ML20079S210
| ML20079S210 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/19/1994 |
| From: | Woodard J SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9410270002 | |
| Download: ML20079S210 (32) | |
Text
Soutnern Nuclear Operating Company Post 0+fice Box 1295 B>rmingham. Alabama 35201 Telephone (205) 868-5086 m
Southern Nudear Operating Company
- a. o. wooe.ra Esecutive Vice President the southem electnc system Octcber 19, 1994 Docket Nos.:
50-348 50-364 (J. S Nuclear Regulatory Conunission ATFN: Document Control Desk Washington, DC 20555 Joseph h1. Farley Nuclear Plant 10 CFR 50 59 Repo_r1 Gentlemen:
As required by 10 CFR 50.59(b)(2), enclosed is a report summarizing changes, tests, and experiments made or conducted at the Joseph h1. Farley Nuclear Plant (FNP). Units I and 2, as reflected in Revision 12 of the FNP Final Safety Analysis Report.
If you have any questions concerning this report or any included items, please advise.
Respectfully submitted, f
Q l
Jack Voodard
.a LLB/ cit:5059rept doc Enclosure ec: Air. S. D. Ebneter h1r. B. L. Siegel hir. T. A1. Ross L.) J _ i a
9410270002 941019 f
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Report Format and Content As required by 10CFR50.59 (b)(2), this re-The " Description" section for each entry port includes a brief description of changes, contains a brief description of the change, tests, and experiments, including a sum-test, or experiment being reported.
mary of the safety evaluation of each, as reflected in Revision 12 of the Farley Nu-The " Safety Evaluation Summary" contains clear Plant (FNP) Final Safety Analysis a summary of the safety evaluation devel-Report (FSAR).
oped to support the determination that no unreviewed safety question was involved.
Each change, test, or experiment is listed This summary was developed solely for use with an " Identifier." This identifier is typi-in this report-no activities important to cally the Design Change number used by safety have been based on these summa-Southem Nuclear to implement the change.
ries.
The identifier may also include the specific revision of the document; for example "R0" This report is sorted alpha-numerically by in the list would indicate " Revision 0." A "B" Identifier. As FNP has a " shared" FSAR, prefix indicates Bechtel Power Corporation this report contains all changes, tests, and as the responsible design organization; "S" experiments affecting either or both FNP indicates Southern Company Services and units.
"P" indicates an FNP site developed change. For procedure changes, the identifier is typically the title of the proce-FoMer ihah me% ag dure, for example FNP-0-AP-16. Other day tea w egeh iWed b mis Identifiers may also be used.
report, contact Southern Nuclear.
G
Identifier Description Safety Evaluation Summary ABN 93-0-0052, Update of calculations associated with These calculations reflect as-built conditions and operating factors and hase no adverse effect R0.
ingestion of diesel exhaust gas in diesci air on diesel generator performance. These changes do not change, degrade or prevent actions intake.
described or assumed in an accident Assumptions previously made m esaluating the rad:ological consequences of an accident are not altered. These changes wdi not result m a different response of safety related systems and components to accident scenanos ABN 93-0-0079, Resolution of DG system P&lD dtscrepancies. Documentation is being revised to accurately reflect the as-built contiguration of the plant _
R0.
This does not change the design or operation of the diesel generator system.
ABN 93-0-0102, Changes to the containment isolation system -
This change provides added assurance that these valves are not inadvertently opened, R0.
seal wiring of emergency bleed valves in the ensuring penetration boundarv integrity. The purge system is not the initiator of any closed position and deenergizing the MOV.
evaluated accidents. This change does not change, degrade or prevent actions described or assumed in an accident, and does not adversely affect the operation of the purge and mini-purge sy stems.
ABN 93-0-0129, Recalculation of Senice Water drainage This change has no impact on safety since it does not adversely afTect the ultimate heat sink.
R0.
basin.
A Probable Maximum Flood analy sis shows no impact on the structural integnty of the Senice Water pond dant This change will not affect the safe shutdown of the plant and has no affect on the frequency of accidents previously evaluated in the FS AR.
ABN 93-0-0170, Clarification of emergency diesel generator This change is made to clarify the use of the FSAR fuel oil specifications as applying to the R0.
fuel oil specifications in FSAR.
purchase of fuel oil. The emergency diesel generator will continue to function and operate as before with no reduction in reliability since the fuel oil will continue to be provided in accordance with existing specifications. There will be no increase in consequences n
accident previously evaluated. The emergency diesel generators will continue to respond to limit the consequences of a loss of offsite power.
ABN 93-1-0156, Assignment of TPNS numbers for chemical This change does not affect any system, structure, or component which has a safety related R0.
injection pump isolation and vent valves and function. There are no direct or indirect impacts to the design basis of the secondary incorporation of these valves in P&lD.
chemical injection systent None of the affected equipment is relied upon to mitigate the consequences of a malfunction of equipment important to safety.
ABN 93-1-0173, Auxiliary building door number corrections.
Revising these door numbers to reflect as-built configuration will not increase the probability R0.
of an occurrence of an accident. No safety related equipment is affected by this change and the fire rating of related fire barriers is not impacted. This change will not affect annunciation for these doors because the drawings that pertain to signaling agree with the as-built condition.
ABN 93-2-0054, Assignment of TPNS number for the 2B DG Assignment of TPNS numbers to the day tanks, and revision of associated documentation, in R0.
day tank.
no way changes the design or operation of the diesel generator system _
ABN 93-2-0157, Assignment of TPNS numbers for chemical This change does not affect any system, structure, or component which has a safety related R0.
injection pump isolation and vent valves and function. There are no direct or indirect impacts to the design basis of the secondary incorporation of these valves in P&lD.
chemical injection system. None of the afTected equipment is relied upon to mitigate the consequences of a malfunction of equipment important to safety.
Joseph M. Farley NucIcar Plant 10CFR50.59 Report Pago 1
Identifier Description Safety Evaluation Summary B REA 94-0523, RCP 1B motor air cooler replacement The presence and quantities of hydrogen producing materials inside containment is not an R0 (1).
evaluation: additional aluminum in accident initiator. His does not change, degrade, or prevent actions desenbed or assumed in containment.
an accident. Assumptions previously made are not altered since the post accident environment pli is not adversely affected and equilibrium sump solution will remain abose the pH required to assure that iodine is retained in the sump solution. This change will not adversely afTect any structure, system, or components used in mitigating the radiological consequences of an accident.
B-86-1-3516, R32. Addition of public address system speaker at The public address speaker added by this change is not safety related and it does not interface the CAS console.
with any other safety related systems or equipment.
B-86-2-3515, R73. Main power block protected area intrusion This modification contains safeguards information and is available on site. Contact Southern detection system.
Nuclear if further information is required.
B-87-1-4341, R0.
Reactor vessel doughnut modification.
Adequate air velocity will still be available past the CRDM operating coils in the CRDM shroud to provide cooling for the CRDM. He new duct arrangement will be seismically supported. Reduction in containment heat sink will not adversely afTect the containment pressure / temperature profile for LOCA, MSLB, or environmental qualification. The modifications do not degrade the associated structure, system, or components' (SSC) reliability nor adversely affect the SSCs that perform safety functions as desenbed in the FSAR.
B-87-1-4341, RI.
Reactor Vessel Doughnut Modification Adequate air velocity will still be available past the CRDM operating coils in the CRDM shroud to provide cooling for the CRDM. This change will not alter the operation of the CRDM cooling system. The new duct arrangement will be seismically supported. De modifications do not degrade the associated structurc, system, or components * (SSC) reliability nor adversely affect the SSCs that perform safety functions as described in the FSAR.
B-87-2-4106, R35. Replacement of SW branch header piping with This piping replacement does not adversely impact the service water systems' reliabihty or stainless steel.
performance. The modifications will not change the operation of the service water system and system flowrates specified in the FSAR will be maintained. He modification of the containment penetration assemblies does not involve any component whose failure can cause initiation of an accident. This change does not change, degrade or prevent actions described or assumed in an accident. Assumptions presiously made in evaluating the radiological consequences of an accident are not altered.
B-87-2-4106, R37. Deletion of check valve Q2P16V073 in This valve does not perform a safety function and will be replaced with HHC class piping.
Senice Water system.
No credit is taken for this valve in the senice water accident analysis, EQ analysis, or response to R.G.1.97. No systems or equipment required to perform dose mitigating functions are adversely affected. This check valve is outside the containment isolation boundarv and redundant isolation capabilities are maintained.
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 2
i; identifier Description Safety Evaluation Summary
[
B-88-1-4958, R0.
Disconnection of micro-switch on plant The disconnection of this switch does not affect the operation of the presently installed security door #453.
security system for door #453. The presently installed system provides adequate security monitoring for the door.
i B-88-1-5406, RI 1.
Pressurizer loop seal insulation installation.
Revision 1 I includes the disconnection of temporary temperature monitoring instruments and removal of associated equipment after one fuel cycle. The disconnection of the temperature instrumentation and the removal of associated canipment after one fuel cycle will not impact the RCS's reliability or performance.
B-90-0-7074, R0.
Revision to Appendix 9B concerning This change involves no modifications to the fire detection or suppression systems and no evacuating the control room in the event of new combustibles are being added, therefore the probability of a fire is no greater than fire.
previously evaluated. This change implements an existing procedure. His is an established procedure which has previously been reviewed and approved. Existing attemate shutdown procedures for safe shutdown of the plant do not involve less reliable modes of operation than safe shutdmvn from the control room.
B-90-1-6986, R6.
Deletion of vent and drains from room coolers Venting and draining operation for the affected room coolers can be accomplished by other in service water system.
means. These modifications do not adversely afTect the operation of the room coolers or the single failure criterion in FSAR Table 9.4-7. The affected room coolers are not discussed in the accident scenarios described in the FSAR. Deletion of the vent and drain connections does not adversely affect any structures, systems or components used in mitigating the radiological consequences of any accident. No assumptions previously made in evaluating radiological consequences are altered.
B-90-2-6442, R0.
