ML20101F029
| ML20101F029 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 03/31/1992 |
| From: | Schnell D UNION ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| ULNRC-2646, NUDOCS 9206240228 | |
| Download: ML20101F029 (44) | |
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Etscruic . Q June 12,-1992 U.S. Nuclear Regulatory Conunission Mail Station PI-137-Washington, D.C. 20555 Attn: Document Control Desk ULNRC- 2646 , Gentlemen: DOCKET NUh1BER 50-483 - CALLAWAY PLANT 10CFR50.59 ANNUAL REPORT SUMA1 ARIES UNION ELECTRIC APPROVED WRITTEN SAFETY EVALUATIONS
Reference:
ULNRC-2419, dated June 6.1991 In accordance with 10CFR50.59, this letter transmits a repon which summarizes written safety evaluations of changes approved and implemented for Callaway Plant since those reponed in the referenced submittal ano through March 31,1992
- All items reponed herein were detennined to not involve an unreviewed safety question.-
If there are any questions, please contact us. Very truly yours,
/ A b Donald F. Schnell GAC/ dis Attachment 9206240228 920331 PDR R
ADOCK 05000483 PDR [/ d/f
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cc: T. A. Baxter, Esq. Shaw, Pittman, 'Potts-& Trowbridge 2300 N. Street, N.W. Washington, D.C. 20037 Dr. J. O. Cermak CFA, Inc. 18225-A Flower Hill Way Gaithersburg, MD 20879-5334
-R. C.. Knop Chief, Reactor. Project Branch 1 U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Bruce Bartlett Callaway Resident Office U.S.-Nuclear Regvlatory Commission RR#1 '
Steedman, Missruri 65077 , L .. R. Wharton (2) Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission !' 1 White Flint, North, Mail Stop 13E21 11555 Rockville Pike Rockville, MD 20852 Manager,. Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102
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i 9 4 10 CFR 50.59 Safety Evaluation Summary Report for Callaway Nuclear Plant J d i Union Electric Company 06/11/92 F T + I. I t
Reference Kev MP # Modification Package (Design Change) e CMP # Callaway Modification Package EMP # Exempt Modification Package RMP # Restricted Modification Package m CN # FSAR Change Notice OL # Technical Specification Change ESP # Engineering Surveillance Procedure ETP # Engineering Test Procedure CTP # Chemistry Technical Proceditre OSP # Operations Surveillance Procedure OTU # Nortral Operating Procedure OTS # Special Operation Procedure RTS # Special Operating Procedu a (Radwaste) - ') RFR # Request for Resolution NMR # Nonconforming Material Report Note: FSAR and Technical Specitication changes are also reported under 10 CFR 50.71 and 10 CFR 50.90 as applicable. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ U
50.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALLAk'AY NUCLFAR PLANT Calc. ZZ-111, Rev0 MSLB/ Fan Cooler Degredation Containment Analysis Analysis performed in anticipation of potential degradation of containment fan coolers heat exchangers to justify total reduction in containment cooler heat exchanger performance of up to 30% in any one containment cooler and up to 25% for all four contain=cnt coolers. Containment coolers perform the safety-related function of maintaining containment atmosphere such that systems and components inside containment properly function during normal operation and of limiting containment atmosphere below the EQ -- temperature limit of 384.9 F and design pressure 60 psig during accident conditions. The consequences of the MSLB inside containment with degraded fan cooler heat exchanger performance have been previously analyzed and shown to be bounded by results presented in F3AR 6.2.1. Radiological consequences presented in FSAR 15.1.5 are not ; affected. For LOCA peak containment pressure and temperature occur within approximately 2 :ainutes after initiation of LOCA. Thus, cooler performance has an insignificant affect on peak containment pressure and temperature during LOCAs. Radiological consequences presented in FSAR 15.6.5 are not affected. CN 90-39 FSAR Chapter 13 Update Various revisions tu the description of the Callaway organization described in the FSAR to incorporate title changes, planned rotations of individuals between positions, editorial changes, and to a6d our commitment to ANSI /ANS 3.1-1981 for Shif t Supervisors , Operacing Supervisors, Reactor Operators, and Shift Technical Advisors. Scme positions are eliminated. The changes do nce significantly alter the duties and responsibilities of the affected departments nor their abilities to support the plant. Minimum qualifications have been maintained. Changes in reporting relationships are of minimal impact and are designed to make the organization more streamlined
. and effective, Title changes have no significant impact upon the affected positions' responsibilitier and minimum qualifications.
The deletion of positions has no significant impact since the responsibilities that they represented were effectively , re-assigned. Report Date 06/11/92 Page 1
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i 50.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALLAWAY NUCLEAR PIANT CN 91-04 Heat Loads to Spent Fuel Pool During Full Core Offload Revise'FSAR Tables 9.2-7:and 9.2 8 to reflect a more realistic accounting of which loads are being carried on component cooling l - water system (CCW) at any' specific point in time. Change required
. based on projected heat loads to spent fuel' pool from a full core
= offload.
i Change represents an ultra-conservative estimte of the earliest the RCPs could be started (i.e., core reloaded and head assembled). .This load combination is conservative because in - order to have both the spent fuel pool load and RCPs running _ completion of refueling would need to-be assumed. Therefore, t loads assumed are very. conservative. No changes to the CCM-lineup or configuration are made and the listed total heat loads on Tables 9.2-7. and 9.2-8 are unchanged. No equipment changes are made. iCN 91-19 i: Revise Description of Emergency Fuel Oil Day Tanks
' Revise FSAR description of emergency fuel oil day tanks and the 2- basis for their designed capacities.
! New descriptions more accurately reflect- as-built configuration of the. day tanks._ Basis of the level setting is more reflective of ANSI N195-1976 as endorsed by Reg. Guide 1.137. Design bases and
. safety analyses in FSAR are unaffected. Minimum required storage volumes in storage and day tanks are unchanged.
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d CN 91 31
- j. Revision to Table 3.11(B)-3, " Equipment Qualifica*fon" Revise'FSAR Table 3.ll(B)-3 to change LOCA and MSLB categories, as i!
applicable, for BM-HV-19, -20, . 21, -22, -35, -36, -37, -38, -65,
-66, 67, -68 and their associated position indication switches from A to C. Add category B mechanical EQ listings for these valves due to containment barrier considerations.
Inconsequential effects on safety functions and accident analyses.
- These valves are not covered in Technical Specifications nor are
, they considered to be containment 3 solation valves since GDC-57 i does not apply. All 12 valves are included in Revision 11 of the , Callaway Plant .Inse rvice Testing Program. Valves (other than 19, 20, 21, and 2' undergo quarterly full stroke exercise and fail
- i 1 safe on loss of power tests and position indication tests at each '
Report Date 06/11/92 Page 2 i
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I 50'.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALLAWAY NUCLEAR PIANT - __ l refueling outage. :The failed open failure; mode is addressed non-mechanistically per NUREG.0588. Harsh environment accidents (MSLB, LOCA, and feedwater line break).are examined. There would be no effect on the mass / energy release data used to address NRC IN 84 90 due to inconsequential _ additional level decrease associated with sample valves failing open. CN 91 41 Post-irradiation Testing of CT Fracture Mechanics Test Specimens c . 1 Revise FSAR to delete post. irradiation testing of 1/2 T (thickness) compact tension (CT) fracture mechanics test specimens. The results of.CT__ testing on Capsule "U" did not , provide _any pertinent'results. Testing methodology has not improved since this time. !
'CT testing is.not included ~in ASTM E 185 82 as a-required test or
- as an optional test. Optional tests include hardness tests condu6ted on Charpy specimens and supplemental *ests. These tests are meant as a backup and supplement to the tension and Charpy testing.
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CN 91 i
~ Recycle Evaporator Feed-Demineralizer Description Revise FSAR to delete " mixed bed" from description of Recycle Evaporator Feed Demineralizers (REFD) to permit use of most effective . type of resin (e.g. , cation resin) based on current
- industry experience.
Change in one domineralizer bed to cation resin will in no way reduce = effectiveness to remove radionuclides or other impurities
.in the_ water being processed. The change should only enhance '
their removal. L t f
'CN 91-54 Recategorizatica of Valves from Electrical EQ to Mechanical EQ Revise FSAR to recategorize various valves from Electrical EQ to -
Mechanical EQ. -Valves consist of the valves, air regulators, , positioners, check valves, converters, and actuators. All of '
-theese items'are mechanical in nature and are totally non-electrical. All electrical appurtenances are listed separately. Affected components are Masonellan supplied control and atmospheric relief valven.
c_ Report Date 06/11/92 Page 3 t ,
50.59. SAFETY EVALUATION SOMMARY REPORT FOR CALIMAY NUCLEAR PLANT ! This recategorization will have no impact on safety as it is only reclassifying the mechanical items as Mechanical EQ and the electrical items as EQ. CN 91 55 Replacement of Manual TLD Reader with Automatic Reader Revise FSAR. to replace obsolete manual thermoluminescent dosimeter . (TLD) reader with an automatic reader. This change does not impact safety-related equipment, components, or. structures. . Organizational structure indicates Assistant Manager, Technical Services -is responsible for the implementation and effectiveness of the ALARA program. This position has been ; 4
' eliminated and the_ responsibility is now on the Superintendent, Health Physics. Both-of these chagnes are administrative in '
nature and do impact safety evaluations presented in the FSAR. 4 CN 91-57 NSSS Recorder Replacement
- Revise FSAR to clarify seismic qualification requirements for non-Class lE NSSS recorders and document general revision to I
several'I&C discussions in the FSAR. Overall protection system performance will remain within bounds of
- the accident analyses since proposed recorder change has no effect on-the operation of the reactor trip system, the 7300 Process Protection System,-the Solid State Protection System,'or any other safety-related . structure, system, or component. Recorders are not modelled in the accident analyses. There are no regulatory *
. requirements or guidance documents that would demonstrate a need to seismically' qualify recorders fed by non.lE power. Change to
[ commercial grade recorders has no effect on accident initiation or [ . progression. Requirement to mount _them as seismic 11/1 will ( ensure that other safety.related control board instrumentation
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will not' be impacted after a seismic event. Recorders are
- - isolated from Class lE portions of instrumentation channels.
