ULNRC-03580, Forwards 10CFR50.59 Annual Rept Summaries for 1996, Summarizing Written Safety Evaluation of Changes,Tests & Experiments Approved & Implemented for Activities at Plant
| ML20141A139 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 05/09/1997 |
| From: | Passwater A UNION ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| ULNRC-03580, ULNRC-3580, NUDOCS 9705140166 | |
| Download: ML20141A139 (61) | |
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1901 Chouteau Avenue s Post Othcn Box 149 St lauts, Missaun 63166 *
- 314 E2I 3222 l
s Uniav nev 9, 1997 ;
Euscmc h3 U.S. Nuclear Regulatory Commission ATrN'. Document Control Desk Mail btation PI-137 Washington, DC 20555 1 Gentlemen: ULNRC 03580 i
l DOCKET NUMBER 50-483 i
! CALLAWAY PLANT !
10CFR50.59 ANNUAL REPORT SUMMARIES UNION ELECTRIC APPROVED WRITI'E_N SAFETY EVALUATIONS
Reference:
ULNRC-03290, dated November 8,1995 In accordance with 10CFR50.59(b)(2), this letter transmits a report which summarizes written safety evaluations of changes,- tests, and experiments approved and implemented for activities at Callaway Plant since j
- those reported in the referenced submittal and through December 31,1996. ;
4 Based on these evaluations, we have concluded:
1 e' The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not increased.
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- That an accident or malfunction of equipment important to safety of a type different than those evaluated previously in the Final Safety Analysis Report has not been created.
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e The margin of safety as defined in tFe basis for any Technical Specification is not reduced.
9705140166 970509 PDR ADOCK 05000483 6 R PDR ll..,,,,,.
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- g. U.S. Nuclear Reguhtory Commission ULNRC-03580 Page 2 l
Therefore, all items reponed herein were determined not to involve an unreviewed safety question, i If there are any questions, please contact us.
Very truly yours, Alan C. Passwater Manager, Licensing and Fuels WBM/jdg i
Attachment 1
cc: M. H. Fletcher Professional Nuclear Consulting, Inc.
19041 Raines Drive Derwood, MD 20855-2432 Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive Suite 400 Arlington, TX 76011-8064 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Kristine M. Thomas (2)
Office of Nuclear Reactor Regulation.
U.S. Nuclear Regulatory Commission 1 White Flint, North, Mail Stop 13E16 11555 Rockville Pike Rockville, MD 20852-2738 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102 l
4 Reference / Abbreviation Key l
1 CN ------- FSAR Change Notice.
1 1
MODIFICATION PACKAGES (Design Changes)
CMP ---- Callaway Modification Package RMP ----- Restricted Modification Package FCN ------Field Change Notice OL ------- Operating License Change (Tech. Specs)
PROCEDURE _S CTP ------- Chemistry Technical Procedure ETP- Engineering Technical Procedure HTP ------- Health Physics Technical Procedure OTG ------ General Operating Procedure OTN ------- Normal Operating Procedure RFR ----- Request for Resolution SOS- Suggestion-Occurrence-Solution TM -- Temporary Modification W --------- Work Request Note: FSAR and Technical Specification changes are also reported under 10 CFR 50.71 and 10 CFR 50.90 as applicable.
SUMMARY
REPORT FOR CALLAWAY PLANT l
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CN 94-002 Inadvertant SI at Power This CN changes the FSAR Chapter 15 discussion of an inadvertant Si event. A 10 minute operator action time is assumed in the analysis to terminate the St. j An additional acceptance criterion was added to the analysis to assure the j pressurizer does not become water solid prior to termination of the St. The 10 ;
minute operator action time is reasonable based on NRC acceptance of other operator action times (Ref: FSAR Sections 6.2.1.4.3.3,10.4.9.2.3, and 15.0.13). Based on the reanalysis, no DNBR limits will be violated, the design pressure limit of the system will not be exceeded, and the pressurizer will not be overfilled during an inadvertant SI. No unreviewed safety question is created by this change.
CN 94-060 Changes to External Dosimetry Description in the FSAR J
This change clarifies the requirements for external dosimetry. National j Voluntary Accreditation Program (NVLAP) accreditation is required by 10CFR20 and '
is added to the FSAR. The change is an administrative type change and does not l cdversely affect plant safety. No unreviewed safety question is created by this i change. '
CN 94-061 Allow Use of Demineralization of Liquid Waste Streams .
This change notice will allow Radwaste personnel the option to either evaporate, !
filter, and/or demineralize the liquid waste stream from the Waste Holdup Tank l cnd the floor drain tanks using existing plant equipment. This change was '
rsviewed for environmental impact. No unreviewed safety question is created by this change, i
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CN 95-006 Tcchnical Specification Relocation FSAR Change ;
This FSAR Change Notice incorporates requirements removed from the Technical l Specifications by amendment request OL-1151 (Amendment No.103). Amendment j request OL-1151 evaluated the changes and obtained NRC approval via Amendment ;
j 103. This FSAR CN simply incorporates the amendment into the FSAR. No unreviewed safety question is created by this change.
Ref: OL-1151 Amendment 103 CN 95-019 Organization Changes to FSAR Chapter 13 This FSAR Change Notice incorporated miscellaneous changes to the plant organizational structure. The major changes are as follows:
- 1. The Quality Control organization will report to the Manager, OA.
- 2. The Superintendent, l&C position will be eliminated and his responsibilities will be assumed by the i&C General Supervisor.
- 3. The Supervising Engineer, Maintenance position is being deleted and Maintenance Engineers will report to Nuclear Engineering.
- 4. The l&C Department will report to the Assistant Manager, Work Control.
These changes do not impact the overall experience base at Callaway Plant nor the commitment to minimum qualifications. This FSAR change does not ad tersely impact operation of Callaway Plant and there are no unreviewed safety questions created.
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CN 96-006 FSAR Description of Normal Offload This FSAR Change Notice makes terminology changes regarding offloading I cctivities. The term " normal offload" is misleading and therefore will be changed to more accurately reflect activities at Callaway Plant. The CN also .
cvaluates the acceptability of completing a full core offload in less than 196.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after plant shutdown. In addition, the CN makes changes to the Component Cooling Water loading table (Table 9.4-7) to reflect that the loading is based on normal operational conditions (non-refueling).
These changes to not adversely impact the spent fuel pool cooling system nor the component cooling water system. The maximum loading on the cooling system is unchanged. No unreviewed safety question is created by this change.
CN 96-007 Revision of Snubber Surveillance Requirements This evaluation is applicable to changes to FSAR Chapter 16 involving the l snubber program. The changes involve the following: 1
- 1. Revision of Figure 16.7-1 to delete the upper " reject"line for the 37 snubber test plan.
- 2. Deletion of the 55 snubber test plan from the functional test plan options. !
- 3. Modification of the functional test plan criteria to permit testing of l snubbers on-line rather than only during shutdown (subject to plant access guidelines and policies).
- 4. Clarification of engineering evaluation criteria for functional test failures and the impact on additional test samples.
- 5. Incorporation of Interpretation 38 into the Bases Section (FSAR 16.'7.2.1.2).
These changes are acceptable and will not impact the existing design basis of the plant. No unreviewed safety question is created by this change.
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CN 96-010 Spent Fuel Pool Seismic Analysis This FSAR Change Notice incorporates the final versions of design analyses cssociated with the spent fuel storage racks. These analyses show that the r cks meet or exceed the structural acceptance criteria for Category I structures described in Standard Review Plan section 3.8.4. The CN simply incorporates the final information that had not been previously incorporated.
Therefore, the change does not involve an unreviewed safety question.
CN 96-011 Spent Fuel Pool Full Core Offload Analysis This FSAR Change Notice approves an increase in the maximum thermal loading of the Spent Fuel Pool (SFP), approves the removal of references to 19G.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> c.fter shutdown for a full core offload, and consolidates / deletes information from Appendix 9.1 A of the FSAR regarding the thermal / hydraulic analysis. The 1 test analysis shows that higher SFP thermal loadings up to 47.8 million Btu /hr will not cause the acceptance criteria to be violated. The deletion of the 196.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for full core offload completion does not impact any analysis ccceptance criterion. The consolidation / deletion of information does not reduce the FSAR level of detail below that required by NRC guidance. No unreviewed srfety question is created by the changes approved by this change notice.
CN 96-049 Correction to Low Temperature Operation Description i
This change notice corrects a statement in FSAR Section 5.2.2.10 which states that the nonoperating charging and safety injection pumps will have power locked out when the primary system reaches 1000 psig (descending) and approximately 425 F. The FSAR is changed to reflect the requirements of Technical Specification 3/4.5.2 and 3/4.5.3. The intent of both the TS and the FSAR s:ction is to provide requirements to protect the primary system from Low Temperature Over Pressure (LTOP) events. The FSAR statement is an error and is corrected to be consistent with TS requirements. No unreviewed safety question is created by this change.
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CN 96-059 Removal of 10% Stroke Testing of Main Steam and Feedwater Isolation Valves l l
This FSAR Change Notice removes references to periodic 10 percent stroke tests I of the Main Steam Isolation Valves (MSlVs) and Main Feedwater Isolation Valves (MFIVs) from the FSAR. This revision is being made to maintain consistency with current inservice Testing (IST) Program requirements and the proposed surveillance requirements associated with the conversion to the Improved Tcchnical Specifications. There are no changes to the physical or operating characteristics of the MSIVs/MFIVs. The operational readiness of these valves is still verified in accordance with Technical Specification surveillance and IST program requirements. The benefit gained as a result of quarterly partial stroke testing does not outweigh the disadvantages associated with the increased potential for a reactor trip or a severe system transient. This change does not create an unreviewed safety question.
CN 96-061 Reduce Minimum Component Cooling Water Temperature This evaluation addresses a change in the minimum allowable Component Cooling Water (CCW) temperature from 60 degrees F to 40 degrees F. This evaluation examined the effects on all major CCW loads and determined that current analyses bound the proposed change. Current maintenance and surveillance practices tssure that any change in equipment performance will be detected. No unreviewed
, safety question is created by this change.