Replacement of R-10, R-11, and R-21.
He equipment covered by this safety evaluation is not the initiator of any accidents. De system measurement range specifications are maintained. These monitors do not perform a post accident dose mitigating function. The reliability of the replacement detectors is equal to or better than the existing monitors. He modification will not afTect the operation of any system important to safety.
B-90-2-6585, RO.
Installation of carpet and fabric wall covering This revision increases the amount of combustibles in fire area 44. This change does not in the TSC.
increase the fire severity in this room. Implementation of this change will not be in conflict with the FSAR and will not decrease the effectiveness of the fire protection program.
B-90-2-6987, RIO. Modifications to vent and drain connections on These modifications will not adversely affect the fu iction or operation of the coolers, nor AFW, CCW, and charging pump room adversely affect the single failure criterion of FSAR Table 9.4-7. The ESF pump room coolers.
coolers are not discussed in the accident scenarios described in the FS AR.
i B-90-2-6987, R7.
Enlargement of drip pan drain for battery his tubing change will not increase the probability of occurrence of an accident. This charging room cooler.
change does not change, degrade or prevent actions described or assumed in an accident.
Assumptions previously made in evaluating the radiological consequences of an accident are not altered. No equipment required to perform dose mitigating functions is adversely affected by this change. There is no adverse impact on system operation.
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 3
s e
identifier Description Safety Evaluation Summary B-90-2-6987, R8.
Auxiliary feedwater pump room cooler These modifications will not affect the room coolers' ability to mamtam room temperatures as modifications.
required, and will not afTect the safe shutdown function of the room coolers. These modifications do not change, degrade, or prevent required actions for any accident No new system performance requirements or failure modes are being introduced, nor will these modifications affect any other equipment important to safety.
B-90-2-7130, Removal / replacement of snubbers in the Evaluation results show that with these pipe support modifications, including the removal of RO&4.
regenerative heat exchanger exclusion area.
5 snubbers, the revised stress levels in the piping and pipe supports are within the applicable allowable limits. The removal of the snubbers and other pipe support modifications will not affect any assumptions previously made in evaluating the radiological consequences of an accident. These changes do not add any new equipment nor do they introduce failure modes of existing equipment which could lead to accidents.
B-91-0-7594, R0.
FSAR change for qualification of H2 The themlocouples are not required for the operation of the Electric Hydrogen Recombiner recombiner thermocouples.
during post-accident operation, therefore they are not required to be qualified So the deletion of the FSAR statement conceming thermocouple qualification does not affect the operation or operability of the recombiners.
B-91-0-7849, R0.
SBO - automatic alignment and remote control This modification provides for automatic and remote control of breakers to eliminate the need of ES02 and ES05.
for local operator action during an SBO event. For design basis events, there is no change to the current alignment of these breakers. The change does not affect any input assumptions used in radiological dose calculations. These changes do not.afrect the reliability of the electrical distribution system. Safety system functioas considered in the FSAR are not impacted. This design maintains the system as class IE.
B-91-1-7687, R0.
Modification oflogic for certain main control The RCP seal standpipe high level annunciator alarm does not perform any control function board annunciators.
nor is there any accident previously evaluated which could be initiated by a malfunction of the RCP seal standpipe. The high level alarm does not perform any dose mitigating function.
This modification will not affect the function of the RCP seal standpipe. This logic change will provide equal or better information regarding the status of seal standpipe level.
B-91-1-7767, R3, Changes to cables and circuit designations The communication system is not involved in the initiation of any accidents evaluated in the RIO,RI1,R13.
utilized in the FTS-2000 system.
FSAR. This design does not change the operational modes, functions, or failure modes of the existing communication system. The FTS-2000 system is designed to be independent of all existing communication systems and therefore does not increase the failuce rate of any i
existing communication channel. No accidents can be initiated by the changes made by this revision.
I i
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 4
f identifier Description Safety Evaluation Summary l
B-91-1-7804, R0.
Re-Identification of auxiliary building No design changes are implemented with this change. Revision of the equipment description chemical injection system components.
and assigning new TPNS numbers will reduce the possibility of tagging, operanons, or maintenance errors. His revision does not change the response to accident scenarios of any mitigating equipment, and will have no adverse efTect on the operation of equipment important to safety except to reduce the hkelihood of mis-operation. No new equipment or failure modes are imolved with this change.
B-91-1-7804, R2.
Re-ldentification of auxiliary building No design changes are implemented with this change. Assigning new TPNS numbers will chemical injection system components.
reduce the possibility of tagging, operations, or maintenance errors. De chemical injection system does not have a dose mitigating function. Assigning these TPNS numbers does not affect the operation of this system or any safety related system. No new equipment, failure modes or credible accidents are involved with this change.
i B-91-1-7850, RO.
Dedication of DG 2C as the alternate AC The proposed changes are limited to one train only and affect systems and components which power source for SBO events.
are not accident initiators. De redundancy provided throughout the plant is not being altered.
All design basis events in Unit I are satisfied with other diesel generators. No new equipment has been added and no new failure modes are being introduced.
B-91-1-7850, RI.
Manual start and automatic loading of EDG The modifications in this PCN revision are minor wiring, labeling and drawing changes.
2C during an SBO.
Rese changes only affect systems and components which are not accident initiators. He redundancy provided throughout the plant is not being altered. These changes do not adversely afTect any structures, systems or components used in mitigating the effects of an accident. These changes do not affect the reliability of the electrical distributien system. No neviequi ment has been added and no new failure modes are being introduced.
2 B-91-1-7878, R0.
SBO - disabling 4KV bus iH sequencer.
He B1h sequencer does not perform any loading function. Disabling of the non-operational circuits and the relocation of the operational circuits does not alter its function. He B1H loading sequencer itselfis not an accident initiator -it responds to plant events that require its i
operation. With the removal of the river water pumps from the B1H loading sequencer, there are no events that require its operation.
B-91-1-7880, R0-SBO - Disabling 4 KV bus 1J sequencer.
This change does not adversely affect the present operation of the B1J sequencer. No 1.
accidents previously evaluated in the FSAR can be caused by the removal of the loading function of this sequencer since the relocation is to the same power source, still within the sequencer. No new failure modes are being introduced. With the removal of the river water pumps from this sequencer, there are no events that require operation of this sequencer.
B-91-1-785 R0.
SBO - Disabling 4 KV bus 1J sequencer -
This change does not adversely affect the present operation of the BlJ sequencer. No i
deletion of the non-operational loading circuits accidents previously evaluated in the FSAR can be caused by the removal of the loading and relocation of the remaining operational function of this sequencer since the relocation is to the same power source, still within the circuit.
sequencer. No new failure modes are being introduced. With the removal of the river water pumps from this sequencer, there are no events that require operation of this sequencer.
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 5
Identifier Description Safety Evaluation Summary t
l B-91-2-7803, R0.
Re-identification of auxihary building No design changes are implemented with this change. Revision of the equipment description chemical injection system components, and assigning new TPNS numbers wi!! reduce the possibihty of tagging, operations, or maintenance errors This revision does not change the response to accident scenarios of any mitigatmg equipment, and will have no adverse effect on the operation of equipment important to safety except to reduce the hkehhood of mis-operation. No new equipment or failure modes are involved with this change.
B-91-2-7803, R2.
Re-Identification of auxiliary building No design changes are implemented with this change. Assigning new TPNS numbers will chemical nyection system components.
reduce the possibility of tagging, operations, or maintenance errors. The chemical injection system does not have a dose mitigating function. Assigning these TPNS numbers does not affect the operation of this system or any safety related system. No new equipment, failure modes or credible accidents are involved with this change.
B-91-2-7836, R0-SBO - containment cooler sequencing.
The changes contained in the design change will not degrade the required containment cooler 2.
availabihty. Hr. handswitches cannot initiate an accident - they respond to manual operator action. He sequencers themselves are not accident initiators - they respond to plant events which require their operation. He sequencing ofonly one containment cooler per train during LOSP events will not increase the consequences of a previously evaluated accident.
Assumptions made in evaluating the radiological consequences of an accident are not altered.
B-91-2-7838, R0.
SBO - load shedding of 600 V load center 21.
These changes will decrease the loads on diesel generator 2B and 2C in some of the design basis events. The equipment affected by this change cannot initiate any accidents previously evaluated. The shedding of the loads connected through this breaker will not increase the consequences of a previously evaluated accident. The loads that are powered through this breaker have no safety function during LOSP or other accidents accompanied by LOSP events. Rese design modifications will not increase the probability of spurious tripping of this breaker.
B-91-2-7844, R0.
SBO -load shedding of fire pump house.
3is change will decrease loads on diesel generators 2B and 2C in some of the design basis events. The equipment affected cannot initiate any accidents. He shedding of the loads connected through this breaker will not increase the consequences of an accident. De loads powered through this breaker have no safety function during LOSP events. Dese modifications will not increase the probability of spurious tripping of this breaker. The sequencers themselves are not accident initiators - they respond to plant events which require their operation Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 6
Identifier Description Safety Evaluation Summary B-91-2-7851, R0-Manual start and automatic k>admg of EDG Although diesel 2C is removed from automatic alignment in all design basis esents, the four 2.
2C during an SBO.
remaimng diesel generators are of sufTicient capacity and redundancy to meet all loadmg requirements of all design basis events These changes are hnuted to one tram only and afTect systems and components which are not accident imtiators. He changes da not adversely afTect any input assumptions used in radiological dose calculation. These changes do not affect the reliabihty of the electrical distribution system. No additional components have been added and no new failure modes are being introduced.
B-91-2-7851, R4.
Changes to FS AR figures and drawings Some of these drawing changes are editorial in nature, and have no physical or functional relative to manual start and automatic loading significance. Other changes were reviewed in previous revisions of this PCN but were of EDG 2C during a SBO.
inadvertently omitted in the associated drawmg. Certain tenninal designations were also changed and drawings rearranged for clarity. A typographical error was also corrected.
B-91-2-7879, R0.
SBO - Disabling 4 KV Bus 2H Sequencer.