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o L CN.91 65 '
.Uydrostatic/ Pneumatic Testinb of D (Augmented) Class Systems Revise FSAR to provide alternative to hydrostatic testing of D Augmented systems. D Augmented systems are non. safety related and '
fall within the scope of ANST B31.1 Power Piping Code wil h provides latitude for an ower to evaluate and determtne, on a case l Report Date 06/11/92 Page 4 P g y vu.., y r .e w *er*.-.v - , y,-v--,-,,r-t-- - m ----.==,,,--w,.-------.-...w. ......-,,,,.--,.em-r, --..-,,,.,---..------,-.---*=-w
50.59 SAFETV EVALUATION
SUMMARY
REPORT FOR CALIAVAY NUCLEAR PLANT by case basin, the retest method to be used. U + proposed change will not eliminate retest requirements for D bg,wented systems. It allows retest assignment to be consistent with the requirements of ANSI B31.1 and with applicable design specification. As a minimum, an initial serce leakage test will be performed on all D Augmented systems / components. This leakage test, coupled with the final visual inspection performed on all D Augmented welds will provide adequate assurance of system integrity. CN 91 66 - Assumed Leakage of Radioactive Fluid From CTMT Recirc Sump pa ECCS 1 sol Vivs Revise FSAR to address leakage from containment recirculation sumps past ECCS isolation valves (that are not leak rate tested) to the kVST and the environment per NRC Information Notice 91 56. This change does not involve a change to the facility or any structure, system, or component. No dcsign changes are associated with this change. No Technical Specifications are involved. Safety evaluation in performed to address additional leakage pathway to environment for radioactive releases after a LOCA, - Radiologicel consequences are investigated; total FSAR doses are not increased. INSERVICE TEST PROC - Inse rvice Testing Progra:n, Rev. 12 Evaluate the changes made to the Inservice Testing Program, Rev. 12, as well as the proposed changes / additions regarding t he implementing test procedures. Changes made to maximum allowable (or limiting value) stroke time for specified valves are all conservative since the values are all lowered. The changes are implemented to adhere to Generic Letter 89 04. No change to t presently approved test method will be made and no unennservative change to plant safety is seen. Removal of relief request #V07 and associated alternate test justification addition does not change the program. Instead this administratP.a change is performed to conform to NRC format requirements since the new alternative replaces the deleted relief request. All specified relief valves to be deleted are classified as pasnive since there is no defined change of position required in :he analyzed accident scenarios. These valves are not included in che original safety evaluation for the plant which indicates that eliminating them from the IST Program does not have any adverse impact on the margin of safety. Adding closure testing for specified check valves and stroke time teating in the closed Report Date 06/11/92 Page 5
50.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALLAWAY NUCLEAR PLANT direction on a quarterly test frequency for other specified valves enhances plant reliability and safety. Other changes are administrative in nature. NES Doc #83A1004 Nuclear Safety Evaluation on Relief Request for Reactor Pressure Vessel Supports Currently the . subject document requires that the NF portion of the Reactor Pressure Vessel supports be examined as part of the ISI program. Relief is requested to_ inspect only visible portions of the supports without removal of the walk plates and insulation. This inspection will encompass the shoe assembly and wear plate to the maximum extent practical. Inspection will encompass the shoe assembly and wear plate to the maximum extent practical. Inspection is scheduled to be performed during the next refueling. Volf Creek performed this visual inspection on their RPV supports and found no degradation. Based J on these findings, it was concluded that "there was no evidence of degradation which would indicate loss of integrity of the inaccessible portion of the supports." The same is expected to be true for Callaway cince the plants have been in operation fer soproximately the same length of time, and they use the same support design. Support shoes and air cool 3 box structures received MT or FT, UT and VT examinations prior to installation. Installation was inspected to site quality control procedures. Stringent ASME Section III, Subsection NF quality assureance programs were utilized in the design, fabrication and installation of these Class 1 components. Section III Examinations confirmed support integrity during construction. Air cooled box supporting the RPV is a rigid structure. Internal structural members with
-inaccessible welds-are in~ compression. The absence of tensile or bending moment loads diminishes the importance of weld inspections. The main i' unction of these welds is to hold the .
members in place rather than direct loads to the building l structure. NRC approved this relief request by letter dated 2/12/92. MP 90 7422 Rance Change on Freeze Protection Recorders of Solid Radwaste System i Expand-range of-freeze protection recorders in solid radwaste
- system. Freeze protection system supports the SRS power design basis to solidify and package concentrated waste solutions from .
the evaporators by preventing solidification of the bottoms of the l evaporators prior to drumming. Solid radwaste system performs no functien related to the safe shutdown of the plant, and its failure do-as not adversely affect i t Report Date 06/11/92 Paga 6
l l l 50.59 SAFETY EVALUATION
SUMMARY
REPORY FOR CALIAVAY NUCLEAR PLANT any safety.related system or component. The SRS has no safety design basis. MP 91 8816 Eliminate Pipe Leakage Problem in SLW System Install pipe caps and nipples at the casing drain connection of the SUV Drain Collector Tank' Pumps (PHF03A/B), Casing drains are used to drain the pump casing when required by maintenance activity. Installation of pipe nipple and cap at casing drain will have no affect on the operation of the pump. Although casing drain valves have been removed, the pump can still be drained when needed for a maintenance activity. Other valves are available both upstream ' and downstream of the pump that would remove most of the water in , ,. the' piping to facilitate maintenance. Remaining water trapped in ' the' casing can be drained from the pump by removing the threaded cap and allowing the water to drain into a conte, aer. Piping and pumps associated with this modification are all non-safety l related, and it will have no affect on any other equipment. ; 1 MP 91 8878 : Install Chain Cate in Lieu of Swinging Cate in Steam Vent Gallery Provide easier opening device in lieu of swing gate at the landing , to door 17011 at the steam vent gallery. Previously identified l personnel safety hazard: Difficulty in opening the swing gate , while a person Js hanging on the adjacent ladder. ' Subject platform and handrail is all non-safety related. Change , to this gate detail is simply to allow' easier--opening while hanging by one hand on a ladder. CMP 84 0801 Remodel Radwaste Control Room to Increase liabitability i Add a suspended ceiling to the Radwaste Control Room to include electrical-and flVAC work, sealing spare penetrations and painting. 4 This modification does not adversely affect any safety-design basis or evaluations for Category I structures contained in the FSAR. Report Date 06/11/92 Page 7 l
[ i 50.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALIAVAY NUCLEAR PIANT l _ t CHP 85 0017A L , Replacement of Velocity Probe in Unit Vent ! Replacement of velocity probe system in the unit vent with a flow measurement system consis-iro n four gas flow sensors in the unit l vent and.tvo low-range pneu a e Jifferential pressure i transmitters with a square utractors in the auxiliary building. The replacement 1. required to provide accurate . measurements of flow, Evaluate non.eelsmic II/I supported l electrical equipment located on auxiliary building. Revise i drawings to delete flow transmitter GTFT0021BB and reroute signal ; cables so that flow transmitter GTFT0021BA will supply input f signals that were supplied by GTFT0021BA and GTFT0021BB. Change ; range of flow transmitter FTf"f0021BA to allow transmitter to see ; entire range of unit vent flows. This flow measurement system provides no safety function, and the . replacement enhanced the system function. (This change has been , partially implemented.) Non-seismic II/I nup}arted electrical ! equipment located on auxiliary building penthouse is not mounted ! above or near any safety related equipment. Flow transmitter ' changes eliminate mismatch flow that has occurred with the dual ; transmitter system. In addition, change deletes work required to i calibrate second transmitter. ! CMP 85 0293 I Add.an Isolation Switch for the Intake Battery [ Add isolation. switch to the battery system at the intake structure. Batteries cannot be isolated from the rectifier that supplies DC power under normal conditions. Batteries need periodic maintenance so they will be ready whenever the intake experiences a loss of AC power. DC power is used to control critical intake components like the intake pump breakers.