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CN 96-020 Delete Requirement to Open Containment Iso Vivs for Hydrogen Analyzers for ILRT This FSAR Change Notice removes the note from Figure 6.2.4-1 sheets 42,69,70, ;
cnd 71 that states the containment isolation valves will be opened during an i ILRT. This is not required since the containment isolation valves are local I leak rate tested and receive an automatic signal to isolate. The valves are i
then opened (via remote manual signal) to monitor containment hydrogen 1 concentration. In the event of a system leak, the Auxiliary Building radiation monitors would alert operators to the problem. This would allow the operators to isolate the affected penetration. The removal of this note from the FSAR l
, figure will not result in an unreviewed safety question. l 1
CMP 89-1045 Rev. B Retire Portions of the Stock Solidification Equipment This modification and associated Field Change Notices (FCN's) approves the installation of a new radwaste solidification system with associated support system changes. In addition, this modification approves the retirement of s:veral parts of the Solid Radwaste System which calls for de-termination of components and mechanically blanking them off from pressurized auxiliary / support systems. The retired equipment is non-safety-related whose failure would not cdversely impact nuclear safety. j 4
Field Change Notice 4 approved the change to the normal valve positions for 2 vnives serving the new RVR-800 Waste Processing System. These ct.anges have been evaluated and found not to adversely impact nuclear safety. There is no direct interface with safety-related plant equipment. No unreviewed safety question is created by this change.
Ref: CMP 89-1045, FCN-04
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CMP 91-1046 Rev. A Isolation Valves on Auxiliary Gas Supply to S/G PORV's and Aux. Feed Valves This modification will install manual isolation valves on the backup gas supply (Nitrogen) to the Steam Generator PORV's and the pneumatic Auxiliary Feedwater i Control Valves. This will enable personnel to individually isolate each valve
. (PORV or Aux. Feed Valve) for maintenance. This will eliminate the necessity of isolating both a PORV and an Aux. Feedwater valve since the are fed from the
- s me backup gas accumulator. The new valves are ASME Section ill, Class 3 components, therefore the probability of pressure boundary failure is not increased. There is no effect on Callaway's SBO analysis and no unreviewed safety question.
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, Ref: CMP 91-104C. FCN-02 !
- 1 CMP 91-1060 Rev. B Deactivation of Halon Fire Suppression for Six Rooms in Control Building This evaluation examines the permanent deactivation of the halon fire suppression system for six switchgear rooms on the Control Building 2016' I clevation. Several new fire detectors are also to be added to these rooms.
. With removal of the system, adequate fire detection will exist to alert operators and sufficient capability will exist to isolate an> manually suppress .
cny fires in affected areas. No unreviewed safety questiu. 's created by this I change.
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CMP 921001 Rev. C Expand SIG Primary Platforms This modification will alter the platforms for steam generators B and C. This will require the relocation of the Seal Water injection tubing and the lube oil ;
drain piping to avoid interferences with the new design.
l The new design will meet the original design codes for the systems. The modification maintains system stresses below the design criteria of FSAR Chapter l 3.6 and 3.9 as well as ASME Section ill, Class 2 requirements. S unreviewed !
safety question is created by this modification. l l
. CMP 92-1041 Rev. A '
Modify ESW System Piping to Support Temporary Containment Cooling
- This modification was intended to install 6" connections to the supply and l return lines of the Containment Air Coolers to facilitate temporary cooling i during refueling outages. After installation on the "A" train ESW supply line, the balance of the modification was cancelled. The design of the completed portion of the modification meets ASME Section Ill, Class 3 requirements and will not adversely impact the integrity of the Essential Service Water System.
. This modification does not create the possibility of an accident different than those previously evaluated in the FSAR and no unreviewed safety question is created.
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CMP 93-1004 Rev. A Add Flow Nozzles in Emergency Fuel Oil Line This modification will install a flow nozzle in the Emergency Diesel Generator
- Fuel Oil line so that flow data can be taken without having to take the D/G out of service. The flow nozzle will be located in safety rela'ed piping and will be safety related for pressure boundary reasons. The few nozzle taps will have
, normally closed, safety related root valves.
The effects of the installation have been evaluated for impact on system flow tnd no adverse impact will result. No seismic concerns are created. This modification does not create an unreviewed safety question.
' CMP 93-1047 Rev. A Add Computer Terminal / Phone Hookups to 2000' and 2016' of Control B!dg CMP 93-1047 will install two terrninal access points in the control building.
One will be on the 2016' level of the Control Building adjacent to the phone in room 3401. The other will be installed on the 2000' elevation on the plant west wl! of the ESF Switchgear Room #1.
The routing of conduit in rooms 3416 and 3301 meet separation requirements.
4 Criculation C-07-39-F Rev. 2, Addendum 5 was performed to qualify 'he supports located on a seismic block wall. No unreviewed safety question is created by this modification.
CMP 93-1060 Rev. A Add Nitrogen-16 Detection System to Main Steam Lines l
This modification will install a Nitrogen-16 (N-16) detection system that will monitor the main steam lines in the turbine building for the presence of j radiation that is indicative of N-16. The presence of N-16 is indicative of a j steam generator tube leak, and this system can be used to identify and quantify 1 the leak. Since the system interfaces will be very limited, the modification l will not affect any safety aspects of the plant and does not result in an I unreviewed safety question.
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CMP 94-1003 Rev. A R: move injector Cooling System on Diesel Generators This modification will remove the injector cooling system for the emergency diesel generators (NE01 and NE02). The injection cooling system is designed for engines that use several different types of fuel to power them. Only diesel fuel is used at Callaway Plant. Extensive testing and evaluation by the engine manufacturer has concluded that the system may be removed without adverse impact on the ability of the generators to start, accelerate to rated speed, and accept load from the emergency bus. No unreviewed safety question is created by this modification.
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CMP 94-1012 Rev. A Remove Control Room Ventilation Rad Monitors Air Sample Return Lines This modification removes the control room ventilation radiation monitors' simple return lines between the sample vacuum pumps' exhaust and the control room supply air ductwork. This will eliminate the noise in the Shift Supervisor's office that is caused by the air pulsations coming from the rcdiation monitors' exhaust. The return lines will be capped at the ductwork, and the exhaust of the sample pump will be vented through an air silencer to the Auxiliary Building. This modification will have no effect on any of the monitors' safety related functions nor will it affect the control room i v:ntilation system's ability tc possurize the control room after a Control Room Vantilation Isolation Lgnal (CRVIC). Therefore, there is no unreviewed safety question as a resut of this modifiction.
Ref: FSAR CN 95-025 Report Date: May 9,1997 Page 10
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_ question since "No Radio Transmission" warning signs are posted throughout the plant and are adhered to by plant personnel. No unreviewed safety question is created by this change.
Ref: FSAR CN 94-052 CMP 94-1019 Rev. A Eliminate Snubbers on Class 1 Piping This modification eliminates most of the snubbers on the pressurizer surge line, excess letdown lines, accumulator lines and seal water lines. Reanalyses of the cffected piping demonstrates that ASME Code, FSAR, and regulatory commitments tre satisfied. No new pipe break locetions are identified, therefore, the existing pipe whip restraint configuration remains acceptable. No unreviewed
, si.fety question is created by this change.
CMP 95-1009 Rev. A Containment Temporary Power This modification provides additional electrical power to the containment building for use during outages. This additional electrical power will be fed from a spare non-safety related load center breaker, through a spare electrical p:netration module, and then to 400 amp safety disconnect switch located in the containment building. Additionally, this modification will route permanent power feeds from from containment building motor control center PG20P breakers to disconnect switches located at the steam Generator secondary platform areas.
The power circuits are non-safety related and do not affect any Class IE cquipment considered in the accident analyses. Equipment is installed as Category ll/l. No unreviewed safety question is created by this modification.
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CMP 95-1012 Rev. A Add Valve to RVLIS This modification adds a manual instrument isolation valve to the upper sensing tubing for the Reactor Vessel Level Indication System (RVLIS). Analyses
- performed for this design ensured that the integrity of the sensing line will be maintained and all postulated failures are bounded by existing accident analyses. No new accidents or malfunctions are created and no unreviewed safety question exists.
CMP 95-1017 Rev. A Install Mechanical Seals on Turbine Driven Auxiliary Feedwater Pump This modification intalls cartridge-type mechanical seals on the turbine-driven auxiliary feedwater pump (TDAFP). The pump manufacturer developed the new d: sign. This change will improve the reliability of the TDAFP. Failure of the n:w seal will result in consequences less severe than failure of the original cssembly. The use of mechanical seals could not credibly create the possibility of an accident of a different type than previously evaluated. No unreviewed safety question is created by this modification.
CMP 95-1025 Rev. A Change Motor Gear Set on TDAFP Trip Throttle Valvo l
This modification changes the motor gear set on the Limitorque operator for the Turbine Driven Auxiliary Feedwater Pump Trip Throttle Valve. The increase will improve the valve thrust performance in the degraded voltage condition. The change will not cause the analyzed stroke time of sixty seconds to be exceeded.
The gear set is a standard set available from Limitorque and is compatible with the operator in every way. No unreviewed safety question is created by the modification.
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CMP 96-1003 Rev. A Install Level Gage on Aux. Feed Pump Turbine Exhaust This modification installs a bypass assembly around the Auxiliary Feedwater Pump turbine exhaust drain steam trap and installs a level indicator on the exhaust h:ader. These changes are being made to aid operations personnel in determining if condensate is accumulating in the exhaust line and in draining it, if n cessary. The installation meets the design, material, and construction standards applicable to the areas impacted. The failure of the subject '
equipment could not create any credible accidents and no unreviewed safety question exists.
CMP 96-1004 Rev. A Install Pressure Gauge in RHR Discharge Piping This modification installs a local pressure gauge in the discharge piping of the Residual Heat Removal (RHR) System. The indicator wlIl provide additional information in identifying leakage into the RHR system. The pressure indicator, valve manifold, and tubing meet existing system design requirements. The system hrs been analyzed as acceptable considering this change. The RHR system is not cdversely impacted and remains capable of performing all of its safety functions, therefore, no unreviewed safety question is created.