The B21I sequencer does not perform any loading function. Disabling of the non-operational circuits and the relocation of the operational circuits does not alter its function The B2H loading sequencer itselfis not an accident mitiator - it responds to plant events that require its operation With the removal of the river water pumps from the B21I loading sequencer, there are no events that require its operation.
B-91-2-7881, RO.
SBO - Disabhng 4 KV Bus 21 Sequencer.
His change does not adversely afTect the present operation of th B2J sequencer. No accidents previously evaluated in the FSAR can be caused by the removal of the loading function of this sequencer. The new power source is equivalent to the old power source. He transfer of power supplies will not create any new credible limiting single failure. No new failure modes are being imroduced. This change does not increase the probability of occurrence of a malfunction of equipment important to safety.
B-92-0-8099, R0-FSAR update for 125 Volt DC batteries.
There are no accidents or other events evaluated in the FSAR which could be afTected by this 2.
change since the batteries are not accident initiators. This change does not change, degrade or prevent actions described or assumed in an accident. Assumptions previously made in evaluating the radiological consequences of an accident are not altered He proposed change does not afTect the ability of the batteries to provide adequate voltage to all safety related components. This change does not adversely affect any structures, systems or components that perform safety related ftmetions.
B-92-0-8109, R0.
Changes pertaining to the spent fuel pool These changes are not expected to increase the probability, consequences or possibility of an (W SECL-91-197, cooling and cleanup system.
accident previously evaluated. This update is not expected to adversely afTect any safety R0) system. There are no physical changes to the SFP as a result of this analysis. This change is not expected to increase the consequences of a malfunction of equipment important to safety, nor have an adverse impact on the integrity and operability of existing safety systems The direct radiation limit for unrestricted access for plant personnel is met.
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 7
Identifier Description Safety Evaluation Summary B-92-0-8275, R0.
Update of tables to reflect containment full Tir revised airborne radioisotopic concentrations values are well below the maximum i
power airborne radiological concentrations and permissible concentrations of the FSAR. Rese changes do not adversely afrect any safety other relevant data.
related system. There are no physical changes made to the containment main or mini-purge systems. The containment mini-purge is not an accident initiator. The NRC SER acceptance criteria remains valid. This change does not adversely affect equipment which has dose mitigating functions. The changes do not affect the containment isolation capabilitics of the containment mini-purge system.
B-92-0-8321, R0.
Update of combustible load calculation for his update increases the fire severity in the Unit I containment from <1 hour to <l.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> FIIA to account for all the combustible loads due to additional cable insulation which has been added. He increase in the combustible that are presently shown in the FSAR.
load does not affect the design requirements for hose stations or the other fire detection systems. No special additional fire protection measures are required and the existing fire protection features are adequate.
B-92-0-8442, R 1.
Addition of fuses to replace the molded case The installation of these components will meet the design requirements and will be scismically circuit breakers for the primary transformer qualified. The seismic response characteristics of the MCCs will not be adversely afrected.
protection on MCCs IF, IG, and IL.
His modification will not cause the systems to be operated outside their design limits. All components will be seismically supported. The changes do not adversely afTect any input or assumptions in radiological dose calculations. These modifications do not diminish the reliability of the electrical distributions system.
B-92-0-8468, R0.
Changes to the fire area hazards analysis.
This change represents a methodology change for calculating the combustible loading of certain cable trays to a method used in other areas of the plant. He increase in fire severity wiil not require any enhancements to existing fire protection features. The results achieved aller this change in calculation methodology will be more conservative but will not afTect the actual combustible loading. The fire suppression systems were reviewed and are not affected by the possible addition of combustibles. This revision does not impact the margin of safety defined in the technical specifications.
B-92-1-7999, R0.
Unit 1 CCTV System.
No accidents of any type could be initiated as a result of the deletion of the Unit 1 CCTV system from the Unit I control room and the TSC. He components associated with the Unit 1 CCTV system are for visual monitoring only and are not used to mitigate an accident. He removal of this equipment has no effect on any other plant equipment, and will not have any effect on the failure modes of any otlier plant systems.
i B-92-1-8143, R0.
Addition of a new relay box to 4kV bus 1F The components used in this modification are class 1E. The 4kV bus IF undervoltage logic sequencer and wiring changes to 4kV bus 1 F which is part of the sequencer will not be afTected by these wiring changes. No accidents sequencer and diesel generator IC logic.
previously evaluated can be caused by these wiring changes. The modifications do not degrade or prevent any other actions described or assumed in an accident. No assumptions previously made in evaluating the radiological consequences of an accident are changed. No new equipment malfunctions have been introduced.
Joseph M. Farley Nuclear Plant 10CFR50.53 Report Page 8
Identifier Description Safety Evaluation Summary B-92-1-X266, R4-Solid State Timers for DG Sequencers B1F &
The replacement timers have better accuracy than the existing relays No accidents 7.
BIG.
previously evaluated can be caused by replacing the existing relays. These sequencer panels are not accident initiators. Changing the time setting will not adversely alTect the diesel operational tinung in the design basis analyses The logic, timing, and functions of the affected circuits have not changed. These changes do not adversely afTect any input assumptions used in radiological dose calculations. There is no impact on the seismic qualification of the sequencers.
B-92-1-8266, R4.
Replacement of Agastat E7000 series timers in The replacement timers have better accuracy than the existing relays. No accidents BIF and BIG sequencers with solid state previously evaluated can be caused by replacing the existing relays. These sequencer panels timers, and certain timing changes.
are not accident initiators. Changing the time setting will not adversely affect the diesel operational timing in the design basis analyses. The logic, timing, and functions of the affected circuits have not changed. These changes do not adversely affect any input assumptions used in radiological dose calculations. There is no impact on the seismic qualification of the sequencers.
B-92-1-8423, R0.
As-built configuration of CCW heat exchanger The "as-is" ductwork does not adversely impact the nonradioactive area ventilation system's room IIVAC ducts.
reliability or perfonnance. The temperature in the CCW heat exchanger room is maintained within the design temperature requirement of 104 degre s F. The ductwork supports are within the design requirements of SS-1102-54 B-92-2-8000, R0.
Unit 2 CCTV System.
No accidents of any type could be initiated as a result of the deletion of the Unit 2 CCTV system from the Unit 2 control room and the TSC. The components associated with the Unit 2 CCTV system are for visual monitoring only and are not used to mitigate an accident. The removal of this equipment has no efTect on any other plant equipment, and will not have any efTect on the failure modes of any other plant systems.
B-92-2-8068, R0-Clarification of amp-hour rating of auxiliary Dere are no accidents or other events evaluated in the FSAR which could be adversely 2.
building batteries and deletion of battery afTected by this change as the batteries are not accident initiators. This change does not manufacturer service testing requirements.
change, degrade or prevent actions described or assumed in an accident. Assumptions previously made in evaluating the radiological consequences of an accident are not altered.
The change does not afTect the ability of the batteries to provide adequate voltage to all safety related components.
B-92-2-8144, R0-DG 1C abgnment and loading logic changes.
The components used in this modification are class 1E. The 4kV bus 2F undervoltage logic 1.
which is part of the sequencer will not be affected. No accidents can be caused by these wiring changes. His change will add reliability to the present system. He modifications do not degrade or prevent any other actions described or assumed in an accident. No assumptions previously made in evaluating the radiological consequences of an accident are changed.
Joseph M. Farley Nuclear Plant 10CFRSO.59 Report Page 9
.---.-----.------------.-----,.,-.-----------------r Identifier Description Safety Evaluation Summary B-92-2-8170, R0.
RIIR vent valve labeling.
Switching the tag numbers of these valves on the P&lD has no potential to cause any accident previously evaluated. Rese valves serve no post accident dose control function D ese valves serve only to isolate the heat exchanger vent. Switching the tags of these valves will not afTect RilR system performance. These valves serve no dose control function.
B-92-2-8447, R0.
Interposing relays for 4 KV breakers.
His modification improves the voltage available at the breaker closing coils and therefore
}
will not increase the probability of occurrence of an accident. -His change does not change, degrade or prevent actions described or assumed in an accident. Assumptions previously made in evaluating the radiological consequences of an accident are not altered. He voltage margin for the operation of the bretker closing coils is being improved and does not adversely affect any safety related structures, systems or components.
B-93-1-8485, R0.
Implementation of black board concept for This change does not adversely affect the systems since additional annunciators are available control room annunciators.
for the CCW and SW which can alert the operators to low flow conditions. His change does not cause these systems to be operated outside of their design limits. His activity does not change, degrade, or prevent actions described or assumed in an accident. His activity does not degrade structure, system, or component (SSC) reliability by deleting nuisance alarms and adding the SW flow indicators. This change does not affect any SSCs that perform safety functions.
B-93-1-8538, RI.
Deletion of unnecessary equipment based on This modification removes certain items that are control grade and perform no protection installation ofmedian signal selector switch.
function. The wiring modifications will not afTect the AMSAC system function or the function of the level loops from which the signal is derived. Neither the items being deleted, nor the AMSAC system perform any radiological mitigating functions.
B-93-2-8486, R0.
Implementation of black board concept for His change does not adversely affect the systems since additional annunciators are available control room annunciators.
for the CCW and SW which can alert the operators to low flow conditions. This change does not cause these systems to be operated outside of their design limits. This activity does not change, degrade, or prevent actions described or assumed in an accident. His activity does not degrade structure, system, or component (SSC) reliability by deleting nuisance alarms and adding the SW flow indicators. This change does not affect any SSCs that perform safety functions.
B-93-2-8683, R0.
Repair of SG 2B level transmitter root valve Addition of this seal cap does not affect the operation of the valve. He manufacturer of the 2
by addition of a seal cap.
valve states that the addition of the seal cap will not adversely affect the pressure boundary l
integrity or the seismic qualification of the valve. He addition of the seal cap has been evaluated from a pipe stress point of view and found acceptable. The seismic qualifications of the valve and the associated piping system are not adversely affected.
r Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 10
i Identifier Description Safety Evaluation Summan r
FNP PC 92-07.
