ML20099A628
| ML20099A628 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 07/22/1992 |
| From: | Gates W OMAHA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LIC-92-206R, NUDOCS 9207290318 | |
| Download: ML20099A628 (24) | |
Text
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..- w Omaha Pubile Power District 444 South 16th Street Mail Omaha, Nebraska 68102 2247 2@
July 22, 1992 LIC-92-206R U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station PI-137 Washington, DC 20555
References:
1.
Docket No. CD-258 2.
Letter from OPPD (W. G. Gates) to NRC (Document Control Desk) dated March 1, 1991 (LIC-91-070R)
Gentlemen:
SUBJECT:
Annual 10 CFR 50.59 Report for Fort Calhoun Station As requirer' by 10 CFR 50.59(b)(2), please find ttached Omaha Public Power District's annual report containing brief descriptions of changes, tests and experiments including summaries of the associated safety evaluations perfc,rmed for the Fort Calhoun Station. This information is for the period of February 1, 1991 through January 22, 1992.
If you should have any questions, please contact me.
Sincerely, AV. $.b4Li W. G. Gates Division Manager Nuclear Operations WGG/sel Attachment c:
tebucuf, Lamb, Leiby & MacRae J. L. Milhoan, NRC Regional Administrator, Region IV R. P. Mullikin, NRC Senior Resident Inspector S. D. Bloom, NRC Acting Project Manager N 00i'r C 7
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r ATTACHMENT As required by 10 CFR 50.59(b)(de)scriptions of changes in th 2, Omaha Public Power District is providing its annual report containing brief changes in the procedures, and changes in tests or experiments conducted ncluding summaries of the associated safety evaluation:; performed for the fort Calhoun Station. This information is for the period since the end of last year's report, February 1, 1991, through January 22, 1992.
Chances in the Facility Packaae Not Descriotion/Ana1Ysis MR-FC-80-104 Radwaste Building Descriotion:
This modification provided additional space for radwaste
)rocessing, packaging and temporary holdup.
The new Radwaste Processing Building facility for the Radwaste Grou(RPB) also provides an office p.
Sa?ety Analysis:
All changes to systems were non-CQE systems.
Interface of new limited-CQE RPB and Auxiliary Building had been evaluated. There was no increase in accident probabilities Piping penetrations installed in the Auxiliary Building shear wall did not cut reinforcing bars, therefore, the wall capacities were unchanged.
The opening in column row U shear wall in the Auxiliary Building was filled with reinforced concrete to provide the wall with the same load carrying capacity as before the opening was made. No design basis of any barriers was affected.
No increased radioactive release paths were created. A new discharge of HVAC exhaust for both the Radwaste Processing Building and CARP Building was created.
This exhaust is HEPA filtered and monitored and will not significantly (effect the' total plant release.
The Design Change Package DCP) establishes that the requirements of 10 CFR 20 & 100 and 40 CFR 190 have been maintained. Failure of limited-CQE RPM would result in an accident that has been previously evaluated in the USAR, and would not result in an accident limiting ability to safely shutdown and maintain control of the reactor.
The Technical Specifications were amended as noted in the DCP.
None of the changes impacted a margin of safety as identified in the Technical Specifications.
During construction, configuration of the plant was in accordance with -Technical Specifications requirements.
There was no reduction in the margin of safety.
During testing of gaseous effluent monitors, systems unaffected by the modification were available to ensure no degradation of nuclear safety occurred. The plant remained in an analyzed condition.
The ' modification did not change the primary pressure boundary or result in reduction of any barriers which prevent release of radioactivity.
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4 MR-FC-80-121 Additional Nuclear Alarms
==
Description:==
This modification was to improve coverage of the Emergency Alarm System (EAS).
For this purpose a gai-tronics mult-tone generator was added to the gai-tronics system, and additional fleshing lights were installed at various locations in the plant. The power feeds to the EAS from Al-42A were removed and replaced with a power feed from MPP-14 which is fed from the security diesel. This decreased the loading of inverter #1. The tone of the tone generator was changed from siren mode to steady tone due to the fact that plant personnel could not tell the difference between the nuclear alarm and the fire alarm.
