all fuel rods removed and placed in unirradiated skeletons. The irradiated skeletons will then be sectioned and transported to Westinghouse for hot cell examination.
This investigation will not increase the probability of occurrence of an accident previously evaluated in the Updated Safety Analysis Report (USAR). The only accident scenario applicable to the scheduled activities is the fuel handling accident covered in Section 15.7 of the USAR. The_ fuel handling accident is postulated to occur after the fuel assembly is moved through the transfer gate and prior to being lowered into the designated spent fuel pool area. This section of the USAR describes the dropping of a fuel assembly which results in 100%
Attachment to ET 97-0017 Page 124 of 209 cladding failure. During the examination phase, a designated number of fuel assemblies will be transferred to alternate Spent Fuel Pool locations under controlled procedures. The breakage of a fuel rod during reconstitution activities is bounded by existing safety analyses and the reconstitution activities are performed using controlled procedures. Inadvertent criticality based on fuel rods positioned in the unirradiated skeleton in a more reactive configuration has been analyzed as well as reactivity effects from moderation effects while removing fuel rods.
This investigation will not increase the radiological consequences of an accident previously evaluated in the USAR. The fuel handling accident analyses contained in Section 15.7 of the USAR utilizes conservative assumptions. In the unlikely event of a fuel handling accident, fission product inventories would be sufficiently below accident analysis assumptions. The fuel handling accident is postulated to occur 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown while the fuel assemblies have actually been in the spent fuel pool several months. Fission products have had adequate time to decay, especially noble gas inventory assumed for offsite release. The failed rod storage rack is placed on an 18" spacer to increase visibility. The failure of fuel rods in the failed rod storage rack in the elevated position would not increase the amount of radioactive gases released. Fission product gas release to the public is minimized by the emergency exhrmst system. One train is required to be running during all fac1 movements.
The probability of occurrence of a malfunction of equipment important to safety will not increase due to the investigation. The change in fuel assembly configuration has no impact to plant equipment. Fuel building emergency exhaust is unaffected by the work being performed in the fuel building. Spent fuel pool cooling is unaffected by this change. Hoists and cranes used during fuel or cask movement will be l limited to the maximum load. Weekly crane and hoist checks will be l performed per procedure.
The only equipment important to safety which would increase the radiological consequences if it malfunctioned during a postulated accident is the emergency filtration. Because the work being performed does not affect this system, the radiological consequences of a postulated fuel handling accident would not increase.
The investigation will not create the possibility of an accident of a different type than previously evaluated. The activities planned for reconstitution and examination take place in the Fuel Building. None of the activities impact plant resources outside of the fuel building with the exception of service air and electrical demand. Electrical I outlets are supplied by non-vital buseb in the fuel building.
Similarly, service air is non-vital and ultimate failure would not affect plant equipment to cause an accident of a different type. All J postulated fuel handling accidents have been previously evaluated.
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4 Movement of irradiated skeleton segments will not create the possibility of an accident of a different type. The dropping and l 1 potential failure of a single rod is bounded by analyses assuming
] failure of.100% of the cladding of 264 fuel rods.
The investigation will not create.the possibility of a different type
} of malfunction of equipment important to safety than any previously evaluated in the USAR. All plant equipment will be used in accordance with plant procedures. Westinghouse equipment will have appropriate l procedures controlling their usage. All Westinghouse procedures will be reviewed and approved. Equipment, parts or tools dropped in the l cask loading pool will not affect operation of the spent fuel pool l
j cooling system. Items in the bottom of the cask loading pool have no means to enter the spent fuel pool. Foreign Material Exclusion 4
procedures will be adhered to throughout the examination.
The margin of safety as defined in the basis for any Technical Specification is not reduced by this investigation. Rebuilt fuel assemblies will have the same burnup as they had prior to relocating i fuel rods. An evaluation performed by Westinghouse concluded that I
there would be no impact to reactivity by relocating fuel rods in
- alternate locations of an assembly. Reactivity effects of removing a
- fuel rod from an assembly from moderation will not occur if soluble boron concentrations are kept above 1000 ppm. This is below the required boron level for the spent fuel pool. Fuel building emergency exhaust will not be affected by the work that will occur in the fuel j building. Boration of the spent fuel pool will not be changed during
] the work activities. Heat loads in the spent. fuel pool will not be increased. The shipping cask movement will be limited to safe zones and mechanical stops and interlocks will be in place when the shipping cask is moved.
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Attachment to ET 97-0017 Page 126 of 209 Safety Evaluation: 59 96-0097 Revision 0 Silica Analyser Changeout This modification provides a stabilized pressure source for the sample flow to the silica analyzers when the demineralizer train is not in service. This will eliminate the numerous nuisance low pressure alarms. The new sample flow source to the silica analyzers is provided by the laboratory demineralizer water pump instead of the regeneration water pump's discharge line. Updated Safety Analysis Report (USAR) Figure 9.2-5 will need to be revised to reflect this change.
There are no design basis accidents identified or evaluated for the non-safety related Makeup Demineralizer (WM) system in the USAR chapters 2, 3, or 15.
1 Since the system's functions are not changed, no credible accidents that could be created are identified.
1 Since the modification would not affect the system's failure modes or i equipment important to safety, no credible malfunctions of equipment I important to safety are identified.
Since this system is not included in the bases of the Technical Specifications, no acceptance limits are identified that could be affected.
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Safety Evaluation: 59 96-0098 Revision 0 Mandrail and Grating Installation in Radwaste Building This modification revises Updated Safety Analysis Report (USAR)
Figures 1.2-3 and 1.2-8 to show the removable handrail and grating to be installed in the Radwaste Building HIC (High Integrity Container) processing area. Specifically, the handrail and grating will be added to the solid radwaste disposal. station and HIC lid storage platforms at elevation 2013'.
-The installation of handrail and grating will not introduce any new
- hazards in the Radwaste Building HIC processing area. The Radwaste Building contains no safety related systems, structures or components.
This modification will not increase the amount of radioactive material, nor will it change or degrade the physical or operational barriers used to confine or mitigate the release of these materials from previous analyses. The function of any safety related plant equipment / components or systems will not be impaired, and.the ability to achieve and maintain safe plant shutdown will not be adversely affected.
There are no design basis' accidents identified because this change has j no effect on the inputs, assumptions or components involved in the 1 accidents evaluated in the USAR.
Since no hazards are introduced by this change and since no safety related equipment is being challenged, no credible accidents that could be created are idancified. Since no new hazards are created, no credible malfunctions of equipment'important to safety are identified. No acceptance limits that could be affected are identified.
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Safety Rvaluation: 59 96-0099 Revision 0 General amployee Training Procedure Revision This revision to procedure AP 30A-001, " General Employee Training" incorporates the guidance of ACAD 93-009, " Guidelines for General Employee Training". This industry standard specifies the minimum level and content of training programs for unescorted access to the protected area - (Plant Access Training, PAT) and for unescorted access to the RCA (Radiological Controlled Area, Radiation Worker Training, RWT). The changes to AP 30A-001 include:
- 2) Adding form APF 30A-001-01 as a requirement for unfettered access.
- 3) Adding a requirement to notify Health Physics upon any RWT requalification examination failure.
- 4) Changed "a written examination" to "an examination" in Section 6.7 to more accurately describe the unfettered access process.
ACAD 93-003 guidance has been implemented throughout the nuclear industry for General Employee Training. It establishes an acceptable level of training for unescorted access which is generally accepted by regulatory agencies. Therefore, there is no reduction in the acceptable level of training and knowledge required for unescorted access.
Updated Safety Analysio Report (USAR) Section 13.2.2.9 describes the General Employee Training program and includes specific descriptions of the content of several topics included in general employee training and radiation worker traiaing. These changes would remove the specific descriptions and replace them with the topical areas covered by PAT and RWT courses in accordance with the industry guidance.
Since these changes incorporate industry guidance that continues to provide an acceptable level of training and knowledge for unescorted access to the protected area and the RCA, and this change does not
-impact structures, systems or components, or affect procedures or processes which could impact the safe operation of the plant, there are no design basis accidents identified which could be impacted by these changes. Nor are there any credible accidents identified that could be created.
Since this change does not impact structures, systems or components, or affect procedures or processes which could impact the safe operation of the plant, there are no malfunctions identified which could be created. Nor are any acceptance limits affected by this change.
Attachment to ET 97-0017 Page 129 of 209 Safety Evaluation: 59 96-0100 Revision 0 Control Rod Drive Mechanism Ductwork Removal e
This change to the Updated Safety Analysis Report (USAR) more accurately reflects the current work practices of removing control Rod Drive Mechanism (CRDM) ductwork in preparation for refueling activities. The USAR currently states that the CRDM ductwork is l t
removed in a maximum of 3 sections. The ductwork is actually removed '
in more than 3 sections. The total amount of ductwork removed does not change.
The most limiting accident for the plant in this condition is dropping the reactor vessel head caused by the failure of the polar crane.
This change will not increase the probability of occurrence of an accident or malfunction of equipment previously evaluated in the USAR 4 since removing the ductwork in more than 3 sections has not increased
- l the possibility of a crane failure.
The radiological consequences of an accident and malfunction of 4
equipment previously analyzed in the USAR will not increase since removing the ductwork in more than 3 sections has been bounded by the reactor vessel head being dropped.
This change will not create the possibility of an accident of a different type than any previously evaluated in the USAR since evaluations have been performed for dropping loads with the polar crane.
- since the operation of the polar crane has not changed, the possibility of a different type of malfunction of equipment important ,
to safety is not created.
This change will not reduce the margin of safety as defined in the
- basis for any Technical Specifications. Removal of the CRDM ductwork 4
is not in the basis for any Technical Specification.
Attachment to ET 97-0017 Page 130 of 209 i
safety Evaluation: 59 96-0102 Revision 0 Health Physics Facilities, Equipment and Methods This change to the Updated safety Analysis Report (USAR) reflects changes to the Health Physics (HP) facilities, equipment and methods.
The change will correct the current USAR by changing locations of the different HP facilities. Also, electronic dosimetry are baing used along with Pocket Ionization Chambers (PICS) to measure radiation exposure and are more accurate than PICS.
The changes do not affect any accidents or analysis of any accidents as shown in the USAR. Therefore, there would be no radiological consequences, no equipment malfunctions or safety concerns.
The changes are strictly administrative in nature and do not affect any equipment or malfunction of equipment as shown in the USAR.
This revision will have no potential for the creation of a new type of ,
unanalyzed event. There is no reduction in the margin of safety as !
defined in the Technical Specifications.
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Attachment to ET 97-0017 Page 131 of 209 Safety Evaluation: 59 96-0103 Revision: 0 Changes to Fences and Building Names This change to the Updated Safety Analysis Report (USAR) will:
- 1) Add a note to USAR Section 1.2.2 to indicate that only " general use" descriptions of structures appear in the Section and change names of buildings on Figure 1.2-44,
- 2) Show the removal of fencing east of Vehicle Maintenance Shop and Administrative Services Shop.
- 3) Remove fencing north of the Potable Water Storage Tank.
- 4) Show the removal of the Air Compressor Building.
- 5) Show the addition of the concrete driveway at the north-west end of the General Office Building (GOB).
- 6) Show the addition of the garbage bin projecting west from the GOB.
- 7) Show the removal of the driveway at the south end of Support Building West and the addition of steps, ramp and sidewalk to the south door.
- 8) Delete the reference to the Communications group in trailer 53 of the Trailer Schedule on Fig 1.2-44.
These changes will have no impact to design basis accidents discussed or referenced in USAR chapters 2, 3, 6, 9 and 15. The name changes to the buildings are strictly administrative changes and have no effect to the " physical" plant site. The removal of fencing and the revisions to driveways / side walks have no effect to the operation of the plant and have no effect to site flooding.
These revisions 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 redt etion in the margin of safety.
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- Safety Evaluation: 59 96-0104 Revision 0
- Dionex Process Analyser Installation I
'his modification installs a Dionex Series 8200 Process Analyzer in the Turbine Building Process Sampling Room to support the secondary
- chemistry program. Installation requires that existing turbine cycle sample lines be extended to the new analyzer. A circuit must be added
, to provide 120 VAC power to the process analyzer. Piping must be
- added to drain sample water from the analyzer to the Turbine Building j oily Waste System floor drain.
4 h The Dionex analyzer will enhance the secondary chemistry program by i providing capability to detect smaller chemical concentrations than a can be presently measured. The analyzer will allow better control of l chlorides and sulfates in the secondary system thereby extending steam j generator life by preventing tube cracking.
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! The portion of the plant sampling system affected by this modification serves no safety related function. No design basis accidents listed in the Updated Safety Analysis Report (USAR) are affected. There is no impact on accidents and malfunctions evaluated as the licensing basis.
The process analy=er modification will not create the possibility of a malfunction of equipment important to safety. The only accidents this modification could create are insignificant, do not affect the release of radiation or plant safety and are not listed in the USAR.
No Technical Specifications' apply to the Process Sampling System for turbine cycle waters.
Attachment to ET 97-0017 Page 133 of 209 Safety Evaluation: 59 96-0105 Revision 0 Main Tool Room and Hot Tool Roosa Data-link Installation This modification installs a high-speed data-link between the Main Tool Room and the Hot Tool Room. This link is necessary in order to maintain a database between the two tool rooms which will be used to l track'the check-out/ return of bar-coded tools. The " Corporate" Local Area Network (LAN) will be used for this link by utilizing the I existing Telephone cable raceway in the Power Block for Fibsr Optic cable routing.
The Telephone System is identified as having a dedicated raceway in the Updated Safety Analysis Report (USAR) Section 9.5.2. This is being changed to identify the telephone raceway as a Telephone / Fiber Optics raceway system.
Trare are no design basis accidents that could be affected by the utilization of the Telephone System Raceway for Fiber Optic cable routing.