CMP 96-1008 Rev. A Add Local Pneumatic Controllers to Steam Generator Atmospheric Dump Valves This modification adds pneumatic controllers to the 'B' and 'C' Steam Generator ctmospheric steam dump valves to provide an alternate means of operating these
. v lves in certain fire scenarios. This will eliminate the need to fire wrap approximately 500 feet of conduit in the Auxiliary Building. The new controllers and equipment will be qualified for safety-related applications.
The installation will not create an unreviewed safety question.
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CMP 96-1009 Rev. A Reroute Conduit to Allow Removal from Appendix R Fire Wrap Requirements This modification reroutes a conduit to Essential Service Water system valve EF-HV-0038. The new conduit routing will assure that a twenty foot separation is maintained between the conduit and the redundant train conduit. With the twenty foot separation, no fire wrap is needed to meet Appendix R guidelines. The rerouted conduit will be the same size as the original conduit, will have approximately the same length, and will be supported using approved conduit supports. Therefore, this change does not have any adverse impact on safety and does not result in an unreviewed safety question.
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l CMP 96-1014 Rev. A l Install isolation Valve on B Train ESW to Auxiliary Feedwater System This modification installs a single stainless steel butterfly isolation valve on the B train ESW supply to the Auxiliary Feedwater System. This modification will provide isolation for piping replacement which will be performed when the ESW system is functional. The new valve has no active safety function and only s::rves as pressure boundary for the ESW system. The installation will meet the original piping codes. There is no impact on the moderate energy line break analysis documented in FSAR Section 3.1.2. No unreviewed safety question is l created by this modification.
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CMP 96-1018 Rev. A Replace Leased Nitrogen Storage Tank With Larger Tank This modification involves the installation of a leased 9,000 gation storage trnk for liquid nitrogen at the site gas storage yard. This tank will replace 4
the existing 1,500 go!!on leased tank. The site gas storage yard is located ssfdy a..cy nom tne power block outside of the protected area. The new tank will be securely anchor bolted to a substantial concrete foundation. The remote location precludes adverse effects on safety-related equipment or control room personnel performance in the event of a tank rupture. No unreviewed safety question is created by this modification.
Rif: FSAR CN 96-048 RMP 89-2004 Rev. A R: move the High TDS Mixers SHF18A and SHF18B This modification approves the removal of the high TDS mixers and installation of a bubbler line that will be used to agitate the contents of the high TDS trnks. The systems involved are non-safety related. Failure of the bubbler system will riot impact any safety related system. No unreviewed safety question is created by this modification.
RMP 92-2002 Rev. A Emergency Eyewash Station Additions This modification approves the installation of two new eyewash / shower statilons and the relocation of a third to enhance personnel safety in areas where hazardous chemicais are stored and used. The new eyewash stations are to be located in the Turbine Building and the relocated station will remain in the D: min Building. These non-safety related system improvements do not affect the probability or consequences of occurrences of accidents previously evaluated in the FSAR and no unreviewed safety question is created.
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RMP 92-2008 Rev A Change the Seal Water Source on the Breathing Air Compressors This modification changes tha seal water source for the breathing air compressors to demineralized water and makes other minor changes to the breathing air piping system. These changes only impact equipment located in the 2000' elevation of the Control Building which is open to the Turbine Building.
The equipment is non-safety related and not credited in any accident analyses.
No unreviewed safety question is created by this change.
RMP 93-2014 Rev. B Replace Lighting Transformer QA15 and Add Seven Receptacles in Hot Lab This modification installs a transformer, panel, and receptacles in HP access cnd the Hot Lab. The changes have been evaluated for seismic impact, load flow impact, and fire protection impact and are compatible with the Callaway service power system. No unreviewed safety question is created.
RMP 93-2024 Rev. A Relocate Demin Water Supply to Vacuum Breaker Water Seals Standpipes This modification relocates the demin water supply valves to the condenser vacuum breaker water seal standpipe to a more accessible location. An overfill line will also be provided to prevent spillage. All systems associated with this modification are non-safety related and have no safety design bases. No unreviewed safety question is created by this modification.
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RMP 93-2033 Rev. A Remove Reactor Makeup Water From Hot Machine Shop Exhaust Scrubbers This modification removes non-safety related reactor makeup water piping and cssociated equipment connected to the hot machine shop exhaust scrubbers.' The reactor makeup water was not needed because the cleaning methods used in the hot ;
machine shop did not use acids or caustics as originally designed. No unreviewed safety question is created.
Ref: FSAR CN 96-028 RMP 93-2034 Rev. A l l
Removal of Circulating Water Mini Corrosion Monitoring Skid !
In order to provide more accuate fouling factor in the high pressure condenser, l connections will be provided in the circulating water system for a sophisticated l monitoring device which will replace the existing mini corrosion monitor. There is no impact on safety related systems and equipment. No unreviewed safety question is created by this modification.
RMP 93-2036 Rev. A Install Restricting Orifices in Compressed Air System This modification provides for the installation of a flow restricting orifice in l c 1 inch air line to prevent excessive air loss in the event of a rupture in the line. The orifice is to be installed in the non-safety related portion of the Compressed Air System. The affected portion of the compressed air system serves no safety function. The installation will not create any adverse system interactions nor any unanalyzed modes of operation. No unreviewed safety question is created by this modification.
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1 RMP 94-2006 Rev. A l l
R: route Neutralization Tank Effluent Line This modification reroutes the effluent of the neutralization tank to the equalization basin. ' This new routing will not involve any safety related buildings or any equipment important to safety. A failure of this piping will not jeopardize the operation of any equipment important to safety. No unreviewed safety question is created by this modification.
R::f: FSAR CN 95-031 RMP 94-2014 Rev. A
] Technical Support Center Modification This modification will change the space allocation within the Technical Support Crnter (TSC) and make minor electrical changes. Some interior walls will be r: moved and partitions will be added to create an Operational Support Center (OSC) within the TSC building. Space allocations will remain within the guidelines of NUREG-0578 and NUREG-0694, therefore, this modification will not
, h:ve any effect on any plant operation or equipment. Provisions are in place to provide any necessary computer generated data to the staff at the TSC should this facility need to be activated during the construction phase of this project. None of the concrete walls or roof will be affected in any way, therefore, their purpose and function remain unchanged. No unreviewed safety question is created by implementing this modification.
- RMP 94-2016 Rev. A
- Main Feedwater Pump Electronics Cabinet Cooling This modification installs a cooling unit in the main feedwater electronics control cabinet which will ensure that the internal temperature does not exceed 90 F. This will reduce the failure rate of the components in the cabinet. This modification does not impact any safety related equipment. No new accdent initiators are created. This modification does not create an unreviewed safety question.
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RMP 95-2001 Rev. A Add a New Intake Deep Well and Crosstie it to Intake Lube Water System This modification will install a 1200' deep well at the intake structure drawing from the Potosi aquifer capable of a 200 gpm supply to the lube water system for the intake pumps.This modification does not affect systems, stuctures, or components that prevent or mitigate accident situations. No unreviewed safety question is created by this modification.
Ref: FSAR CN 95-008 FSAR CN 97-001 RMP 95-2008 Rev. A isolation Valves for Steam Generator Nitrogen Blanketing This modification installs isolation valves on the nitrogen supply header for crch steam generator. The lines are non-safety related and are not part of any cccident analysis. The installation will meet original design, material, and construction standards. No unreviewed safety question is created by this modification.
RMP 95-2011 Rev, A Replace LP Heaters 2A and 2C This modification will replace the 2A and 2C Low Pressure Feedwater heater tube bundles with a new U-tube design. This design will replace a straight tube d sign which had fixed tube sheets on both ends. This modification will not affect or interface with any safety related components or systems and does not h;ve any safety design bases associated witn the portions of systems affected by the modification. No unreviewed safety question is created by this modification.
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RMP 95-2015 Rev. A R: move Low Flow Switches on Plant Heating System This modification removes the the plant heating system pump discharge flow switches. Removal of the switches will cause some increased probability of pump .
4 d: mage due to low flow, however, multiple flow paths in the system make the possibility of deadheading the pumps very low. The plant heating system serves no safety functions. No unreviewed safety question is created by this change.
Temporary Modification 95-M020 approved blocking the switches with a similar safety evaluation. This modification will allow removal of TM-95-M020.
RMP 96-2002 Rev. A Ultimate Heat Sink Heaters and Controls This modification will change the controls for the Ultimate Heat Sink sump h aters by moving the thermostats out of the sumps and onto a nearby wall in the cooling tower basin. The modification will also eliminate the level controls for the heaters. The sump heaters will also be lowered to below thre technical specification minimum UHS level and a thermostat bypass switch will be added.
These changes will enhance the reliability of the sump heating system. The sump h aters are not safety related but are powered from safety related MCCs. The safety related MCCs are isolated by the breakers at the MCC. The function of the heaters is unchanged. No unreviewed safety question is created by this modification.
OL-1110 Diesel Generator Fuel Oil Storage Tank / Day Tank Minimum Volumes This amendment application would revise Technical Specification 3/4.8.1.1, "A.C.
Sources Operating," and 3/4.8.1.2, "A.C. Sources Shutdown," to change the minimum required storage volume of the Emergency Fuel Oil (EFO) storage tank from 85,300 gallons to 80,400 gallons and the EFO day tank from 390 gallons to 510 gallons. This change was approved by NRC via Amendment 100 dated July 6, 1995.
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OL-1116 Miscellaneous Technical Specifications Bases Changes i
This amendment application requests changes to the Bases for several Technical l Specifications. The revised Bases are in accordance with previously approved evaluations. No unreviewed safety question is created by this change.
Approved by NRC letter dated February 1,1996.
1 l
OL-1140 Tcchnical Specification Bams Change for 3/4.1.1.3 This amendment requeu changes the Bases for Technical Specification 3/4.1.1.3, Moderator Temperabre Coefficient. This change does not involve an unreviewed safety question.
1 Approved by NRC via letter dated February 1,1996. '
I l
OL-1148 I l
Technical Specification 3/4.9.4 Bases Change i
Change the Bases for Callaway Technical Specification 3/4.9.4, Containment Building Penetrations. The Bases for the Containment Purge Isolation Signal (CPIS) setpoint from containment purge radiation monitors GT-RE-22 and GT-RE-33 is changed to correct a description error carried over from a calculation. No s tpoints are changed.