Security Certificates of Destruction.
Changing the description in the FSAR of how revised or obsolete secunty procedures are processed is an administrative type activity and will not affect any Condition 11,111, or IV accident in FS AR Chapter 15, nor affect 'any accident initiator. This change will not increase radiological consequences for any accident and will not affect any plant equipment important to safety. This change will not create a credible accident that could involve any plant structure, system, or component important to safety.
FNP Security Plan. Replacement of manned defensive positions Review of the FSAR revealed that this change will not increase the probability of an accident with closed circuit television as primary as previously evaluated, increase the consequences of an accident, afTect the operation of any surveillance method.
equipment related to safety, or increase the consequences of any safety related equipment malfunctions. Review of the accident types evaluated in the FSAR indicates that this change will not create the possibility of a difTerent type accident. This change does not relate to any equipment important to safety of a difTerent type than previously evaluated in the FSAR.
FNP-0-AP-16, R23 Procedure Change: Conduct of Operations -
This change reduces the normal shiR complement by one system operator, however, Operations Group. Reduces nonnal shiR experience has shown that two operators are not required to meet all requirements of the complement by one system operator.
position. Reduction of the shiR complement by one system operator will not prevent the remaining crew members from meeting all operational requirements required for safe operation of the plant. In addition, the minimum shift complement specified in the Technical Specifications is met.
FNP-0-AP-45, RI 1 Change to Farley Nuclear Plant Training Plan Changes to 10CFR55, Part 59 and NUREG 1021 require FSAR update and changes to AP-
- 45. The proposed changes are administrative in nature and no unreviewed safety question exists as a result of these changes. The changes bring training center administrative procedures and the FSAR in line with current regulations and guidance.
FNP-0-AP-59, R4 Voiding administrative procedure concerning This revision deletes this interface procedure between the APCO Construction Department the Use of Construction Work Requests on and the APCO Nuclear Generation Department. Voiding this procedure would require all Licensed Units.
work (except for a Major Modification or Addition) to be performed using the OQAPM and existing plant procedures. Voiding this procedure constitutes no unreviewed safetv question.
FNP-0-AP-76, R7 Procedure Revision: Conduct of Operations -
The potential for corrosion will not be increased by the use of ETA in the secondary system.
Chemistry and Environmental Group to allow Secondary system materials of construction will not be adversely impacted by the use of the use of ethanolamine (ETA) rather than ETA. The change in chemistry does not affect the function of the condensate, feedwater and morpholine for pH and crosion control.
steam systems at FNP. The ability of the condensate, feedwater, and steam systems to perform their intended functions is not adversely afTected. The use of ETA does not cause the initiation of any accident nor create any new failure mechanisms.
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 11
Identifier Description Safety Evaluation Summary FNP-0-ETP-3657, Initial issue of test procedure for intermediate This entire test evolution is conducted within established FNP operatmg knuts and technical R0 Range Rod Shadowing Measurement.
specifications, and does not place the plant in an unanalyzed condition at any time. This test procedure incorporates appropriate cautions to ensure the Bank D rod insertion limit is not violated during manipulation. Since rods are used to compensate the reactivity effects of dilution, the average net reactivity addition during the test is approximately zero. No changes to any plant system are made or required.
FNP-0-M-015, Voiding of Master Training Plan The Master Training Plan (M-15) is redundant to curriculum guides, administrator's guides, R17 and certain information in AP-45. This manual is only updated annually, while other controls are updated as conditions warrant. This manualis duplicative and unnecessary. As this manual does not affect plant operations, this change will have no effect on the plant's ability to mitigate accidents and does not increase the consequences or possibility of radiation exposure to the general public in the event of an accident.
FNP-0-M-11, R13 Revision to Offsite Dose Calculation Manual The ODCM Revision 13 changes and supporting Efiluent Management System (EMS)
(ODCM) due to changes in 10CFR20 and software are consistent with the requirements of the new 10CFR20 and other applicable various other changes including utilization of regulatory requirements governing radiation protection, and do not involve an unreviewed EMS software.
safety question, contingent on NRC approval of the proposed technical specification changes referenced in the safety evaluation and successful completion of the site acceptance test program for the EMS computer software.
FNP-0-M-30, RIO Revision of Process Control Program (PCP)
The Process Control Program Revision 10 changes are consistent with the requirements of the due to changes in 10CFR20 and various other new 10CFR20 and other applicable regulatory requirements governing radiation protection, changes.
and do not involve an unreviewed safety question, contingent on NRC approval of the proposed technical specification changes referenced in the safety evaluation.
FNP-0-M-49, TCN Change to the Chemical Control Program to The potential for corrosion will not be increased by the use of ETA in the secondary system.
23G allow the use of ethanolamine (ETA) rather Secondary system materials of construction will not be adversely impacted by the use of th:m morpholine for pH and erosion control.
ETA. The change in chemistry does not afTect th: femction of the condensate, feedwater and steam systems at FNP. The ability of the condensate, feedwater, and steam systems to perform their intended functions is not adversely affected. The use of ETA does not cause the initiation of any accident nor create any new failure mechanisms.
FNP-1-ARP-1.8, Change to Annunciator Response Procedure to Defeating this input will ensure that the operators will be alerted of a high level condition R18 defeat the high level input to the MCB alarm should one develop on the other reservoir. High oil level input to this alarm alerts operators for RCP oil level when operating with a known of potential CCW leak into oil system. Diverse indications of such conditions are available, high oil level condition.
such as RCP bearing temperature, CCW surge tank level, and CTMT sump level. Complete loss of the affected RCP or on service train of CCW is the extent to which this actisity can affect any equipment important to safety; the consequences of this eventuality have been fully evaluated in the FSAR.
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 12
l l
Identifier Description Safety Evaluation Summary FN P-1 -SO P-23.0, Temporary change to Component Cooling This wdi reduce the amount of waste water generated as well as the solume of treated TCN 34A Water System operation to provide a closed makeup water required. Since no new system mterconnections are created, this change can I
loop from the CCW pump discharge through only affect the probabihty of a leak in the affected CCW train All components will be the temporarv demineralizer.
appropnately pressure tested, and the system will be monitored. All protective features of the CCW system remain available. This is not an initiating event for any accident evaluated in the FSAR.
FN P-2-ARP-1.4, Change to Annunciator Response Procedure His change in RCP shutdown sequence improves the capability of the No. 2 seal to R14 concerning RCP shutdown sequence.
withstand RCS pressure during cooldown and depressurization. Since these actions are taken l
in response to an RCP No. I seal failure, they cannot affect the probability of occurrence of the failure. Since the actions taken are already assumed to occur, the probability of occurrence of any other accident is unaffected.
FNP-2-ARP-1.8, Change to Annunciator Response Procedure to Defeating this input will ensure that the operators will be alerted of a high level condition f
RIO defeat the high level input to the MCB alarm should one develop on the other reservoir. High oil level input to this alarm alerts operators for RCP oil level when operating with a known ofpotential CCW leak into oil system. Diverse indications of such conditions are available, high oil level condition.
such as RCP bearing temperature, CCW surge tank level, and CTMT sump level. Complete loss of the affected RCP or on service train of CCW is the extent to which this activity can affect any equipment important to safety; the consequences of this eventuality have been fully i
evaluated in the FSAR.
FNP-2-SO P-23.0, Temporary change to Component Cooling This will reduce the amount of waste water generated as well as the volume of treated TCN 31E Water System operation to provide a closed makeup water required. Since no new system interconnections are created, this change can loop from the CCW pump discharge through only affect the probability of a leak in the affected CCW train. All components will be the temporary demineralizer.
appropriately pressure tested, and the system will be monitored. All protective features of the CCW system remain available. This is not an initiating event for any accident evaluated in the FSAR.
FNP-2-SOP-3.0, Change in operation of the Boron Hermal His change allows parallel operation of the BTRS and normal letdown such that there is less TCN 13C Regeneration System to allow using an of a transient on the BTRS and CVCS and the possibility oflosing letdown is lessened. Also, unsaturated BTRS demineralizer for RCS up to three demineralizers may be bypassed. His change does not challenge the design of the dilution without the use of the BTRS chillers.
BTRS or CVCS beyond normal operation of these systems. This change in procedure does not increase the probability of occurrence of an accident previously evaluated, nor increase the dose to employees or the public. The change does not cause the CVCS system to be operated outside ofits normal operating parameters.
FSARC 92-27, R0 Changes to record type identifiers for quality Quality documentation continues to be identified and filed in a manner that supports all (S) records.
activities required by the FNP operating license. De method ofidentifying and filing FNP quality documentation is not relied upon to mitigate the consequences of any accident. Dere will be no increase in the probability of occurrence of a malfunction of safety related equipment as a result of this change. His new scheme complies with the quality documentation guidelines of ANSI N45.2.9.
i Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 13
identifier Description Safcy fvaluation Summary FSARC 92-95, R2 Revision of SWIS batterv mmimum design The batteries have been shown by calculation to be capable of perfornung all their safety (S)(ES 92-2215) temperature and seismic quahfication.
functions in case of a design baas accident, when their minimum temperature is 35 degrees F The reduced design temperature will rot increase the probabihty of an accident of any kmd Purchase of new battenes to IEEE 344-87 will not result in degradation of the seismic performance of the batterie.s. Le reduced design temperature reduces the capacity of the battenes somewhat, but there is still capacity available to perfonn 100% of the design ftmetions of the batteries following an accident.
P-90-2-7126, RI.
Installation of 2 additional phone imes to the These new phones will be located in the TSC and will enhance the communication TSC for the NRC.
capabilities within the TSC. His change does not alter the operation or function of any existing communications equipment within the TSC. His change does not alter any safety related equipment.
P-90-2-7126, R2.
Installation of 2 additional phone lines to the These new phones will be located in the TSC and will enhance the conununication TSC for the NRC.
capabilities within the TSC. This change does not alter the operation or function of any existing communications equipment within the TSC. This change does not alter any safety related equipment.
P-91-2-7722, R0.