If a failure were to occur on this system, no equipment important to the safety of the plant would be affected. Tool issuance could be controlled by manual log. For equipment important to safety, no credible accident could be created by failure of this equipment.
The communications system contains no equipment important to safety.
There are no credible malfunctions of equipment important to safety which may be affected by routing of Fiber Optic cable within the Telephone System raceway.
Since no acceptance limits were identified that could be affected, the margin of safety is not affected.
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Attachment to ET 97-0017 Page 134 of 209 Safety Evaluation: 59 96-0107 Revision 0 Operations Organisational Changes This change revises Updated Safety Analysis Report (USAR) Chapter 13.1.2.2.1 to change the Wolf Creek organizational Chart for the Manager operations. This is being done to reflect the di cot reporting of the Shif t Supervisor and his direct reports to the Ma ager Operations.
This change, since it is only a change to personnel, will not affect equipment, procedures nor test or experiments. This change does not affect or create accidents as are or could be described in the USAR.
This change does not affect malfunctions of equipment or acceptance limits for the Technical Specifications.
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Attachment to ET 97-0017 Page 135 of 209 Safety Ivaluation: 59 96-0108 Revision 0 Organisation Change to Reflect a New Chief Operating Officer and a Accountability, Comunitment. and Excellence Of ficer This change only affects organizational descriptions, the Resume for the Chief Operating Officer in the Updated Safety Analysis Report i (USAR) and the Organization Chart (Fig. 13.1-1). C.C. Warren has been assigned to the position of Chief Operating Officer. O.L. Maynard has been assigned to the position of Acountability, Commitment and Excellence Officer. Also, positions reporting to the Chief Business Officer have been corrected.
This change will not affect equipment, procedures, tests or experiments. Accidents as described in the USAR are not affected or created. Since the change is to personnel, it does not affect malfunctions of equipment or acceptance limits for the Technical -
Specifications.
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Attachment to ET 97-0017 Page 136 of 209 Bafety Evaluation: 59 96-0109 Revision 0 Main Feedwater Temperature Reduction This Unreviewed Safety Question Determination (USQD) evaluates a reduced main feedwater temperature in an effort to reach the Wolf Creek Generating Station's (WCGS) full rated thermal power of 3565 Mwt. In an effort to determine the feasibility of the feedwater temperature reduction, the safety Analysis Group has performed an evaluation demonstrating the acceptability of the change with respect to plant safety analyses. This change will allow the high pressure feedwater heaters to be partially bypassed which will result in a decreased main feedwater temperature. The main feedwater temperature will be allowed to decrease until the plant achieves 100% rated thermal power (RTP, 3565 Mwt) or until the feedwater reaches a lower bound of 420*F.
The Loss of Coolant Accident (LOCA) and non-LOCA safety analyses, including steamline break mass and energy release calculations confirm the acceptability of plant operation with the main feedwater temperature reduced to a lower bound of 420'F. The WCGS LOCA Updated Safety Analysis Report (USAR) analyses have been evaluated for conditions which bound the reduction in feedwater temperature. The justification for non-LOCA transients is based on evaluation of the current licensing basis. It is confirmed that all analyses continue to meet their respective acceptance criteria, demonstrating the acceptability of the reduction in main feedwater temperature.
The reduction in feedwater temperature has been evaluated with respect to plant safety analyses. The evaluation demonstrates that the condition is bounded by the Wolf Creek licensing basis. The reduced feedwater temperature has also been considered with respect to plant equipment. Based on this evaluation, the test will not result in the degradation of any plant equipment. Therefore, the reduction in feedwater temperature will not increase the probability of occurrence of an accident previously evaluated in the USAR.
Based on the fact that the evaluation of the licensing basis demonstrates that the test condition is bounded, the reduction in feedwater temperature will not increase the radiological consequences of an accident previously evaluated in the USAR.
Plant equipment important to safety will continue to be operated within the deeign basis and the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR will not be increased. .
Based on the fact that the evaluation demonstrates that the probability of a malfunction of equipment important to safety does not l
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i Attachment to ET 97-0017 Page 137 of 209 4 increase and that the analyzed primary plant operating parameters are not impacted, the test condition will not increase the radiological
- consequences of a malfunction of equipment important to safety
. previously evaluated in the USAR.
The feedwater system is designed such that it may be operated with the feedwater heaters either partially or fully bypassed. Therefort, i
based on the fact that the condition is bounded by the licensing basis, and the fact that the feedwater system is not_being operated outside its design, the condition will not create the possibility of an accident of a different type than any previously evaluated in the 2 USAR.
1 The feedwater temperature reduction will not result in a plant condition outside the designed operating conditions of plant equipment important to safety. Therefore, this change will not create the
] possibility of a different type of malfuncLion of equipment important to safety than cry previously evaluated in the USAR.
The current licensing basis bound the reduction in feedwater temperature. Therefore, plant safety analyses are demonstrated to continue to meet the acceptance criteria set forth in NUREG 0800 for l the applicable ANS Condition events assuming a feedwater temperature j of 420*F. Therefore, the feedwater temperature reduction does not reduce the margin of safety.
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I Safety Evaluation: 59 96-0110 Revision: 0 l 1
} Corporate Quality Manual Deletion I This revision to the Updated Safety Analysis Report (USAR) deletes 3 reference to the Corporate Quality Manual (COM). The COM was a sub-
, tier document to USAR Chapter 17.2, the operating quality program, and '
s basically duplicated the requirements therein. Therefore, the COM was deleted.
Since the requirements identified in the COM are identified in chapter
. 17.2 of the USAR and in Wolf Creek Generating Station procedures, the i
, deletion of the COM does not impact accidents or malfunctions i evaluated as the licensing basis. Nor is the potential for the creation of a new type of unanalyzed event created by this revision.
l Deletion of the CQM does not reduce the margin of safety as defined in ,
the basis for any Technical Specification. I 9
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Attachment to ET 97-0017 Page 139 of 209 Safety Evaluation: 59 96-0112 Revision 0 l Replace Fiber Glass Insulation With Armaflex Insulation This modification evaluates the use of an alternate insulation material, Armstrong Armaflex, to prevent piping corrosion from vapor condensation. Chilled water lines, to and from Component Cooling Water Pump Room Coolers, SGLO7, SGL11A, SOL 11B, SGL17A, and SGL17B are i presently installed with fiberglass insulation. The fiberglass l insulating blanket is covered with a vapor barrier and an aluminum jacketing. The vapor barrier is easily damaged and allows condensation to form which accelerates pipe corrosion. I 1
l The insulation's weight difference as compared to the total weight of the pipe and water is negligible and will have insignificant impact on the seismic analysis. The increase in fire load, while not l negligible, is clearly insignificant compared to the capability of required plant fire barriers. Therefore, the use of Armaflex insulation on the specified piping will have no impact en accidents l and malfunctions evaluated as the licensing basis. There is no i potential for the creation of a new type of unanalyzed event. Margin of safety is not impacted.
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Attachment to ET 97-0017 Page 140 of 209 Se.faty Evaluation: 59 96-0113 Revision 0 i
epening of Wolf Creek Lake l This change to the Updated Safety Analysis Report (USAR) will allow access to the Wolf Creek Cooling Lake for fishing by the public. The lake areas near the Ultimate Heat Sink, the Essential Service Water pump house,-the circulating Water screen house and the circulating Water discharge structure will be restricted from public use.
The USAR changes:
. Revise statements to recognize public use of the lake, and state restrictions
. Revise statements to include public fishing as an activity, unrelated to plant operation, which may occur within the exclusion area Rename Wolf Creek Cooling Lake, Wolf Creek Lake throughout the USAR The initiators of design bases accidents, as currently described in the USAR, do not include activities related to fishing, therefore previously evaluated accidents are not affected.
Activity on the Wolf Creek Lake will not create new failure modes of )
plant equipment or affect the operation of plant equipment important l to safety. l l
Activity on the Wolf Creek Lake will not create any new or different type of accidents.
There is no reference to lake activities in the bases for any Technical Specification. Lake activities cannot influence the margin of safety simply because of the remote location of boating relative to plant systems.
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- Safety Evaluation: 59 96-0115 Revision 0 Revision to Procedure Process to Delete Program Descriptions !
This revision to Procedure AP 15C-001, " Procedure Writer's Guide,"
affects references to Program Descriptions in Updated Safety Analysis Report (USAR) Chapter 17.2. It has been administratively decided to delete Program Descriptions and a reference to AI 15C-003, " Action Verbs, Abbreviations, and Acronyms," from the Wolf Creek Nuclear Operating Corporation procedure program.
Deletion of Program Descriptions has no impact or accidents and malfunctions of equipment important to safety. Nor is there any potential for the creation of a new type of unanalyzed event. The margin of safety as defined in the basis for Technical Specifications is not affected.
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Attachment to ET 97-0017 Page 142 of 209 Safety Evaluation: 59 96-0116 Revision 0 Minimum Roost Temperature Evaluation This revision to Updated Safety Analysis Report (USAR) Table 3.11(B) -
1 " Plant Environmental Normal Conditions" indicates that the Minimum Normal Operating Temperatures are nominal design values and not operability or equipment qualification limits. They do not represent the minimum temperature that may occur during abnormal operating conditions. Two rooms in the Auxiliary Building were 7bserved to be unusually cool, approximately 52*F, during the extreme cold spell of January 1996. This temperature was lower than the stated minimum normal value(s) given in the USAR. Plant areas were identified that may at times be as low as 45'F without threatening equipment operability. Although these rooms can be cooler than 60*F, it is normally desirable to maintain them at or above the originally identified 60*F minimum. This is the desired minimum room temperature in most areas for personnel comfort. The design basis documents will continue to ludicate a normal minimum design temperature of 60*F.
There are no accidents in the USAR or nelfunctions of equipment important to safety that are initiated by low room temperature in the plant room areas. Thus, the probability of accidents or malfunctions previously analyzed in the USAR could not be increased.
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A lower initial operating temperature for the affected rooms will not
-affect radiological consequences. No analyzed accidents create consequences of low room temperature in the affected rooms.
Any malfunctioning equipment postulated resulting from temperatures of 45'F in the affected rooms would not increase the radiological consequences of the malfunctioning equipment.
A review of equipment, important to safety, by Engineering was performed to assure that no accident or failure could be postulated from lower room temperature.
There are no Technical Specification bases that deal with a low room temperature limit for the affected rooms.
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Attachment to ET 97-0017 Page 143 of 209 Safety Rvaluation: 59 96-0117 Revision 0 i l
Installation of Essential Service water system Temperature Instrumentation The scope of this modification is to install non-safety related Resistance Temperature Devices (RTDs) at different locations at the Essential Service Water (ESW) pump house piping and lake. It has been determined that additional instrumentation is required to detect the unique conditions that promote the formation of Frazil Ice at the ESW pump house. Temperatures will be monitored by the Nuclear Plant Information System (NPIS) computer at the control room.
Installation of RTDs to measure the ESW temperatures do not increase the likelihood of accidents previously evaluated in the Updated Safety Analysis Report (USAR). The RTDs are for operator information only.
Since the design change does not result in interaction with safety related equipment, the probability of occurrence of a malfunction of equipment previously evaluated in the USAR is not increased.
This change does not affect the Engineered Safety Features which are safety related systems and components designed to mitigate the consequences of a design basis accident. This change does not affect the design basis of the ESW system. Therefore, there is no increase of. radiological consequences of an accident or malfunction previously evaluated in the USAR.
The design, material, installation and the operation of temperature indications does not create the possibility of an accident of a different type than any previously evaluated in the USAR. Nor will the change create the possibility of a malfunction of equipment important to safety, since no safety related equipment is being affected.
The ESW temperature instrumentation is a design enhancement and does not change the design bases of the ESW system. Therefore, the margin of safety is not affected.
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l Safety Evaluation: 59 96-0118 Revision O Relocation of Discharge Pressure Gauges for Circulating Water Pumps This modification provides for the relocation of the discharge l pressure gauge for each circulating water pump from the main pump to the front enclosure near the discharge valve to allow operators to read the discharge pressure gauge as they manually manipulate the discharge valve hand wheel. Also, heat tracing for the 1/2" pipe line that supplies each discharge pressure gauge will be increased in ;
wattage and surface area to prevent these lines from freezing. The I heat tracing will also be increased for the line that supplies the pressure transmitter for the control room indication.
The heat tracing system is non-safety related and is not analyzed for any design basis accidents. The relocation of discharge pressure indication gauges will not affect the main circulating water pumps or i motors for the circulating water system. Therefore, this revision I will not increase the probability of occurrence of an accident previously evaluated in the USAR.
The only accident analyzed for the circulating water system is a pump trip which in enveloped by the turbine trip analysis. The relocation of the local pressure indication gauges will not affect the circulating water pump or motor. Therefore, this revision will not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR.
Radiological consequences of an accident or malfunction of equipment important to safety will not be increased by this revision.
The heat tracing syster is non-safety related and independent of all other plant systems. The relocation of a pressure indicating gauge after the root valve will not affect the main pump or motors for the Circulating Water system. Therefore, there is no potential for the creation of a new type of unanalyzed event.
The heat tracing and circulating water systems are not associated with any Technical Specifications.
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l Safety Evaluation: 59 96-0120 Revision 0 l Conduit Re-route for Thermo-Lag Resolution This modification revises the Updated Safety Analysis Report (USAR)
Appendix 9.5B, Section C.1.7.2 to remove information about the 1-hour rated fire barrier for the conduit associated with valve EFHV38, which is an Ultimate Heat Sink isolation valve for Train B. The conduit is being re-routed to achieve more than 20 feet of spatial separation between the conduit and its redundant counterpart as required by 10 CFR 50 Appendix R. Moving the conduit eliminates the need for I wrapping the conduit with a 1-hour rated fire barrier. The conduit is l currently wrapped with Thermo-Lag fire barrier. This conduit re-route l is part of the Thermo-Lag Resolution project. Additional revisions I include a statement about room 3104 containing separation group 4 l cables only and USAR Tables 9.5B-3 and 9.5B-4 showing that room 3104 l does contain safe shutdown cables. I The overall function of the valve (EFHV38) remains the same and is not affected by the conduit re-route. As a result, there is no change in the overall performance of the affected equipment. Therefore, this change will not increase the probability of occurrence of an accident previously evaluated in the USAR.