The Bases change does not involve an unreviewed safety question because it does not involve any design changes nor does it affect the probability of any event initiators. The change is, in essence, an editorial change to make the verbal d;scription of the setpoint basis match the calculational input which is not in question.
Approved via NRC letter dated February 1,1996. ,
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OL-1151 Tcchnical Specification Split This amendment request changes the Technical Specifications (TS) to implement NRC's Final Policy Statement on Technical Specifications Improvements for l Nuclear Power Reactors published in the Federal Register on July 22,1993 (58 FR l 39132). These improvements involve focusing the TS on those requirements that tre of controlling importance to operational safety by screening each TS section using the criteria provided in the policy statement. The purpose of the cmendment request is to relocate the specifications that do not meet any of the four criteria on regulatory requirements related to inclusion in the TS. The relocated specifications will be moved to Chapter 16 of the Final Safety Analysis Report.
These changes were approved by NRC via Amendment 103 on October 20,1995.
Ref: FSAR CN 95-006 OL-1153 Removal of Containment Isolation Valve Table from Technical Specifications I l
This amendment request revises Technical Specifications (TS) 3/4.6 by: )
(c) removing TS Table 3.6-1, Containment Isolation Valves, from the TS and placing it in the Final Safety Analysis Report in accordance with NRC Generic ,
Lctter 91-08, " Removal of Component Lists from the Technical Specifications;"
(b) extending the allowed outage time from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for motor operated isolation valves associated with component cooling water flow to the reactor coolant pumps; and (c) allowing containment penetration check valves to be used as isolation d vices, consistent with NUREG-1431, " Standard Technical Specifications -
Westinghouse Plants."
These changes were approved by NRC via Amendment 113 on June 28,1996. !
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OL-1154 l
R vise Technical Specifications /COLR -Implement Burnup Dependent Penalty Factor This amendment application revises Technical Specifications 4.2.2.2,4.2.2.4, i
- cnd 6.9.19 to incorporate a penalty in the Core Operating Limits Report (COLR) l to account for heat flux increases greater than 2% between monthly surveillance l measurements. This change was approved by NRC via Amendment 101 c.7 July 20, i 1995.
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OL-1156 Technical Specifications-Move ORC /NSRB to FSAR Chapter 13/ Organization l Changes This amendment request changes the Technical Specifications (TS) to relocate the review and audit requirements of the On-site Review Committee (ORC) and Nuclear Safety Review Board (NSRB) contained in TS 6.5.1,6.5.2, and 6.5.3 to the Operating Quality Assurance Manual (OQAM). In addition, TS 6.2.3 is changed to d:lete reference to the Manager, Nuclear Safety and Emergency Preparedness. -l These changes were approved by NRC via Amendment 107 on December 26,1995.
OL-1162 Auxiliary Feedwater Surveillance Frequency I This amendment decreases the frequency of testing the Auxiliary Feedwater pumps l from once per 31 days to once per 92 days. This change also provides an l cxemption from Specification 4.0.4 for entry into Mode 3 for response time l t: sting and auto-start testing of the Turbine Driven Auxiliary Feedwater Pump (TDAFP) and removes the requirement to perform TDAFP auto-start tests only during shutdown. The Bases for 3/4.7.1.2 are also changed to clarify Auxiliary Feedwater Pump performance parameters.
Many of the changes are endorsed by NRC Generic Letter 93-05 and NUREG-1431, Westinghouse Standard Technical Specifications. The changes have been evaluated end no unreviewed safety question is created by the Technical Specification changes. ,
These changes were approved by NRC via Amendment 108 on March 11,1996.
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OL-1165 Relocate Technical Specification Tables 3.3-2 and 3.3-5 to FSAR Chapter 16 This amendment relocates Technical Specification tables 3.3-2, Reactor Trip System Instrumentation Response Times," and 3.3-5, " Engineered Safety Features Response Times," to FSAR Chapter 16, section 16.3. The Bases discussion
, specific to table 3.3-5 is also to be relocated to FSAR Section 16.3.
The NRC has already implemented this line-item Technical Specification improvement in the new Standard Technical Specifications (NUREG-1431 for l Westinghouse Plants). This proposed amendment is taken directly from NRC l Generic Letter 93-08.
Overall protection system performance will remain within the bounds of the accident analyses since no changes to the response times or measurement interval are proposed. No unreviewed safety question is created by this amendment.
This change was approved by NRC via Amendment 104 on November 22,1995.
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OL-1167 1
Modifications to Overpower Delta-T and Overtemperature Delta-T l
This amendment application proposes changes in the Total Allowance (TA), K4, K6, and Allowable Value in Technical Specification Table 2.2-1 in order to reduce i repeated alarms and partial reactor trips associated with the C-4 interlock and l OP Delta-T system. These changes have been justified through the use of available margin. Overall protection system performance will remain within the bounds of the accident analyses documented in FSAR Chapter 15. The Safety Analysis Limit for OP Delta-T remains unchanged, thus preserving the overpower safety margin.
1 The editorial change to the Overtemperature Delta-T equation in Technical Specification Table 2.2-1 represents the manner in which these signals have always been electronically processed. The Bases changes discuss the correlation b: tween time constants for lag cards in the 7300 Process Protection System and the total lag time assumed in safety analyses which credit Overtemperature Delta-T and Overpower Delta-T for Protection.
These changes do not create an unreviewed safety question and were approved by NRC via Amendment 102 on August 21,1995.
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OL-1168 Control Room Emergency Ventilation System This amendment application includes changes to the Surveillance Requirements of Tcchnical Specification 3/4.7.6 that would reduce the upper limit on the flow rtte through the control room filtration system and would adopt ASTM D-3803-1989 cs the laboratory testing standard for control room filtration and control building pressurization charcoal adsorbers. The Bases are revised accordingly.
Overall protection system performance will remain within the bounds of the tecident analyses documented in FSAR Chapter 15, WCAP-10961-P, and WCAP-11883 since no hardware changes are proposed. The change to the control room filtration flow rate is consistent with the original licensing basis and will ensure an average atmospheric residence time of greater than or equal to 0.25
- seconds. Revisions to the Surveillance Requirements for the CREVS will ensure that the control room dose assumptions made in support of OL Amendment 96 are vnlid.
. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of these changes. The new charcoal adsorber sample testing laboratory prorocol is more stringent than the current tasting practice and more accurately demonstrates the required performance after a design basis LOCA. The proposed amendment does not involve an unreviewed safety question.
This change was approved by NRC via Amendment 106 on December 20,1995.
i OL-1169 Reduced Surveillance Testing at Power (GL 93-05)
This amendment application revises the Technical Specifications (TS) to implement the recommendations of NRC Generic Letter 93-05, "Line-Item Technical Specifications improvements to Reduce Surveillance Requirements for Testing During Power Operation." In addition, TS 4.1.1.1.1 and TS 4.1.1.2 related to shutdown margin and TS 3/4.1.3.1 related to movable control assemblies are rcvised to implement portions of NUREG-1431, " Standard Technical Specifications
- Westinghouse Plants." These changes were approved by NRC via Amendment 105 on December 7,1995.
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OL-1170 l Emergency Diesel Generator Technical Specifications Change l
This amendment application revises Technical Specification 3/4.8.1 and the
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cssociated Bases to improve overall emergency diesel generator reliability and 1 l
cvailability. These changes incorporate recommendations and suggestions from NRC Generic Letter (GL) 93-06, "Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements During Power Operation;" NUREG-1431,
" Standard Technical Specifications - Westinghouse Plants;" Regulatory Guide 1.9, Revision 3, " Selection, Design, Qualification, and Testing of Emergency Diesel 4 Generator Units Used as Class 1E Onsite Electric Power Systems at Nuclear Power Plants;" and GL 94-01, " Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators from Plant Technical Specifications." l These changes were approved by NRC via Amendment 112 on June 17,1996. )
OL-1171 SFP Region 1 Criticality Analysis
. This amendment request revises the Technical Specifications (TS) to aliow for storage of fuel with an enrichment of up to 5.0 weight percent U-235, subject to c:rtain integral fuel burnable absorber (IFBA) requirements or discharge exposures. Plant operation using the higher enriched fuel will be demonstrated i to be acceptable by cycle-specific reload safety evaluation prior to each fuel
! loading. These changes were approved by NRC via Amendment 109 on April 30, i
1996.
Ref: FSAR CN 95-042 OL-1172 Appendix J, Option B This amendment application revises the Technical Specifications (TS) to permit implementation of 10 CFR 50, Appendix J, Option B. This change adds the
" Containment Leakage Testing Program" to the TS. These changes were approved by NRC via Amendment 111 on May 28,1996.
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OL-1173 Containment Air Locks This amendment application revises the Technical Specifications (TS) to allow the containment personnel airlock doors to be open during core alterations and movement of irradiated fuel in containment. This change was approved by NRC via 4- Amendment 114 on July 15,1996.
OL-1175 ZlRLO Cladding for Fuel Rods i
This amendment application revises Technical Specification 5.3.1 to allow the -
' use of ZlRLO clad fuel rods and ZlRLO filler rods. The ZlRLO fuel is described in topical report WCAP-12610, " Vantage + Fuel Assembly Reference Core Report."
This change was approved by NRC via Amendment 110 on April 30,1996. ,
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OL-1176 MSFIS Modification Technical Specification Change This amendment application revises Technical Specification 3/4.3 to support a 4 modification to replace the digital portions of the main steam and feedwater isolation system (MSFIS) with digital processor equipment and authorizes the revision of the FSAR to include a description of the MSFIS modification. These changes were approved by NRC via Amendment 117 on October 1,1996.
- These changes were implemented via Callaway Modification Package 92-1038, Rev.
A.
Ref: CMP 92-1038 Rev. A FSAR CN 96-023
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OL-1178 Steam Generator Tube Sleeving (Westinghouse)
This amendment application revises Technical Specification 3/4.4 and the associated Bases to address the installation of laser welded tube sleeves in the steam generators. The laser welded sleeves are designed by Westinghouse Electric Corporation. The details of sleeve qualifications are discussed in rcport WCAP 14596, " Laser Welded Elevated Tubesheet Sleeves for Westinghouse '
Model F Steam Generators." This change was approved by NRC via Amendment 116 on October 1,1996.