- Modification involves electrical equipment at the intake structure and the intake has no safety.related functions. This modification does not affect any safety-related equipment or increase the challenge frequency to protective or engineered safeguards '
features. l CMP 86 0017A ! Installation of Overtemperature Alarms for Firewater Storage Tanks Install firewater tank heater overtemperature alarm. Alarm alerts control room that fire water freeze protection heaters fail to , l turn off. Report Date 06/11/92 Page 8 I~
50.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALIAWAY NUCLEAR P1 ANT Modification provides overtemperature alarms to a local panel and to the control room in the event fire water heaters fail to turn off. The alarm allows operator action to prevent damage to the-fire water heaters due to overtemperature. Modification has no adverse effect on the fire protection system. CMP 87 1012A Replacement of Pumps PilC01 and PilC06 Replace evaporator bottoms transfer pumps with new punpa that feature integral pump bearing and externally flushed m*x hanical seal-design. Vork scope includes entending reactor makeup piping to pump seal housing, suction and discharge piping rework, and solenoid actuated flush water. Pumps do not maintain seal face alignment leading to shaft leakage and an area contamination problem. New pumps eliminate significant shaft deflection and ensure proper seal operation. New pumps are sized using current capacities, materials and service conditions, Therefore, the solid radwaste system as described in FSAR remains unchanged. Current power feed capacity is not exceeded by increased mccor horsepower. Also, piping support system capacity remains unchanged even though suction line stantion supports are being modified and pump pedestal load is increased due to slight increase in new pump weight (primary only), CMP 87 1072A Install MCB Annunciator for Low Pressure / Loss of Flow in ESW System , i Install an annunciator in the main control room to alert for low ! pressure _and loss of flow in the essential service water system. FSAR Table 9.2-6 was revised to add the header pressure alarm to a list of ESW system alarms. The annunciator design is achieved by wiring a contact multiplying relay into the control circuit of both ESW pump control circuits. The relay added is Class 1E qualified and is mounted within the switchgear cubicles with a seismic approved mounting. This change , does not alter or add any control functions to the ESW pump control circuits. L ............................................................................... i l Report Date 06/11/92 page 9 i __.a~.__,___~_,,- , _ . . _ . _ _ _ _ _ - _ - _ . _ , _ , _ . - _ . . . _ . _ _ ....a. _ _ _ _ _ -
__ . _ _ . _ _ . . _ _ _ _ _ . _ _ __-__________.m. 50.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALIAWAY NUCLEAR P! ANT l CMP 87 1072B r ESV Pump Auto Start Interlock ! Connect differential pressure switch to rnonitor essential service , water (ESW) flow to containment coolers and to various room ! coolers to the circuitry which automatically starts an ESW pump on undervoltage to the opposite train switchgear. Switch will ; replace existing switch in same circuitry which monitors pressure ; in the main ESW header. During testing of CMP.1072A, upon loss of voltage to one of the safety.related busses, cross. connect valves - between service water and ESW systems would close, but leaked ; sufficiently to maintain pressure on ESW header. This prevented [ pressure switch from tripping and thus prevented automatic pump { start. Pump start interlock changed to activate on ESV flow i instead of pressure. Since this is the critical parameter for the ; ESW system, this will provide a more meaningful basis for the pump ; start. Also change engraving and inputs to annunciator windows
- 54A, and 55A, B, and C. ,
Change c'. lows ESV flow to be-maintained to equipment which would l otherwise bel isolated from its supply when an undervoltage condition occurs on the opposite train. This eliminates dependence on operator action to start an ESV pump after a partial ! loss of AC power. Addition of flow switch into pump start ; circuitry introduces only one conceivable malfunction which could affect safety: If switch is flooded or if a fire occurs in the area of the switch, it could cause a ground on the DC control ! p power at the switchgear. If another ground-then occurred ; elsewhere in-the circuit, excessive current through the control power fuses could occur. This would blow control power fuses, disabling the closing circuit for the ESW pump breaker, and the 3 ESW pump would become inoperable. To prevent this, switches and i cables are designed and installed according to standards and specifications required for safety.related instrumentation as listed in FSAR Table 7.1 2. To prevent-inadvertent grounding ; switches are installed above flood levels. One. amp fuses are used ; with proper coordination so that if an overcurrent condition I occurs in cales.or switches, fuse will blow before the 15. amp . i control cable fuses. This provides diverse means to meet commitments to 10CFR50 Appendix R. Material installed complies , with all requirements for safety.related equipment, and the function of the design meets original design intent. ! CMP 88 1012A Spent Fuel Pool Bridge Crane Load Cell; Digital Load Display to Electric lloist Replace lead cell on spent fuel pool bridge crane with a strain gage type load cell and an electronic load processor. Load ,
- i. processor along with three electro. mechanical auxiliary relays and 4 a digital panal meter will replace all of the functions of the !
Report Date 06/11/92 Page 10
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50.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALIAVAY NUCLFJJt P1 ANT load cell and the spring scale previously used to monitor the load suspended from the hoist's book. Change will provide hoist operator with a continuous digital
. readout of the load suspended from the hook of the spent fuel bridge crane on the pendant control station. The function of the lights on the control station are not affected, but their setpoints change. New load cell witt not provide any lateral support for the hoist. Since the replaced load cell also provided no lateral support, this change will not affect hoist movement during a seismic event. New electrical enclosure and conduit have
-no effect on the seismic response of the bridge crane.
Modification will reduce the probability of fuel handling accidents occurring by providing the operator with better and more - meaningful information about the load that the hoist is carrying. This will allov_the operator to respond to abnormal indications earlier than previously possible. This modification has no effect on the consequences of a fuel handling accident should it occur. CMP 88-1018A Finer Hesh Filters for the RCS and Additional Control Room Alarm Provide finer mesh filter cartridges in RCS filter FBG06 and install a control room alarm for filter high differential pressure. The change in filter sizes has no offect on the pressure boundary. The new cabling / conduit for the remote alarm is nonsafety, not routed over safety components, and does not require II/I installation. - CMP 88 1023A Revise Limitorque Valve Operator Breaker Setting Increase several motor operator valve (MOV) breaker instantaneous trip setpoints. The increased setpoints avoid unintentional tripping when the MOV reverses travel. Components affected by the increased instantaneous trip setting were evaluated: power cables, motor, penetration assemblies, and breaker coordination. Penetration assemblies protection is not degraded by~the increased instantaneous setpoint. Report Date 06/11/92 Page 11
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i l i 50.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALIAWAY NUCLEAR PIANT CMP 88 1038A Installation of Vibration Monitoring System for ESU Pumps Installation of vibration monitoring system for essential service water pumps. . Calculation EF.18 verifies vibration probe bracket design adequate for this II/I application. No adverse impact on operation of motor / pump shaft coupling on pump. i A j CMP 88 1044A Modification of Essential Service Water System ! Install tubing and instrumentation on existing flow elements of essential service vater system. Installation extends ESW system pressure boundary. , Pressure boundary integrity not compromised. Mechanical hardware meets requirements of ASME Section XI. Stresses were verified to be below Code allowables. Instrumentation utilized performs no active safety function. Indicators and switches are seismically ; qualified to IEEE 344. CMP 88 1049A' _ Isolation / Voltage Drop Resistors in 125VDC Inputs to ESF Status Indication Panel l Install 3000 ohm-resistor in the 125 VDC ESF status indicating system inputs from the main steam /feedwater isolation system . panels to isolate the control circuit fuses of these inputs from r inadvertent short. circuiting during trouble shooting to prevent blowing these fuses. ; I 3000 ohm resistors are sameoas those used on main control board i indicating lights. Small weight and positive mounting method t _ precludes any seismic concern. Minimal additional heat load will [ not adversely affect existing devices in these panels. Status ; i indicating panel (FSAR 7.5.2.2.1 and 7.5.2.2.2) is safety.related by' association not by function. Modification will improve operation of the status indicating system and reduce possibility
- of any adverse interaction with other safety-related equipment. ,
l Resistors provide circuit isolation, reduced relay coil heating, and will not' cause a reduction of safety. Report Date 06/11/92 Page 12 J
s 50.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALLAVAY NUCLEAR PLN4T CMP 88 1056 Improve Reliability of MCC Supply Breakers Replace 100-amp main transformer supply breakers with 125. amp circuit breakers. Main transformer supply breakers protect the main transformer cooling fans and oil pumps from damage dur.to high current. New breakers are rated for 125 amps at 50 C nabient. Rating increases to 144 amps at 40 C. Cables connecting the MCCs to the transformers are 3. conductor 1/0 AVC triplex cables. This size cable is rated for 179 amps at 40 C, thus the new breakers will protecc the existing cables. The cooling fans and oil pumps are - protected from overcurrent by circuit breakers located in the main transformer control panel. The main transformers and the MCCs are non safety related and serve no safety function. The MCCs are located in the turbine building so there is no seismic concerns. No new cable or conduit is being installed so no separation violations will be created. CMP 88 1061A Room Cooler Vent and Drain Line Galvanic Corrosion Replace safety-related room cooler vent and drain piping damaged by galvanic corrosion. Installation of these components falls under general requirements of ASME Section XI, Article IVA 7000 Replacements. Subarticle IVA.7400 specifically exempts piping valves and fittings 1 inch nominal pipe size and less, except that material and primary - stress' levels shall be consistent with the requirements of the applicable Construction Code. Stress evaluations indicated that stress levels are maintained below Code allowables during static
-conditions as postulated seismic-conditions. Copper and stainless steel materials utilized in this modification are close in galvanic properties to the cooling coils thereby eliminating the failures caused from galvanic corrosion. Replacement tubing provides approximately the same flow area as the original piping, Therefore, the ability to vent.or drain the cooler will not be significantly affected by this modification. Pressure retaining
-components utilized in this modification meet or exceed the requirements of the original components. The design bases of the affected room coolers remain unaffected.
Report Date-06/11/92 Page 13 i
__ __ _ . _ . _ _____-_ >._._ ____ _..--.__--.___.m._.- _ 50.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALIAWAY NUCLEAR PIANT CMP'88 1064A Remove Vendor Wiring to Eliminate Incorrect " Isolate" Indication I Correct improper energization of the containment cooler fan 8 MCC :
" Isolate
- handswitch indicating light on the ESF status indicating l panel. When fan is in fast, the light will be on regardless of l the MCC handswitch position. Vendor wiring between the spare j 42F/a contact and the associated terminal block points in MCCs -
NG02T and NG04T to correct problem. ' The only terminations to be lifted are associated with the spare 42F/c contacts. No other terminations are revised. Spare contact , is not intended to have any function in any circuit and will not < be a part of any circuit af ter the modification. Modification does not have any impact on safety-related equipment and does not ' cause any safety concerns. CMP 89-1016A Relief Valve on the Air Start Skid ', the DischarSe of Compressor Install relief valve downsteam of the pulsation dampener of air start skid for the emergency diesel generators. This will provide ; overpressure protection without the expense of increased wear from > valve chatter when the compressor is running. Relief valves on '
-discharge of the compressor are relieving from the pulsation shock waves produced from the positive displacement air compressors.