Ventilation for new spectrometer.
This system is not required to be operable and will not afTect the safe operation of the plant.
His system does not initiate any accident evaluated in the FSAR. This modif: cation will not adversely afTect any other plant system, and will have no adverse efTect on the ventilation flowrate from this room or any other safety related parameter.
P-91-2-7750, Addition of tubing from the pressurizer cavity This change meets the applicable design, material, and construction standards and will not RO&l.
to stairwell #3 to support RCS level affect normal RCS operation. The proposed change will not afTect radiological barrier monitoring during outages.
performance and will not affect dose consequences of an accident. The proposed change will not degrade the performance of the RCS or any safety system assumed to function in the accident analysis.
P-92-0-8129, R0.
Modifications to secondary chemistry lab air The systems afTected by this change cannot initiate any accident previously evaluated since llVAC.
the operation of the air conditioner or duct heater is not required in any accident analysis and the service water in the turbine building is isolated on a safety injection signal. None of the systems affected by this change has equipment or supports equipment important to plant safety. This change does not change the function of any affected equipment, it only increases the cochng capacity of the A/C and removes the duct heater.
P-92-1-8318, R0.
Modification of auxiliary building door locks.
Administrative control of the doors will be maintained. Locking of these doors plays no role in mitigating the consequences of an accident. No credit is taken for locking these doors in reducing the probability of an equipment malfunction. Locking of these doors plays no role in mitigating the consequences of a malfunction of equipment important to safety.
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 14
Identifier Description Safety Evaluation Summary P-92-1-8318, RI.
Modification of door locks for selected Administrative control of the doors will be maintained Locking of these doors plays no role auxiliary and turbine building doors.
in mitigating the consequences of an accident. No credit is taken for locking these doors in reducing the probability of an equipment malfunction. Locking of these doors plays no role in mitigating the consequences of a malfunction ofequipment important to safety.
P-92-1-8371, R0.
Additional test connections adjacent to steam This change only provides a means to provide double isolation valves for test safety purposes chest to allow data acquisition by plant and will not revise the bases of any previous FSAR evaluation on the probability of an personnel.
accident. This change will not introduce conditions in the main steam system whicri did not previously exist and will not revise the bases of any previous FSAR evaluation of the probability of the consequences of an accident. No acceptance limits will be exceeded.
P-92-1-8387, R0.
Capping of SW line to chlorination building.
This activity does not afTect senice water flow rates, temperature, etc. to any safety related component. Excavation activity will be per site requirements. This change afTects only the service water flow to the circulating water chlorination system which is a non safety related load on the SW system.
P-92-2-8319, R0.
Modification of door locks for selected Administrative control of the doors will be maintained. Locking of these doors plays no role auxiliary and turbine building doors.
in mitigating the consequences of an accident. No credit is taken for locking these doors in reducing the probability of an equipment malfunction. Locking of these doors plays no role in mitigating the consequences of a malfunction of equipment important to safi:ty.
P-93-0-8552, R0.
Modifications to TSC document room and This room modification will not revise the bases of any previous FSAR evaluation, and is not kitchen.
related to any consequence of an accident previously evaluated. These modifications have no efTect on any equipment important to safety.
P-93-1-8649, R0.
Installation of ultrasonic flow measurement The installation of this hardware is not intrusive and will not affect the service water system instrumentation on SW system.
pressure boundary. Operability or reliability of the senice water system and components will not be affected by this modification. This hardware will function independently from other plant systems and components. No systems or components that mitigate accidents or accident conditions will be affected by these modifications.
P-93-2-8650, R0.
Installation of ultrasonic flow measurement The installation of this hardware is not intrusive and will not afTect the service water system instrumentation on SW systent pressure boundary. Operability or reliability of the service water system and components will not be afTected by this modification. This hardware will function independently from other plant systems and components. No systems or components that mitigate accidents or accident conditions will be affected by these modifications.
P-93-2-8681, R0.
Scaling of 2B cooling tower bypass nozzles.
The cooling tower bypass lines are provided for cold weather operation. A failure of any portion of the circulating water system will not affect any systems necessary for the safe operation or safe shutdowr. of the plant.
Joseph M. Farley Nuclear Plant 10CFRSO.59 Report Page 15
Identifier Description Safety Evaluation Summary REA 93-007 (S)
DG dynamic criteria (Regulatory Guide 1.9)
The performance of the diesel generators and the powering of ;afety related loads assumed in FSAR accident analyses will riot change. There are changes m the allowable magmtude of voltage variations during diesel loadmg, but the loads wntinue to perfmm their safety functions. These events do nix cause an accident, but are in response to it. This change does not alter the performance, design intent, or reliabihty of any equipment important to safety. It provides for a more accurate assessment of equipment performance during accident conditions.
S-81-0-1058, RI.
IIVAC redesign in diesel generator switchgear Controls for cycling the fans and opening the supply dampers will remain unchanged. This modification will not reduce the capacity of the ventilation system. Changing the position of rooms.
the exhaust fans will have no efTect on the structural integrity of the diesel generator building.
This modification will only improve the system's reliabihty and performance and will in no way increase the chances or consequences of an accident.
S-81-1-1031, R10.
Installation of a difTerential pressure indicating The new switch has seismic and performance specifications which meet or exceed those of the switch in the senice water line upstream of existing switch. The function perfomied by this switch is alarm actuation only and does not each diesel generator heat exchanger.
initiate any automatic actions. The only safety related function of this equipment is maintenance of the senice water pressure boundary. The new switch will be purchased and installed a.; Class 1 E, seismic category I equipment; associated valves and tubing will be purchased and installed to seismic category I requirements. Accordingly, the pressure retainmg capability of the system will not be degraded.
S-81-2-2066, R14.
Installation of a differential pressure indicating The new switch has seismic and perfonnance specifications which meet or exceed those of the switch in the senice water line upstream of the existing switch. The function performed by this switch is alarm actuation only and does not diesel generator heat exchanger.
initiate any automatic actions. The only safety related function of this equipment is maintenance of the senice water pressure boundary. The new switch will be purchased and installed as Class 1E, seismic category I equipment; associated valves and tubing will be purchased and installed to seismic category I requirements. Accordingly, the pressure retaining capability of the system will not be degraded.
S-82-0-1260, R0-Update oflightning protection design for the Addition oflightning protection to the low level radwaste building (LLRB) will reduce the 1.
low level radwaste building.
chance of an accident due to lightning stnkes and therefore will not increase the probability of occurrence of evaluated accidents. The LLRB is not a safety related stmcture; the addition oflightning protection will not afTect any plant safety rel.ted systems or components.
S-84-1-2914, R36.
SW 2 inch and under pipe replacement with This modification, including the vent and drain lines addition, meets applicable standards and stainless steel has no adverse effect on the system, structure, or components. Stainless steel piping provides adequate strength while mimmizing corrosion. This change will not degrade the operational reliability or availability of the senice water system, nor increase the probability of occurrence of malfunction of equipment important to safety. Scismic requirements will continue to be met.
Joseph M. Farley Nuclear Plant 10CFRSO.59 Report Page 16
t i
identifier Description Safety Evaluation Summary S-84-1-2914, R37.
Replacement of Senice Water carbon steel This change meets applicable standards and has no adverse efTect on the sy stem. stmcture, or piping with stainless steel; addition of a flange components. The addition of this flange will not degrade the operational rehabihty or connection.
availabihty of the service water system.. He function of the service water system will not be adursely affected by this mothlication. Seismic qualification and Code requirements will continue to be met. He single failure analysis remains valid his will have no adscrse effect on any radiological accidents.
S-84-2-2915, R23.
Replacement of all 2 inch and under senice Stainless steel piping will casily provide adequate strength for system piping while water system carbon steel piping with 3:a;nless minimizing the corrosion and crud accumulation problem. Plant reliability will be improved.
steel piping.
He use of 600# valves versus 1500# valves improves the seismic capacity of system piping due to the lesser weight. The 600# valve is rated well beyond t' maximum sersice water system design conditions.
S-85-1-3286, R1.
Addition of new branch sprinklers in the This change has been evaluated for its effect on other systems and has been found to not auxiliary building.
interfere with or degrade other systems. The design is in accordance with applicable codes and standards and with FSAR and licensing commitments. Nn smoke detection components will be added. deleted, or relocated. only wiring changes will be made.
S-85-2-3324, R3.
Removal from senice of battery charging his change does not have any impact on core heat removal by secondary systems, nor does it room cooler DP Indicators.
have any impact to the reactor coolant system. No new failure modes are introduced by this change. The change does not increase the probability of failure of the coolers. or any other safety related equipment, nor does it involve new failure modes of a higher probability or an increase in failure modes for the equipment being changed. This change does not increase the consequences of an equipment failure.
S-88-0-5476.
Primary access point modifications.
This modification contains safeguards information and is available on site. Contact Southern Nuclear if further information is required.
S-88-1-5545, RO.
Update of drawings for BIT sample lines.
His drawing change will not alter the function of the ECCS or the boron injection systems.
He valves were added such the original function or classification of the sample lines is not altered. The function and reliability of the ECCS system will not be affected. He piping and valves meet the requirements of a seismic installation. The change will have no effect on equipment in the system imponant to safety.
S-88-2-5221, R0.
Relocation of RCDT compressed gas Rese drawing changes have been evaluated and will not reduce the effectiveness or reliability cylinders.
of any system, stmeture or component relied upon to prevent an accident. His change has no adverse effect on any of the accidents or transients that may have radiological consequences. No acceptable radiation dose limits for the plant will be exceeded. The relocation of the compressed gas cylinders farther away from the containment / auxiliary building will not increase the consequences of an accident. He affected system is not safety related.
1 1
S-89-0-5983, R0.
Secondary Access Point Modifications.
His modification contains safeguards information. If more information is required, contact Southern Nuclear.
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 17
identifier Description Safety Evaluation 2:mmary S-89-1-5677, RI.