No previously evaluated accidents identified in USAR Chapters 2, 3 or 15 are affected by this change, so there are no increased radiological consequences.
The safe shutdown function of valve (EFHV38) remains unchanged.
Removing the existing 1-hour rated fire barrier and re-routing the conduit has not increased the probability of occurrence of a malfunction of the valve.
The form, fit or function of the safety related valve required to mitigate a design basis accident is unaffected. Therefore, the radiological consequences due to a malfunction of the valve are not increased.
The failure modes of the conduit for either configuration are identical. The removal of the Thermo-Lag material and the new conduit route does not affect the safe shutdown scheme or failure modes for room 3101. There are no new credible failure modes that can arise from re-routing the conduit.
l The Technical Specifications were reviewed and remain unchanged.
Therefore, the_ margin of safety is not. reduced.
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Attachment to ET 97-0017 Page 146 of 209 Safety Evaluation: 59 96-0121 Revision 0 Gaseous Radweste System Updated Safety Analysis Report Clarification This change provides clarification to the BASES of Updated Safety l Analysis Report (USAR) Chapter 16.11.3.1.2. It allows swapping between gas storage tanks, in the Gaseous Radwaste System (HA), )
without requiring sampling of the tanks, if the tank being swapped to was previously demonstrated to have a curie content below the required USAR limit. l This change is a clarification to the USAR bases only, with no change allowed in volumes or processes with a potential for release.
Therefore, this change will have no impact on accidents or i malfunctions evaluated as the licensing basis and there is no i potential for the creation of a new type of unanalyzed event. There l is no reduction in the margin of safety.
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Safety Evaluation: 59 96-0122 Revision 0 High Pressure Feedwater Heater Bypas.s Throttling Operations i This Unreviewed Safety Question Determination (USQD) evaluates the basis for implementation of the high pressure feedwater heater bypass, in order to achieve 100% reactor pcwer, through a decrease in j feedwater temperature through the et.d of Cycle 9. Procedure SYS AE- l 125, "HP FW Heater Bypass Throttling Operations," has been developed to facilitate long term implementation assuming a lower bound feedwater temperature of 400*F at 100% power. The reduction in feedwater temperature has been evaluated with respect to the plant's safety analyses. The evaluation demor.strates that the condition is bounded by licensing basis analyses. The reduced feedwater temperature has also been considered with respect to plant equipment. i Based on this evaluation, the test will not result in the degradation j of any plant equipment important to safety. Therefore, the reduction l in feedwater temperature will not increase the probability of occurrence of an accident previously evaluated in the Updated Safety !
Analysis Report (USAR). Based on the fact that the evaluation of the l licensing basis demonstrates that the test condition is bounded, the reduction in feedwater temperature will not increase the radiological consequences of an accident.
The reduction in feedwater temperature will not result in the primary plant operating parameters exceeding their analyzed values and the feedwater isolation valves will continue to be operated within their designed temperature range. The feedwater system is also designed such that it may be operated with the feedwater heaters either partially or fully bypassed. Therefore, plant equipment important to safety will continue to be operated within the design basis and the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR will not be increased.
Based on the fact that the evaluation demonstrates that the probability of a malfunction of equipment important to safety does not increase, and that the analyzed primary plant operating parameters are not impacted, the test condition will not increase the radiological consequences of a malfunction of equipment important to safety previously evaluated in the USAR.
The reduction in feedwater temperature is bounded by the licensing basis Therefore, based on the fact that the condition is bounded by licensing basis analyses and the fact that the feedwater system is not being operated outside of its design, the condition will not create i the possibility of an unanalyzed event, There is no reduction in the I
margin of safety as defined in the basis for any Technical Specification.
Attachment to ET 97-0017 page 148 of 209 Safety Evaluation: 59 96-0123 Revision 0 Margin to Saturation Caution Annunciation Setpoint This modification changes the setpoint for the Margin to Saturation Caution annunciation to 5'F subcooled (Trip Breakers Closed) and.25'F subcooled (Trip Breakers Open) in order to eliminate nuisance alarms but still provide the operator adequate time to respond to the warning condition. Annunciator 56B "RCS < 50 SUBCOOL" annunciates frequently during steady-state full power operations when no abnormal conditions are present. The annunciation is caused by Train B " Thermocouple Margin to Saturation caution," which is generated by the Train B Thermocouple / Core Cooling Monitor (T/CCM).
The core cooling monitor compares both channels of core outlet thermocouple temperatures, and hot and cold leg temperatures with the saturation temperature based on the lowest of three pressure signals.
Two levels of alarms are provided for the core cooling (Test) monitor
, function, " Caution" and " Alarm". These alarms provide an early warning to plant personnel that core conditions are approaching a saturation condition. The Thermocouple (T/C) Margin to Saturation caution setpoint does not affect any Updated Safety Analysis Report (USAR) accident analysis. No credit is taken for this setpoint in the Operator response of any USAR accident. Any accident in the USAR which results in inadequate core cooling is not affected by revising the setpoint for the caution annunciation. Therefore, this change does not increase the probability of occurrence of an accident previously evaluated in the USAR.
Since the T/C Margin to Saturation caution setpoint does not affect any USAR accident analysis, no radiological consequences will be affected by revising this setpoint.
The T/C Margin to Saturation caution setpoint has no safety related functions. It is used for operator warning only and has no impact on any plant equipment. The caution indication does not trigger operation of any piece of equipment, so there is no increased chance of malfunction and no radiological consequences will be changed.
Revising the setpoint will not create a different type of accident or malfunction than those presented in the USAR.
The T/C Margin to Saturation caution alarm serves only as a caution to alarm operators of potential inadequate core cooling. Since only the value at which the caution annunciator will sound is changing, no margin of safety is affected.
i Attachment to ET 97-0017 Page 149 of 209 i
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- Safety Evaluation
- 59 96-0127 Revision 0 Addition of Warning Siren to Support Opening of Wolf Creek Lake This modification revises the Alert and Notificiation System (ANS) to provide emergency notification to Wolf Creek cooling lake users. The warning siren presently locate $ at Sharpe is being relocated directly south of the Education Center and is being renamed WC1. In addition, a new siren is being added at the north end of Saddle Dam IV (WC2).
These modifications will provide a minimum of 60dBC sound level coverage to all areas of the lake where there is expected to be transient populations.
The ANS is not connected to any Plant System, safety related or otherwise. The siren installations have no impact on accidents identified in Updated Safety Analysis Report (USAR) chapters 2, 3 and 15.
If a failure were to occur on the ANS system, no equipment important to safety would be affected. No credible accident could be created by failure of the siren equipment. There are no credible malfunctions of
, equipment important to safety which may be affected by siren installations or a failure on the ANS system.
No acceptance limits as defined in the Technical Specifications are affected.
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Attachment to ET 97-0017 Page 150 of 209 i
Safety Evaluation: 59 96-0128 Revision 0 Steam Dump C-9 Permissive Setpoint Change This revision to the Updated Safety Analysis Report (USAR) increases the C-9 Steam Dump Permissive setpoint from 4.9 in Hga to 5.0 in Hga.
USAR Section 10.4.4, Turbine Bypass System, specifically states this setpoint in the description of the system's operation. However, the General Description of the Turbine Bypass System in Section 10.4.4 states that 5.0 in Hga is an administrative limit imposed on turbine operation by the turbine-generator manufacturer. Therefore, this change will make the two descriptions consistent.
$ The Turbine Bypass System and the associated C-9 permissive serves no safety function and has no safety design basis. There is no safety related equipment in the vicinity of the Turbine Bypass System.
Operation of the Turbine Bypass System is not taken credit for in the USAR Chapter 15 Accident Analysis. There are no design basis accidents identified which involve changes to this administrative setpoint.
Since this change to an administrative setpoint is bounded by a higher administrative setpoint, no credible accidents that could be created 1 are identified. l l
Since there are no equipment important to safety associated with the Turbine Bypass System, there are no credible malfunctions of equipment important to safety identified, s There are no acceptance limits associated with this system, therefore the margin of safety is not affected by this change.
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1 Safety Evaluation: 59 96-0129 Revision 3 0 Operations Quarterly Tasks This revision to Procedure STN OQT-001, " Operations Quarterly Tasks," )
allows for cleaning the boron crystals off of Fuel Pool Components using demineralized water from the Reactor Makeup Water System (RMWS). Cleaning boron crystals off of pool level float balls and other pool components will maintain optimal mechanical movement of these devices and enhance the house keeping around the Spent Fuel Pool l (SFP), the Fuel Transfer Canal (FTC) and the Cask Loading Pool (CI.9) .
. The Fuel Handling Accident (FHA) in the Fuel Handling Building (FHB) was identified in Updated Safety Analysis Report (USAR) Section 15.7.4 and reviewed for impact due to procedure change actions. No impact I j resulted and no other accidents were identified. Therefore, the l probability of occurrence and the consequences of all previously l
- evaluated accidents remain as evaluated.
4 No credible malfunctions of equipment important to safety are j i identified. {
l Dilution of the SFP/FTC/CLP has been evaluated and the subject procedure changes will not compromise this evaluation. No other credible accidents besides dilution could be created or identified.
Therefore, the potential for a unique accident has not been created.
l The margin of safety provided by Technical Specifications 3.9.11 and 5.6 are maintained and not reduced by the procedure actions. .
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l Page 152 of 209 Safety Evaluation: 59 96-0130 Revision 0 Radiological Emergency Response Plan Revision 51 l This revision to the Radiological F,mergency Response Plan (RERP)
Emergency Action Levels, which is part of the Updated Safety Analysis Report (USAR), makes a classification based upon mode dependent equipment rather than only if equipment is affected by fire.
This revision does not impact design basis accidents or create any new credible accidents. This revision more clearly states the conditions I for which a classification would be impacted by a fire affecting Plant safety Systems. Therefore, there is no impact on accidents and !
l malfunctions evaluated as the licensing basis. There is no potential for the creation of a new type of unanalyzed event. The margin of safety is not affected by this revision to the RERP.
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Safety Evaluation: 59 96-0131 Revision 0 Organization Changes for the Reporting of System Engineering and J Integrated Plant Scheduling This change to the organization includes the realignment of the Operations Organization to exchange System Engineering and Integrated Plant Scheduling (IPS) reporting. System Engineering will now report directly to the Vice President Plant Operations. IPS will report to '
the Manager Operations.
l There are no accidents discussed in the Updated Safety Analysis Report I (USAR) that rely on organization to mitigate. A change to the l organization will not create new accidents nince all functions l
continue to be performed and the personnel *)erforming the functions J meet the ANSI qualifications. For the same reasons that no accidents i are impacted or created, no malfunctions of equipment important to -
I safety are affected. There are no acceptance limits affected by this I change since no functions have been deleted.
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Attachment to ET 97-0017
- Page 154 of209 Safety Evaluation: 59 96-0133 Revision 0 Operations Changes to Updated Safety Analysis Report This revision to the Updated Safety Analysis Report (USAR) Section 13.1.3.2, Qualification of Plant Personnel, adds resumes for new.
individuals holding the position of Shift Supervisor, moves a current Shift Supervisor to the position of Supervisor Operations Training and removes the resume of the current Supervisor Operations Training.
There are no effects due to this activity. The revisions are administrative in nature. The candidates for the promotions are fully qualified and meet the minimum qualifications for the position they are being promoted to as required by the USAR.
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Attachment to ET 97-0017 Page 155 of 209 Safety Evaluation: 59 96-0134 Revision 0 Thermal Performance instrumentation Identification This modification approves the as-built conditions for test instrumentation installed on non-nuclear safety related portions of the Main Turbine (AC), Feedwater (AE), Feedwater Heater Extractions, Vents and Drains (AF) and Circulating Water (DA) systems. These instruments will be shown on the appropriate P& ids, all of which are reflected in the Updated Safety Analysis Report (USAR).
The instrumentation added by this modification all have a passive non-nuclear safety related function. No design bases accidents have been identified which would be impacted by these additional instruments.
No credible accidents that could be created are identified. There are no credible malfunctions of equipment important to safety identified.
Since there are no safety related acceptance limits associated with the non-nuclear safety related portions of the affected systems, the margin of safety is not affected.
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Attachment to ET 97-0017 Page 156 of209 safety Evaluation: 59 96-0137 Revision 0 Security Plan Revision 28 Revisions to the Wolf Creek Physical Security Plan (PSP) are as follows:
- 1) Administrative changes to' include revising the designation for certain buildings within the protected area barrier; title change from Manager Plant Support to Director Site Support; and delete location of switchboard in the Security building. These changes are administrative in nature and do not alter any plan commitments.
- 2) The Physical Exercise Performance Test (PEPT) was changed in the
- last revision to the Training and Qualification Plan. The change was
, found unacceptable to the NRC due to the fact that personnel were no i
longer required to wear their duty equipment. This change adds a 4 -
i 5 lb. weight requirement for the PEPT 220 yard run to simulate normal duty equipment. This change enhances the PEPT.
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- 3) Changes to the Access Authorization program commitments were made j in the last Physical Security Plan. The NRC found certain aspects of the change unacceptable. The phrase "as approved by the NRC" was a
removed from the text. This change enhances and clarifies the Plan 1 text.