OL-1182 Charcoal Test Methodology for Emergency Exhaust System Filter Adsorbers This amendment application revises Technical Specifications 3/4.7.7 and 3/4.9.13 and associated Bases. Specifically, the acceptance criteria for the testing of carbon samples from the auxiliary / fuel building emergency exhaust system filter adsorbers would be changed to adopt ASTM D-3803-1989 as the laboratory testing I standard. These changes were approved by NRC on November 13,1996.
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4 CTP-BB-06320 Rev. O Chemical Addition to Pressurizer This procedure allows the addition of hydrazine into the pressurizer (PZR) to control and scavenge the oxygen that is entrained in the PZR liquid prior to plant startup. This procedure revision is necessary to allow the plant to climinate water-solid operation. Use of this procedure wik not adversely affect any plant system response to design basis accidents. No unreviewed safety question is created.
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CTP-EJ-06040 Rev. O Chemical Addition to the Residual Heat Removal System This procedure allows the addition of hydrazine into the Residual Heat Removal (RHR) system to control and scavenge the oxygen that is entrained in the RHR system liquid prior to plant startup. This procedure also allows the addition of hydrogen peroxide into the RHR system to provide chemical degas and forced oxidation of the reactor coolant during plant cocidown. This procedure revision is necessary to allow the plant to eliminate water solid operation. Use cf this procedure will not adversely affect any plant system response to design basis cccidents. No unreviewed safety question is created.
ETP-AE-ST004 Rev. O High Pressure Feedwater Heater Bypass Test This special test will determine plant response to operating with the high pressure heater bypass valve throttled open. Data taken during the test will be used to evaluate whether operation in this manner should be incorporated into normal plant procedures.
After completion of this test, RFR 16653 Rev. A approved long term operation in the new configuration. It has been concluded that extended operation in this I manner will not introduce any additional failure modes or increase the consequences of any accidents analyzed in the FSAR. No unreviewed safety )
question is created by this change.
Ref: RFR 16653 Rev. A
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ETP-AE-ST008 Rev. O R: pair and Retest of Feedwater Isolation Valve AE-FV-0042 This procedure provides guidance to repair and retest the standby train hydraulic manifold on one of the four feedwater isolation valves while the plant is in Mode 1. This procedure will require entry into the Action Statement of Technical Specification 3.7.1.6. This activity does not permanently modify the configuration of the valve. The current accident analysis as documented in the FSAR remains valid. The work will be performed within the requirements of a Technical Specification allowed outage time and will be a short duration. This activity does not create an unreviewed safety question.
ETP-BB-01339 Rev. O Steam Generator in-Situ ' Pressure Testing This procedure provides guidance for in-situ pressure testing of steam generator tubes. This testing verifies that defective tubes meet the minimum acceptable structural limits imposed by Regulatory Guide 1.121. After testing, affected tubes are repaired in accordance with Technical Specification 4.4.5. The repair of the tubes will restore them to a structurally acceptable configuration, therefore, no unreviewed safety question is created.
ETP-BG-ST015 Rev. O Letdown Heat Exchanger Flow Test This evaluation allows a flow test to gather data and measure performance of the Letdwn Heat Exchanger. The test will have no adverse affect on plant equipment required to safely bring the plant to hot or cold shutdown. During the test, CCW will still be operated within its normal temperature range. If a train of CCW is rendered inoperable by throttling one of the CCW heat exchanger isolation v:lves, the train can be restored to operable status by the Equipment Operator stationed at the heat exchanger. This test will not cause any unanalyzed event.
No unreviewed safety question is created by this test.
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ETP-GL-ST004 Rev. 0 Auxiliary Building Special Negative Pressure Test l This evaluation examines the conduct of a special negative pressure test with I the Auxiliary Building and Fuel Building suction dampers in the open position to confirm operability of the system in the specified configuration. No automatic
, functions will be impacted by the special test. Should an SIS or FBVIS occur, cll components would align to their accident positions. Both the Fuel and ,
Auxiliary buildings will be maintained at a negative pressure during the test. I No unreviewed safety question is created by this procedure. i ETP-MB-ST003 Rev. O Main Generator Excitation 100% Load Stability Test ,
1 This test is performed to determine the stability of the main generator
- cxcitation control system in response to step disturbances at or near full load.
l No generator or excitation relaying or protection will be disabled for the tcsting and no operating limits for the generator will be exceeded. During all
- t: sting, all changes to the operating voltage level of the exciter will be made
. by a licensed operator stationed at the main control board. Since no changes to protective relaying is being made, the possibility of a generator / turbine trip are not increased. No unreviewed safety question is created by this change.
4 ETP-ZZ-01310 Rev. 7 Steam Generator Nozzle Dam Installation and Removal 4
This procedure allows the installation of nozzle dams in all four steam g:nerators simultaneously. In 1992, an evaluation of the use of nozzle dams in all four steam generators was documented under procedure ETP-ZZ-01310, Rev. 4.
This was reported to NRC in the June 12,199210 CFR 50.59 Summary Report as "ETP-ZZ-01210, REV4" (sic). The safety evaluation for ETP-ZZ-01310, Rev. 7 consisted of a review of the procedure and the previously reported safety evaluation for applicability. It was concluded that the safety evaluation performed for Revision 4 of the procedure also applied to the latest revision and that no unreviewed safety question is created.
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ETP-ZZ-ST006 Rev. 4 B:nk Reactivity Worth Measurement (Rod Swap Method)
This evaluation takes credit for a similar evaluation performed to support R vision 1 of the procedure. Shutdown Margin during all phases of rod swap will l always be greater than the lowest shutdown margin during conventional bank worth measurements (boron dilution). It is not possible for rod swap testing to involve a bank configuration which represents inadequate shutdown margin. No I unreviewed safety question is created by this procedure.
t l i ETP-ZZ-ST019 Rev. O Plant Re:dio Tests l
This evaluation ellows plant radio tests to be conducted to determine if Health Physics Spread Spectrum 900 Mhz radios adversely impact plant equipment. All radio testing, except one test, will be performed with the core offloaded and both SSPS trains in " test". The radio test for the Rod Contro! Motor Generator 1 S:t Room. This test must be performed when the rod drive mechanism is energized. This portion will be performed when rod drop testing is being p rformed. During that test, only one rod bank at a time pulled, the plant is not critical, and sufficient shutdown margin exists to keep the plant shutdown.
No unreviewed safety question is created by this test.
4 1 HTP-EC-00001 Rev. 0 '
Decontamination of Fuel Pools With Skimmer System This evaluation provides a means for underwater vacuum of the Cask Loading Pit, the Fuel Transfer Canal (FTC), the Spent Fuel Pool (SFP), or the Refuel Pool (RP) using a hose connected to the Fuel Pool Skimmer System. This hose and associated attachments will be used to vacuum the floor of the respective pool.
The concern with this evolution is with the potential to drain the pool and expose irradiated fuel. This evaluation shows no increased potential for pool draining. Personnel will be in constant attendance during the vacuuming evolution. No unreviewed safety question is created by this procedure.
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i Liquid Radwaste Release Permit i
i: This evaluation applies to_a procedurally controlled temporary modification which installs a braided stainless steel hose on a radiation monitor drain which 1
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is routed to the suction side of a Discharge Monitor Tank pump. This i
- installation involves the Liquid Waste System. The affected portion of the system is non-safety-related. Failure of the installation will not result in
1::ak rates greater than those evaluated in the FSAR. No unreviewed safety
. question is created by this temporary modification.
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OTG-ZZ-00001 Rev. 20 Plant Heatup - Cold Shutdown to Hot Standby This procedure revision is primarily associated with drawing a steam bubble in the pressurizer (PZR) prior to starting the reactor coolant pumps (RCP's), thus avoiding solid plant operations. Drawing a PZR steam bubble prior to RCP operation lowers the potential for overpressurization of the primary system with the added benefits of reducing RCP starts and the total time required to startup the plant. These operating changes do not adversely impact plant safety and no j unreviewed safety question is created. l OTG-ZZ-00006 Rev.17 '
Plant Cooldown - Hot Standby to Cold Shutdown This procedure revision is primarily associated with maximizing the pressurizer (PZR) steam bubble, thus avoiding solid plant operations. Maximizing a PZR steam bubble lowers the potential for~overpressurization of the primary system with the added benefits of reduced radwaste generation and increased operational ficxibility. These operating changes do not adversely impact plant safety and no unreviewed safety question is created. !
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OTN-BB-00006 Rev. 2 R: actor Coolant Degassification This procedure change allows the addition of nitrogen into the pressurizer (PZR) to dilute hydrogen that will come out of solution during a normal plant cooldown without requiring water-solid conditions in the RCS. The vacuum degassification skid is to be used to remove the gaseous mixture from the PZR. Use of this procedure will not adversely affect any plant system response to all assumed
. d: sign basis accidents. Therefore, use of the procedure will not create an i unreviewed safety question. I l
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l RFR 04272 Rev. D l I
Operability of Electrical Penetration Room Equipment With Cooler Out of S::rvice '
This RFR evaluates the operability of equipment in the electrical penetration rooms during periods when the associated room coolers are out of service. The '
t:mperature limit of FSAR Chapter 16 (106 Degrees F) applies to the Equipment ,
Qualification life of the affected components while the maximum temperature of H 131 Degrees F is an immediate operability limit. Calculated peak accident t:mperatures remain less than or equal to the FSAR limit of 131 Degrees F. If the 106 Degrees F limit is approached while the room coolers are out of service, cfforts will be undertaken to provide additional sources of cooling to the room
- and/or reducing the room heat load. This change does not create an unreviewed safety question.
- RFR 09780 Rev. F Allow 6" Penetration Between Auxiliary and Control Buildings This RFR evaluates the acceptability of having a 6" penetration between the Auxiliary and Control buildings to support the installation of conduit for Modification 92-1014 (Swing Battery Charger Modification). An analysis showed j that the 6" diameter penetration could be made without impacting the operations of the HVAC equipment for the Auxiliary or Control Buildings nor causing excessive leakage of unfiltered air into the Control Building. No unreviewed safety question is created by this change.