Installation of a new balance-of-plant control There will be no impact of FSAR evaluated accidents. This system is a proven design, system.
widely utdized m the utihty industry. Failure af any control component will result m the same failure mode of the previous systen. and no decrease in accuracy or rehabihty is expected. The replacement system will have the same efTect on the plant process -s as the previous system. The new system's failure modes are no difTerent from the old system and there are no interfaces of this system with safe-shutdown equipment or NSSS protection systems. No new failure modes are created S-89-1-6346, RO.
Installation of a dike at the door of the Turbine This modification will enhance the capability of the FNP Fire Protection Program to perform Building Oil Storage Room.
its intended ftmetion. The addition of this dike will prohibit the spread of combustibles to other areas of the building in the event of a spill in the room. The dike will be designed to contain the maximum amount of combustible liquids which might reasonably be expected to spill in an accident plus the amount of water the room's water suppression system is expected to discharge in a ten minute period.
S-89-2-5678, R0.
Replacement of the analog Leeds and There will be no impact on FSAR evaluated accidents. Failure of any control component will Northrup BOP instruments with digital result in the same failure mode of the previous system and no decrease in accuracy or Westinghouse instruments.
reliability is expected. De addition of new handswitch controls will not affect accident probabilities. No new failure modes have been created and there will be no increase in the consequences of a previously evaluated accidera. Here are no interfaces of this system with safe shutdown equipment of NSSS protection systems.
S-90-0-6413, R0-Removing RW pumps' auto start from pond This system does not function as a precursor to any accident, and the deletion of this function 3.
level.
will involve no safety related equipment. This change does not alter the function of the river water pumps and the remaining redundant level controls ensure adequate pond volume. No safety related equipment is affected by this change nor will the non safety related equipment afTect any safety related equipment due to this change. There is no impact on pond level, thus no equipment important to safety is impacted.
S-90-1-6607, R9.
Replacement of main steam drain carbon steel This portion of the main steam system is non safety-related and is bound by accidents piping with stainless steel.
previously evaluated in the FSAR. This change is in accordance with applicable specifications. Installation of stain! css steel piping will provide resistance to crosion/ corrosion.
S-90-1-6695, RI.
Installation of a new CO analyzer.
This design change will not alter the function of the compressed air system or the breathing air system. The carbon monoxide analyzer is not safety related. Replacing and relocating the CO analyzer will not increase the consequences of an accident. This change does not afTect the function of any safety related equipment. Seismic II/I concerns have been evaluated and no efTects exist. Existing conditions are not altered by this change.
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 18
Identifier Description Safety Evaluation Summary S 40-1-6695, R2.
Installation of new CO analyzer This design change will not alter the function of the compressed air sy stem or the breathmg air system. The carbon monc xide analyzer is not safety related Replacing and relocatmg the CO analyzer will not merease the consequences of an accident. This change does not affect the ftmetion of any safety rel:ted equipment. Seismic II/I concerns have been evaluated and no effects exist. Existing conditions are not altered by this change.
S-90-2-6395, R0.
Change non-rad sump pumps primary seal The safety function of equipment in the CCW room is maintained through operator action.
water supply from RMW to demineralized Water tight walls / doors and drainage systems insure that two AFW pumps remain available.
water.
These systems are not required to mitigate radiological consequences of an accident. His change does not increase the challenges to equipment served by the non-rad sump pumps.
S-90-2-6395, RI.
Changing line number of piping that supplies Changing the line number has no adverse effect on the system, structure or components. This demineralized water to non-rad sump pump has no adverse effect on any of the accients or transients that may have radiological motors.
consequences. There will be no change in the acceptable radiation limits for the plant.
Changing the line number does not adverseiy affect any item performing a safety related function. There are no direct or indirect impacts to the design basis of the system, structure or components.
S-90-2-6746, R0.
Increase in size of feedwater heater drain pot This modification is an improvement of the original design. Eliminating the continuous high drain to the condenser.
level alarms during nonnal operation will alleviate the additional burden on the control room operator. No credit is taken for this system in any accident analysis and interface with safety-related systems is not affected. This modification will occur in a non seismic area and is not safety-related. Rese changes have no adverse effect on equipment in the system which are important to safety.
S-91-1-7661, R0-Modification of the RIIR pumps to a coupled The proposed modifications do not adversely affect RIIR system or CCW system design, 1.
configuration.
operability, or perfomiance. This design meets the required design, material, and construction standards and has no adverse affect on the system, structure, or component. The impeller modification does not impact RHR pump hydraulic performance. He safety function of the RHR or CCW systents has not been altered. His change has no adverse effect on any of the accidents or transients that may have radiological consequences.
S-91-1-7735, R4 Change of setpoint for SW Mini Flow Valves.
None of the operational or failure modes of these valves are considered to be precursors of an accident evaluated in the FSAR. Changing the opening setpoint of the valves will have no adverse impact on the service water system and no impact on the consequences of previously evaluated accidents. The service water design flows will not be affected by lowering the pump minimum flow setpoint. He function of the valves remains unchanged. The valves will perform the same fiinction as prior to the modification.
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 19
f Identifier Description Safety Evaluation Sm nary S-91-1-7758, R0-Feedwater flow venturi DP digital indication.
No credit is taken for the reactor trip initiated by the SF-FF mismatch in mitigating the 1.
consequences of any design basis accident or transient. De median selector switch meets the requirements ofIEEE 279-1971. The main feedwater delta-P signals will continue to be r
handled, routed, and treated in all electrical design as class 1 E signals. The new transmitters will provide all the informaticn needed and the new transmitters are at least as accurate and I
reliable as the currently installed transmitters. This change does not alter any assumptions previously made in evaluatin~g the radiological consequences of an accident.
S-91-1-7760, R0.
Installation oflocal pressure transmitters for This design does not provide any control functions and does not impact any equipment the low pressure turbine exhaust to previously evaluated in the FSAR. This change does not impact any equipment which could condensers.
affect safety, or equipment important to safety.
S-91-2-7736, R2.
Change of setpoint for SW Mini Flow Valves.
None of the operational or failure modes of these valves are considered to be precursors of an accident evaluated in the FSAR. Changing the opening setpoint of the valves will have no adverse impact on the service water system and no impact on the consequences of previously evaluated accidents. The service water design flows will not be affected by lowering the pump minimum flow setpoint. He function of the valves remains unchanged. The valves will perform the same function as prior to the modification.
S-91-2-7757, R0-Addition of feedwater flow venturi DP digital This modification does not affect any portion of the MFW flow loops which proside input to 1.
indicators.
the reactor trip on steam flow /feedwater flow mismatch in coincidence with low water level in any steam generator. The seismic design of the new PDTs is similar to the existing local pressure indicators. This change does not alter any assumptions previously made in evaluating the radiological consequences of an accident. The design change does not affect any fission product barriers nw does it affect any equipment that plays a direct role in mitigating the consequences of a accident.
l S-91-2-7759, R0-Installation of pressure transmitters at the his design does not provide any control functions and does not impact any equipment l
3.
condenser near the low pressure turbine previously evaluated in the FSAR. His modification will not increase the probability of a exhaust.
malfunction of equipment important to safety;and does not impact any equipment important l
to safety.
S-91-2-7782, R0-Replacement of extraction steam to MSR ist Neither the function nor the operation of MSR 1 A or its associated systems is affected by this 2.
stage piping with erosion-resistant materials.
change, and the reliability of the main steam supply system is not adversely impacted. This portion of the main steam supply system is isolated during an accident and not relied upon to mitigate the consequences of an accident. This change will not adversely affect any equipment that has been previously evaluated and noted as important to safety in the FSAR.
This is not a safety related change.
S-92-0-7977, R0.
Tray and conduit details and notes drawing Adding raceway voltage class numbers will not increase the probability of occurrence of an revision.
accident, will have no effect on the operation of equipment associated with raceway or the raceway, and will not increase the consequences of a malfunction of equipment associated with the raceway or on the racewa_v.
3 i
Joseph RR. Fariey Nuclear Plant 10CFR50.59 Report Page 20 m
we-..s
1 Identifier Description Safety Evaluation Summary
[
S-92-0-8040, R0-Replacement for the hot well emergency fill His change has no adverse efTect on the system, structure, or components. Here is no 2.
control valve.
adverse efTect on any of the accidents or transients that may have radiological consequences.
This change does not adversely affect any structure, system or component from performing its safety related function. Here are no direct or indirect impacts to the design basis.
S-92-0-8053, R0.
Water Treatment Plant filtered water tank This change will not change the original design parameters for maintaining tank level. The level control.
water treatment system does not serve any nuclear safety function and no accident analyses would be afTected. The water treatment system is not discussed in any postulated accident j
scenarios, nor is it required for any accident mitigation. The function and reliability of the plant water treatment system will not be afTected by this change. This change will have no l
cfTect on equipment important to safety.
S-92-0-8151, R0.
Revision to clarify the operation of TSC This documentation change does not increase the probability of the occurrence of an accident, I
and does not increase the. consequences of an accident. This change does not increase the probability of occurrence of a malfunction of equipment important to safety.
S-92-0-8198, RO.
Change in the alarm setpoint for the low This change will not alter the function of the diesel generator starting system or the air pressure switches on the 1C and 2C diesel compressors. The failure of the diesel generator is not a precursor to a postulated accident.
generator air reservoirs.
The switches will perform the same alarm function as before but at a higher setpoint. He higher setpoint will help assure the air reservoirs have enough air for the required 5 starts.
This change does not remove or replace any equipment. His change does not affect the function of the low pressure switches and will have no adverse impact on the air reservoirs or the diesel starting system.
S-92-0-8258, R0.
Removal of fuel oil tank vent dryer and Neither the vent modifications nor the Code clarification revise or degrade the capabilities or installation of vent line, and Code clarification. operational characteristics of the diesel fuel storage tank. Basically, this change returns the i
tank vent to its original design configurations. Neither the proposed vent modifications nor the Code clarification will reduce the required availability or effectiveness of the emergency diesel fuel system. A maximum of one storage tank could be lost due to missile damage, however 5 tanks exist and only four operative tanks are required. This change does not alter the basic design of the existing installation.