- 4) This change clarifies that the metal detector search exemption applies to only Armed security officers on official duty. The text was revised to correct a typographical error in the previous revision. This change clarifies that only armed security personnel on official business are exempted from the metal detector requirements.
In the original submittal, the word = armed" was inadvertently omitted., This change is in accordance with recent NRC guidance.
- 5) In the previous revision to the PSP, a new search exemption c (category VI) was submitted. This exemption was referred to NRR for their review. This change is being withdrawn from the PSP until completion of that review. This change removes a search exempt category that is not currently approved as a 50.54(p). This change ensures that the plan accurately reflects only approved commitments.
- 6) This change requires Camera Callup (CCTV) observation by a member of the security force whenever Gate G-10 is open and unsearched material remains in the cargo search area. This change clarifies and enhances the PSP requirements for Gate G-10.
- 7) This change incorporates recent Organizational changes that include Title change from Vice President Plant Operations to Chief Operating Officer; Addition of a position titled Plant Manager; Elimination of
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Attachment to ET 97-0017 Page 157 of 209 text that is duplicated in E-plan commitments; and Changes to the Call Superintendent position. This change delineates the current management organization at Wolf Creek.
- 8) The intrusion detection hardware for door 33031 has been removed.
This door is located on the Northwest corner of the Communications i Corridor. This door is not a Vital or Protected area boundary. The continual maintenance and testing of hardware which is not required or utilized is an ineffective and inefficient use of resources. Door 33031 was previously identified as a " passive" door in which the hardware was available but kept off-line and not monitored. The removal of the hardware does not reduce the effectiveness of the PSP as it does not alter the Vital or Protected Area boundary or alarm i response requirements.
l These revisions 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 l the margin of safety. )
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Attachment to ET 97-0017 Page 158 of 209 l
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Safety Evaluation: 59 96-0139 Revision 0 Moisture separator / Reheater 2nd Stage Main Steam Bypass valve ABRV-56 3
& 57 Enhancement This modification removes the motor operators on valves AB HV 56 and 57, Moisture Separator / Reheater (MSR) 2nd Stage Main Steam Supply Bypass Valves, and replaces them with manual handwheels.
There is no change in plant line up or operation except the valves will be operated locally instead of from the Control Room. Main Steam piping pressure boundary will not be affected so the probability of occurrence of an accident previously evaluated in the Updated Safety
- Analysis Report (USAR) has not increased.
] All of the accidents in USAR section 15.1 have been analyzed. The most severe radiological consequences will result from the main steam line break accident. This modification is bound by the accident analyses of the main steam break in section 15.1 therefore, the radiological consequences of an accident are not affected.
AB HV 56 and 57 serve no safety function and have no safety design basis. There is no safety related equipment in the vicinity of the valves. Since no malfunctions are identified, the probability of occurrence and the radiological consequences of a malfunction of equipment important to safety is not affected by this modification.
There is no potential for the creation of a new type.of unanalyzed event since no credible accidents or malfunctions are identified.
Steam to the MSRs are not in the Technical Specifications or any of the bases. No acceptance limits are identified that could be affected.
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Attachment to ET 97-0017 Page 159 of 209 Safety Evaluation: 59 96-0140 Revision 0 Addition of Blow off valve to Y Strainer on Air Compressor CKA01A & B This modification adds blow off valves, KAV1518 and KAV1519, to Updated safety Analysis Report (USAR) Figure 9.3-1-01. Air compressors CKA01A and CKA01B have a Y strainer in their cooling water supply lines to prevent debris from entering the air compressor and potentially blocking cooling passages. These blowoff valves are added to each strainer cover. Normally, the blow off valves will remain
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closed and capped. The presence of these valves will cause no change in operation or function of the Y strainers or the cooling water system or the air compressors.
i The cooling water strainers are minor components in the compressed air system, and their function is not specifically described in the USAR section 9.3.1. The air compressors themselves are powered from a 1E
- source, and they are load shed on receipt of a safety Injection (SI) signal.
The initiators of design basis accidents are not influenced by the presence or absence of blowoff valves on the cooling water strainers
- to the air compressors. Previously evaluated accidents are not affected.
The blow off valves are passive components, normally closed, and they do not influence the radiological consequences of an accident or malfunction of equipment important to safety.
The blow off valves do not change the function of the Essential Water System (ESW) or the air compressors, and therefore they have no
, influence on any components whose function is to mitigate radiological i events. Use of the blow off valves will ensure design cooling water flow is delivered to the air compressors.
The blow off valves are located in a non-safety related branch of the ESW system. They are passive valves which when opened may cause the ESW to air compressor isolation valves to close on high differential
- pressure, but this is by design to prevent loss of Ultimate Heat Sink (UHS) inventory. Therefore, no accident of a different type than previously evaluated is created.
4 The blow off valves are manual valves, and they cannot influence the function or malfunction of any equipment important to safety.
The margin of safety is not affected by the addition of blow off valves to the Y strainers.
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Page 160 of209 l
l Safety Rvaluation: 59 96-0141 Revision 0 -
Local control of steam Generator Atmospheric Relief Valves '
This modification provides for the addition of a local' pneumatic controller assembly and tubing in the existing pneumatic controls for the Steam Generator Atmospheric Relief Valves ABPV0002 and ABPV0003 to provide local operation at the valves. Emergency lighting unit A-79 will also be modified to accommodate two additional light fixtures for the new controllers. The effect of this modification is to enhance the existing design to allow complete local manual control of the valves. 1 I
The new controllers provide enhanced local operation of the I atmospheric relief valves (ARVs) after a design basis fire and a passive role (flowpath) for design basis accident mitigation.
Therefore, the addition of the new controllers does not increase the frequence of occurrence of an accident previously evaluated in the Updated Safety Analysis Report (USAR).
The new controllers and tubing connect to existing actuator control air and do not create any new release pathway nor affect other systems providing accident mitigation. Therefore, this modification does not- i increase dose rates assumed as the result of any accident or I malfunction of equipment.
-The new controllers and tubing serve a passive control air flow path I function and do not interact with any other equipment important to safety. Therefore, this modification does not increase the frequency j of malfunction of equipment important to safety. '
The controllers and tubing serve only a passive function with regard i to the Atmospheric Relief Valves except in the event of a fire causing their use and therefore, do not create the possibility of an accident than any previously analyzed in the USAR.
The function associated with the existing valve pneumatic controls during accident conditions have not been altered due to the addition of'the new local manual controllers and tubing. Therefore, the possibility of a different type of malfunction of equipment important to safety is not created.
Since the addition of the controllers does not affect the operability of the ARVs or their safety related function, acceptance limits as defined in the bases of any Technical Specification are not affected.
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I Attachment to ET 97-0017 Page 161 of 209 Safety Evaluation: 59 96-0142 Revision 0 P&ID Drawing Deficiencies This modification corrects P&ID drawing deficiencies. The existing P&ID does not show some valve designations like fail open or fail
- close. This is an administrative change only, not a technical change. The following valves from the Feedwater Heater Extraction Drains & Vents system need a fail close (F.C.) or a fail open (F.O.)
designation:
AF LV-0105 and AF LV-0135 need F.C. designation. (M-12AF02)
AF FV-0072B and AF FV-0073B need F.O. designation. (M-12 AF01)
AF LV-0053 is listed incorrectly as an F.C. valve. It should be F.O.
(M-12AF01)
The following valves from the Main Turbine system need a F.C.
designation: AC PV-0186A, AC PV-0186B, AC PV-0186C and AC PV-0186D. (M-12ACO2)
This is an administrative change to reflect the as-built configuration of the valve and actuator. Therefore, there will be 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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1 Attachment to ET 97-0017 l Page 162 of 209 Safety Evaluation: 59 96-0143 Revision:0 Auxiliary Feedwater in the Function of Residual Heat Removal This change revises Updated Safety Analysis Report (USAR) Section 3.1.3 to reflect the Auxiliary Feedwater in the function of residual l
heat removal and provide reference to the Auxiliary Feedwater System description section.
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Additional information is being provided by this change. The description of its residual heat removal function removes the implication that the function of the steam generators, atmospheric relief valves and Auxiliary Feedwater is a redundant function, when in fact, for higher temperatures, these systems are the only safety 4 related systems intended to perform this function. Therefore, this i change does not increase the probability of occurrence of an accident j previously evaluated in the USAR or a malfunction of equipment. '
This change has no effect on the systems, operations or programs which l could affect the radiological conseqences of an accident or malfunction of equipment.
This change is a USAR clarification only. The design basis and operation of the system are not changed, thus the potential for the creation of a new type of unanalyzed event is not created.
No limits are exceeded and the margin of safety is not reduced.
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Attachment to ET 97-0017 Page 163 of 209 i
Safety Evaluation: 59 96-0144 Revision 0 Radiological Emergency Response Plan Revision 52 This revision of the Radiological Emergency Response Plan (RERP) incorporates changes necessary to allow control of access to the Wolf Creek Lake, which is being opened to the public. Opening Wolf Creek Lake for recreational fishing resulted in Wolf Creek Lake becoming a new subzone with an evacuation time equal to that of John Redmond Reservoir. To accommodate notification of people on the lake, the siren previously located at Sharp was moved to the north end of the j lake and another siren was added to the south east corner of the lake, l for a total of eleven sirens. j Additional changes were made to reflect current county population data, an improved direction for evacuation routes and provision for the Duty Emergency Director to assume responsibility without Technical Support Center activation, and other editorial corrections. l This revision of the RERP does not impact design basis accidents nor are any new credible accidents created. Credible malfunctions of l l
equipment important to safety are not created. Acceptance limits of Technical Specifications are not affected.
Attachment to ET 97-0017 Page 164 of 209 Safety Evaluation: 59 96-0145 Revision 0 Modification to Security Door This modification will make two figure changes in the Physical Security Plan (PSP) and a change to the alarm point listing in the PSP. Door 33031 is an exterior hollow metal double door that provides access between the Communication Corridor 2000' level and the west yard. It has no fire rating nor is it currently being used as a security door. This door, however, still contains certain security hardware as it was once available to be used to control access. The changes deal with alarm points, and the types of alarms on the door.
The door will be removed from the alarm point list and certain security devices will be removed.
Door 33031 is not a protected area door or a vital door, therefore j
- there are no regulatory requirements which require this door to be
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alarmed. This modification will not decrease the overall level of the security system's performance, and will not reduce the effectiveness '
, of the PSP.
There is no impact on design basis accidents identified in Updated j Safety Analysis Report (USAR) Chapters 2, 3 or 15 associated with this modification. The location of the door closures, and the removal of the security devices will have no impact on the identified accidents.
If a hardware failure would occur on the door, no equipment important to the safety of the plant would be affected. No credible accident or i malfunction of equipment could be created by the failure of the door I hardware.
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This modification will not reduce the margin of safety as defined in j the basis for any Technical Specification.
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Attachment to ET 97-0017 Page 165 of 209 Safety Evaluationt 59 96-0147 Revision 0 Changes to Existing Site Structures The revision'to the Plant Site Section 1.2.2 and Figure 1.2-44 of the Updated Safety Analysis Report (USAR) are administrative in nature and have no effect on the physical plant site. This revision, 1) adds an existing building in the cable reel yard area, 2) adds the existing Cathodic Protection Building south of the plant, 3) adds the existing X-Ray Building west of the NDE/ Civil Test Center, 4) adds the existing lean-to buildings to the NDE/ Civil Test Center and the Vehicle ,
Maintenance Shop, 5) deletes the Drum Storage Building, and 6) changes f the name of the Warehouse to the Materials Center. l These changes will have no impact on accidents or malfunctions evaluated as the licensing basis and there is no potential for the creation of a new type of unanalyzed event. There is no reduction in the margin of safety.
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Attachment to ET 97-0017
- page 166 of 209 Safety Evaluation
- 59 96-0148 Revision 0
, 1 Revises Scaffolding Procedure to Delete 10 Day Seismic Qualification This change to the Updated Safety Analysis Report (USAR) reinstates changes created by USAR Change Request ('USARCR)94-047 which provided reduced seismic accelerations for evaluation of plant temporary conditions such as temporary rigging, lead / radiation shielding, scaffolding, freeze plugging and temporary alteration of supports or .
boundary conditions. USARCR 94-047 was reported previously by USQD 94-0121 and provided an alternative to the full. seismic evaluation that was being performed for plant temporary conditions.
This_ change reinstates the original design basis and the requirement l for an engineering evaluation of each scaffolding constructed in safety-related structures by deleting the reduced seismic accelerations methodology. This change will have no impact on I
accidents and malfunctions evaluated as the licensing basis. There is ;
no potential for the creation of a new type of unanalyzed event. No !
margin of safety defined in the Technical Specifications are affected.
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Attachment to ET 97-0017 page 167 of 209 Safety Evaluation: 59 96-0149 Revision 0 Installation of Temporary Piping Supports This temporary modification installs temporary supports to support piping while permanent seismic supports are temporarily disconnected.
The bend of the riser downstream of valve isolation BG-8388 is fabricated with an angle of 87 to 88 degrees instead of 90 degrees.
'When this pipe was rotated for maintenance work the riser became skewed. This skewed riser also offset the BG-8388 isolation valve which is installed on the horizontal run pipe. valve BG-8388 is used for isolation for maintenance work. Downstream of the valve, the system is in service. Two seismic supports, BG01-C027 and BG01-C028, are installed on two sides of the isolation valve BG-8388. To eliminate offset of the riser and isolation valve, the above seismic supports are to be temporarily disconnected from the pipe. Temporary supports will be installed and are capable of supporting the design loads. After correcting the piping layout, the seismic supports will be reinstalled and the temporary supports will be removed. There will be no adverse effect on the piping in service.
Since the seismic supports are replaced by temporary supports which will be capable of supporting the design loads, there is no potential impact on design basis accidents discussed or referenced in U pdated Safety Analysis Report (USAR) chapter 2, 3, 6, 9 or 15.