Safety Analysis:
Tfns modification involves only non-CQE equipment which has no adverse impact on other plant systems or components. The response of the EAS will be ecual to or better than the response prior to the modificat' ion.
There is no interface between this system and any system or com)onent that could cause any previously analyzed malfunct< on to equipment important to safety or-increase the consequences of a malfcnction of equipment imgortant to safety.
Failure of any component of either the cAS or Gai-Tronics system would not result in an accident that could impact the ability to shutdown and maintain the reactor or control the release of radioactivity.
During conctruction and post maintenance testing there were no Tednical Specification requirements for any component or system in this modification. The test configuration is identical to the final configuration.
MR-FC-83-004A Remaining VA-66 Flow Problems - Part 1 (Mechanical)
==
Description:==
This mouification installed the VA-121 booster fan to increase flow to VA-66 with associated duct and dampers This modification was abandoned in place and not hooked up i
l electrically FC-83 0048) based on revised fuel handling accide(under nt for FCS. The duct work, fan, and dampers will be maintained in original configuration for flow, with d:mpers locked open.
t I-l Safety Analysin l
The ductwork, fan and dampers meet the design and l
construction standards of seismic installation cnd leak l
testing.
The system performance was not changed nor were the characteristics changed by leaving the equipment in l
place and inoperable.
Leaving VA-121 in place and inoperable does not increase radiological consequences because the analysis shows that conservatively ignoring VA-66 charcoal filter during a fuel handling accident keeps radiological consequences well within 10 CFR 100 limits.
l The three dampers w.11 all be locked open to provide for least resistance flow and leaving the VA-121 fan components in place will not introduce a different possibility of an accident, or change safe shutdown systems or radiological
4 consequences.
The vendor states that leaving the fan blades, motor and bearings to rotate is preferable to no rotation.
No new failure modes have been introduced of a different type. No credit is taken for the system during a fuel handling accident. The system remains in its Technical Specification rec uirement configuration.
The as-built system is surveillance tested on a monthly basis for VA-66 i
per Technical Specification 3.5 aiJ the safety margins are not affected.
Post modification testing on the installed duct and dampers was equivalent to original testing and has been successful since 1984.
Flow testing is not to be performed becarse the VA-121 fan will not be made operable.
The fan is not needed and does not impact the existing system.
The initia leak test of the installed duct was ruccessful. No testing will be performed so there will not be a configuration which would result in a change to the primary pressure boundary or the reduction in any barriers which prevent the release of radioactivity to the environment.
MR-FC-85-049 Security System Upgrade
==
Description:==
This modification provided for 1) a security computer, 2)4) a security video switcher, 3) a security system simulator, security consoles, 5) riverbank stabilization south of the Intake Structure, 6) expanded Protected Area fences / Jersey barriers 7
a security diesel enerator and ower distribu[ ion) system, 8) & security bninterruptable hower Supply (UPS) system, 9) a manhole and duct-bank network with cable trench system on the Protected Area perimeter, 10)
Protected Area lighting (high mast & wall mounted)facil)ity
, 11 a
Protected. Area access portal (Warehouse), 12) improvements for the new computer, video switcher and simulator, 13 halon fire-
)rotection systems in the Warehouse and ) Security Build'ng for computer & console equipment,14) closed circuit television towers and closed circuit televisions,15) a microwave link detection system, ala)rm only door,17)/ lights),
16 access control hardware for each access control door and Very High Radiation Area door local alarm boxes (horns 18)19) radio communications remodeled Security equipment, antennas and tower, Building Communications Room, 20) site drainage north of Warehouse, northeast of Maintenance Shop and the entire southeast corner of Protected Area.
Safety Analysis:
The-security system and the modification do not involve CQE equipment. The modified system cannot initiate an accident.
No adverse interactions with other plant systems were identified. The modified system does not form a portion of the primary pressure boundary or a portion of a system designed to assure fuel integrity or mitigate the consequences of an accident.
The modification does not interface with safety systems.