Modification.will not affect the safety-related portions of the air skid. Although the modification will raise the setpoint pressure of the air compressor relief valve and install a relief valve downstream of the pulsation dampener, this will not modify the safety.related air receiver internal pressure. Modification does not introduce a condition where the pressure on the non-safety related components are exceeded above their safe limits of operation. ; CMP 89 1041A Desensitize the Ground Detection Relays for 125 VDC Battery System ' Install handswitch and resistors to be used as a desensitizing switch for the PK01 and PK02 125 Volt DC battery system. The ;' switch will densensitize the ground detector alarm in the control room. The ground detector relay is used to alerr plant operators that a ground has occurred on the 125 VDC bus, i Installation of the desensitizing switch does not increase the probability'of an accident because without the densensitized position, the alarm would be in all the time after initial annunciation. With the new setpoint the alarm may reannunciate to Report Date 06/11/92 Page 14
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REPORT FOR CALLAVAY NUCLEAR PIN 3T give the operators an indication that the ground is decreasing in resistance and increasing in severity. This design reduces the probability of an accident by giving the operators indication of a ground condition that could cause inadvertent equipment operation. This modification only charges the annunciator setpoint which has no direct affect on the probability of inadvertent operations; it only indicates when the DC system has lost isolation, and that a second ground could cause inadvertent operation or the loss of the DC system. Because safety systems are designed to shut down the plant independent of non. safety systems, no unreviewed safety question exists. CMP 90-1006A Modification to CTMT Spray Additive Tank Level Transmitters and Reference Leg Add swagelock tee and unistrut support to the top of the reference legs of level transmitters (ENLT0017 and ENLT0019) used to monitor and maintain the containment spray additive tank level. Also change fluid in reference leg from Na0li to domineralized water. Spray additive tank has no initiating role in any analyzed accident. Supports added are standard details; therefore, no new mechanical failure mode is created. Use of demin water in reference leg has been accounted for and the setpoints are modified. Change is well within range of transmitters. Therefore, no new electronic failure mode is created. Vapor pressure _of water at maximum temperature of reference leg is within the cover pressure range so evaporation would be suppressed or slowed such that no adverse interactional failure modes are created. CMP 90 1011A-Install Additional Phones, Gaitronics, Receptacles Add new phones, Caitronics, and computer terminals and relocate receptacles in Health Physics Access. I L Addition of new Gaitronics stations and new phones do not adversely affect either system. All cable added is routed in , conduit to preclude any separation violations. Conduits are supported using standard support details so no seismic concerns
- are created.
l L Report Date 06/11/92 Page 15
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REPORT FOR CALIAWAY NUCLEAR PLANT CMP 90 1012A Change RCP Thermal Barrier Heat Exchanger Temperature Indication Configuration Rearrange configuration of instrumentation associated with temperature indicator (BBTI0009) for component cooling water (CCW) out of the reactor coolant pump (RCP) thermal barrier heat exchangers. Function of instrumentation is unchanged. Indicator monitors same parameters as before, and the specific pump is still selectable from BBHS0009. However, since selector switch selects a voltage signal rather than a resistance signal, the switch contact resistance has no effect on the reading observed on the meter. - Therefore, this change provides more accurate and reliable indication on BBTIO009. All components associated with this change are non. safety related, and since this-instrument loop provides indication only, this modification will have no impact on any other systems -or components which provide a safety-related function. New installed cable is in separation group 6; however, two of the RTDs connected to it are in separation group 5. The
.two channels.of non. safety related cables and raceways require no specific separation. Separation for non. safety related cables of different load groups is not necessary for plant safety.
CMP 90 1016A Rewire Generator Field Current Annunciator Circuit Rewire general field current annunciator circuit from a parallel - circuit to a series circuit. Circuit indicates the generator field temperature. Circuit provides input to the plant computer and the generator monitoring system to alert the plant operators when a generator field overcurrent condition exists. Modification relands wires in series inside the turbine generator panel' Modification does not affect plant generator or turbine contrSis. Annunciator, plant computer input and generator temperature monitor system are not safety.related. CMP 90 1040A Disconnect Motor' Compartment Space Heaters in Limitorque Operators of FCHV312
-Disconnect space heaters in motor compartment in Limitorque operator for valve FCHV0312.
NRC IE Notice 86-71 identified potential for energized heaters to damage wiring in close proximity to the heater. Limitorque advised that heaters were provided only for use during storage and are not required for installed operators. Space heaters are Report Date 06/11/92 Page 16 i
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REPORT FOR CALLAWAY NUCLEAR PLANT l intended only for use during storage and were not used during L Limitorque's environmental qualification testing program. Callaway's Limitorque operators are qualified for both normal and ; accident plant conditions with the space heaters deenergized. r t CMP 90 1042A Concrete Pavement en East Side of Fuel Building Install concrete laydown area outside the fuel building and a groundwater monitoring well and dewatering point between fuel building and reactor building. l Clay blanket is intended to provide protection to the structural backfill from surface water seepage. Structural backfill provides ; lateral support to Category I structures and to protect the foundation material. Concrete slab provides same level or higher + protection from surface water seepage; therefore, structural backfill ramains. unchanged. Structural fill remains undisturbed and in its' current condition. 1 CMP 90-1049A Install Conduit and Terminal Boxes Required to Install New Phone System j Install new telephone system. Existing conduit system is not adequate for new system, thus new conduit and terminal boxes are installed. Gaitronics. telephone interface units will be installed in the service building telephone room. All conduit and terminal boxes added will be field located so as ' not to create any separation violations. Conduit in control and auxiliary buildings will be supported using pre. approved seismic II/I supports to ensure no seismic concerns are created. Ductbanks will be installed in accordance with existing ductbank detail drawings. Interface units will not adversely affect the operation of either the telephone system or the Gaitronics system. ' CMP 90-1054A Instrument Sump Pumps Vent Line to CTMT Normal Sumps j Add stainless steel tubing from the instrument sump pump header vent valve and route to containmer.t normal sumps to ensure , instrument sump pumps will establish flow out of the sumps when sump level calls for the pumps to run. Tubing added in vicinity of containment noraml sumps is supported with II/I supports and will not be available for potential sump j Report Date 06/11/92 Page 17
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REPORT FOR CALLAWAY NUCLEAR PINIT t l blockage. Instrument sumps do not interface with any j safety-related system and will not degrade a safety related : system. Maximum flow rate does not adversely impact containment ! sump ~1evel measurement system. Tubing provides a full time vent { and enhances the availability of the pumps. Loss of the instrument sumps does not affect the operation of any 'Q" equipment. Tubing material is compatible with the system. i CMP 90 3003A Add-Tubing and Drain Valve for Air Receiver Tanks [ Extend the drain tubing for the air receivers associated with the ! switchyard breakers MDV41, 43, 45, 51, 53, 55, and 85. Chunge is required for personnel safety. In order to drain the air j receivers the operator must reach through wires to operate the drain valve. By extending the tubing and adding another drain valve, this operation can be accomplished without fear of electrical shock to the operator. . The operation of the system remained unchanged; therefore, no Ladverse impact to plant systems is expected. The new valve in series with the existing valve will perform the same isolation service in place of the existing valve. ; CMP 91-1011A Repair Water Leak in Pipe Penetration of Control Building Remove " Link-Seal" located on interior side of control building exterior wall. Install barrel.like seal around pipe, made of ; carbon steel and galvanized steel' welded to pipe and penetration sleeve. Existing silicone rubber foam, in conjunction with boot seal on the exterior of the control building wall, is intended to prevent ground water from seeping into the basement of the control ! building. Removing existing " Link-Scal" and installing new closure does not ' affect the safe operation of the plant. The penetration closure is not required to be fire of air barrier since the west side of the wall is soil. Function of ESW system is to provide cooling to - plant components required for safety shutdown following an accident. Welding a plate to the outside of the return piping ! does not prevent the system from performing its. intended _ function nor will it add any significant stresses to the pipe during an SSE l or DBA, Based on the physical separation, in conjunction with the ' fact that there is:no radiation source in line with the penetration which could cause streaming, the mod'ficatien is considered acceptable'from ALARA consideration. he i configuration of the-barrel.like seal prevents . from being dislodged from the pipe / sleeve. Therefore, it could not impact , any safety.related items in the area. The mass of the vent and Report Date 06/11/92 Page 18 i
L 50.59 SAFETY EVALUATION
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REPORT FOR CALIAVAY NUCLEAR PIANT drain lines is insignificant compared to any safety.related lines l in the area. Therefore, no II/I concerns exist. ; i 1 1 CMP 91 1027A Replacement of Gas Decay Tanks Common Sample lleader with Individual Sample Lines l l Replace common sample header with individual sample lines for each l vaste gas decay tank. 4 Caseous radwaste system is not safety related and has no safety function. Affected portion of the system is downstream of the ; sample isolation valve, and, therefore, is not group D Augmented. Tubing does not need to be designed to the seismic design criteria l of FSAR Table 3.2-5. Modified portion of system is normally isolated from the gas decay tanks. Technical Specifications , 3/4.3.3.10, 3/4.11.2.5, and 3/4.11.2.6 are not affected by this modification. CMP 91 1028A Install 3 10" Penetrations in Control Building and RWST Valve House Install three penetrations in control building and RWST valve house to allow access for temporary c. les or lines without need t to have. security posting. Penetrations, covers, and potential cut rebar were reviewed for seismic loads, tornado wind pressure and tornado missiles. , Modification does not affect any operating systems in the plant. Penetration covers will serve as pressure boundary and fire and missile barrier. . EMP 88 3004A Service Building Addition Construction of a 3. story 60,000 square feet Service Building i t addition, a 72 square feet Main Access Facility Addition, and related miscellaneous site work The enlarged Service Building will allow for the consolidation of selected plant support staff under one roof. Many of these personnel are presently located in temporary construction t buildings. Power to Service Bldg addition-will be supplied by a , new 1500 kVA transformer and duplex primary load break switch. L . Input power to this transformer will be from either the PA02 bus or from the safeguards transformers. Neither of these sources ' will be overloaded. Power to transformer will be delivered through the site 13.8 kV power distribution systen. The 90 deg C temperature rating of the conductors will not be exceeded. Report Date 06/11/92 Page 19
i c 50.59 SAFETY EVAL.UATION
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REPORT FOR CALLAWAY NUCLEAR PLANT Sp1' ices in the electric manholes will be made and fireproofed. [ Fuse in the duplex primary load break switch will be sized to trip on a phase to ground fault on the transformer secondary. Transformer secondary power feed to building switchgear is sized : per IPCEA standards. Load currents remain the limits of the , protective relay settings for the P00209 bus and the safeguards tranformers . Modification will provide separate fire main for building add 8 tion. Building will have a post indicating valve for isolation purposes; however, isolation valve for Technical Support ! Center will also isolate the fire protection line for the building addition. ANI has approved this configuration. Transformer will 7 be partially enclosed by fire walls to protect nearby buildings from damage due to fire. A fire in the Service Building addition will be alarmed in the control room. Main Access Facility additon will eliminate blind spot in security system caused by the Service Building addition. Membrane roofing used on the addition is seamed together-into one unit and is held in place by aggregate - to meet requirements for 100. year recurrence for wind. It is highly unlikely that during extreme wind, the aggregate will be removed such that a large enough portion of the membrane can be torn apart, thrown into the ESW pond, sink and block both intake. , A similar occurence is also highly unlikely to block the diesel generator intake louvers. EMP 88 3018A Provide Replacement-of Motor Temperature Scan ; Replace obsolete intake, circulating, and service water namp motor stator winding temperature scanners. Remove trip circuit that trips intake, cire. and service water pump motors on high stator temperature to eliminate a pump trip due to an "open" RTD. Rewire RTDs'to allow for propor operation of the scanners. New scanners are capable of scanning the same 6 motor RTDs on each pump as the existing scanners and can provide individual alarm and trip setpoints.for each RTD input. Only the alarm setpoints is retained on the cire,_ service, and intake -water pumps. The high stator temperature trip is removed to eliminate a pump trip due to t an "open" RTD. Motor failure due to overheating will not impose a more sever transient on the primary plant than a motor trip. . Plant reliability will not be adversely imp;cted. All RTDs are ! rewired for new scanners; existing viring is incompatible with new , equipment. Form, fit, and function cf the stator temperature monitoring _ systems remain unchagned with the exception of the removal of the pump motor trip. The intake, circulating, and service water systems are not safety-related. a l r Report Date 06/11/92 Page 20
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REPORT FOR _.CALIAWAY NUCLEAR PUtNT . EMP 88 3020A Retirement. Remove All Wiring from Relay 94XAC9
~ Remove all terminations at relay 94XAC9 and for cable 25ABYO4AA, Relay contacts are both of which become installed spares.