There are no credible accidents or malfunctions of equipment that this temporary modification could create.
There are no acceptance limits contained in the bases for Technical Specifications that could be affected.
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Attachment to ET 97-0017 Page 168 of 209 safety Evaluation: 59 96-0150 Revision 0 supplier / Material Quality and Material Engineering Reorganization Updated Safety Analysis Report (USAR) Chapter 13.1 and 17.2 are being revised to reflect the reorganization of Supplier / Material Quality and Material Engineering into the Purchasing and Material Services organization. This change is administrative in nature, as it is only a realignment of responsibilities, with all commitments and requirements previously existing still being satisfied.
f There are no design basis accidents identified because these changes
- do not change any information identified or discussed in these chapters of the USAR. No credible accidents or malfunctions of equipment will result. No acceptance criteria is affected.
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Attachment to ET 97-0017 Page 169 of 209 Safety Evaluation:
59 96-0152 Revision 0 Normal Charging Pump Pre-Operation Flush This temporary modification allows suction isolation valve BG-8394 be cracked open to allow water, to suction header, from the in service charging pump the suction of the new Normal Charging Pump (PEG 04)to be used for flushing The flush water will enter the empty but sealed pump casing and be routed to a floor drain through one of the pump casing drains. Fiber optics will then be used to verify suction pipe cleanliness using an unfinished instrument flushing. line tie-in point which will be temporarily sealed during Use of BG-8394 for flushing essentially opens the in service charging pump suction header supplied by the Volume Control Tank (VCT) to an unfinished Control) and out of service section of the BG (Chemical and Volume system.
The water source for the rest of the flush uses BL water from HBV 353 (reactor makeup water transfer pump to demineralizer degasifie r drain) by temporary connection.
Normal Charging Pump (NCP) This water source is the closest to the pressure water suitable for flushing. room and supplies high quality, high The use of BG-8394 as a flush source and the use of HBV-353 as a source of reactor makeup water for flushing does not affect the initiating events of accidents and hazards discussed in Updated Sa f ety Analysis Report (USAR) Chapters 2, 3, 6, 9 or 15.
BG-8394 is fully under operator flushing control and can be closed immediately to terminate if required.
HBV-353 supplies water from a non-safety related water source to an out of service portion of the BG system and cannot cannot affect accident initiation. Therefore, this modification increase the probability of occurrence of an accident previously evaluated in the USAR.
The radiological consequences of using BG system water to flush into a closed but unfinished system is easily bounded by the limiting release described in USAR Section 15.6.2.
The Chemical and Volume Control system single failure analyses are not affected the unfinished by the use of BG-8394 and HBV-353 as flush water sources into NCP.
All failure analyses remain valid since the flush is under operator control and flush hoses are not allowed to pass over or be attached to equipment important to safety. Therefore, the probability safety is not of.increased.
occurrence of a malfunction of equipment important to Sir.ce the malfunctions described in USAR Tables 6.3-6, , 6.3-5 and 9.3-
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) Attachment to ET 97-0017 Page 170 of 209 10 are not affected, the radiological consequences are not affected.
No new types of accidents are created by this flushing operation. The
, . operation of the NCP suction valve BG-8394 does cause an indicated leakrate of approximately 10 gpm. However, this is easily within the makeup capability of the reactor makeup system to the VCT and is fully under control of personnel at the valve.
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, No new types of potential failures have been created.
Isolation valve
- BG-8394 is fully operable and is able to isolate charging pump suction header pressure from the NCP. Likewise the use of non-safety related valve HBV-353 and associated flush hoses does not create any new types of failures as long as.the flush hoses are not allowed to interfere
, with the function of any equipment important to safety.
. The controlled flush is not considered leakage as defined by Technical 3 Specification Definition 1.15 and does not violate the limits of i Technical Specification 3.4.6.2b or d since the usage of VCT volume is not considered RCS leakage as defined in USAR Section 5.2.5.2. For i
this reason, the margin of safety assumed by the use of these limits is not reduced.
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Attachment to ET 97-0017 Page 171 of 209 Safety Evaluation: 59 96-0153 Revision 0 Service Water Strainer Isolation Valve Improvement This revision to the Updated Safety Analysis Report, Plant Site Section 1.2.2 and Figure 1.2-44 are administrative in nature and have no effect on the physical plant site. This revision, 1) adds an existing building in the cable reel yard area, 2) adds the existing Cathodic Protection Building south of the plant, 3) adds the existing X-Ray Building west of the NDE/ Civil Test Center, 4) adds the existing lean-to buildings to the NDE/ Civil Test Center and the Vehicle Maintenance Shop, 5) deletes the Drum Storage Building, and 6) changes the name of the Warehouse to the Materials Center.
These changes will have no impact on accidents or malfunctions evaluated as the licensing basis and there is no potential for the creation of a new type of unanalyzed event. There is no reduction in the margin of safety.
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Attachment to ET 97-0017 Page 172 of 209 1
I Safety Evaluation: 59 96-0154 Revision 0 i
Chemical and volume control Systesa Drawing Corrections This change removes notes from the BG (Chemical and Volume Control) system P& ids which pertain to locked valve positions. The change is only being made to avoid document discrepancies between plant operating procedures and the P& ids. Operating procedures are in place i
, to align the plant and to administratively control the locked I positions of components during all modes of operation. These P&ID changes do not change any work activities, nor do they affect the operation of any system, structure or component as described in the Updated Safety Analysis Report that are important to safe and reliable i
operation.
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! There are no design basis accidents identified because these document 3 l
4 changes do not affect any procedures or administrative controls that govern work activities. No credible accidents or malfunctions of equipment important to safety are identified. No acceptance limits are identified that could be affected by this change.
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Attachment to ET 97-0017 Page 173 of 209 Safety Evaluation: 59 96-0155 Revision 0 Clarification of Corrective Action Program This revision to the Updated Safety Analysis Report (USAR) provides j additional information to clarify that Wolf Creek Nuclear Operating i Corporation's (WCNOC) corrective action program meets the requirements l of Regulatory Guide 1.144, " Auditing of Quality Assurance Programs for 4
Nuclear Power Plants". Under the Wolf Creek corrective action program, Performance Improvement Requests (PIR) receive a scheduled completion date prior to the responsible group receiving it. However, I the internal Quality Evaluation (QE) initiated PIR has a different I scheduled completion date than a non-QE generated PIR. QE initiated PIRs, because of ANSI N45.2.12, " Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants" wording, have a 30 day completion date, while other PIRs have a different completion date assigned per management's expectations. Due to an interpretation of the ANSI requirement, in 'ractice this date was and could be changed to address a realistic completion date.
I USAR, Appendix 3A "Conformance to NRC Regulatory Guides" is revised to j clarify WCNOC's commitment to Regulatory Guide 1.144. This will allow i for the scheduled date to be assigned prior to the responsible organization receiving the PIR. This change has been determined to be I only a clarification and is not a reduction in the Quality Program or I a change in commitments to the NRC. l This change is issued to provide additional information that Wolf Creek's corrective action program meets Regulatory Guide 1.144 for .
findings identified during the audit process and does not affect l accidents, malfunctions, unanalyzed events or margins 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 l l the margin of safety. l
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Safety Evaluation: 59 96-0156 Revision 0 Meteorlogical Tower Recorder Replacement This modification revises Updated Safety Analysis Report (USAR) section 2.3.3.6.2 to delete a reference to Leads & Northrup chart recorder for monitoring temperature. This is a recorder replacement
- and is intended to allow use of the more reliable Yokogawa recorder.
1 The modification does not affect the channels monitored, and all recorders allowed, analog or digital, continue to provide backup to the NPIS (Nuclear Plant Information System) information.
The design basis accidents identified in Chapter 2 of the USAR have been reviewed. The replacement of referenced recorders will have no impact on identified accidents.
The recorder cabinets or any device connected to the recorders, such
- as the analog inputs, is not considered as equipment important to safety. There are no credible malfunctions of equipment important to safety which may be affected by the replacement of the recorders or the relocation of the analog inputs.
The replacement of these recorders and the relocation of the analog inputs cannot cause any effect on any acceptance limits. 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 4 safety.
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1 Safety Evaluation: 59 96-0157 Revision 0 i
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Steam Generator Slowdown Flash Tank Liquid Relief Valve Replacement l-
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.This modification replaces an existing steam relief valve, BMV0319 l (Steam Generator Blowdown Flashtank Liquid Relief Valve) , . with a.
l liquid relief valve. Presently, BMV0319 valve is a steam relief valve
- installed on.the liquid relief line. A steam relief valve is unstable
- for liquid applications. Inertia of liquid can produce a chattering
- action that can damage the valve disc and the spindle. The chattering can cause a leakage through the valve and a continuous leakage can i damage the valve internals. Therefore, the existing steam relief valve is being. replaced by a liquid relief valve.
i j There are no design basis accidents discussed or referenced in Updated j Safety Analysis Report ('USAR) chapte'ts 2, 3, 6, 9 or 15 that are
- affected by this modification, nor nre there any types of credible-accidents created.
- There are no credible malfunctions of equipment important to safety
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which may be affected by this change.
There are no acceptance limits contained in the bases for the ,
Technical Specifications that could be affected. I l I 1
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Safety Evaluation 8 59 96-0158 Revision 3 0 Chief Administrative Officer Organizational Change This personnel change reflects the fact that O.L. Maynard has been re-assigned to the position of Chief Administrative Officer to replace R.
N. Johannes. This is a personnel change only and has no effect on the l organization. This change only affects the Resume for the Chief l
- Administrative Officer in the U pdated Safety Analysis Report (USAR). I It will not affect equipment, procedures nor test or experiments.
All requirements continue to be met. This change does not affect nor l create accidents as described in the USAR. Since the change is to l
. personnel only, it does not affect malfunctions of equipment or j margins of safety as defined in the basis for Technical Specifications.
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Attachment to ET 97-0017 Page 177 of 209 Safety Evaluation: 59 96-0159 Revision 3 0 Containment Isolation Phase & Roset Switch Lifted Leads This temporary modification eliminates the potential of a switch contact failure from preventing an Automatic and/or Manual initiation of the A Train Containment Isolation Signal Phase B (CISB) by lifting the wires to SBHS0052 (Containment Isolation Phase B Reset Switch) thereby simulating the normally open contacts. It was identified that the contact block for handswitch SBHS0052 was cracked and that a potential existed for the normally open contacts of the switch to remain closed if the switch was pushed. Also, it was considered credible that during a seismic event the contact block could fall apart and wires could short to one another or to ground. Either failure has the potential to result in the Reset for the A Train CISB being constantly held in. This would prevent an Automatic and/or Manual actuation of an A Train CISB from being initiated.
Lifting the wires to the A Train CISB Reset handswitch does not affect the initiating events nor create a new initiating event for accidents described in the Updated Safety Analysis Report (USAR). Therefore, :
implementation of this temporary modification will not increase the l probability of occurrence of an accident previously evaluated in the I USAR.
The CISB Reset handswitch is not utilized in mitigation of design basis accidents as described in the USAR therefore, radiological consequences of an accident previously evaluated are not increased.
There are no credible malfunctions of equipment important to safety which may be affected by this temporary modification.
There are no acceptance limits associated with resetting of the CISB.
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Attachment to ET 97-0017 Page 178 of209 Safety Evaluation: 59 96-0160 Revision 0 Temporary Change to Allow for the Injection of Hydrogen Peroxide into Secondary Liquid waste Monitor Tank This temporary procedure provides instruction to inject Hydrogen Peroxide (H202) into the Secondary Liquid Waste Monitor Tank A (SLWMT A) to remove Organics (oil) from the fluid. The H202, an acid, is injected via a temporary pump and tygon tubing through the Monitor Tank Discharge Pump Sample Vessel Purge Line (quick disconnect) located in the SJ144 sample panel.
According to the Material Safety Data Sheet database, the H202 is an approved chemical. Addition of H202 to the Secondary Liquid Waste Monitoring Tank A to treat the waste will not change the function cf the SJ (Nuclear Sampling) or HF (Secondary Liquid Waste ) system as described in the Updated Safety Analysis Report (USAR), because the Radwaste System is designed to receive Acid or Caustic as necessary to treat the waste for pH adjustments.
There is no safety related equipment within the vicinity of the temporary pump or hoses. Therefore, any water from a ruptured hose will not damage any safety related equipment. Since the system's functions are not changed, no credible accidents that could be created are identified.
The SJ144 panel and all the temporary equipment are non-safety related. Therefore, no malfunctions of equipment important to safety will be affected.
Since there are no acceptance limits identified in USAR sections 9.3.2 and 10.4.10, the margin of safety as defined in the basis of the Technical Specifications will not be affected by this temporary modification.
1 Attachment to ET 97-0017 Page 179 of 209 Safety Evaluation: 59 96-0161 Revision 0 Temporary Diesel Engine Driven Fire Punip Replacement This temporary modification provides alternate pump and piping to support fire protection needs of the plant by maintaining system design and function in accordance with the Fire Protection Manual while permanent diesel engine driven fire pump (1FP001PB) is being worked on. 1FP001PB has been found incapable of supplying the flow and pressure requirements. The Fire Protection Manual (FPM) requires ;
two fire pumps to remain operable when the plant is in Modes 1-4.
With one pump out-of-service, the FPM requires the out-of-service pump be returned to service within 7 days or an alternate pump of equal or l greater capacity and pressure be provided. )
The probability of a fire occurrence has not been increased by this temporary modification because it does not increase the variables of fire creation or occurrence.
The capability to fight a fire is maintained because the system design flow and pressure is maintained and shall be verified by surveillance testing prior to declaring the alternate system operable. Therefore, this modification does not increase radiological consequences of an accident. l l
The probability of occurrence of a malfunction of equipment important to safety is not increased because, in the event of a fire, system physical separation distance between redundant or diverse equipment important to safety is maintained.