J MR-FC-86-ll6 Chemical and Radiation Protection (CARP) Building - Locker Facility, Office / Cafeteria Addition D m riot"on:
This modufication provided a new lab facility to replace a muct smaller lab in the Auxiliary Building; a locker facility for both OPPD personnel and contractors; and office space for Chemistry and Radiation Protection personnel. An additional facility (office, cafeteria building)/ cafeteria was added to the scope of the CARP facility.
The office includes a 10,000 square foot building adjacent to the CARP Building.
A large kitchen cafeteria and several offices were built to provide additional cafeteria space and office space.
Safety Analvsis:
The Limitec-CQE structure met design functions for Limited-CQE structures listed in USAR.
The modification did not increase the probability of an accident.
Piping penetrations installed in the Auxiliary Building wall did not cut any reinforcing bars, therefore, wall capacities remained unchanged. An opening in column / row 9 shear wall in the Auxiliary Building was filled with reinforced concrete to provide the wall with the same load carrying as before the opening was made. The CARP Building is Limited-CQE but does not adversely affect the pressure boundary or a radiological barrier. No new release paths were created.
Release of HVAC exhaust is through the Radwaste Processing Building.
Interface with CQE structures Auxiliary Building) has been evaluated in the Design Package (.
Failure of the Waste Disposal (WD line between the CARP and the Auxiliary buildings would) allow the contents of the line to flow to the Auxiliary Building. A liquid radwast spill in the Auxiliary Building has been previously analyzed in the USAR. Total volume of the water and contamination levels in the WD line is insignificant. No reduction in the margin of safety resulted.
During construction, precautions were taken to ensure that any activities involving potential interface with CQE structures did not impact nuclear safety.
MR-FC-88-009 RM-065 VIAS Actuation
==
Description:==
This modification eliminated the need for Operators to l
manually activate RM 065 in an accident situation.
- Also, this modification relocated RM-065 due to the changes in the Control Room per modification MR-FC-81-51.
Safety Evaluation:
n diation monitor RM-065 does not have a safety-related l'
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function and its performance was not changed.
All H
components were seismically analyzed to preclude "2 over 1" concerns and electrical separation is maintained.
The new interface between RM-065 and the Ventilation Isolation Actuation Signal (VIAS) is electrically isolated to prevent l
challenges to the VIAS components. The modification has no direct or indirect interactions with the fuel, the pressure
- boundary, or containment berriers which limit the consequences of an accident as defined in the USAR. Failure of D.H 065 will not challenge CQE systems or components which interface with these barriers.
MR-FC-88-049 Instrument Air Dryer Installi. tion Description-The instalTation of a new hr.aterless air dryer CA-12 along with prefilters CA-llA/B and CA-13A/B. This new air dryer will be utilized in parallel with the existing air dryer CA-4.
Safety Analysis:
The modification does not involve CQE equipment. Instrument air is not referenced in USAR Chapter 14 as an accident initiator. No adverse int.eractions with other plant systems were identified.
The modification does not involve or impact any of the barriors which limit the consequences of an accident.
CQE equipment which utilizes instrument air are designed to assurre their ' fail safe' position upon losses of instrument tir.
Addition of new t.quipment will increase reliability of the. instrument air system. Limited CQE pipe supports are seismically designed to maintain their integrity.
The instrument air system is not addressed by the Technical Spect fications, therefore, no technical specification margint, of safety are reduced.
MR-FC-89-061 Normal and Emergency Lighting Upgrade Descriotion:
This modification provided for additional AC-DC emergency lighting with self contained battery supplies. Additional normal 119hting fixtures were also added in the Diesel Generator Rooms-(63 & 64).
S"afety Analysis:
I,ii, modification only involves non-CQE equipment which has no adverse impact on other plant systems or components.
There is no interface between this system and any safety-related system or component that could cause any previously analyzed malfunction to equipment important to safety. The response of thit modification will provide adequate lighting to comply to the requirements of 10 CFR Appendix R.
- Also, it will provido adequate normal lighting in both diesel rooms.
The Emergency 1,ighting System is non-CQE and there are no interactions between any equipment or systems that could increase the consequences of a malfunction of equipment important to safety.
The failure of the system would not result in any accident limiting the ability to safely shutdown and maintain the reactor or control the L
release of radioactivity.