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l RFR 10372 Rev. B I 2
Replace Manual Head Vent Valves with New Type This evaluation approves the replacement of two manually operated 1" diameter
, globe valves located in the reactor head vent piping with an alternate model.
)
The valve replacement is expected to eliminate problems previously encountered '
when venting low pressure nitrogen cover gas during refueling operations. The !
I replacement valves are ASME Section lli Class 1 valves and meet existing design requirements. The piping system has been reanalyzed and found acceptable with this change. The replacement valves will continue to perform their passive pressure boundary function during an accident condition, therefore, no
, ur reviewed safety question exists.
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RFR 10753 Rev. C Evaluate Fire Brigade Storage Room in Communications Corridor This RFR approved the addition of a fire brigade storage room in the Communications Corridor on the 2061'_6" elevation. The room will provide j storage space for equipment and supplies necessary for fire fighting. This l change will result in a change to the combustible loading of the area. The i combustible loading will remain less than 50,000 Btu / square foot and will not adversely affect the adjacent Control Building. There is no safety related i equipment in this area of tha plant. This change does not create an unreviewed sifety question.
RFR 13877 Rev. B Cavity Return Temperature Alarm Setpoint This evaluation approves the change in the alarm setpoint for the cavity return temperature alarm from 240 F to 300 F. This evaluation covers the effects on i
the concrete only. Concrete c ass limits will not be exceeded. No unreviewed i snfety question is created by tr,:s change.
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e RFR 13877 Rev. C i
Evaluation of increased Cavity Temperatures on Excore Instrumen:ation !
This evaluation addresses the increase in the alarm setpoint for the reactor I cavity return temperature sensor End its effect on power range nuclear instrumentation. The increase to 275 F due to the trend of increasing temperctures has no effect on the power range detectors. The trend is not reflected on any of the detector well eWaust duct temperature indicators. The consequences of the increasing temperature trend at the cavity exhaust (GN-TE-0049) is not demonstrable on any of the power range detectors or their l cssociated triaxial connectors and no unreviewed safety question exists.
RFR 13877 Rev. D l Cavity Return Temperature Increase This RFR raises the alarm point for GN-TE-0050 from 150 F to 160 F and lowers the alarm points for GN-TE-0051,52,53, and 54 from 140 F to 130 F. The rcquirements for aligning ESW to the containment coolers in OTN-GN-00001 will now be based on a GN-TE-0050 temperature of 165 F or a temperature of 135 F from either GN-TE-0051, 52, 53, or 54. The increase in the alarm piont for GN-TE-0050 due to an increasing air temperature trend has no detrimental effects on the concrete stresses. No unreviewed safety question is created by these changes.
RFR 14063 Rev. B Evaluate Replacement of Obsolete Seismic Instrumentation This equivalent material evaluation approves the replacement of existing obsolete seismic instrumentation with a more modern, computerized system that uses RAM and computer disks for data storage and recording. The seismic instrumentation is only used to evaluate the strength of seismic activity at Callaway Plant. The equipment does not interface with any safety related cquipment. The equipment will be capable of performing all of the functions of the existing system. No unreviewed safety question is created by this change.
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RFR 14132 Rev. B l l
Evaluate Equipment Storage in Hot Machine Shop .
1 This evaluation examines the storage / staging of transient combustibles in the Hot Machine Shop. The additional combustible loading will remain low and will J not affect any safety related equipment. No seismic concerns are raised by this change and no unreviewed safety question is created.
Ref: FSAR CN 96-016 RFR 15218 Rev. A Incorporate FSAR Change Notices into CEL EQ Data This evaluation involves the correction of room numbers and their associated Accident Categories and Environments for several safety related components listed in FSAR Table 3.11(B)-3. No change to the physical locations of any equipment were made and the equipment is environmentally qualified for its installed locations. No unreviewed safety question is created by this evaluation.
RFR 15455 Revs. A & B Add Cooling Water to RWST This evaluation allows the use of chilled water to cool the RWST during warm weather operation. The chilled water is to be routed through existing heating coils on the exterior of the RWST via hoses from the Integrated Leak Rate Test after cooler. This does not affect the safety related portion of any system and will not prevent any safety related system from performing its function. No unreviewed safety question is created by this evaluation.
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I RFR 15642 Rev. B l Chemical Degas of the RCS During Cold Shutdown ,
This evaluation examines the effects of chemically degassir g the Reactor Coolant System during plant shutdown. This is performed in Mode 5 and involves the reduction in RCS hydrogen concentration through reaction with hydrogen peroxide.
Using this method to remove dissolved hydrogen does not adversely affect the RCS or any of its supporting safety related systems. Plant response to a design basis accident is also unaffected. Therefore, no unreviewed safety question is created by this activity.
RFR 15847 Pcv. 5 Revision to Cycle 6 COLR to Account for Axial Offset Deviation This safety evaluation applies to the Cycle 8 COLR, Rev. 2, which incorporates additional W(z) factors to be used when the axial offset is more than 3% more 4
positive than predicted. The cycle 8 Reload Safety Evaluation remains valid for this COLR revision. The cycle 8 reload design satisfies all of the applicable safety parameter limits and acceptance criteria and has been evaluated using j standard reload design and approved fuel rod design models and methods. No j unreviewed safety question is created by this change.
- RFR 16159 Rev. A Substitution of Alternate Readout for Failed Wide Range Temperature Channel This evaluation approves the use of the core cooling monitor wide range temperature indication for use as wide range T-hot and T-cold Post Accident Monitoring Indication (TS 3.3.3.6 - Table 3.3-10). The core cooling monitor d
- sign meets the qualification requirements for Category 1 design. The core cooling monitor is already used for readout of category 1 information required by item 17 on the same Technical Specification Table. There are no reductions in Technical Specification margins. A human factors review concerning the readout location, type, and accessibility during an accident has been conducted.
No unreviewed safety question is created.
Ref: Procedure OSP-SH-00001
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RFR 16196 Revs. B & C Remove Insulation on Residual Heat Removal Pipe Flanges This evaluation examines the effects of permanently removing the insulation from the Residual Heat Removal (RHR) pumps and approximately an approximately 8 ft ,
srction of piping. The* additional heat load in each RHR pump room can be cccomodated by the existing room coolers and will not result in exceeding Technical Specificatic.1 temperature limits. The impact on the capacity of the Ultimate Heat Sink is negligible. Plant response to a design basis accident is unaffected. Therefore, this change does not result in an unreviewed safety question.
RFR 16297 Rev. C Tcmperature Correction Chart for Feedwater Isolation Valves This RFR approves the use of a " minimum operable pressure" graph and table for the FWlVs. This graph was supplied by the Valve / Actuator vendor. The approval of the graph does not change the functional or operational requirements of the
. FWlVs. All current accident analyses remain valid. Use of the information will not cause any unanalyzed event or reduction in safety. No unreviewed sifety question is created by this change.
RFR 16337 Revs. A & B Sediment in Diesel Generator Fuel Oil Storage Tank These evaluations address on-line filtering of the diesel fuel oil storage tanks to satisfy the technical specification requirement to remove accumulated stdiment from the storage tanks. These evaluations clarify that when the i storage tanks are drained per the 10-year surveillance requirement, an internal inspection and cleaning (sediment / microbiological growth) are performed if required. On-line filtering does not satisfy the TS requirements to drain the i
trnks and clean for microbiological growth. No unreviewed safety question is created by this change.
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l RFR 16363 Rev. A Evaluate the Main Turbine Vacuum Instrumentation l This RFR revises the C-9 (condenser available interlock) setpoint and evaluates the main turbine vacuum instrumentation. This evaluation takes credit for a similar evaluation performed for CMP 84-899A. This change does not affect any s fety design and will not result in an unreviewed safety question.
1 Ref: CMP 84-899A l
RFR 16366 Rev. A l Draining of CUNO Filter Housing This evaluation allows for the installation of a design change to the CUNO filter housing (FHF10) to use Service Air as the motive force and transfer water 3
out of the filter housing prior to a filter changeout. Using service air will not lead to overpressurization of the Discharge Monitor Tanks. The maximum cctivity will not exceed that of the tanks previously analyzed in the FSAR.
3 This change does not impact any safety related system and no unreviewed safety question is created.
RFR 16370 Rev. A Acid Metering Pump Removal This avaluation allows a change in the method that acid is injected into the Discharge Monitoring Tanks (DMTs). The existing pump requires significant cttention because of acid solidification in the check valve. No safety related equipment nor equipment associated with any accident analyses are affected by this change. No unreviewed safety question is created.
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RFR 16470 Rev. F Evaluation of Alternate Fasteners for RCP Seal Housing This evaluation examines the acceptability of replacing carbon steel fasteners on the Reactor Coolant Pump (RCP) number 1 seal housing with inconel fasteners.
The inconel studs will be tensioned within the code allowable stresses and will provide better corrosion resistance in a boric acid environment. This change meets all ASME requirements and no unreviewed safety question is created.
RFR 16475 Rev. A l
Containment Fire Protection During Outages !
This evaluation determined that for a limited scope outage with limited work in i the reactor building, it is permissible for the Fire Protection Engineer to rsview the work scope and determine the fire protection requirements in lieu of rcinstalling all extinguishers and fire hoses in containment. This does not conflict with FSAR Section 9.5B Hazard Analysis or FSAR Section 3.1.3 Criterion 3 - Fire Protection. No unreviewed safety question is created by this change. l RFR 16532 Rev. A Use Studs and Nuts to Secure RCP Main Flange Joint The proposed change allows the use of studs, nuts and washers in lieu of bolts to close the main flange joint on the reactor coolant pumps. The material and cross sectional area of the studs is the same as the bolts. A standard heavy h::x nut is utilized in lieu of the bolt head. Hardened washers will be utilized under the nut to aid in tensioning and to minimize preload losses. The use of washers is allowed by the ASME Code. No unreviewed safety question is created by this change.
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RFR 16569 Rev. A Remove Valve HC-V-0226 and Cap Line.
This evaluation approves the removal of valve HC-V-0226 and capping of the line. l The valve is in a non-safety related system and was designed to be a flush / test !
connection. The cap will be welded in place of the valve. Since the valve is no longer used, it can be removed. No unreviewed safety question is created by this change.