connected to open turbine drain valves ABPV23, 25, 27, 29 and ACHV134, 135, 136, 137 when turbine andisautomatic tripped. Manual operation open signal is of these drain valves is provided, not needed. Drain valves provide means for removing water from main steam lines during turbine startup and cooldown. Main steam line drain valves are not safety.related; however, they can affect RCS temperature control during cooldown following turbine trips. Rt al of the automatic "open" signal eliminates the RCS cooldown d .iculty. EMP 89-3002A Portland Composite River Vater Sampler Replace sample pump with a sample pump of the same size and capacity that is designed to handle abrasive ortee clear to water. face Rotate the composite river water sampler inlet downstream of the sample pump upstrcam and install a 6" teeInstallation allowa river water to flow past pointing downstream. the sample pump reducing the potential for pump to become silted in. Also, utilizing a pump designed to handle abrasive liquids gives better and .nore reliable performance for pumping river water. Portland composite river sample obtains river water smples downstream of the plant discharge line fer radiological analysis. The 6" pipe that the sample pump is installed in acts as a sample well in the river and provides for pump protection. The sample pump designed for handling abrasive liquids has the same capacity and performance curves that the existing sample pump screen has. The sample inlet and outlet is covered with a 1/4" mesh (inlet) and a 1/2" screen mesh (outlet) to prevent debris from plugging pump suction. The flow thru design will allow representative river water samples to be taken for radiological I analysis. The Portland composite river sampler has no safety design basis nor does it serve any safety related equipment. EMP 89-3003A Replacement of Bulk Oxygen Bottle Storage System Replace bulk oxygen bottic storage system consisting of two banks of 32 bottles with a vendor's portable high Service pressureGasoxygen oxygen trailer and a reserve bank of 8 bottles. Storage and Transf er System stores bulk quantities of oxygen and l transfers it to the radwaste building for use in the hydrogen l Page 21 l Report Date 06/11/92 _ J
50.59 SAFETY EVALUATION
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REPORT FOR CALIAVAY NUCLEAR PLANT recombiners. Continual change-out of the bottles is a safety hazard to the personnel handling the bottles and the large number ' of threaded connections has caused large expenses of manpower to , troubleshoot leaks. L Modification will provide same quality of oxygen and double the [ storage volume. Reserve bank of 8 bottles will be maintained to provide a sufficient' quantity to supply the plant's needs while the trailer is being refilled. Bulk oxygen storage system is
, approximatelv 350' south of power block; a fire in the gas yard !
. will not pree hazard to systems required for_ safe shutdown of-ne oxygen storage and transfer system is not plant. safe rf.related and is not required to perform any safety.related fruction during an accident. Modification complies with ANSI B31.1_and' appropriate DOT specifications. The evaluation of potential accidents-in FSAR 2.2.3 is not affected because it assumes only one chemical (hydrogen or propane) is involved, and , oxygen-is nonflammable. t EMP 89 3015A Provide Safety Access to Check Levels on OWRT and SAST Provide means to safely access sulfuric acid storage tank (SAST) ; and oily waste reclaim tank (OWRT). Personnel safety concern existA when equipment operators view the level gauges on these , tanks. Operators must view the level gauges on a regular frequency. Inclement weather can create hazardous conditions when accessing these areas. These access structures are considered non-safety related. Their purpose is to provide a safe means to access level instruments on , the above tanks. A II/I concern does not exist as the above access structures cannot fall or impair the function of safety-related components during a seismic event. l , EMP 89-3017A Install Flow Rest'rictor in Lube Water Filter Back Flush Line ( l Install orifice in the common backflush header downstream of the service water pump lube water filter backflush manual isolation valves to restrict' filter backwash flow rate. Service _ water pump lube water filter _is to provide filtered water to the service water _ pump bearings for lubrication. Filter is desf 6ned to be . backwashed on line to remove collected debris while providing ' clear filtered water to the pump bearings. Per vendor specification, filters are designed to process 100 gpm of-filtered water. During backwash cycle, filter will provide 75 gpm filtered water the system and 25 gpm for filter backwash to waste. Service water. system has no safety design basis and serves Report Date 06/11/92 Page 22
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REPORT FOR CALIAWAY NUCLEAR PIANT ' no safety-related function. Restricting the service water pump ! lube water filter backwash flow rate to design flow of i approximately 25 gpm will not have any adverse impact on plant safety or equipment reliability. t EMP 89 3026 , Pressure Sentry Valves on Condensate Polisher ! Remove pressure control valves and install relief valves to solve ; rep'sted problems with the preLugre control valves in the cot ensate polisher sodium analyzer skid. Pressure relief valves will provide additional sodium analyzer protection. Modification does not affect any accidents in the FSAR and does not degrade the. standards of design of the non safety relrted l skid. - Modification does not alter the design or operation of the ! skid which could affect the probability of any analyzed accidents. Skid is not used for accident mitigation or detection. Skid does not interact or inter-tie with any equipment that is safety-related. Output of sodium analyzer is local with no automatic actions. EMP 90 3004A - Retirement - Replacement Part for Condensate Hydrazine Recire. Pump Replace condensate hydrazine recirculation pump (PAQ07A) which is damaged but original equipment is now obsolete. PAQ07A promotes mixing of a hydrazine solution added to the condensate system to scavenge dissolved oxygen. The condensate hydrazine recirc pump facilitates secondary chemistry control. Failure of this system does not compromise other safety-related systems or affect safe shutdown of the plant. EMP 90-3007A ; Replace Deaerator Orifice Bypass Valve ' Replace deaerator orifice bypass valve (FBV0215) with a back
. pressure control valve to control deaerator operating pressure
- during. low load operation.--_Since deaerator vent will admit more steam / gas mixture, vent will be rerouted outdoors. Deacrator is
. provided to remove noncondensibles from the feedwater of the auxiliary steam system. Degassed feedwater is necessary to minimize corrosion of the auxiliary boiler.
- Deserator is designed for-chemistry control of feedwater entering !