This mcdification dot. not increase the margin of safety defined in the Technical Specifications.
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t iiR-FC-89-085 '
Diesel Generator (DG) Air Inlet Damper Control
==
Description:==
This modification provided for the replacement of six(6) solenoid operated valves (SOVs) for DG-1 and DG-2 fresh air inlet dampers.
Safety Analysis:
The S0Vs that were replaced have the same form, ff *., and function as the originals The new S0V's have a mud, wider temperature operating ran e, enhancing the availability of the Diesel Generators (DG.
Fuel integrity was maintained by the assuranc2 that the Gs would be available when called upon.
The DGs remained in a condition allowed by the Technical Specifications.
Redundant systems will be available during tes+ m Should the SOVs fail, the louvers fail open allowing wi '. ion and cooling air flow to the DGs.
MR-FC-90-018 Waste Drai. Line for "B" Steam Generator
==
Description:==
This modification provided for 9e removal of a dead leg from a 4 inch drain header conti.iing a 100 REM "H0T SP01 and installation of a flanged flushing connection to facilitate removal of future waste collection in this line by flushing or. removal of the flange and cleaning.
Safety Analysis:
This modification had no operational effect on Fort Calhoun Etation.
It was performed to reduce radiation exposure to plant personnel. Addition of the flush connection will not 1.tcrease the probability of an accident.
Any lines which were isolated during installation were isolated in accordance with -approved tagout and Shift Supervisor appioval.
The irstallation did not put the plant in a condition which could degrade nuclear safety.
No special post modification testing was done. A hydrostatic leak test of the affected piping was not required by B31.1-1986, Section 137.3.2, since the line is open to the atmosphere and drains downstream of the last shutoff veive.
- MR-FC-90-027 RWP Building Room 29 Tie-ins Descriotion.1 Inis modification provided for MR-FC-88-121 to install piping tie-ins for the Radwaste Building.
This includes piping tie-ins which route through the Volume Control Tank Room, Room 29. Room 29 becomes uninhabitable when the Plant is on line. To make the Radwaste Building functional prior
'to the 1992 Outage, Room 29 needed to be completed.
ety Analysis:
i ftimically supporting the capped extension does not change probability of occurrence of a previously evaluated accident. The system interaction checklist looked into all interactions and found no new initiators.
There is no contact with any equipment important to safety, except for
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. proximity, therefore no. changes to the Technical Specifications are required.
-MR-FC-90-035:
TM-89-E-007-and-TM-89-M-53 Closure
==
Description:==
'lhis modification provided for the closure of Tomporary Modification 89-E-007 which was written to remove the dual setpoint switch on RM 061. This modification also provided for the-close out=of Temporary-Modification 89-M-053 which was written to' install an attenuator plate on RM-063H.
Safety Analv h+. -
RM-06SH is non-CQE. Although RM-061 is CQE, removal of the setpoint switch returns the-monitor to its original configuration.
The modification also removed a potential
~ missile hazard' from RM-061.
The installation of an attenuator plate on RM-063H provided the proper overlap with RM-063M.
The modification does not affect containment boundaries or radioactive release paths.
The system impacted tby this modification are not covered by the Technical Specifications.
MR-FC-90-049 SIGMA Meter Replacement 7
Osscription:
Inis modification provided for replacement of existing variable ower trip power margin indicating meters A/JI-JI-007X,Y, C Also 007X,Y, ication adds /JI-007X,Y, and D/JI-a7X,Y.
this mod resistors in order to compensate for the impedance difference with the original meters.
Scfety Analysjn The modification did not change the purpose or function of the circuit ~.
All work and testing was done dcring hot or cold = shutdown.
Testing was done in accordance-to the PRC approved testine proceoure and did not -affect nuclear safety. The modification changed the output circuit of the APD calculator-to facilitate proper impedance matching
-between the calculator and the new meters. The function of the calculator will not be affected.
MR-FCa90-074 CEDM #9 and #13 Modification-Descriotion:
This modification provided for the removal of the upper housings of CEDMs #9 and 813. These housings were replaced with blind flanged joints, which included metal o-rings, bolts, nuts, and washers.