I RFR 16675 Rev. A i install Blind Flange Downstream of LE-RE-0059 l This evaluation approves the installation of a blind flange downstream of ,
radiation monitor LE-RE-0059 to prevent an inadvertant unmonitored release of l unmonitored effluent. The flange is to be installed in the oily waste system which is not safety related. The portion of the system is not used. This change does not create an unreviewed safety question.
RFR 16678 Rev. A Install Bypass Line in Secondary Waste System This RFR approves the installation of a bypass line and isolation valve around a check valve in the Secondary Liquid Waste (SLW) system. This would allow the transfer of the SLW collection tank fluid to the Hi TDS tanks for neutralization in the event that the SLW fluid pH is out of tolerance. These systems are non-safety-related and failure of these systems is bounded by current analyses.
Failure of these systems does not impact any safety-related system. No unreviewed safety question is created by this change.
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RFR 16737 4
Excess Letdown Replacement i
This RFR addresses the changes made to the excess letdown line to correct the
, configuration that led to an October 1995 leak. Revisions B and C of the RFR 4
cddress changes to the configuration of the excess letdown piping and approve .I
- operation in all modes with the new configuration. The new configuration has i
- been reanalyzed and stresses were found to be acceptable. No unreviewed safety question is created by this change.
RFR 16775 Rev. A Eliminate Seat Drains on Trip and Throttle Valve 4
This evaluation addresses the acceptability of eliminating the above and below
, stat drains associated with the Turbine Driven Auxiliarv Feedwater Pump Trip and Throttle Valve. Both drains are not necessary to assure reliable operation of j the Auxiliary Feedwater Pump Turbine. Elimination of the drains will reduce the l load on the turbine exhaust drainage system. No unreviewed safety question is l created by this change. l 1
1 RFR 16790 Rev. B Provide Guidance on Disabling P-12 Interlock This evaluation provides guidance on defeating the P-12 interlock, allowing use of all 12 steam dump valves for cooldown prior to placing RHR in service. The plant must be in Mode 4 or lower with the RCS borated for cold shutdown conditions. All potential effects are bounded by existing analyses, including the main steam line break analysis. No unreviewed safety question exists.
Ref: OTG-ZZ-00006
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RFR 16896 Rev. A MOV Program Changes This evaluation addresses changes to the Callaway Motor Operated Valve (MOV) program to:
- 1. Extend periodic verification of operability test intervals to 4 cycles /6 years for MOVs that meet current program requirements such that periodic differential testing is not required.
- 2. Extend intervals for perforn ante of preventive maintenance evolutions on Limitorque operators to 4 cycles (72 months).
- 3. Accept the use of suitable alternative testing equipment and methods for periodic verification of operability.
These changes are based upon operational and testing experience gained. No reduction in the ability of the MOVs to perform their assigned safety functions results from these changes, therfore, no unreviewed safety question exists.
RFR 16916 Rev. A Thermo-Lag Appendix R Fire Wrap Evaluations This RFR evaluates the Appendix R fire wrapped raceways in 12 fire zones in the plant. The RFR results in FSAR design basis and fire hazards review methodology r visions for FSAR Appendix 9.58. These changes do not result in an unreviewed s fety question. License Condition 2.C.(5)is not impacted because the proposed changes have no adverse impact on the ability to achieve and maintain safe shutdown in tha event of a fire.
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RFR 16916 Rev. C ,
R: place Thermo-Lag with Darmatt Material i The replacement of Thermo-Lag with Darmatt KM1 material in fire zones A-1, A-6, l A-24, C-35, and RB-10 has been evaluated and found to be acceptable to meet Appendix R requirements. The replacement on junction box 1JJ043 has been found .
ccceptable to meet IEEE-384-74 requirements. Use of existing Thermo-Lag to l 4
srtisfy the requirements of IEEE-384-74 has been reviewed and found acceptable. l License Condition 2.C.(5) is not impacted because the changes have no adverse 3 impact on the ability to achieve and maintain safe shutdown in the event of a l Gre. No unreviewed safety question is created.
RFR 16916 Rev. F Removal of Appendix R Thermo-Lag This RFR evaluates the impact of installed Thermo-Lag on the seismic qualification of affected equipment. The underlying components would not have failed catastrophically during a seismic event based on an evaluation of ultimate stresses. This RFR also approves the removal of all 10 CFR 50 Appendix ;
R Thermo-Lag. The Appendix R Thermo-Lag was previously abandoned-in-place per l RFR 16916 Rev. A. The removal of the Thermo-Lag does not adversely impact !
s:ismic analyses. No unreviewed safety question is created by these d; terminations.
RFR 16978 Rev. A Elimination of Reactor Cooiant System RTD Changeout This evaluation examines the removal of the requirement to replace one Reactor Coolant System RTD with a newly calibrated RTD at alternating refuelings to cssure that the RTDs are not drifting in a mono-directional fashion. This commitment was made in response to Amendment 57 to Facility Operating License NPF-30 and its associated Safety Evaluation Report (Ref: ULNRC-2368, dated February 20,1991).
This evaluation shows that the Callaway RCS RTDs do not drift in a monodirectional fashion and therefore, the commitment to replace one RTD at alternate refuelings can be eliminated. No unreviewed safety question is created by this evaluation.
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RFR 17005 Rev. A 4
Operability of the Accumulator Safety injection System with Nitrogen Buildup This RFR evaluates the operability of the Safety injection (SI) System with nitrogen enriched water accumulation in the discharge piping. The nitrogen rich w;ter has been introduced into the system from backleakage past the Si system-to-RCS check valves from the Si accumulators. Calculations indicate that there cre approximately 300 gallons of water in the SI system discharge piping. This cmounts to 1.6% of the three accumulators assumed to be intact during a LOCA.
The head vent valves, which are designed to vent off non-condensible gases in the reactor vessel, are capable of venting-off this insignificant amount of q cdditional nitrogen. This problem does not cause the SI system to be j inoperable, and no unreviewed safety question exists.
a j
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- RFR 17052 Rev. A i
, Change Calorimetric for Steam Flow and RCS Flow Calibration This evaluation allows the following:
i l 1. Use of the NSSS computer indication of reactor power to be used for beginning-of-cycle (BOC) RCS calorimetric flow measurement.
1
- 2. Elimination of temporary steam pressure gauge installation.
- 3. Elimination of direct feedwater temperature and flow measurement.
This evaluation shows that measurement uncertainties remain bounded by the cssumptions in Callaway's accident analyses and Improved Thermal Design Procedure (ITDP) analyses. No unreviewed safety question is introduced by this change.
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RFR 17096 Cycle 9 Core Design R vision A of this RFR applies to the Cycle 9 reload design and is applicable for operation in Mode 6. Revisions B and C of this RFR are applicable for l operation through an exposure of 30 Effective Full Power Days. Revision D of ;
l this RFR approved operation through the end of Cycle 9. The Cycle 9 reload l
l d:: sign satisfies all of the safety parameter limits and acceptance criteria and h:s been evaluated using standard reload design and approved fuel rod design models and methods. These evaluations conclude that there is no unreviewed safety question related to core design for operation through Cycle 9.
1 _
RFR 17231 Rev. A Evaluate Frazil Ice Concerns This evaluation addresses the concern with frazil ice accumulation on the Essential Service Water inlet grilles. Through analysis, it has been determined ,
that if frazil ice forms on the Ultimate Heat Sink (UHS) pond, it will not be drawn down through the inlet grilles. Therefore, the frazil ice can not !
cccumulate on the grilles and cause blockage and the grilles can remain in J place. Therefore, the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR will not be created by leaving the grilles in place. No unreviewed safety question exists.
RFR 17378 Rev. A Allow Removal of Pressurizer Spray Line Restraints in Modes 5 and 6 This RFR approves the removal of several piping restraints from the pressurizer spray line when the plant is in Mode 5 or lower. Based on analysis, the piping stresses in the altered configuration are maintained below Code limits and d: sign basis requirements. No unreviewed safety question is created by this configuration.
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RFR 17439 Rev. B Disconnection of Containment Cooler Space Heater This RFR approves operation of the B containment cooler with its space heater disconnected. This allowance does not impact the safety-related function of the containment cooler. The space heater is designed to prevent the formation of condensation during periods when the motor is idle. During normal operations, the containment cooler is operated continuously. The only time the coolers are deenergized is during refueling outages. During that time, conditions inside containment do not change significantly to create a condensation buildup inside the motor. Therefore, the cooler will operate as designed during normal and cccident conditions. No unreviewed safety question is created.
SOS 95-0079 Hot Leg Streaming Evaluation This SOS evaluates the effect of hot leg temperature streaming at Callaway Plant. Temperature streaming is caused by low leakage loading patterns which give rise to large temperature gradients at the core exit. When reactor coolant cnters the hot legs, it is not thoroughly mixed. This typically results in a measured value of hot leg coolant temperature (Thot) that is greater than the true value. This results in a higher than actual measured value of average coolant temperature (Tavg). Another phenomenon, " upper plenum flow crossover",
clso effects loop operating temperatures. This is the result of flow sharing between adjacent loops involving fluid streams at different temperatures. Upper plenum crossover causes loops to operate at different power levels and t:mperatures.
Systems and analyses that depend on Thot and Tavg were identified and evaluated to ensure that applicable limits would not be exceeded when actual operating t::mperaturo profiles were assumed in lieu of the original assumptions. These evaluations concluded that no unreviewed safety question is created by these phenomena.
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SOS 95-0610 !
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Radioactive Contamination of the Auxiliary Boiler during Refuel 7 '
The operation of the radiologically contaminated Auxiliary Boiler during Refuel 7 at the levels identified in SOS 95-0610 have been evaluated. Dose 3 calculations with conservative assumptions show that the potential doses to the ,
public are well below the limits of 10 CFR 20 and 40 CFR 190. Previous and l continued operation does not constitute an unreviewed safety question. )
i j SOS 95-1228 l Continued Operation of the Sewage Treatment Plant with Trace Contamination This evaluation addresses the radioactive contamination of the Sewage Treatment l 4 Plant (STP) Sludge Holding Tank and the continued operability of the STP per NRC 1 IE Bulletin 80-10.