S-92-1-7931, Updating of drawings associated with steam Changing position of the valves to normally closed on the P&ID complies with procedure RO&l.
generator feed pump seal water.
FNP-1-SOP-21.0 He system design and performance will not be afTected as a result of correcting the P&lD. Correcting this drawing will include showing the correct position of non safety related vahrs. He function and reliability of the system will not be affected by correcting the drawing.
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 21
t I
identifier Description Safety Evaluation Summary I
l S-92-1-8234, R0.
Replacement of a portion of 8" senice water His modification will enhance the ability of the service water system to provide backup piping.
water to the suction of aux feed pump A and to the pump room cooler. The modification will not adversely impact the ability of any system or structure to perform its safety related function. Here are no direct or incirect impacts to the design basis on any system or structure. There is no adverse effect on the previously evaluated malfunctions of equipment important to safety.
S-92-1-8234, R I.
Replacement of a portion of 8" ser ice water These modifications do not adversely impact the ability of any system, structure, or piping.
component to accomplish its safety related function (s). Here are no direct or indirect impacts to the design basis of any system, structure or component. Dere are no adverse efTects on previously evaluated malfunctions of equipment important to safety.
S-92-1-8257, R0.
Removal of fuel oil tank vent dryer and Neither the vent modifications nor the Code clarification revise or degrade the capabilities or installation of vent line, and Code clarification. operational characteristics of the diesel fuel storage tank. Basically, this change returns the tank vent to its original design configurations. Neither the proposed vent modifications nor the Code clarification uill reduce the required availability or efTectiveness of the emergency diesel fuel system. A naximum of one storage tank could be lost due to missile damage, however 5 tanks exist and only four operative tanks are required. This change does not alter the basic design of th: existing installation.
S-92-1-8277, R2.
Modification to the isolation joint between the The flammability of t he material being used has been reviewed and found to be acceptable.
auxiliary building and containment and No safety related eqripment is affected by this change. This change does not afTect the extension of fuel transfer chase drain.
operation of any equipment nor does it affect the function of the isolation joint. De original design intent is not altered. No adverse effects are possible.
S-92-1-8285, R0.
Change to waste processing system P&lD to The change to reflect the correct and as-built piping configuration will enhance plant reflect as-built condition.
operations and help improve plant safety. There will be no impact to the accidents which have been previously evaluated. This change has no impact on system function and will not affect the safety related equipment or increase the probability of an equipment malfunction.
Dere will be no increase in the consequences involving equipment important to safety. This is a change to represent the correct and as-designed configuration.
S-92-1-8293, R0.
Addition of flush lines for the condenser tube There is no increase in probability of occurrence of an accident. This change meets plate seal water system.
applicable standards and has no adverse efTect on system, structure, or components. There is no adverse effect on any of the accidents or transients that may have radiological consequences. There will be no change in the acceptable radiation limits for the plant. There are no direct or indirect impacts to the design basis of the system, stnicture, or components.
Dere is no adverse efTect on the previously evaluated malfunction of equipment important to safety.
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 22
identifier Description Safety Evaluation Summary S-92-1-8294, R0-Condenser tube plate seal water piping his modification meets applicable standards and has no adverse efTect on the system, 3.
modifications.
structure, or components. Stainless steel piping provides adequate strength while minimizing corrosion This change has no adverse efTect on any of the accidents or transients that may have radiological consequences. Here will be no change in the acceptable radiation limits for the plant. This change does not adversely affect the ability of affected items to perform their intended ftmetions. Here are no direct or indirect impacts to the design basis of the system.
S-92-1-8389, RO.
Removal of valve internals in service water He removal of the check valve internals does not prevent sepantion of the service water supply line to control room liVAC condensers. systems for the two units. This function can be performed by the manually operated isolation valves. His change has no adverse efTect on any of the accidents or transients that may have radiological consequences. He modifications do not impact the basis of any existing FSAR evaluation of potential consequences of an accident. These changes do not adversely impact the operations and reliability of the service water or control room flVAC systems.
S-92-1-8409, R0.
Update of documentation to show existing His change meets applicable standards and has no adverse efTect on the system, structure, or control valve in senice water supply to HVAC components. This change has no adverse effect on any of the accidents or transients that may in turbine building.
have radiological consequences. The control valve does not increase flow requirements in the turbine building. The senice water supply to the turbine building is isolated during accident conditions. This change does not adversely affect the system's, structure's, or the component's ability to perfonn their safety related function. There are no direct or indirect impacts to the design basis of affected items.
S-92-1-8431, R0.
Liquid waste disposal drawing update.
Dere is no increase in probabihty of occurrence of an accident. This change will have no efTect on the operation of this system. This change has no adverse effect on any of the accidents or transients that may have radiological consequences, and no change in the acceptable radiation limits for the plant. There are no direct or indirect impacts to the design basis of the system, structure, or components.
S-92-2-8259, R0.
Removal of fuel oil tarb vent dryer and Only one storage tank is postulated to be lost from missile damage to a modified vent. Such installation of vent line, and Code clarification. loss will not reduce the number of tanks below minimum requirements. Clarification of the vent line Code cannot reduce the number of available tanks. This arrangement does not create a new failure mode. The proposed modifications do not reduce the margin of safety.
S-92-2-8278, R2.
Modification to the isolation joint between the The flammability of the material being used has been reviewed and found to be acceptable.
auxiliary building and containment and No safety related equipment is afTected by this change. His change does not afTect the extension of fuel transfer chase drain.
operation of any equipment nor does it afTect the function of the isolation joint. The original design intent is not altered. No adverse effects are possible.
i l
Joseph M. Farley Nucioar Plant 19CFR50.59 Report Page 23
Identifier Description Safety Evaluation Summary S-92-2-8390, R0.
Removal of valve internals in senice water The removal of the check valve intemals does not prevent separation of the service water supply line to control room iWAC condensers. systems for the two units. His function can be performed by the manually operated isolation valves. This change has no adverse elTect on any of the accidents or transients that may have radiological consequences. De modifications do not impact the basis ofany existing FSAR evaluation of potential consequences of an accident. These changes do not adversely impact the operations and reliability of the senice water or control room HVAC systems.
S-93-1-8519, R0.
Conversion of aux feed to the 24V power This modification will reduce the probability of challenges to the reactor protection system by supplies in the rod control power cabinets reducing the probability of reactor trips caused by lightning strikes. He CRDM control from site Jistribution system to the rod drive power supplies are not required for safe shutdown. This change does not alter the function, MG sets.
operation or reliability of the rod dnvc MG sets or the control rods.
S-93-1-8570, R0.
Temperature monitoring system for the Installation of the RTDs and RTD cables on the exterior of the pressurizer will not impact the pressurizer.
RCS reliability or performance. No credit is taken for the pressurizer temperature monitoring equipment for performing any safety-related function. The temperature monitoring instrumentation does not have any dose mitigating functions nor does it affect equipment which does have dose mitigating functions. This change does not result in any original design specification being altered.
S-93-1-8690, RO.
Changes to auxiliary building battery charger His change improves the reliability of the battery charger A.C. supply for design basis feeder breaker instantaneous trip settings to accidents. Proper coordination is maintained with upstream protective devices to prevent loss avoid trippmg on inrush currents.
of other plant equipment for overloads or faults in the battery charger circuit. The proposed change does not alter the capability of the battery charger to perform its safety function.
Probability of charger malfunction is not increased by the changes.
S-93-2-8497, R0-This change provides flush lines for the This modification will allow flushing the condenser tube plate seal water periodically to 1.
condenser tube plate seal water system.
minimize degradation of the tubesheet seal performance. This change meets applicable standards and has no adverse effect on the system, structure, or components. There is no adverse effect on any of the accidents or transients that may have radiological consequences, and no adverse effect on the previously evaluated malfunction of equipment important to safety.
S-93-2-8497, R2.
Assignment ofline number and re-numbering The design remains the same except for the line number and TPNS numbers. This change of valve T'/NS numbers for condenser flush has no effect on any of the accidents or transients that may have radiological consequences.
lines.
Here will be no change in the acceptable radiation limits for the plant. His change does not adversely affect the systems', structures', or components' ability to perform their functions.
There are no direct or indirect impacts to the design basis. Here is no effect on the previously evaluated malfunction of equipment important to safety.
l 1
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 24 t
t identifier Description Safety Evaluation Summary S-93-2-8498, R0-Changing a portion of the condenser tube plate His modification will help to reduce steam generator contaminant levels. This change has no 2.
seal water lines from carbon to stainless steel.
adverse effect on the system, structure, or component. Here is no adverse effect on any of the accidents or transients that may have radiological consequences. There will be no change 1
in the acceptable radiation limits of the plant. There are no direct or indirect impacts to the design basis of the structure, system, or components. There is no adverse effect on the previously evaluated malfunction of tquipment important to safety.
S-93-2-8498, R3.
Assignment ofline number and re-numbering he design remains the same except for the line number and TPNS numbers. This change of valve TPNS numbers for condenser flush has no effect on any of the accidents or transients that may have radiological consequences.
lines.
There will be no change in the acceptable radiation limits for the plant. His change does not I
adversely affect the systems', structures *, or components' ability to perform their functions.
There are no direct or indirect impacts to the design basis. Here is no effect on the i
previously evaluated malfunction ofequipment important to safety.
S-93-2-8549, RI.
Installation details for zinc addition and This installation in no way changes the operability or reliability of the systems in which they monitoring system.
are installed, or systems important to safety. The electrical load addition will not adversely
(
impact the operation of the plant electrical system.
i S-93-2-8549, R5.
Installation of zinc addition and monitoring The ZAMS skid components installed via this change are non safety related. The connections system (ZAMS) components.
to these systems are in accordance with applicable codes and standards, and are of a quality consistent with the associated systems. These connections have been designed such that the capability of these systems to perform their safety functions is not adversely impacted. The seismic aspects of this modification have been reviewed and found to be acceptable. The combustible loads added do not exceed the current fire severities.
S-93-2-8678, RO.