Safety ' Anal ysi s: -
The CEDM pressure housings that were removed were unused.
. These housing. locations do not contain any mechaaism or other active component.
These spare locations were originally designed for possible future use.
The unused CEDM housings consequently do not serve to mitigate the consequences of any accident or malfunction of equipment important to safety.
The safety analysis report remains unchanged after removal of #9 and #13 CEDM.
The integrity 7
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.!of the: Joint is maintained by 8: studs', washers, and nuts.
Removing #9 and #13 unused CEDMs. and installing the CEDM-blind flange' did not reduce the margin of-safety defined in the basis for the reactor. coolant leakage <11mits..The leakage limits remained the same to provide the same degree of assarance--that the chance of a crack or seat failure would not progress to an unsafe conditions without detectia and proper evaluation ~. It is considered that the margin of
- safety will :likely increase ra6ner than decrease.
The reactor coolant-system remained in a mode as governed by the Technical Specifications resulting in a previously analyzed plant condition. There was no increase in the probability of can -occurrence and no impact-on other plant systems.
There 'was no breach to the barriers for release 7f radioactivity and no affect on post accident modes to other CQE systems. The plant was in cold shutdown and the RCS was
' tagged out.'. CIDMs #9 and #13 - were leak tested under Reactor Coolant System operating conditions already described in the USAR.
The modilication did not have the credible potential for directly/inorectly damaging CQE
- systems _or components.
MR-FC-91-027-HELB in Room 57=P'rt I a
==
Description:==
his codification provided for removal of the 2 inch tie-in-piping to the 8 'nch Auxiliary Stea;c Header. in Room 81.
This removed the steam supply-through Room 57 and AS-533 in Rooit 81.
Safety Analysin This modification effectively removed the existing capability to heat the diesel generator rooms. This is not a safety concern because engineering calculations (FC05765) indicated the diesel generator rooms will be maintained above 15'F (lower DES limit due-to. jelling of fuel oil).
nuclear safety.- -The gh energy line in Room 57 increases Elimination of-the hi new tap-in-Room 82 has no effect on nuclear safety because equipment in that room has no safety
-function.
The Lauxiliary steam pip ng was isolated and vented prior to cutting, the supply p pe was cut before the relief valve-pipe in-Rooms (AS-533) was removed an the remaining capped 57, 64,.63, 65 and 66 were left vented to L
prevent pressure-fluctuations due to temperature changes.
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These precautionary steps addressed overpressure protection and flooding issues during-installation.
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1 4
t Chanaes in Procedures /Temocrary Modifications (TM)
Procedure /TM No.
Qgigtlption/Analysi s A0P-22 Reactor Coolant Leak Descriotion:
The purpose of this procedure change was to:
protect the charging pumps from air bindin when e
switching from the Boric Acid Storage Tank (BAST)gto the Safety Injection Refueling Water i.nk (SIRWT) ensure that containment isolation criteria / requirements e
are met during a LOCA e
make steps 44 and 45 of A0P-22 consistent with similar steps in E0P-03 (Rev. 15 - Step 3.50 & 3.51) e correct an incorrect reference in step 44.C.
Safety Analveis:
A0P chan apply only after an accident has already occurred.gesOperator actions in the AOP are not initiating events for any equipment malfunction or accident evaluated in the USAR. The A0P change ensures that penetration M-3 is operable ?.nd capable of preventing containment leakage once the charging pumps are stopped. Therefore, the consequences of an accident (LOCA) have not increased. Operator actions in the AOP change prevent containment leakage via penetration M-3.
As concluded containment integrity is maintained and penetration M-3 Is operable. The margin of safety as defined in the basis for Technical Specification 2.6 is not raduced by the A0P change.
CH-SMP-MI-00ll Non-Routine Sampling
==
Description:==
This is a new procedure for performing non-routine sampling while ensuring compliance with Standing Orders (50) S0-0-1 and S0-0-44.
Safety Analysis:
The new procedure precludes the loss of containment integrity and ensures compliaoce with appropriate administrative controls (Standing Crders 0-1 and 0-44) as required in the USAR. Compliance with the procedure ensures proper conduct of operation of a valve for sampling to ensure no increased consequences of an accident.