the auxiliary boiler and does not perform any safety.related function nor does it interact adversely with any safety system. Report Date 06/11/92 Page 23 f V g yt~$"w ^W9N--r'** T fg Mt-W* 'T- g-'w' ywet w- w' ay-r' 6 ='mg'4 *T- '-'
-wap=D4-e-e- - * - * -**h 9 1r' 14*r*r 7me4- we e e W *mer Ew-= +-' - **9- -** * * ' * -*~"E'-*D-
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REPORT FOR CALLAVAY NUCLEAR PIRIT EMP 90 3034A Remove Existing Grating and Install a Handrail in Sludge Water Pumphnuse Remove grating and install handrail around pit to the sludge pumps in the Sludge Water Pumphouse. Railing prevents personnel from falling into the pit once grating has been removed. ; l Handrailing is not safety-related and provides no safety function l with respect to safe operation of the plant. Removal of the grating and installation of a handrail around the pit to the sludge pumps poses no operability concerns. , EMP 91 3013B Install Encapsulation Chamber Around AFV0918 for Furmanite Repair Encapsulate root valve (AFV0918) for high level alarm for 2d stage reheater drain tank and inject encapsulation chamber with , furmanite to repair body.to. bonnet leak. Level switch fed from this valve provides an alarm only. Valve being encapsulated does nto have a safety-related function. Downstream components that may be affected by this modification are not safety.related, nor affect the operability of a safety-related component. To minimize chemistry impact, limits on total halogens, low melting metals, and total low melting point metals are limited to same level as approved for use of lubricants, gaskets, and packing and can, therefore, be assumed to minimize any adverse chemical affects on the components. The compound will be injected into the encapsulation chamber, not into the valve internals. The amount of compound expected to enter the i system is minimal. EMP 91-3015A Install Bypass Line Around Flow Indicators on Circulating Water Pumps ! Install bypass line around flow indicators (FIDA2101A, B, and C) which are flow indicators for the lube water supply to the bearings for the applicable pump and motor coolers. Equipment is non. safety t%12ted and has no safety function. , Circulating water pump lube water supply does not interface or affect any safety-related equipment, nor does it interface or affect any equipment important to safety. Installation of bypass lines on the cire pump lube water flow indicators will not affect
-the design acceptance limits of any equipment, or comporents i
referenced in the Technical Specifications. Design acceptance limits remain unchanged. Report Date 06/11/92 Page 24
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REPORT FOR CALIAWAY NUCLEAR PIANT RMP 87 2026A Eliminate Potential Slip or Fall Hazards Add handrails on specified plant building roofs at locations where l surveillances and preventive maintenance activities pose safety i hazards at roof edges. Also add two laddres to the Turbine and ESWS Pumphouse roofs to provide access to security cameras. Each handrail and/or ladder is attached to the plant roofs and are located such that, in the event of structural failure of the connection, the handrail / ladder will not affect the function of safety.related systems. Modification does not affect the Technical Specifications or the design basis for the safe shutdown margin for the plant. The added loads on the buildings from the addition of the handrails / ladders is considered minimal and does not adversely impact the design considerations for the buildings. RMP 89-2010A Add MCB Digital Meter for Indication of Switchyard Bus Voltage Add digital meter to main control board for accurate indication of the switchyard bus voltage. Voltage input'is supplied by the potential transformer on the Mentgomery.Callaway 7 transmission line in the switchyard. This configuration will provide accurate indication of the switchyard voltage and an input to annunciator window-for alerting the operators of high switchyard voltage conditions to protect the main generator step.up (CSU) transformer from switchyard voltages in excess of the transformer rating. Instrumentation associated with the CSU transformers and the generator have no safety design bases. Modification will allow the operators to accurately monitor the switchyard voltage and reduce the generator's output during high switchyard voltage. Added combustibles to the lower cable spreading room and the contrel room is under fifty pounds allowed'by the existing fire protection program and accounted for per procedure. -Added raceway , fill _to the instrumentation trays has been walked down to verify that the new cables can be added without extending above the side railes of the tray. . The added weight of the cable and the meter to the MCB is negligible. All cables and wires will be installed per. applicable installation procedures. Location and the window engraving.of the annunciator window and the meter location and size were review and appraved by. human factors personnel. Circuit is not safety-related and will have no interaction with safety-related circuits. Report Date 06/11/92_ Page 25
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REPORT FOR CALIAWAY NUCLEAR PLANT r 9
+
RMP 90 2002A _f-Add MCB Annunciator Panel to Airborne Radiation Monitor Air Sampling Hest Trace - Add local annunciator panel to airborne radiation monitor air sampling heat trace panel (0J156) to replace low temperature alarm indicating lights. Annunciator panel provides local acknowledgement capability, enabling MCB window 61F to be cleared after control room acknowledgement and then to alarm subsequent low heat tracing temperatures. Modification makes no physical or functional change to the monitor sample line heat tracing. Heat tracing low temperature alarms do not have any safety design basis and modification of the alarms has no impact on the safety design bases of the airborne radiation monitors. c RMP 90-2008A Disconnect Rod Control Interlock of Loose Parts Monitoring System Discor.nect interlock from rod control to loose parts monitoring system (LPMS) to allow LPMS to be available all the time and never disabled. Permits performing Technical Specification surveillance of daily channel checks without concern. LPMS is not required for any accidents considered in the FSAR. _ LPMS still performs as designed with all the channels being enabled and available for daily channel checks. RMP 90-2010A Remodeling of BOP Computer Room Relocate equipment in BOP Room and remove equipme7t no longer in use. There are no safety-related equipment located within or cables that pass through POP Computer Room. Fire protection provided by this modification is at least equal to the previous level of protection. New sprinkler system is installed in accordance with NFPA 13. Fire protection boundary is maintained, detection / protection of the areas is maintained. RMP 90-2011A Replace Center Bar Fall Arrest System Install safety device (fall arrest system) to ladder extending down into the north end of the fuel building refueling canal. Report Date 06/11/92 Page 26
ii } t L 50.59 SAFETY EVALU^ TION
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REPORT FOR CALLAWAY NUCLEAR PLANT No design basis or licensing documents are affected by this l modification. Materials selected for this modification are stainless steel and are therefore compatible with the borated e water in the refueling canal. No loose or moving parts are present in the design. Physical location of the device allows ; free movement of the fuel transfer cart. RMP 90 2014A I Replace Lead. Lag Compensator to Hi T. ave Signc1 for Rod Control Replace Westinghouse 7300 system NMA card with NLL card for i BBTYO412P in order to improve reliability and reduce manhours expended in troubleshooting and calibration. BBTYO412P provides
- dynamic compensation to tha auctioneered high T-AVE signal for development of a-rod demand signal for control of the control
. rods. NLL card is functionally equivalent to the NMA card configured as a lead. lag compensatior but is more reliable because it exhibits less drift. Increased impedance of the NLL card is still within the specified limits of the associated signal source. No credit is take in FSAR Chapter 15 for 7300 system control circuits. N11 card provides control rather than protection function, is . . functionally equivalent to the NMA-in this application, and is more reliable than the NMA. RMP 41 2001A Connections for Temporary Cooling Tower
-Installed non-safety related connections on the non-safety related i
. portion of.the Service. Water (SV)_ system.
In the event of an accident, this portion of the SW system is isolated from the safety-related Essential Service Water (ESW) system. This prevents this modification from affecting any ' safety-related equipment. Electrical power is provided by i non. safety related power and is not routed over any safety.related equipment.
-RMP-91 2005A- y Modification to Leakoff Lines of Auxiliary Feedwater Turbine Relocate steam trap and add an atmospheric vent for the stem lenkoff lines of valves FCFV0313 and FCHV0312 to eliminate backpressure on valves and ensure steaming does not continue to occur in the basement of the auxiliary building.
Change in location of steam trap, addition of vent line, and Report Date 06/11/92 Page 27 l I.
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REPORT FOR CALIAVAY NUCLEAR PLANT changes and additions to various hangers do not impact safety nor i the operation of safety equipment. Pipe stress. analysis indicates stresses in all areas of the modification are within the allowable levels. Affected hangers are within allowable limits for seismic and also for the AISC code. Routing of vent line off of the stem leakoff lines has no impact on operation of FCHV0312 and DCFV0313. Line is. sloped to allow any condensation to drain to the LRW , drain. Failure of steam trap would not impact the operation of auxiliary feedwater pump turbine (KFC02). Performance of the turbine will not be compromised. RMP 91 2011A Modification of DV Supply Conn. to Cold Lab Milli-Q Water Filtering Sys. Change location of demineralized water supply to cold lab Milli.Q i pure water system.so supply line can be routed away from hot
- sample lines. P Plant sampling system serves no safety function and has no other safety design basis (except for containment isolation which does not apply in this situation). Rerouting of domin water supply within the cold lab will affect non. safety related equipment nor any accident analysis addecsed in FSAR Chapter 15.
4 RMP 91 2019A ; Install check Valves on Second Stage Condenser Drain
~
Install check valve in tubing on 2d-stage line to have drain valve operate properly without lifting-the relief of the 2d stage drain line. Previous design included an internal check valve that
. prevented backflow into 2d stage drip line; new valves do not include this feature. Automatic drain valve allows draining of moisture from the discharge of the air compressors for the diesel generator air start system.
Compressors are not safety-related and are not required for the operability of the diesel generators. Failure.of air compressor will not result in failure of air start system. Failure of check valve will result in a failure of the air compressor, but not the air start skid. Since air compressors are not safety-related, the , accident evaluations in FSAR do nct take credit for operability of
- these compressors.
i-
. Report Date 06/11/92 Page 28
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SUMMARY
REPORT FOR CALIAWAY NUCLEAR PLANT NMR/DR 91-100137 Detached Sample Tube of Station Battery NK12 Battery cell #54 of station batter FX12 has a detached sample tube which is floating freely in the cell This is one of two tubes i which extend into the electrolyte so that specific gravity samples j may be obtained at 1/3 of1 the electrolyte depth. Second tube is i intact and can still be used for samples. Sample tube is normally submerged in the cell's electrolyte and is , made of a material that will not degrade in the acid. The tube will not affect the charge or discharge characteristics of the - cell. It has no significanc affect on electrolyte level. A
- representative specific gravity sample can still be obtained from the second sample tube. The detached tube will not affect the '
seismic characteristics of the cell. The tube weighs only a few ounces and.is restricted to fractions of an inch for movement. E OL 1089 Containment Integrity,_ Leakage, Isolation Valve Revise Technical Specifications 3/4.6.1.1, 3/4.6.1.2, 3/4.6.3, and Bases 3/4.u.1.2 which address containment integrity, containment leakage, and containment _ isolation valves. Changes' maintain consistency with existing Technical Specifications by providing an action statement for containment
-leak rate testing in modes 1 through 4. No design change is made that would create possibility for an accident or malfunction of ,
- equipment. Partially approved by NRC via Amendment 62, 9/11/91.
OL 1101 tite Review' Committee Membership Revise Technical Specifications to delete specific title , -designations from Onsite Review Committee (ORC) membership.
- Change is administrative only. Requirements of ORC composition for quorum, for representation at management level, and for specified areas of expertise remain unchanged. Approved by NRC via Amendment 63, 10/8/91.
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l t y __ _ : Report Date 06/11/92 Page 29 s 4-w. , + . A , , - - , _ . . , - - , . . . . - . . , . . . , - . - - . . - , - ~ ~ - ~ . _ .
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l 50.59 SAFETY EVAUJATION
SUMMARY
P.EPORT FOR CALLAMAY NUCLEAR PlRIT
- h. #
OL 1106 Allowable Out-of-Service Times for Analog Channels of ESFAS Revise Technical Specification and associated Bases to extend all< W ble out of. service times (A0Ts) and surveillance test intervals (STIs) for the analo6 channels of the Engineered Salety Features Actuation System (ESPAS) and for the ESPAS actuation logic and actuation relays of the Solid State Protection System (SSPS). l There may be a slight increase in the probability of core damage accidents and a slight increase in core damage frequency (CDF) due to increase ESFAS unavailability. Small potential increase in ' accident probability has been accepted by the NRC Staff when compared to the range of uncertainty in the-CDF and.to the net : benefits to be gained by these changes. The Staff also previously concluded that actual CDF increases for individual plants from the proposed A0T and STI changes are expected to be substantially less than 6t. The Staff considered this CDF increase to be acceptably small when compared to the range of uncertainty in the CDF analyses. Additionally, the Staff. concluded that a staggered test strategy need not be implemented for ESPAS analog channel c' esting
- and is no longer required for.RTS analog channel testing.
l Approved by NRC via Amendment 64, 10/9/91. OL 1107 ' Maximum ' Allowable Leakage of RCS Isolation Valves Revise Technical Specifications to change allowed leakage linit for reactor coolant system pressure isolations valves (RCS PIVs) and to correct valve numbers and descriptions in Table 3.4-1. The change revises the acceptable leakage criteria of the PIVs to l values based on valve size Change does not affect operability requirements of the RCS PIVs or the ability of these valves to perform their intended safety i ' functions. No plant design chagnes are involved and the current
- practices and procedures for monitoring valve leakage are p unchanged. Approved by NRC via Amendment 66, 1/24/92.