P; 9togicc' consequences of a malfunction of equipment important to satety are not affected because the operability of this equipment is not affected.
In accordance with 10CFR50 Appendix R, only one fire is postulated to occur at any one time. Multiple fires are not postulated. Installing
, the alternate fire pump and piping connections does not introduce the possibility of creating multiple fires. Therefore, the possibility of an accident of a different type than any previously evaluated in the Updated Safety Analysis Report is not created.
This temporary modification does not modify, remove, or change the operation or configuration of equipment important to safety, so it does not create the possibility of a different type of malfunction.
The margin of safety as defined in the. basis for Technical Specifications are not affected.
Attachment to ET 97-0017 Page 180 of 209 Safety Evaluation: 59 96-0161 Revision: 1 Temporary Diesel Engine Driven Fire Pusp Replacement Revision 1 of this Unreviewed Safety Question Determination (USQD) provides a source of priming water from the Circulating Water System for the temporary pump. Utilizing Circulating Water has no adverse effect on the Circulating Water System. The capacity of just one circulating Water pump is over 160,000 gpm. Since the maximum priming flow out of the common Circulating Pump header will be less than 150 gpm, this amount of flow loss will have no adverse effect on condenser heat removal capability.
This temporary modification will have no impact on accidents or i 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.
- Revision 0 of USQD 96-0161 is show below.
Alternate pump and piping is provided to support fire protection needs of the plant by maintaining system design and function in accordance with the Fire Protection Manual while permanent diesel engine driven fire pump (1FP001PB) is being worked on. 1FP001PB has been found incapable of supplying the flow and pressure requirements. The Fire Protection Manual (FPM) requires two fire pumps to remain operable when the plant is in Modes 1-4. With one pump out-of-service, the FPM requires the out-of-service pump be returned to service within 7 days or an alternate pump of equal or greater capacity and pressure be provided.
The probability of a fire occurrence has not been increased by this temporary modification because it does not increase the variables of fire creation or occurrence.
The capability to fight a fire is maintained because the system design flow and pressure is maintained and shall be verified by surveillance testing prior to declaring the alternate system operable. Therefore, this modification does not increase radiological consequences of an accident.
The probability of occurrence of a malfunction of equipment important to safety is-not increased because, in the event of a fire, system physical separation distance between redundant or diverse equipment important to safety is maintained.
Radiological consequences of a malfunction of equipment portant to safety are not affected because the operability of this equipment is not affected.
I Attachment to ET 97-0017 4 i
Page 181 of 209 i In accordance with 10CFR50 Appendix R, only one fire is postulated to occur at any one time. Multiple fires are not postulated. Installing !
the alternate fire pump and piping connections does not introduce the i possibility of creating multiple fires. Therefore, the possibility of an accident of a different type than any previously evaluated in the ,
Updated Safety Analysis Report is not created. !
This temporary modification does not modify, remove, or change the i operation or configuration of equipment important to safety, so it l does not create the possibility of a different type of malfunction. 1 l
The margin of safety as defined in the basis for Technical Specifications are not affected. .
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Page 182 of209
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Safety Evaluation: 59 96-0161 Revision:2 Temporary Diesel Engine Driven Fire Pump Replacement l This Unreviewed Safety Question Determination (USQD) was originally i described in USQD 96-0161 Revision 0 and Revision 1. Revision 2 l removes the 150 psig limit on pump operation and evaluates the maximum pump speed condition that will not exceed 200 psig, which is the hydro test pressure for the fire protection piping. The pump shall be tested to find the maximum speed with the pump dead-headed which will result in a pressure that will not exceed 200 psig. This test conditic will set the maximum engine speed allowed. This maximum speed wixi be with +he ball valve closed. 1hese actions will ensure that over pressure concern- uf plant piping are not possible.
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.
This temporary modification, as described in prior revisions of USQD 96-0161, is shown below.
This revision 1 provides a source of priming water from the Circulating Water System for the temporary pump. Utilizing Circulating Water has no adverse effect on the Circulating Water System. The capacity of just one Circulating Water pump is over 160,000 gpm. Since the maximum priming flow out of the common Circulating Pump header will be less than 150 gpm, this amount of flow loss will have no adverse effect on condenser heat removal capability.
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.
This temporary modification, as described in Revision 0 of USQD 96-0161 is show below.
The permanent diesel engine driven fire pump, 1FP001PB, has been found incapable of supplying the flow and pressure requirements. The Fire Protection Manual (FPM) requires two fire pumps to remain operable when the plant is in modes 1-4. With one pump out-of-service, the FPM requires the out-of-service pump be returned to service within 7 days or an alternate pump of equal or greater capacity and pressure be provided. This temporary modification.provides the alternate pump and piping to support the fire protection needs of the plant by maintaining system design and function in accordance with the Fire Protection Manual while 1FP001PB is being worked on.
Attachment to ET 97-0017 l Page 183 of 209 l
The probability of a fire occurrence has not been increased by this temporary change because it does not increase the variables of fire creation or occurrence.
The capability to fight a fire is maintained by this modification l because the system design flow and pressure is maintained and shall be l verified by surveillance testing prior to declaring the alternate system operable. Therefore, this change does not increase ;
radiological consequences of an accident. l The probability of occurrence of a malfunction of equipment important to safety is not increased because, in the event of a fire, system physical separation distance between redundant or diverse equipment important to safety is maintained.
Radiological consequences of a malfunction of equipment important to safety are not affected because the operability of this equipment is not affected by this change.
In accordance with 10CFR50 Appendix R, only one fire is postulated to occur at any ona time. Multiple fires are not postulated. Installing the alternate fire pump and piping connections does not introduce the j possibility of creating multiple fires. Therefore, this change does I not create the possibility of an accident of a different type than any previously evaluated in the Updated Safety Analysis Report, j This temporary modification does not modify, remove, or change the operation or configuration of equipment important to safety, so it ;
does not create the possibility of a different type of malfunction.
The proposed change does not reduce the margin of safety as defined in ]
the basis for any technical specification. l I
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I Attachment to ET 97-0017 Page 184 of209 Safety Evaluation: 59 96-0161 Revision: 3 Temporary Diesel Engine Driven Fire Pump Replacement i
This Unreviewed Safety Question Determination (USQD), was originally described in Revision 0, Revision 1"and Revision 2. Revision 3 reflects the fact that there are other means to prime the pump if a loss of Circulating Water Pumps occurs due to a loss of electrical power. If Circulating Water flow is lost due to loss of electrical power as a result of a fire, the pump prime would most likely not be lost due to the back flow from the condenser. If the pump prime is lost, other means are available to prime the pump such as the belt j driven priming pump off the engine, gas powered booster pump or other
- means.
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.
Revision 2 of USQD 96-0161, is shown below.
I i Revision 2 removes the 150 psig limit on pump operation and evaluates
- the maximum pump speed condition that will not exceed 200 psig, which
, is the hydro test pressure for the fire protection piping. The pump shall be tested to find the maximum speed with the pump dead-headed which will result in a pressure that will not exceed 200 psig. This test condition will set the maximum engine speed allowed. This maximum speed will be with the ball valve closed. These actions will 4
ensure that over pressure concerns of plant piping are not possible.
i This temporary modification will have no impact on accidents or l 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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Revision 1 of this USQD is described below.
Revision 1 provides a source of priming water from the Circulating Water System for the temporary pump. Utilizing Circulating Water has no adverse effect on the circulating Water System. The capacity of just one Circulating Water pump is over 160,000 gpm. Since the maximum priming flow out of the common Circulating Pump neader will be less than 150 gpm, this amount of flow loss will have no adverse effect on condenser heat removal capability.
This temporary modification will have no impact on accidents or i malfunctions evaluated as the licensing basis and there is no potential for the creation of a new type of unanalyzed event. There
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l Attachment to ET 97-0017 Page 185 of209 is no reduction in the margin of safety.
Revision 0 of USQD 96-0161 is, described below.
I Alternate pump and piping is provided to support fire protection needs of the plant by maintaining system design and function in accordance with the Fire Protection Manual while permanent diesel engine driven fire pump (1FP001PB) is being worked on. 1FP001PB has been found 4 incapable of supplying the flow and pressure requirements. The Fire !
Protection Manual requires two fire pumps to remain operable when the plant is in modes 1-4. With one pump out-of-service, the manual requires the out-of-service pump be returned to service within 7 days l or an alternate pump of equal or greater capacity and pressure be provided.
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The probability of a fire occurrence has not been increased by this- l temporary modification because it does not increase the variables of l fire creation or occurrence. I l
The capability to fight a fire is maintained because the system design flow and pressure is maintained and shall be verified by surveillance testing prior to declaring the alternate system operable. Therefore, !
this modification does not increase radiological consequences of an ,
accident. '
The probability of occurrence of a malfunction of equipment important i to safety is not increased because, in the event of a fire, system '
physical separation distance between redundant or diverse equipment -l important to safety is maintained.
Radiological consequences of a malfunction of equipment important to I safety are not affected because the operability of this equipment is i not affected.
In accordance with 10CFR50 Appendix R, only one fire is postulated to I occur at any one time. Multiple fires are not postulated. Installing
~ j the alternate fire pump and piping connections does not introduce the -
possibility of creating multiple fires. Therefore, the possibility of an accident of a different type than any previously evaluated in the Updated Safety Analysis-Report (USAR) is not created.
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This temporary modification does not modify, remove, or change the I operation or configuration of equipment important to safety, so it does not create the possibility of a different type of malfunction.
The margin of safety as defined in the basis for Technical j Specifications are not affected. -
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Attachment to ET 97-0017 Page 186 of 209 1
Safety tvaluationt 59 96-0162 Revision 0 l 1' I i
Domestic Water supply Isolation valve Configuration This modification revises the Updated Safety Analysis Report (USAR)
Section 9.2 to reflect the actual as-built and required configuration of valves KDV133 and KDV559 (Potable Water Isolation Valves). KDV133 provides service to the rest rooms on level 2000' of the Turbine building; KDV559 provides isolation to an eye wash station.
There are no design basis cecidents where the KD (Domestic Water) system is referenced or discussed. There are no credible accidents or malfunctions to equipment important to safety created. There are no acceptance limits in the Technical Specifications for potable water, inside the restricted area, where this modification is imposed.
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Attachment to ET 97-0017 Page 187 of 209 Safety Evaluation: 59 96-0164 Revision 0 organizational Changes kelated to Chief Business officer This Unreviewed Safety Question Determination is being performed as a result of an organizational change that affects Chapter 13.1 of the Updated Safety Analysis Report (USAR).
The Chief Business Officer's title has been changed to Comptroller.
His responsibilities remain the same with the exception that Purchasing and Material Services will now report to the Chic!
Administrative Officer, the Coordinator Audit and Compliance will now report to the Manager Performance Improvement and Assessment and the Coordinator Strategic Planning will now report to the Vice President Engineering. All functions will continue to be performed and all requirements continue to be met.
4 As this change is administrative in nature there will not be any effect on accidents or equipment and acceptance limits are not affected.
i This revision to the USAR will have no impact on accidents or 1 malfunctions evaluated as the licensing basis and there is no l potential for the creation of a new type of unanalyzed event. There is no reduction in the margin of safety.
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4 4 Safety Evaluation: 59 96-0170 Revision 0 containment Cooler Modification This modification removes one coil of twelve in service on containment cooler SGN01C. The inlet and outlet header connections will be blind flanged. To make room for the blind flanges, the inlet and outlet nozzles to the coil may be cut away. The coil is abandoned in place
- by this modification, and is to be replaced no later than Refuel 9.
The function of the containment cooler is not affected by the replacement of one coil with blind flanges. The initiators of design basis accidents are not influenced by the presence or absence of one I
cooling coil, therefore there is no impact on accidents and malfunctions evaluated as the licensing basis.
The passive nature of the blind flanges precludes any possibility of their creating a new or different accident or malfunction. The margin of safety as defined in the basis for any Technical Specification is not affected.
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, Attachment to ET 97-0017 Page 189 of 209 Safety Evaluation: 59 96-0170 Revision 1 containment cooler Modification This Revision 1 of Unreviewed Safety Question Determination (USQD) 59 96-0170 removes a maximum of three out of twelve cooling coils from service on containment cooler SGN01C. Revision 0 removed only one coil as is described below.
Revision 0 of this modification removes one coil of twelve in service on containment cooler SGN01C. The inlet and outlet header connections will be blind flanged. To make room for the blind flanges, the inlet and outlet nozzles to the coil may be cut away. The coil is abandoned in place by this modification, and is to be replaced no later than Refuel 9.
The function of the containment cooler is not affected by the replacement of one coil with blind flanges. The initiators of design basis accidents are not influenced by the presence or absence of one
, cooling coil, therefore there is no impact on accidents and malfunctions evaluated as the licensing basis.
The passive nature of the blind flanges precludes any possibility of their creating a new or different accident or malfunction. The margin of safety as defined in the basis for any Technical Specification is not affected.
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Safety Evaluation: 59 96-0175 Revision 0 Turbine Driven Auxiliary Feedwater Pump Actuation Time Delay Acceptance Criteria This revision to Procedure STS AL-104 (TDAFWP EST Response Time Test) is made to reflect the auxiliary feedwater flow actuation time delay assumptions provided in the NRC approved transient analysis l methodology for non-LOCA (Loes of Coolant Accident) transients and the Westinghouse ECCS (Emergency Core Cooling System) Evaluation Model for l small break LOCAs. l This revision changes the way the actuation response time for the TDAFW (Turbine Driven Auxiliary Feedwater ) pump is determined. This response time is currently determined based on the measurement of turbine throttle valve stroke open time. The revision will require that the time required for the turbine driven auxiliary feedwater pump to start and reach its full operating speed be used as an acceptance criterion. By ensuring that the TDAFW pump reaches its rated speed of 3850 rpm within 60 seconds, it is demonstrated by the pump performance curve, M-021-148-01, that the pump will be capable of delivering the flow required by the safety analyses.