The procedure only allows valve operation for sampling, with appropriate approval, which could not cause a malfunction of any equipment. Operation of a sample valve could not cause an increase in the consequences of a malfunction or reduce the margin s
- safety.
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.t CH-SMP-PR-0011 Safety Injection and Refueling Water Tank Sampling
==
Description:==
Chemistry Manual Procedure (CMP) 2.1 has been upgraded by Project 1991 to comply with Fort Calhoun Station Writers Guide. CH-SMP-PR-00ll is 1 of 22 procedures that superseded CMP 2.1 Rev. 14.
Safety Analysis:
There is no tecident evaluated in the USAR that is applicable to tais evolution.
The activity has no safety related functions.
Even though the sampling valves described in the procedure are CQE the activity will not decrease the design specifications or the reliability of the sampled system.
This is a sampling procedure manipulating CQE valves and will not affect the function of the sampled system. The sampling activity is not likely to cause SIRWT level to fall below minimcm requirements as stated in the Basis of Technical Specification 2.3.
The tank volume is routinely maintaiaed at a
level significantly above Technical Specifications requirements.
A potential malfunction would be that the valves were left open.
However, the change in level of the Safety Injection and Refueling Water Tank has been previously evaluated.
':0P-03 Loss of Coolant Accident
==
Description:==
The purpose of this procedure change was to incorporate steps ensuring the charging header is kept pressu.'ized by the High Pressure Safety Injection (HPSI) p'. imps during a loss of Coolant Accident (LOCA).
Safety Analysis:
The procedure change applied to a procedure for dealing with an accident previously evaluated in the USAR.
Since the accident has already occurred, the change cannot increase the probability of an occurrence of that accident. Critical post-LOCA paranei.ers such as peck fuel clad temperature and containment peak pressure are still within limits with this change.
Fuel integrity and containment integrity were not impacted, so the consequences of a LOCA were not increased.
A single failure of safety-related equipment with the change will not impact containment integrity or fuel integrity.
Therefore, this change will not increase the consequences of a malfunction of safety-related equipment. The alignment of HPSI to charging will not cadse a different type of accident in addition to the LOCA, as component ratings would not be exceeded.
Therefore, a different type of ace Ment not previously analyzed in the USAR could not occur. Jlso, the I
possibility of a different type of equipment malfunction was not created by this change.
HPSI and Chemical and Volume Control System will still oe capable of performing their respective safety functions without a reduction in the margin of safety for fuel clad temperature or containment pressure.
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E0P-20 & TBD Functional Recovery Procedure
-E0P-03 & TBD Loss of Coolant Accident
==
Description:==
The p rpose of this procedure change was to protect the charg ng pe.ps from air binding by switching suction from the B ric Acid Storage Tank (BAST) to the Safety injecti1n Refueling Water Tank (SIRWT), and to ensure that containment isolation requirements are met during a LOCA.
Safety Analysis:
E0P changes apply only after an accident has already occurred.
Operator actions are not initiating events for USAR evaluated equipment malfunctions or accidents. The E0P change ensures that penetration M-3 is operable and capable of preventing containment leakage once the charging pumps are stopped.
Therefore, the consequences of an accident (LOCA) are not increased.
As concluded, the containment integrity is maintained and penetration M-3 is operable.
With containment integrity maintained, the margin of safety as defined in the basis of Technical Specification 2.6 is not redated by the E0P chang).
01-CC-1 Component Cooling System Normal Operation
==
Description:==
This procedure was changed to eliminate the option of having cMy one heat exchanger in service.
This ensured that a minimum of 2 heat exchangers have component cooling water (CCW) flow at all times during normal operation.
Safety Analysis:
The change to the operating instructions eliminates the less restrictive alternative in the operating mode, thereby increasing the conservatism of operation.
The general operating philosophy was not changed.
The required temperaturc range for the system was not changed.
The consequences of an enuipment malfunction did not channe since the procedure clange did not affect or chang, tfie number of components required to function during an accident. The fail safe mode of the CCW inlet / outlet valves to the heat exhangers is open. This ensures that at least 2 heat exchangers (4 valves), (3 heat exchangers when river water temperature meets or exceeds 70*F) are in their fail safe position during normal operation.