Aux building battery chargers 600V feeder The proposed changes improve the reliability of the battery charger A.C. supply for design breakers current sensor change for basis accidents. Protective devices will maintain equipment protection for electrical instantaneous trip.
overloads and short circuits. Proper coordination is maintained with upstream protective devices to prevent loss of other plant equipment for overloads or faults in the battery charger circuit. These changes do not alter the capability of the battery charger to perform its safety function in mitigating an accident.
SCS SECL (S)
Clarification conceming submersible MOVs his revision is for clarification only and does not represent a physical change to the plant.
(FSAR section 9.2.1.3).
There is no requirement for the MOVs located in the service water valve boxes to be equipped with submersible motors. Additionally, there are no "other areas" which are subject to external flooding which require submersible motors on MOVs. His change does not increase the probability or consequences of an accident, create the possibility of a new or different accident or decrease the margin of safety.
1 Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 25
Identifier Description Safety Evaluation Summary SNC FP SECL, R2 199310CFR50.46 model assessments for The PCT assessment has no cfrect on probability ofoccurrence of an accident since the (S) large break and small break LOCA.
calculation of PCT is perfomted in response to an assumption that the LOCA has occurred.
Model changes which more accurately predict initial conditions and allocating the calculated PCT margin will have no physical effect on the plant. The model changes and allocation of 4
PCT margin have no efTect on assu uptions specifically made for LOCA dose calculations.
System, component and functional integrity in response to a LOCA will be unafTected.
SNC SECL (S)
Executive management organization of SNC in The organizational changes ar: administrative in nature and involve no physical alteration of effect as of 06-01-93.
the plant or changes to setpoints or operating parameters. He physical operation, maintenance, or testing of the plant will not be changed, nor will any changes be made to any current procedures that deal with plant operations. The accident analysis in the FSAR remains unchanged since the organizational changes are administrative in nature.
SNC SECL (S)
Spent fuel pool thermal analysis.
All design criteria, including the standard review plan acceptance criteria are met. The SFP thermal analysis is not considered an initiator for any FSAR transient. No additional fuel failures will result from this change. This change will not create the possibility of an accident of a different type than any evaluated previously. No new failure modes have been defined l
for any system or component important to safety nor has any new limiting single failure been identified.
SNC SECL (S)
Use of Pre-Tect 7000 as pli and erosion-Use of this product is within the limitations ofindustry accepted guidelines, and will not corrosion control agent in the secondary increase the potential for corrosion of secondary side materials, including steam generator system.
tubes. Secondary system materials will not be adversely affected. De change does not adversely affect the overall performance and operation of the affected systems. Use of this product does not alter any of the conditions or assumptions in the FSAR accident analysis.
1 Here is no adverse impact on the performance of any safety related equipment. His change does not alter the design functions of these systems.
SNC SECL (U2)
Cycle 10 reload safety evaluation - Revision 1, This Cycle 10 redesign meets all applicable design criteria and ensures that all peninent Modes 5 and 6 only.
licensing basis acceptance criteria are met. He Cycle 10 redesign is bounded by the current FSAR analysis for the radiological consequences of any accident. De lead rod average bumups do not exceed the lead rod average burnup limits of WCAP-10125-P. No additional mass release or fuel failures should result from this redesign. No new performance requirements are being imposed on any system or component such that any design criteria will be exceeded.
Joseph M. Farley Nurf ear Plant 10CFRSO.59 Report Page 26
identifier Description Safety Evaluation Summary SNC SECL (U2)
Cycle 10 reload, R2.
His Cycle 10 redesign meets all appbcable design criteria and ensures that all pertinent licensing basis acceptance criteria are met. Overall reactor system perfonnance is not adversely afTected by the reked redesign. The core reload is not considered an imtiator for any FSAR transient. The core redesign does not have a direct role in mitigating the radiological consequences of any accident, and does not affect any of the current bases for the current analyses as desenbed in the updated FSAR. No additional mass release or fuel failures should result from this reload.
SNC TS SECL (S) Executive management organization of SNC in The organizational changes are administrative in nature and involve no effect on 12/13/93.
the plant or changes to setpoints or operating parameters. The physical operation, maintenance, or testing of the plant will not be changed, nor will any current procedures that deal with plant operations. Rese changes do not increase the potential for a dose increase above 10CFR100 limits. Dere is no dose increase as the accident analysis remains unchanged.
SNC TS SECL Operation with the excess letdown heat Failure ofletdown or excess letdown does not initiate the occurrence of any accident. He (U2) exchanger isolated excess letdown flowpath is normally isolated to ensure RCS integrity. Manual isolation of the line will not increase the consequences of any accident. He excess letdown flowpath is not considered necessary for safe shutdown or operation of the plant. Manual isolation of the excess letdown flowpath will reduce the consequences of failure of the air operated containment isolation valves on this flowpath from the RCS.
SNC TS SECL, Cycle 13 reload.
He Cycle 13 reload core redesign meets all applicable design criteria and ensures that all R1 (Ul) pertinent licensing basis acceptance criteria are met. He fuel design changes have no adverse impact on fuel rod performance or dimensional stability nor will the core operate in excess of pertinent design basis operating limits. Overall reactor system performance is not adversely afTected. He core reload is not considered an accident initiator. His redesign does not have a direct role in mitigating the radiological consequences of any accident.
W SECL-92-321, Charging flow controller setpoints.
Increasing the charging flow controller limit and the high charging flow alarm limit does not R2 (S) (B-92 change the assumptions used in previously evaluated safety analyses. There is no change to 3287/-2-8288, RI) any assumption used in dose calculations. He magnitude of this change is well within the range of acceptable operation of the CVCS components. All equipment important to safety will continue to function as designed.
TV SECL-93-036, 199210CFR50.46 model assessments for The PCT assessment has no efTect on probability of occurrence of an accident since the R1 (S)
calculation of PCT is performed in response to an assumption that the LOCA has occurred.
Model changes which more accurately predict initial conditions and allocating the calculated PCT margin will have no physical efTect on the plant. The model changes and allocation of PCT margin has no effect on assumptions specifically made for LOCA dose calculations.
System and component and functional integrity in response to a LOCA will be unafTected.
Joseph M. Farley Nuclear Plant 10CFR50.59 Report Page 27
Identifier Description Safety Evaluation Summary W SECL-93-125, Non-continuous operation of the waste gas The revised operation does not adversely afTect the system performance and will not affect R1 (S) system (WGS).
any system or component connected with the GWPS or any other safety related equipment.
The response of the plant to postulated accident conditions is not afTected For a steam generator tube rupture, the most limiting event relative to the concem for increased RCS activity, doses will increase by approximately 50% with this change; however, the calculated doses will remain well below applicable limits. For accidents involving core damage, the increase in coolant activity is insignificant.
W SECL-93-167 Elimination of the low FW flow reactor trip Elimination of this trip function does not affect any FSAR Chapter 15 analyses since this (S) via implementation of the median signal function is not explicitly assumed in any accident analysis. Since the low feedwater flow selector (MSS) in the steam generator water reactor trip is not explicitly credited in any accident analysis, the current level of functional level control system.
diversity will still exist after implementation of the MSS and elimination of the low feedwater flow reactor trip.
W SECL.-93-177 Modification of the anticipated transients The AMSAC is not identified as an accident initiator. The AMSAC only responds to ATWS (S) without SCRAM mitigation system actuation events. The proposed AMSAC change replaces control grade steam generator water level circuitry (AMSAC).
input signals with isolated protection grade steam generator water level input signals. He change does not make any changes to the reactor protection system. He proposed AMSAC change does not make any functional changes to the AMSAC nor is the functional '
performance of the AMSAC adversely afTected. The AMSAC is not a safety-related system and is not credited in any accident analysis.
W SECL-93-192, Modification to RCP 1 A/lB bearing oil pot &
This change will improve level sensing performance. The result of the modification is neither R1 (U1)(PCN B-aluminum inventory reduction.
an accident initiator or mitigator. This modification will not detract from the ability of the oil 91-1-7866, RO) pot to supply oil to the lower bearing during all plant conditions. The modified oil pot will retain structural integrity, and will not influence any accident scenario assumptions. This change will not alter the original seismic evaluation of the pump. His modification will not affect any safety related equipment. No input assumptions used for dose calculations are changed.
W SECL-93-255, 10CFR50.46 assessments for SBLOCA.
The PCT assessments have no effect on probability of occurrence of an accident since the R0 (S) calculation of PCT is performed in response to an assumption that the LOCA has occurred.
PCT assessments and code model assumptions will not affect the probability of an accident.
The dose consequences for the LBLOCA dose analysis in the FSAR are not changed. The model changes and allocation of PCT margin will have no physical effect on plant equipment performance or design criteria.
Joseph M. Farley NucIcar Plant 10CFR50.59 Report Page 28
i identifier Description Safety Evaluation Summary
[
W SECL-93-260, Reanalysis of LOI1TT with increased purge This reanalysis was performed to demonstrate that all Condition il criteria continue to be met.
R0 (S) time ofloop seal for pressurizer safety valves.
This standard reanalysis is performed in response to a postulated event and does not increase the probability of that event. 'Ihe LOI1TT is not a limiting accident for radiological releases.
Previously calculated dose consequences remain bounding. All equipment important to safety will continue to function as designed. No new challenges to pressurizer safety valves are predicted by the reanalysis. No fuel damage is predicted; no new challenges to containment integrity occur.
W SECL-94-005, Zine addition progrant
'Ihe positive effects of Zinc addition (lower RCS radiation fields and reduction of stress R4 (U2) corrosion cracking) are gained with no associated detrimental effects on the wetted surfaces of components in the RCS. As such, the Zinc addition program represents a condition that equates to an enhancement to the original design basis of the plant. LOCA analyses are not impacted. Heat transfer for LOCA accidents is not impacted. Zinc addition will have no significant impact on post accident Hydrogen generation. All previously analyzed accidents remain below 10CFR100 limits.
Joseph M. Farley Nuclear Plant 10CFRSO.59 Report Page 29
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