Because the basis of the change is the fact that current licensing basis accident analyses already considers a condition consistent with the change, the licensing basis analyses are not impacted. Therefore, this revision does not increase the probability of occurrence of an accident previously evaluated in the Updated Safety Analysis Report (USAR). Also, the revision does not increase the radiological consequences of an accident previously evaluated in the USAR.
This revision only revises the acceptance criteria associated with the response time testing of the TDAFW pump in a manner consistent with the licensing basis accident analyses. The revision does not result in a change to the operation of the TDAFW pump or any other plant equipment or procedure. Further, the revision does not result in any physical modification to the pump or any other plant equipment.
Therefore, this revision does not increase the probability of occurrence of a malfunction of equipment or the radiological consequences of a malfunction of equipment important to safety previously evaluated in the USAR.
There is no potential for the creation of a new type of unanalyzed event since this revision does not affect the operation of any plant equipment or modify any plant equiptnent.
Based on the fact that this revision to the response time testing acceptance criteria is consistent with the NRC approved transient methodology (SER provided in letter 93-01432) as well as the current
1 Attachment to ET 97-0017 1
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licensing basis accident analyses, ensuring that the pump will be i operating at 3850 rpm within 60 seconds, will also ensure that the l pump will be available to provide the flow required by the safety analyses. Therefore, the margin of safety is not reduced.
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Attachment to ET 97-0017 Page 192 of209 i
- , safety avaluationt 59 96-0176 Revision 0 i
tn a on s e h a g function of circulating water warming line air release valves, which are shown on ,
4 Updated safety Analysis Report (USAR) Figure 10.4-1-2. The purpose of )
l this modification is to determine if air binding is a credible cause i
! of a loss of circulating water warming line flow that caused the plant I
! shutdown in January 1996 due to the loss of condenser vacuum. This event actually demonstrated the worst case scenario, which is a loss I of warming line flow during frazil ice conditions that can lead to the previously evaluated turbine trip under section 15.2.3 of the USAR.
This modification provides guidance for validating that air binding can cause circulating water warming line flow reductions. A test will shut off the vent path for the air release vents for the circulating I
water warming line. Provisions are in the test instructions to
- l restore the air venting flowpath immediately if-the circulating water inlet temperatures are below 34*F. This will provide ample margin to i prevent an ice blockage event similar to the January 1996 event Therefore no previously evaluated accidents are affected. l The sole purpose of the warming line as designed and described in the USAR is to prevent ice blockage of the traveling screens and trash racks at the circulating water screen house. Therefore, there is no potential of creating any new type of accidents not previously evaluated.
There are no malfunctions of equipment important to safety affected by this modification. Nor are any Technical Specification limits affected.
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d Attachment to ET 97-0017 f Page 193 of 209 I
Safety Evaluttion: 59 96-0177 Revision 0 ,
l Organization Change in the Operations Department i
- This revision to Updated Safety Analysis Report (USAR) section 13.1.1.2.3.1 provides for a change in personnel assignments in the !
position of Manager of Integrated Plant Scheduling. Also, the responsibility for Chairman of the Plant Safety Review Committee (PSRC) is deleted from the Assistant to the Chief Operating Officer responsibilities.
These changes are administrative in nature and do not affect any systems, structures or components. No organizational functions have been deleted. No accidents are created nor affected as described in the USAR. Neither malfunctions of equipment important to safety nor acceptance limits for the Technical Specifications are affected.
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Attachment to ET 97-0017 j Page 194 of 209 Safety Evaluation: 59 96-0178 Revision 0
! Nitroger. Accumulator Pressure Clarification This change to Updated Safety Analysis Report (USAR) Section 9.3.1.2.2
. and Table 9.3-1 notes the range of operation of the Main Steam s
Atmospheric Relief, Main Feed Control and Auxiliary Feedwater Control Valves Nitrogen Accumulators from 650 to 750 psig. The USAR currently notes operation at 750 psig. A range is allowed during normal l operation. Table 9.3-1 is also being corrected to note the above j accumulators are not actually vertical tanks but are horizontal. This
, is a minor change that does not affect any of the design or operating capabilities of the tank.
, None of the design bases accidents discussed or referenced in Chapters i 2, 3, 6, 9 or 15 are affected by this USAR clarification. No credible accidents are created. No credible malfunctions of equipment important to safety are affected by these corrections and
, clarifications of the USAR.
The Teclnical Specification requirements for operability of the Main Steam Atmospheric Relief and Auxiliary Feedwater Control Valves Nitrogen Accumulators are not affected. The 650 psig minimum normal operating pressure has been evaluated to be acceptable given a 0.9 3
psi / min (54 psi /hr) maximum allowable unknown leakage rate. The Main Feedwater Control Valves Accumulator is non-safety related, therefore the range of operation does not require the same level of evaluation i nor is it required to perform a safety-related function. Therefore, these changes do not affect acceptance limits as defined in the basis for any Technical Specifications.
4 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 Evaluation: 59 96-0184 Revision:0 Temporary Modification to Remove Power Load Unbalance Circuit Cards This temporary modification removes the Power Load Unbalance (PLU) circuit cards 1PU2-A001 and 1PU3-A005 to eliminate the possibility of a turbine trip / reactor trip due to the failure of the watt transducer 1PU1-B201. The effects of removing these circuit cards disables the
" anticipatory" overspeed protection function as described in Updated Safety Analysis Report (USAR) Section 10.2.2.3.2 and Vendor Manual M-800-0231. This " anticipatory" overspeed protection functions to close l the Control Valves (CVs) and Intercept Valves (IVs) upon a loss of !
load from the Main Generator. If the electrical load returns quickly, the PLU circuit will reset, and the CVs will reopen to the position called for by the Speed Error Signal and the Load Reference which is j decreased by a small amount because the Load Set Motor has run back l for only a short period of time. The IVs will automatically reopen l after 1 second regardless of PLU duration. With this overspeed )
protective function disabled, there are two overspeed trip devices which will trip ALL of the valves closed. These overspeed trip devices (Mechanical and Backup overspeed trips) are set at 110% and 111% of normal operating speed (1800 rpm) . These overspeed trip devices will remain in operation to preclude an overspeed event.
There is only one accident previously evaluated in the USAR to which this change is applicable. That event is a Main Generator load ;
rejection which has the potential to result in a turbine / generator overspeed condition. However, the PLU circuit has no relationship to the events or conditions which can lead to a Main Generator load rejection. Therefore, this change does not increase the probability of occurrence of a an accident previously evaluated in the USAR.
By disabling the PLU circuit, two overspeed trip devices remain in )
operation to preclude a turbine overspeed event. Therefore, this {
change will not increase the radiological consequences of an accident i previously evaluated in the USAR. l This change does not affect any equipment important to safety !
therefore, this change does not increase the probability of occurrence of a malfunction of equipment important to safety or the radiological j consequences previously evaluated in the USAR.
Since no credible accidents or malfunctions were identified, there is j no potential for the creation of a new type of unanalyzed event.
Since no acceptance limits were identified that could be affected, the l margin of safety is not affected by this change. j f
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Attachment to ET 97-0017 Page 196 of209 Safety Evaluation: 59 96-0191 Revision 0 10-Year Reactor Cooling Pump Notor Flywheel Inspection Plan This revision to the Updated Safety Analysis Report (USAR) restores the commitment to Regulatory Guide 1.14, " Reactor Coolant Pump Flywheel Integrity," present during the NRC's review and approval of the operating license, through the time that USAR Change Request 95-003 was approved. This USAR change corrects the previous change (USAR-95-003) and adds wording to correctly reflect the 10-year Reactor Cooling Pump (RCP) motor flywheel inspection in the Inservice Inspection (ISI) Program Plan . The USAR currently states that the inspections occur in conjunction with the schedule established for motor refurbishment rather than the ISI Program Plan (which it had previously reflected). Correction of this requirement does not create any test or experiment not already described in the USAR.
This change does not affect flywheel integrity and its ability to provide the inortial rotation for pump coastdown as described in USAR Section 5.4.?. 3.2. Therefore, this change will not increase the probability of occurrence or radiological consequences of an accident previcasly evaluated in the USAR.
There are no credible malfunctions of equipment important to safety created by this change. Reduced coastdown due to a failed flywheel l has been evaluated and is bound by the locked rotor analysis. l There are no accidents or malfunctions of a different type created than previously evaluated in the USAR.
No margin of safety was identified as associated with this change to the inspection frequency of the RCP motor flywheel.
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safety Evaluation: 59 96-0192 Revision 0 Auxiliary Steam System Piping Replacement Due to Flow Accelerated Corrosion This modification replaces existing worn carbon steel piping components with low allow steel (2 1/4 Cr-1 Moly) to mitigate abnorwal pipe-wall thinning due to Flow Accelerated Corrosion (FAC) located on line AD-026-HBD-4, between valve ADV266 (Condensate System) and LP Condenser nozzle No.133.
The pipe replacement does not change the cross sectional properties, i mechanical properties or the geometric configuration. The change does '
not adversely impact the existing safety margins or structural I integrity of the affected piping system. The pipe replacement is an enhancement to the original design by providing an increased resistance to FAC. Therefore, this change will not increase the probability of occurrence of an accident or create the possibility of an accident of a different type than was previously evaluated in the Updated Safety Analysis Report (USAR).
The pipe replacement does not adversely affect any system, component l or procedures required to mitigate the consequences of an accident ;
previously evaluated in the USAR. This change will restore a degraded section of the affected piping system to its original design I configuration (piping geometry, cross section, support location, l fittings). All functions will continue to be safely performed, )
therefore the consequences of accidents previously evaluated in the '
USAR will not be increased.
1 Ductile fracture, corrosion, erosion / corrosion, loss of mechanical properties, excess strain, mech $nical creep etc., are credible failure modes for which the piping replacement has been evaluated. It was concluded that a new credible failure mode is not introduced. !
Therefore, the probability of vecurrence of a malfunction of equipment important to safety previously evaluated in the USAR is not increased, nor is the possibility of a different type of inalfunction of equipment important to safety created.
Since this modification is a design enhancement which does not adversely affect equipment important to safety, the radiological consequences of a malfunction of equipment important to safety will not increase.
This modification will have no effect on margins of safety as defined in the basis for any Technical Specifications since all system functions will continue to be performed as designed and no safety-related system or component is involved.
Attachment to ET 97-0017 Page 198 of 209 i
. Safety Evaluation: 59 96-0193 Revision 0 l 'Feedwater Heater Extraction Drain Piping Replacement Due to Flow Accelerated Corrosion This modification replaces the existing section of line AF-061-GBD-10 (Feedwater Heater Extraction Drains and Vents), between valve AFLV13 and Heater Drain Tank Nozzle C-1, with low allow steel (2 1/4 Cr-1 Moly) to mitigate abnormal pipe-wall thinning due to Flow Accelerated Corrosion (FAC).
The pipe replacement does not change the cross sectional properties, mechanical properties or the geometeric configuration. The change does not adversely impact the existing safety margins or structural integrity of the affected piping system. The pipe replacement is an enhancement to the original design by providing an increased resistance to FAC. Therefore, this change will not increase the -
probability of occurrence of an accident or create the possibility of an accident of a different type than was previously evaluated in the Updated Safety Analysis Report (USAR).
The pipe replacement does not adversely affect any system, component or procedures required to mitigate the consequences of an accident previously evaluated in the USAR. This change will restore a degraded section of the affected piping system to its original design configuration (piping geometry, cross section, support location, fittings). All functions will continue to be safely performed, therefore the consequences of accidents previously evaluated in the USAR will not be increaseo.
Ductile fracture, corrosion, erosion / corrosion, loss of mechanical properties, excess strain, mechanical creep etc., are credible failure modes for which the piping replacement has been evaluated. It was concluded that a new credible failure mode is not introduced.
Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR is not increased, nor is the possibility of a different type of malfunction of equipment important to safety created.
Since this modification is a design enhancement which does not adversely affect equipment important to safety, the radiological consequences of a malfunction of equipment important to safety will not increase.
This modification will have no effect on margins of safety as defined in the basis for any Technical Specifications since all system functions will continue to be performed as designed and no safety-related system or component is involved.
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- Safety Evaluation
- 59 96-0194 Revision: 0 Low Pressure Heater 2A to 1A Piping Replacement Due to Flow Accelerated Corrosion This modification replaces the existing section of line AF-388-MBD-8 4 (Feedwater Heater Extraction Drr. ins and Vents), between valves AFLV97 and AFV266, with low allow steel (2 1/4 Cr-1 Moly) to mitigate j abnormal pipe-wall thinning due to Flow Accelerated Corrosion (FAC) .
1 The pipe replacement does not chang; the cross sectional properties, mechanical properties or the georSceric configuration. The change j does not adversely impact the existing safety margins or structural integrity of the affected piping system. The pipe replacement is an d
enhancement to the original design by providing an increased l resistance to FAC. The.'re f ore , this change will not increase the l
. probability of occurrence of an accident or create the possibility of an accident of a different type than was previously evaluated in the
, Updated Safety Analysis Reoort (USAR).
The pipe replacement does not adversely affect any system, component or procedures required to mitigate the consequences of an accident previously evaluated in the USAR. This change will restore a degraded section of the affected piping system to its original design configuration (piping Seometry, cross section, support location, i fittings). All functions will continue to be safely performed, therefore the consequences of accidents previously evaluated in the J USAR will not be increased.
i Ductile fracture, corrosion, erosion / corrosion, loss of mechanical properties, excess strain, mechanical creep etc., are credible failure modes for which the piping replacement has been evaluated. It was concluded that a new credible failure mode is not introduced.