01-RM-1 Radiation Monitoring - Normal and Accident Operation p
Descriotion:
The purpose of this procedure change is to correct instructions and diagrams for loading filter paper for RM-050/051 and RM-061/062.
Sifsty Analysis:
Correct installation of filter paper ir. the process 'nonitors will result in a reductinn of accident consequences by raising the sensitivity of the monitor. The probability of an accident is unaffected.
Correct installation of the filter paper will enhance the operation of equipment important to safety.
It will also have no effect on the probability of equipment malfunction.
Enhanced monitor sensitivity reduces the consequences of an accident by providing an earlier warning of the accident condition, and increases the margin of safety.
01-WDG-1 Waste Gas Disposal System Normal Operation
==
Description:==
lhe purpose of this procedure change was to upgrade for the Procedurd Upgrade Project, Project 1991, and to comply with the FCS Writers Guide and to enhance human performance.
Safety Analysin Although the Waste Gas Compressors (WGC)l mode the pressure and Wate Gas Decay Tank (WGDT) are o]erated in the manua control valve will still automatically shut at 100 psig.
Therefore, the probability of a WGDT rupture accident is nnt affected by operating the WGCs and WGDTs in manual. There is no effect on the consequences of a WGDT rupture accident, because operation of the WGCs and WGDTs in manual has no effect on the concentration of combustible or radioactive gases or maximu.a operating pressure in the WGDT.
The revision does not affect any safety related equipment previously evaluated in the USAR. The revision limits the compressor's cycling on and off which reduces the possibility of a compressor failure.
The Technical Specifications do not discuss required normal operation equipment lineup for the WGCs and WGDTs, and there are no margins of safety affected by operation of this equipment in manual.
SS-ST-CONT-0001 Surveillance of Containment Prestressing System Dercription:
The purpose of this procedure change was to incorporate changes to Technical Specification 3.5(7) per Amendment 139.
Safety Analysis:
According to USAR 5.3.3 the containment is designed with enough margin to compensate the loss of 5 to 10 wall tendons and 2 to 3 dome tendons.
Only one tendon was rendered inoperable at a time, therefore, containment integrity was maintained.
Containment prestressing tendons do not interface with any other system.
Since containment integrity was maintained, the consequences of a LOCA or Main Steamline Break were not increased.
The probability of malfunction of equipment important to safety was also not increased.
The testing of the prestressing tendons is considered in the USAR.
This change did not change the methods used to test the tendons, only the number of tendons to be lift off tested and the number to be detensioned for 1
f inspection.
Therefore, this activity did not create an additional possibili+y of an accident of a different type, ror the same reasons, this activity did not create a different type of malfunction of equipment ortant to l
safety than previously evaluated in the USAR. imfestir.g the tendons oer these changes increased the reliability of data obtained thus getting a better assessment of the remaining safety margin.
The limits as defined in USAR Fig. 5.10-3, as referenced in Technical Specification 3.5(7) were not changed.
TDB-V.9 Shutdown Margin Worksheet
==
Description:==
The purpose of this procedure change was to make fermat changes and to make the worksneets consistent with other 108 figures.
It al so increased the minimum required boron concentration from 1800 ppm to 1900 ppm and corrected mis-spellings.
~
Lafety Analysis:
Increasing the boron concentration to 1900 ppm does not change the probability of occurrence of an accident as no new modes of operation occur.
No new event initiators are created.
1900 ppm is conservative with respect to the current Technical Specification value of 1800 ppm. The use of the new boron concentration is conservative in the dilution event since it increases the time to critical
^
resu' tant. No changes in equipment operation are required, thus, no increase in the probability of occurrence of a malfunction of equipment important to safety was realized as no new event initiators were created.
In a boron dilution event, it is the operator of the equipment which is the initiating event, not the boron causing a piece of equipment (or system) to respond in a manner that has not been previously evaluated in the USAR.
The boron dilutien incident is analyzed in USAR Section 14.3 No other accidents are impacted by the refueling boron concentration.
No different types of accidents are created and no new or different modes of operation are created.