OL-1109 i L Surveillance Requirements for Maximum and Minimun ECCS Flowrates l-Revise Technical-Specifications to change charging and safety injection pump flows and to revise requirements for performing a flow balance test on an ECCS subsystem. The design of the ECCS piping, valves, and pumps has been reviewed and found adequate to support operation with increased flow, keport Date_06/11/92 Page 30
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M.59 SAFETY EVALUATION
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REPORT FOR CALIAWAY NUCLEAR PLANT Callaway Safety Analysis has been evaluated based on the proposed changes to the ECCS flow criteria. Consequences of any accident or malfunction of equipment has not increased. Performing a flow balance test on just the affected ECCS subsystem has no effect on e' any accident as the intent of the Technical Specifications is being met. No plant design changes are involved. The current practices and procedures for operating the ECCS system will not change. Approved by NRC via Amendment 68, 3/24/92. OL 1111 - Snubber Visual Inspection Intervals and Corrective Acsions Revise Technical Specification and associated' Bases to change snubber visual inspection intervals and corrective actions. Proposed revisions are consistent with guidance of Generic Letter 90-09. Changes' provide an alternative inspection interval for visual inspection that main the same confidence level as the previous schedule and generally allow the performance of visual inspections and corrective actions during plant outages. Changes do not impact reliability nor availability of plant equipment. Approved by NRC via Amendment 67, 3/5/92. OL 1122 Mode Changes with Control Room Ventilation Tech Spec Action Statement in Effect Revision to Technical Specifications regarding Engineered Safety Features Actuation System Instrumentation and Control Room Emergency Ventilaticn System (CREVS) to take exception to Specification 3.0.4 which prevents entry into an operational mode unless'the conditions for the Limiting Condition for Operation (LCO)-are met. The change allows operational mode changes in
- nodes 5 an 6 while operating in accordance with existing actions which allow continued operation for an unlinited time, ,
Revision is consistent with guidance provided in Generic Letter , 87-09. Changes-do not alter the design or method of operation of i the CREVS or its actuation-instrumentation. Revised actions would allow operational changes while operating in accordance with existing actions which. allow continued operation for an unlimited
-period of time after the system has-been placed in its emergency-(recirculation)-mode-of operation. Oparational-mode changes within the bounds of the action would not degrade the capability .
of-the CREVS to mitig.tesan accident. Changes only alter restrictions on making operational mode changes and the acceptance , criteria for any design basis accident is unchanged. Approved by j NRC via Amendment 69, 3/26/92. Report Date 06/11/92 Page 31 l l _ - . _ . . . . _ _ _ _ _ _ _ . . _ _ . . _ _ . _ _ . _ . _ _ . - - _ - . , _ . . . . . _ ._._...._.._..-,_.,m.-.-_._,,,_,
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I 50.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALIAWAY NUCLEAR PLANT I h OL 1123 i Diesel Cencrator Load Reject Testing Revise Technical Specifications to remove value for largest single ' load required to be rejected for emergency diesel generator -
. testing. Change states that the Essential Service Water (ESW) ,
pump is the largest single emergency
- load. '
Change is administrative in nature and does not involve any design changes or hardware modifications. Surveillance test performed meets the reequirements of the regulatory guides and represents , the worst case loss of a single load. Approved by NRC vis Amendment- 65, 1/13/92. ; ETP.AQ-ST002, REVO Morpholine Test Program. . Procedure to control use of morpholine for pH control of secondary ; water chemistry instead of ammonia to reduce elevated erosion / corrosion rates that have been experienced in the F two-phase regions at Callaway. Procedure objective is to l
- determine. optimum concentration of morpholine required to '
satisfactorily reduce erosion / corrosion of secondary system piping while-maintaining secondary system chemistry parameters within ; chemistry control bands. i Evaluetion of capatibility fo morpholine with secondary side materials is based on available information from laboratory corrosion' studies-and plant experience. A synergistic effect has
- not been ooserved other than its contribution towards the ,
achievement of target pH levels _to-control corrosion.- Jith ; respect to_ contaminant levels, the use of morpholine does not require a change in the guidelines established for secondary water
~
chemistry after extensive study of impact of secondary water chemistry on the potential for corrosion of steam generator tubes and turbine components. Small chagnes in other water chemistry control parameters may be required. Secondary. side treatment with morpholine does not create corrosive condtions for steam generator , materials or feedtrain materials; instead, observations from
. laboratory studies and plant experience indicate morpholine offers better protection to feedtrain material than ammonia. Degradation i of rubber components,11f any,Lis expected to be gradual and is not
- -w- expected to-result-in sudden nignificant loss-of pressure boundary 2
integrity. The condensate chemical addition skid and associated ; valves, pumps and piping were evaluated for compatibility with 40% ! morpholine mixture, and all materials were found to be acceptable. ! The_ use of morpholine in the secondary chemistry is not expected ! to have- an impact on the steam generator thermal. hydraulic l analyses, operating parameters, design transients, or analyses of l postulated accident conditions, nor adversely affect post accident !
. radioactive release calculations.
Report Date 06/11/92 Page 32
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REPORT FOR CALLWAY NUC1. EAR PLANT ETP-3B.01322 REVO Remote Welded Plug Installation Contingency steam generator tube repair method for Refuel 5. A remote weldtug process consisting of a remote weld tool capable of interfacing with a remote manipulator, a plug, and a welding process. Structural integrity of remote wel, d plug indicates that the plug and wald are acceptable for insts C Stress analysis sho7s _.nimum scceptable weld throat is e atantially smaller than those coa;ured in procedure qualification. Failure of one plug is bounded in the plant techncial specification guidelines for safe shutdown due to tube leak. Veld has been qualified and analyzed to insure that the weld is large enought to insure that the plug vill not eject or rupture during any design transient. Plug material was chosen because of its superior corrosion resistance. ETP-BB-01325, REVO Remote Shot Peening of Steam Generator Tubes Shot peening of the ID of the steam generator tubes as a remedial preventive measure against pure water stress corrosion cracking (PWSCC). Vendor has performed a number of tests which demonstrate that shot peening enhances the resistance to PWSCC of Inconel 600 tubing. They have also shown that, under the controls they impose on the - pee-ing process, secondary side stresses are not increased such that they increase the tube susceptibility to secondary side stress corrosion cracking. This remedial treatmu.c of Inconel 600 tubing has previoc31y been performed at several plants throughout the country. ETP.GN-00001, REVO Containment Cooler Performance Test Provides instruction to verify performance of containment coolers and guidelines to predict adverse trend with the heat removal ability of the coolers. Allows RTDs to be installed in thermowells provided to test the containment coolers and cables to be reouted to terminal boxes normally used for GP system computer points. Instrumentation will be removed once required operating data on containment coolers is obtaine6. No change in ESW thermowells is performed by this procedure. Test instrumentation will not affect intended function of GP system i Report Date 06/11/92 Page 33 i 1
50.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALIAWAY NUCLEAR . PLANT during ILRTs and will be removed before next ILRT. . GP system has , no normal or accident design function. Cables will be routed to maintain separation criteria from safety.related cables and routed outside combustible free zones. Additional combustible loading of
- cable insulation is insignificant. Cable is IEEE-383 qualified, fire. resistive cable. Weight of cable is insignificant compared to capacity of components to which cabic will be tied. . Cable will be secured such that there is no potential for it to enter RHR sumps. Weight of RTDs is insignificant compared to capacity of ESV pipe. RTDs en not affect inventory of aluminum or zinc.
F ETP.ZZ.01210, REV4 Steam Generator Nozzle Dam Installation and Removal Nozzle' dams are installed in the primary channelheads of steam generators to allow the RCS to be brought above the channelhead elevation in order to allow refueling activities to take place coincident with primary side steam generator work (i.e., eddy current testing). A nuclear safety evaluation should be performed
-if nozzle dams are to be installed in all four steam generators simultaneously. -This is the case in Refuel 5.
' Union Electric's response to NRC Generic Letter 8817 described actions to be.taken upon' loss of RHR at reduced inventory with nozzle dams installed in all steam generators. These actions
~
- include use of safety injection pumps to provide cooling. Based '
on this, if all hot legs are blocked simultaneously, two safety injection pumps are available taking suction from the RWS1 and injecting into the hot legs when all four legs are blocked simultaneously and no sent path is provided. Two safety injoction pumps provide adequate flow and cooling to prevent pressurization
~fo the upper plenum and prevent core uncovery.
i ETP.ZZ.01320. REV3 Mechanic 11 Plugging of Steam Generator Tubes Procedure' controls installation of mechanically installed roll expanded plug used to remove a steam generator; . tube from service by plugging both its inlet and outlet. Tubes may be plugged following eddy current indication or for preventative reasons. The resules of the vendor's Westinghouse Model F steam generator roll plug test program confirm the leak and structural adequacy of the ' plug.to. tube joint The rolled plugs are fully qualified for ' installation 11n all Westinghouse Model F steam generators. The use of a rolled plug will not affect any safety systems or a system important to safety. I Report Date 06/11/92 Page 34
l 50.59 SAFETYLEVALUATION
SUMMARY
REPORT FOR CALLAWAY NUCLEAR PLANT ETP.ZZ.ST007, REVO Pressure Reactivity Test-Procedure _ raises RCS pressure to 2247_psig and lowers it to 2225 psig for'several hours in order to obtain data to determine whether RCS pressure has-significant effect on core reactivity. , The maximum pressure for this test was selected to assure that, when pressure is : raised, it will be bounded by the assumptions of - the accident analysis'in the FSAR. Technical Specification Table
-3.2-1 requires that indicated pressurizer pressure be maintained
- greater than or equal to 2220 psig. This is above the minimum assumed in FSAR 15.0.3.2.- . The re fore , a low pressure of 2225 psig was selec: %*L ibis test. ' Pressure will be varied by leaving the mastm .