Continued Operation of the STP as slightly contaminated with low levels of radioactive material will not impact the environment or the public since no j rcdioactive material will be released to the environment. All slightly I contaminated sludge will be proceced as low level radioactive waste. No unreviewed safety question is created.
SOS 95-1340 1
i Radioactive Coiltamination of the Auxiliary Boiler Continued operation of the auxiliary boiler once per month for a total of not greater than 56 hours6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br /> will not adversely affect the environment. All releases of radioactivity from the auxiliary boiler will be reported in the Annual Radiological Effluent Release Report for 1995. No unreviewed safety question is created.
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SOS 95-1341 Install Tube Inserts into CCW Pump Room Cooler SGL11B Two tubes of the "B" CCW Pump Room Cooler, SGL11B, will be replaced under a Section XI repair plan using either Swagelok or brazed couplings. The material used will be equivalent or better than the as-built material currently used in the room cooler. Any degradation in heat removal capacity will not compromise the design heat removal capability of the room cooler. No change in the seismic or environmental conditions will occur that will compromise the qualification of SGL118. No unreviewed safety question is created by this change.
SOS 95-1936 Evaluation of Failed Check Valve Disk Arm This evaluation approves the use of similarly designed check valve internals to r: place internals that had failed in a 6" RHR check valve. A unique combination of factors including outage flow conditions, duration of flow, valve location near a turbulent area, and disk arm dimensional characteristics contributed to the failure. This combination of factors would not be applicable at the time the valve is called upon to perform its safety function during an accident condition, therefore, an unreviewed safety question does not exist.
SOS 95-2033 Radioactivity Detected From Sludge at Sewage Treatment Plant This evaluation addresses the radioactive contamination of the Sewage Treatment Plant (STP) Sludge Pit and the continued operability of the STP per NRC IE Bulletin 80-10.
Continued Operation of the STP as slightly contaminated with low levels of radioactive material will not impact the environment or the public since no radioactive material will be released to the environment. All slightly contaminated sludge will be processed as low level radioactive waste. No unreviewed safety question is created.
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SOS 96-0090
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Radioactive Contamination of the Auxiliary Boiler This evaluation addresses the operation of the auxiliary boiler during 1996 with cxisting low levels of radioactive contamination. Calculated public doses are a v ry small fraction of the Radioactive Effluent Controls (REC) dose limits. No unreviewed safety question is created by continued operation of the auxiliary boiler.
SOS 96-0422 l l
Repair of Penetration Room Cooler Leaks Two tubes of the 'B' Electrical Penetration Room Cooler will be replaced per an ASME Section XI Repair Plan using either swagelok or brazed couplings. The material will be equivalent or better than the "as-built" material currently )
used in the room cooler. Any degradation in heat removal capacity will not i compromise the design heat removal capability of the cooler. No change in s:ismic or environmental conditions will occur that will compromise any ;
qualification of the cooler. No unreviewed safety question is created by this l change. '
i i
SOS 96-0876 Contaminated Sludge in Sludge Holding Tank This evaluation addresses the radioactive contamination of the Sewage Treatment Plant (STP) Sludge Holding Tank and the continued operability of the STP par NRC IE Bulletin 80-10.
Continued Operation of the STP as slightly contaminated with low levels of rcdioactive material will not impact the environment or the public since no r dioactive material will be released to the environment. All slightly contaminated sludge will be processed as low level radioactive waste. No unreviewed safety question is created.
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TM 95-E006 Removal of Loop BBT-0413B Inputs to Cold Overpressure Protection Circuits This temporary modification removes the input from a failed temperature channel to the cold overpressure protection (COPS) circuit for Train B (valve BB-PCV-0456A). This is to prevent spurious power operated relief valve actuations from the failed channel, leaving COPS functional. This helps prevent a loss of coolant accident through a PORV due to a spurious temperature signal from the friled temperature channel. It is recognized that the COPS train is inoperable during the period that this temporary modification is installed. No unreviewed sifety question is created by this temporary modification.
l TM 95-M012 l Turbine Torsional Test Collars Installed at Bearings T-4, T-6, and T-8 This temporary modification approves installation of torsional test instrumentation on the main turbine for the duration of cycle 8. The instrumentation consists of 4 strain gages mounted on the turbine shaft adjacent l to turbine bearings T-4, T-6, and T-8. The strain gages are secured at each l location by a 360 degree telemetry collar. A power supply unit, antenna and cabling is also installed at each bearing location.
Each telemetry collar's halves are bolted together and secured with epoxy, Each I power supply unit and antenna are secured by bolting. All bolted connections l s: curing these items are tack welded to prevent any loosening. The main turbine s rves no safety function and has no safety design basis. The changes approved by this temporary modification do not effect the results of the turbine trip and loss of electrical load analyses of FSAR Chapter 15.2. No unreviewed safety question is created by this temporary modification.
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l TM 95-M015 l l
Temporary Potable Water Distribution System !
This evaluation approves the installation of a temporary potable water distribution system which will be used to supplant the existing system during maintenance activities. The installations will be made with piping / plumbing 1 that meet the service condition requirements for potable water. The j installation will remain in place for approximately 3 weeks to support system maintenance and modifications. The system serves no safety function and no
. unreviewed safety question is created.
TM 95-M016 Install Portable Gas Analyzer System to Waste Gas System This temporary modification installs a portable gas analyzer system to the waste g:s system. This will be used to monitor Hydrogen and Oxygen in the Waste Gas System. The maximum activity or volume will not exceed that of the tanks .
previously analyzed in the FSAR. The skid will operate in parallel with plant !
equipment, therefore, Technical Specification 3.3.3.10 (which is now FSAR j s:ction 16.3.3.7.A) will be met with having at least one hydrogen and both inlet end outlet oxygen explosive gas monitoring instrumentation channels operable.
No unreviewed safety question is created by this temporary modification.
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TM 96-E010 l Remove Power to DP Gauges on Class IE and Control Room A/C Units
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This temporary modification removes safety related power from non-safety related instruments on the Class IE and Control Room A/C Units to prevent a possible failure of the air conditioner control circuits due to a malfunction of a non-s-fety related component. This condition was identified and documented by SOS 96-0877.
The instruments (gauges) serve no safety related or controlling functio'n for the AIC units, therefore a loss of power to the indicator and removal of the alarm will not affect the operability of the equipment. No unreviewed safety question is created by this change. l Operability of the affected equipment is documented in SOS 96-0877. RFR 17206 l Rev. A was initiated to pursue permanent correction to the problem.
Ref: SOS 96-0877 l RFR 17206 Rev. A TM 96-M007 Install Regulating Valve in SI System Test Line This temporary modification installs a regulating valve downstream of EM-V-0040 cnd a pressure gauge downstream of EM-V-0139. The valve will provide pressure control in the RHR system that is built up due to ECCS check valve leakage. The :
gauge will be installed to provide local pressure indication on the Si test header line. This temporary modification is installed downstream of containment isolation valves which receive a closure signal during design basis accidents.
No unreviewed safety question is created by this change. ,
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TM 96-M010 Remove TDAFP Turbine Exhaust Steam Trap Internals This temporary modification removes the internals from steam trap FC-ST-002 for troubleshooting of a condensate buildup problem. The removal of the internals will make the system similar to the original design. This will allow condensate to more freely drain. Resulting changes to steam flow rates have been evaluated cnd found acceptable. No unreviewed safety question is created by this temporary modification.
TM 96-M011 Install Test Duct With Damper in HVAC System This temporary modification installs a test duct with an adjustable knife damper in a non-safety-related section of the Auxiliary / Fuel building HVAC systems.
This section of duct is isolated following a Safety Injection Signal (SIS) or a Fuel Building Ventilation Isolation Signal (FBVIS). This change will enhance 4 HVAC effectiveness and mitigate the spread of noble gas during refueling operations. Seismic concerns have been evaluated and found acceptable. The l n::gative pressure in the Fuel Building will be maintained during normal !
operations. No unreviewed safety question is created by this temporary ,
modification. l 1
TM 96-M012 Allow Vendor to Provide Laundry Cleaning During Refuel 8 This temporary modification approves the installation of temporary water and air supplies to operate vendor laundry trailers during Refuel 8. RFR 10604B previously approved the method of using ozone injected into the laundry water to clean the PC's. Possible leakage from the associated storage tanks has been evaluated and found not to result in an unreviewed safety question.
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- TM 96-M014 Install Flanges on Containment Cooler Ca!!
This temporary modification will install blank flanges on the inlet and outlet fitnges of the third cooling coil from the bottom on the plant West side of
. containment cooler SGN01 A due to a tube leak identified by SOS 96-0938. The
, d: sign basis cooling capacity is maintained in the proposed configuration since this less than the maximum number of passes will be blanked-off by the flanges.
The Code pressure boundary, seismic, and environmental qualification of the cooler is maintained. No new initiators or failures are are introduced by blanking-off one coil of SGN01 A. No unreviewed safety question is created by this temporary modification.
TM 96-M020 Isolate Air Supply to Standby Train Actuator From Active Train on FWlV
- This temporary modification blocks off air and hydraulic crossover lines between the active and passive trains for feedwater isolation valve (FWlV) AE-FV-0042.
This temporary modification impacts a non-safety-related portion of the cctuator. This evolution is necessary to allow repairs to a leak in the standby train while enabling the active train to remain operable. Failure of the air or hydraulic plugs would only cause a fluid leak. The active train of the actuator would remain operable due to the check valves in each system (air / hydraulic).
This installation does not create an unreviewed safety question.
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W575186 Repair of 2 Inch Branch Line from RCS Loop D A cracked weld in excess letdown piping connected to the RCS was documented on SOS 95-1891. This identified flaw in a 2-inch diameter branch line from the main reactor coolant loop 4 crossover leg will be prepared by peening and weld overlay for a permanent repair. These preparations are performed while the plant is in Mode 5, with the RCS depressurized. The repair is in accordance with the ASME Code Section XI, alternative rules established by ASME Code Case N-504 as approved in NRC Regulatory Guide 1.147, and the approved ASME Section 2
XI Repair / Replacement Program. This repair will restore the structural integrity of the pressure boundary to required allowab' . stresses for normal
- operating conditions. No unreviewed safety question is created by this
- activity.
Ref: SOS 95-1891
. Records printed: 150 4
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