Therefore, the probability of occurrence of a malf unction of equipment
- important to safety previously evaluated in the bSAR is not increased, nor is the possibility of a different type of malfunction of equipment important to safety created, i
Since this modification is a design enhancement wh1 a does not adversely affect equipment important to safety, the radiological consequences of a malfunction of equipment important to safety will not increase.
t This modification will have no effect on margins of safety as defined in the basis for any Technical Specifications since all system functions will_ continue to be performed as designed and no safety-related system or component is involved.
Attachment to ET 97-0017 Page 200 of 209 I
Safety Evaluation: 59 96-0195 Revision 3 0 Low Pressure Heater 25 to IB Piping Replacement Due to Flow Accelerated Corrosion This modification replaces existing worn carbon steel piping components with low allow steel (2 1/4 Cr-1 Moly) to mitigate abnormal j pipe-wall thinning due to Flow Accelerated Corrosion (FAC) of line AF-389-HBD-8 between valves AFLV128 and AFV268 (Feedwater Heater Extraction, Drains, and Vents system).
The pipe replacement does not change the cross sectional properties, mechanical properties or the geometeric configuration. The change i does not adversely impact the existing safety margins or structural i integrity of the affected piping system. The pipe replacement is an enhancement to the original design by providing an increased resistance to FAC. Therefore, this change will not increase the probability of occurrence of an accident oc create the possibility of an accident of a different type than was previously evaluated in the Updated Safety Analysis Report (USAR).
The pipe replacement does not adversely affect any system, component or procedures required to mitigate the consequences of an accident previously evaluated in the USAR. This change will restore a degraded section of the affected piping system to its original design configuration (piping geometry, cross section, support location, fittings). All functions will continue to be safely performed, therefore the consequences of accidents previously evaluated in the USAR will not be increased.
Ductile fracture, corrosion, erosion / corrosion, loss of mechanical properties, excess strain, mechanical creep etc., are credible failure modes for which the piping replacement has been evaluated. It was concluded that a new credible failure mode is not introduced.
"herefore, the probability of occurrence of a malfunction of equipment incortant co safety previously evaluated in the USAR is not increased, not is the par *ibility of a different type of malfunction of equipment importan'. to safety created.
Since this modification is a design enhancement. which does not adversely affect equipment important to rafety, the radiological consequetees of a malfunction of equipment impo tant to sarety will not increave.
This modification will have no effect on margins of safety as defined in the basis for any Technical Specifications since all system functions will continue to be performed as designed and no safety-related system or component is involved.
Attachment to ET 97-0017 i
Page 201 of 209 i l
Safety EEaluation: 59 96-0196 Revision:0 Low Pressure Heater 2C to 1C Piping Replacement Due to Flow Accelerated Corrosion This modification replaces existing worn carbon steel piping components with low allow steel (2 1/4 Cr-1 Moly) to mitigate abnormal l pipe-wall thinning due to Flow Accelerated Corrosion (FAC) of line AF-390-HBD-8 between valves AFLV159 and AFV270 (Feedwater Heater Extraction, Drains, and Vents system).
l The pipe replacement does not change the cross sectional properties, j mechanical properties or the geometeric configuration. The change does not adversely affect the existing safety margins or structural integrity of the affected piping system. The pipe replacement is an enhancement to the original design by providing an increased resistance to FAC. Therefore, this change will not increase the l probability of occurrence of an accident or create the possibility of l an accident of a different type than was previously evaluated in the Updated Safety Analysis Report (USAR).
The pipe replacement does not adversely affect any system, component or procedures required to mitigate the consequences of an accident ,
previously evaluated in the USAR. This change will restore a degraded l section of the affected piping system to its original design l configuration (piping geometry, cross section, support location, I fittings). All functions will continue to be safely performed, l therefore the consequences of accidents previously evaluated in the l USAR will not be increased. l Ductile fracture, corrosion, erosion / corrosion, loss of mechanical properties, excess strain, mechanical creep etc., are credible failure modes for which the piping replacement has been evaluated. It was concluded that a new credible failure mode is not introduced.
Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR is not increased, nor is the possibility of a different type of malfunction of equipment important to safety created.
Since this modification is a design enhancement which does not adversely affect equipment important to safety, the radiological consequencap of a malfunction of equipment important to safety will not incres4*.
This modification will have no effect on margins of safety as defined in the basis for any Technical Specifications since all system functions will continue to be performed as designed and no safety-related system or component is involved.
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Page 202 of 209 i
i Safety Evaluation: 59 96-0197 Revision 0 Add Loop Power Supplies to Chestical and Volume Control System Drawing This modification revises P&ID M-12BG03, " Chemical and Volume Control 1 System," to add existing components BGPQYO115 and BGFQYO121 in their
{ respective instrument loops. Currently, the subject instruments are shown only on the Westinghouse supplied documentation. BGPQYO115 is the loop power supply for pressure transmitter BGPT0115, which is
- interlocked to BGPCV0115 and is designed to control pressure in the 4
VCT (Volume Control Tank) through the gaseous vent line where the discharge is expelled to the gaseous radwaste system. BGFQYO121 is the loop power supply for flow transmitter BGFT0121, which is l interlocked to BGFCV0121 and controls charging flow from the centrifugal charging pumps.
This modification to the P&ID does not change any procedures, work activities or administrative controls, nor will it change or affect the operation of any system, structure or component as described in the Updated safety Analysis Report (USAR).
, There are no design basis accidents identified because this I administrative document change will enhance the P&ID. There are no
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credible accidents or malfunctions of equipment important to safety identified.
4 There are no acceptance limits identified that could be affected by d
this modification.
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Safety Evaluation: 59 96-0199 Revision 0 j Partial Seismic Instrumentation System Replacement
) This modification documents the replacement of portions of the seismic Monitoring and Alarm (SG) System housed in cabinet SG058 wi?.h
, functionally equivalent components supplied by the original system I manufacturer. Tht new componenta maintain the original design basis, f qualifications cd functions.
The Updated Safety Analysis Report (USAR) (3.7 (B) .4.1c) refers to the use of recording tape for event recording. The new components use solid state memory chips to accomplish the same function. The SG system has no safety related functions or II/I requirements.
The function of the SG system is to monitor for and alarm the occurrence of a seismic event. It does not initiate any plant function other than a Main Control Board Annuciator window.
Therefore, this change does not increase the probability of occurrence of an accident previously evaluated in the USAR. This modification cannot create a new accident type.
Seismic event detection has no connection to radiological consequences and can not initiate a radiological release.
The SG system does not interface or connect with any-safety related system physically or electrically so the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR is not increased. Since there is no interconnection with equipment important to safety, the possibility of a different type of malfunction of equipment is not created. j The replacement components are more sensitive and more capable than the components they replace. They maintain, at a minimum, the current margins of seismic detection. The SG system is not a Technical Specification 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. j M
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, Attachment to ET 97-0017 l
. i Page 204 of209 I safety Evaluation: 59 96-0200 Revision: 0 Organizational Change in Operations Department l This change revises Updated Safety Analysis Report (USAR) section 13.1.2.2.1 references to the Supervisor Water Treatment and Radwaste Supervisor positions which have been combined into one position.
Procedure AP 17C-013 Rev. O, " Supervisor Treatment Systems Qualifications and Responsibilities," which supersedes procedure ADM 02-013 Revision ,3 " Supervisor Radwaste," has been written to conform to the Writer's Guide format, reflect title changes, and reflect
] additional responsibilities of the Supervisor Treatment Systems, who is responsible for both the water treatment systems and radwaste systems.
These procedural changes do not change any administrative controls which would reduce the level of qualification of personnel, nor does -
3 it affect any structure, system or component. These procedural l 4
changes also do not change the performance of activities that are 1 important to safe and reliable operation. Therefore, there are no design basis accidents identified.
These changes 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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0 Attachment to ET 97-0017 Page 205 of 209 l
Safety Evaluation: 59 96-0203 Revision 0 J
Auxiliary Building HVAC System Drawing Changes Updated Safety Analysis Report ('USAR) Figure 1.2-9, " Equipment Location Reactor and Auxiliary Building Plan Basement El 1974' O.(M-4 1G020-6)"," and Figure 1.2-19, " Equipment Location Reactor and Auxiliary Building Sections E, F &G, (M-10030)" are being issued to update the existing USAR figures with as-built information. USAR
, Figure 9.4-3, " Piping and Instrument Diagram Auxiliary Building HVAC,"
i is being issued to remove valves GLV0741 and GLV0742 from the drawing as they do not exist. The valves were originally installed for isolation of existing pressure differential indicator GLPDIO102 in the i event a repair or replacement was required for the indicator. The effects on cooler SGLO7 due to the removal of these valves were
- evaluated. SGLO7 is a non-safety related room cooler that provides cooling to the Normal Charging Pump (NCP) Room. These valves are not j required for operation of the room cooler or the BG system (Chemical &
i Volume Control System). The valves do not perform any safety related function. The removal of these valves will not affect the desisc.
bases of the cooler or its associated BG system.
l Since these changes will not affect the design parameters previously evaluated for accidents in the USAR, the probability of occurrence of any of these accidents will not increase.
1 These changes will not adversely affect the integrity of radiological
- barriers or the ability of the system to mitigate a radiological dose
- to the public that falls outside the acceptable limit.
These changes will not adversely affect equipment protective features, system redundancies or frequency of operation of the related safety j
' systems. Therefore, the probability of occurrence of a malfunction of '
equipment important to safety previously evaluated in the USAR will not be increased.
The radiological consequences of a malfunction of equipment important to safety are not increased because the ability of the involved systems to perform their safety related functions is not affected.
The changes will not create a condition in the involved systems that falls outside the approved limits of the design criteria, therefore the possibility of an event other than one that has been previously l evaluated in the USAR is not increased. I l
Failure modes of safety related equipment are not changed, therefore the possibility of a different type of malfunction other than one that has been previously evaluated in the USAR is not increased.
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i j -Attachment to ET 97-0017 Page 206 of 209 i
? The changes will not exceed or reduce the previously approved limits j .. established in the Technical Specifications of the involved systems.
Therefore, the margin of safety is not reduced, d
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Safety Evaluation 59 96-0205 Revision 0 l l
Reanalysis of Pipe Stress Values Due to Reduction of Snubbers From '
chemical and Volume control System This change revises Updated Safety Analysis Report (USAR) Table 3.6-3 to reflect revised pipe stress values which have been reanalyzed due to the reduction of snubbers from the BG (Chemical & Volume Control) l system.
The new pipe stresses are within the allowable limit of 'High Energy Pipe Break' as indicated on USAR Table 3.6-3 so there is no impact on j accidents and malfunctions evaluated as the licensing basis. ;
f The analysis did not create any new pipe break locations. Therefore the re is no potential for a new type of unanalyzed event.
Pipe atresses are within the ASME (American Society of Mechanical Engineers) code allowable limits and also within the allowable limit of 'High Energy Pipe Break' as indicated on USAR Table 3.6-3.
Technical Specifications have been reviewed. Since the level of qualification has not changed and there is no effect on the system, structure or component, no acceptance limits are identified that could be affected.
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 Attachment to ET 97-0017 Page 208 of 209 a
d Safety Evaluation: 59 96-0209 Revision 0 Revision 28 to the Wolf Creek Physical Security Plan ,
. This change to the Physical Security Plan adds a new category of l search-exempt cargo. This change allows cargo to be exempt from 4 search under the following conditions: 1) The cargo is sealed by another licensed nuclear facility prior to leaving that licensees Owner Controlled Area, 2) The cargo is searched and sealed at another
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location by a member of the Security Force, 3) The cargo is searched at another location and escorted by a member of the Security Force, 4) the cargo is sealed prior to leaving the Protected Area and is stored within the owner Controlled Area.
1 This change improves the efficiency in controlling and searching outage related cargo from vendors. By allowing sealed containers to
.1 be stored in the Owner Controlled Area, the Protected Area is less congested and reduces the potential for concealment of an adversary.
- Lighting and surveillanceLis further enhanced by eliminating shadows and visual obstructions.
- There will be no impact on accidents or malfunctions evaluated as the I
- licensing basis and there is no potential for the creation of a new l
- type of unanalyzed event. There is no reduction in the margin of i safety. i i
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l Attachment to ET 97-0017 Page 209 of 209 Safety Evaluation: 59 96-0211 Revision 0 Temporary Removal of 3/8" Drain Line and Valve GLV766 This temporary modification changes the piping configuration, as shown on Updated Safety Analysis Report (USAR) Figure 9.4-3-05, to remove an approximate one foot section of drain line and associated isolation
, valve, and install a cap at the 3" GA (Plant Heating) system line i fitting.
A GA system leak was identified in a drain line near a fitting at the inlet to SGLO1 (Auxiliary Building Normal Supply Fan) heating coil.
, The leak was located at the fitting where a 3/8" copper tube fastened to the 3" GA system pipe. The leak was repaired by removing the drain line and valve and installing a cap directly on the existing pipe fitting. The cap will be left in place until corrective maintenance is performed on the system.
The GA system, associated piping and SGL01 are non-safety related and are not discussed in any design basis accidents. They have no impact on the safety systems or any safety-related system, structure or component (SSC) for the plant.
The function of this drain line and valve is to provide a convenient method to drain the SGLO1 unit heating coil when the coil is isolated and draining is required. Draining the heating coil can still be accomplished by removing the cap. Temporarily removing this drain line and valve from the GA piping has no impact on equipment important to safety. Since there are no effects to any safety-related SSC no credible accidents that could be created are identified.
There are no credible malfunctions of equipment important to safety which may be hffected by the removal of this drain line and isolation valve. Since there are no identified effects on any SSC, no credible malfunctions of equipment important to safety are identified.
During a design basis accident these systems are automatically shut down and are not required to operate. There are no identified acceptance limits affected by this modification.
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