'l montroller in automatic and adjusting the setpoicf. Th. a (ffect' the setpoint of valvo BBPCV0455A whose -f purpr is r- s pressure to prevent the safety valves from Jopeni_ setpoint of BBPCV0455A during this test will be-21_. , well below the 2485 psig setpoint of the
, fpressut - thus, BBPCV0455A will be able to perform its functic - rr an of the test which lowers pressure will lower _ths .. of BBPCV0455A which also will not impair the 4 ability of this valve to perform its function. Although Technical Specifications require this valve remain operable, no setpoint is specified, Therefore,-it can be concluded that it will be
- operable throughout the test. .This test only varies pressure within the' range assumed in the FSAR.
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'OTN.BN-00001,.REVO Borated Refueling Water Storage System Install mechanical- bypass from Containment ILRT system providing Central ch'.11ed Water to heaters on exterior of RWST.
' Central chilled water system ser ves no safety function. Temporary cooling must be installed to maintain RWST below 100 deg F; thus, preventing' plant shutdown. Temporary cooling does not affect any safety.related portion of any system. It does not prevent any of
- the affected systems from performing their design function.
Switching. conducting medium from steam to chilled water will pose
' no problem operationally for heat system after cooling is removed. '
Cooling water is not in contact with tank contents, thus preventing'in dilution of contents. Cooling will enhance RWST and ECCS to ensure peak ' clad temperature during LDCA is 'below limits. Connection of hoses will affect seismic qualification of tank Temporary cooling will ensure compliance with Technical Specifications. 4 i
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Report Date 06/11/92 Page 35 4
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- 50.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALIAWAY NUCLEAR PLANT OTS.KE.00028,-REV2-NL1 1/2 Spent Fuel-Shipping Cask Handling and Loading Procedure The top of NLI.1/2 cask, which holds six fuel rods, is to be raised to the edge of the. cask loading pit. An' alignment pin will ; be removed from an incorrect position, installed in the-proper position, and then the cask will be lowered back to the cask loading pit floor. -The cask lid will not be installed during this operation. .The configuration is an operation that is outside the FSAR description. In the' event that a cask in this configuration were to drop from the top of the. cask-loading pit to the_ floor, the resulting consequences are bounded by FSAR 15.7.4 and 15.7.5 analyses. The NLI.1/2 cask only contains six fuel rods rather than 264 rods that exist in an assembly. The fuel rods were discharged in 1987 which results in a fission product inventory significantly less than that- assumed in the' fuel handling accident analysis. Although only approximately 2 feet of water are over the fuel rods, due to the significantly smaller number of' rods, the consequences of less water are more'than offset. . In addition, due to use of approved plant procedures, the probability of a cask drop is not increased. RFR 07755A-
< Revise Setpoint for'BM-PIC-0072 (Blowdown Flash Tank)
Revise setpoint listed in Callaway Equipment List (CEL) for ; pressure indicator controller located on blowdown flash tank outlet vent to atmosphere. 1 No physical setpoint changes are required. The CEL is revised to match the'present plant setpoint. Revised setpoint in conjuction with the procedural method of operation is consistent with design
- requir6ments.
,1 i RFR 08718A l- Signal Source'for Phase Reference Generator Assembly Drawings M22AC01 and M22ACO3 which appear in the FSAR are revised
-to correctly depict as. built condition. Signal source for phase reference generator assembly is.shown to be on bearing #2 of the j main turbine, but the correct signal source is bearing #3.
Plant drawings are merely' revised to correctly depict as-built condition of the plant. No changes to any plant equipment , results. Whether component is mounted to bearing #2 or bearing #3 L does not change-its intended function. A detector at either I
-position will provide an adequate reference for the phase angle measurement. The component is not safety-related, and its
- . Report Date 06/11/92 Page 36
- 50.59 SAFETY - EVALUATION
SUMMARY
- REPORT FOR CALLAVAY NUCLEAR PIANT-function does_not enhance or interfere with nuclear safety. "The turbine generator has no safety function and has no safety design basis."-(Ref: FSAR 10.2.1.1) > RFR 08728B - Revise - Secondary System pH Alarm Setpoint Revise upper and lower alarm setpoint on pH recorder for following sample points: condensate pump discharge, Icw pressure feedwater, steam generator feedwater, steam generator blowdown. Current alarm setpoints for these sample points results in pH alarms 1 coming in while chemistry pH is within the control band. Chemistry control bands and operating limits are unchanged. Chemistry technician will continue to have valid pH alarm indication. No change'to any chemistry parameters in the secondary system. l RFR 08911B Install Corrosion Monitor of Service Water and Circ Water Systems Permanently mount RCS-8 corrosion monitor on rack where it is currently used to monitor water conditions in Circulating and Service water systems. Also route conduit and phone line to allow
-remote operation via modem.
. Circulating and Service water systems perform no safety function.
- Modification does not affect any safety.related system, structure, or component RFR 09034A Evaluation of Emergency Diesel Fuel Oil Day Tank Level-Instrument Scaling Revise scaling and setpoints of Emergency Diesel Fuel Oil day tank
' level transmitters to make scaling more consistent with specific gravity of fuel oil that-actually exists in the tank. This will also require recalculation of the setpoints for the level switches which receive signal from these transmitters.
- New scaling values and setpoints were calculated using methodology (J.UGEN) to determine. instrument uncertainties and even. included
. an allowance:for transmitter response time which will affect the setpoints. Bases for new setpoints are same as for existing setpoints, and since all instrument uncertainties and process uncertainties are accounted for, this change will have no affect on function of any component. Since change only. involves adjustments to bistables in these instrument loops, the possibility of an accident different than any already evaluated in
___ i Report Date 06/11/92 Page 37
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~
50,59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALIAWAY NUCLEAR PLANT the FSAR is not_ created. _ Indicator reading used to verify compliance with' Technical Specification will-become more accurate
.as a result of this change.
RFR 09076B : Install Safety Line on Radwaste Bridge Crane Install personnel _ tie-off safety line to bottom of bridge girders ; on the radwaste building bridge crane. j I . No safety.related systems, equipment, components, or structures are affected by this modification. Modification may be installed and operated Without increasing personnel radiation exposure beyond normal operating and maintenance activities. Safety line i does not afCect operation of the crane. RFR 09130A 1 Change UHS Low Level Alarm Setpoint Change low level alarm setpoint for the ultimate heat sink (UHS) from plant elevation 1992' 0" to 1993' 2" to maintain the minimum water volume in required in the UHS by Technical Specification t 3.7.5. The existing low level alarm warns the reactor operator that the UHS level has been reduced to below the automatic makeup point and L -is approaching the Tech Spec limit. Since low level switch performs no other function than to control an anntnciator, no control function is affected by this' change. Also, since the ,
-alarm'setpoint i is~being increased, it will-actuate at a more conservative volume than what exists- now. Therefore, the minimum required volume is more certain to be maintained.
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'RFR 09154B Replace ' Insulation on Roof of Condensate Storage Tank p
Replace insulation on condensate storage tank with sprayed urethane foam and acrylic rubber. More heat. loss'can be expected out the top of tank, but this will l be minimal and will not affect the ability of the heating coils to
. maintain the 50'F tank temperature. Condensate storage and transfer system serves no safety function and has no safety design basis. Volume of water. In tank will not be affected. Exposure fire in area of condensate storage tank will not prevent safe
- shutdown of plant.
' Report Date 06/11/92 Pac, 38 i
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i 50.59 S AFETY EVALUATION
SUMMARY
REPORT FOR CALIAWAY NUCLEAR PLANT . RPR 09430A' Change Sulfuric Acid Day Tank Refill Level Setpoint ' Change setpoint for automatic makeup by sulfuric acid transfer pumps to, sulfuric acid day tank. Acid is gravity drained via-a hose in the event both metering pumps are out of service. Reduced operating range of tank level will reduce undesirable flow variations in acid flow rate due to changes in tank driving head. [ No part of any system associated with this setpoint change is
. safety related, and no part of this system is required for safe shutdown of the plant. Change in setpoint will have little or no impact on consequences of failure of the system. No part of any system associated with this setpoint (i.e., Circulating Water .
Chemical Control) is described in the FSAR, considered in accident analysis or covere'd by Technical Specifications. RFR 09637A Rescaling of Water Treatment Plant Sleeve Valve Controller Rescale output circuit cards for cooling tower basin level control. System design requires controller's output control only , one valve to supply the necessary makeup water to the basin. Cooling water system serves no safety function and has no safety design basis. Water treatment plant has-no safety design basis. . - Cooling tower basin level control is not necessary or taken credit for any FSAR accident analysis. Circuit cards and valves that ~ control the water. flow to the cooling tower basin are all classed
- non. safety.
RFR 09834B
- Revise Label on Annunciator Window Delete annunciator window F-65 ("ERFIS IN'.TIATE") from ERFIS logic. - Window will be reused for new ope rator " prog rammable" alarm and will be relabeled "0PERATIONAL PARAMETER 3ETPOINT."
- ERFIS INITIATE function has been duplica ed by the " EVENT- "
field on the computer terminal-status line an.d is unnecessary.
-This window can be used to satisfy operators' request for a
" programmable" alarm. Alarm will be programmed from a template '
- display on the plant computer allowing the reactor operator to L enter his selection of computer points and alarm values. This
. allows operator to use plant computer to monitor selected .
. parameters at setpoints deemed important.
l Annunciator nor plant computer is safety-related and does not
- perform a function other than indication. Modification does not l involve any equipment important to safety. ERFIS nor annunciator i Report Date 06/11/92 Page 39 i
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50.59 SAFETY EVALUATION
SUMMARY
REPORT FOR CALLAWAY NUCLEAR PLANT window are addressed in Technical Specifications, Small ; additional load placed on' plant computer to process operator alarm will not reduce'its ability to process the extr. ting Technical Specification functions. J t 1 i 4 4 4 A , Report Date 06/11/92 Page 40
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