The margin of safety is maintained.
TDB-EOP 20 & E0P-20 Functional Recovery Procedure A0P-22 Reactor Coolant Leak
==
Description:==
The purpose of these procedure changes were to incorporate steps ensuring the charging header is kept pressurized by the HPSI pumpe during a LOCA.
Safety Analysis:
The procedure change applies to a procedure for dealirg with an accident previously evaluated in the USAR.
If the accident has already occurred, tae procedure change cannot increase the probability of an occurrence of that accident.
Fuel integrity and containment integrity are not impacted, so the consequences (f a LOCA are not increased. Affected 1
,a safety-related equipment (HPSI, Charging) will not be more likely to fail or malfunction.
This change will not increase the consequences of a malfunction of safety-related
]
equipment.
When the changed steps are performed, a has already occurred.
previously analyzed accident (LOCA)l not cause a different The alignment of HPSI to charging wil type of accident than any previously analyzed in the USAR.
The alignment will also not cause the ratings of any component to be exceeded, as the lineu3 is much like that already used in the E0Ps for Hot Leg In;ection. Therefore, the possibility of a different type of equipment malfunction is not created by this change.
HPSI and CVCS systems will still be capable of performing their respective safety functions without a reduction in the margin of safety for fuel clad temperature or containment pressure.
TDB-III.26.A Technical Data Procedure - DC Output Power vs. Ambient Temperature
==
Description:==
The purpose of this procedure change was to revise the operability temperature limits of the diesel generators following the installation of exciter cabinet air conditioners and enhanced maintenance on the jacket water radiator.
Safety Analysis:
There is no affect on diesel generator safety function. The figures quantify the engine output as a function of outside air temperature.
Engineering Ansi.ysis (EA) EA-90-062 R2
" Diesel Generator Upper Ambient Air Temperature Limit" demonstrates that the diesel generator output under the restrictions of the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating curve is greater than the demand due from the highest expected Post LOCA loads.
The predictions on jacket water system performance and turbo air inlet temperatures contained in EA-90-062 R2 were supported by test data taken 06/26/91 with an outside ambient air temperature of 95'F for Diesel Generator DG-1.
These parameters define the rating of the diesels.
-t.
,J Enaineerina'Chanae Notices (ECN)
ECN 90-43 RE-055 Descriotion:
)
The-purpose of this ECN was to evaluate Temporary Modification 1M-88-E-021 in order to implement it as a permanent change under the ECN process.
TM-88-E-021 disconnected cable 8622 at RM-055, 8622 is the signal cable for RM-055 which has already been taken out of service.
Although RM-055 has not been physically removed, RM-055A has been installed to replace the function.
Safety A.n_plysis:
The equipment involved is not CQE. Monitoring of releases is still. adequately performed by RM-055A and/or by grab sampler during releases.
Since alternate methods of monitoring exist, the probability of releases in excess of the limits is not increased. Also, the consequences do not changSAR. Radioactive effluent releases are accounted for in e.
the t.
No new type of accident has been identified that would result from this temporary modification.
The Technical - Speci fications allow for several ways of monitoring releases.
The margin of safety is not compromised.
ECN 90-088 Technical Support Conter Radiaticn Monitor, RM-093 Descriution:
The radiation monitor currently in use, Victoreen Model No.
- 808D, is being deleted from the radiation protection inventory. The Dosimeter Model No. DCA-3090 has been chosen to replace the existing Victoreen Model.
Spfety Analysis:
The new Dosimeter radiation monitor meets or exceed:: the design, material, and construction standards of the existing Victoreen monitor.
The Technical Support Area Radiation Monitor does not alter the radiological consequences of any of the accidents described in the USAR.
It only measures area radiation and alarms at unacceptable levels. The Area Radiation Monitor does not interact with or have an impact on any other equipn.ent, Replacing the existing Victureen radiation monitor with a comparable Dosimeter model will not increase the radiological consequences of a malfunction of
. any safety-related equipment. The Area Radiation Monitor is non-CQE.
There are no new possibilities of an accident created by changing the model of radiation monitor used in the Technical Support Center.
There is no safety-related equipment associated with radiation monitor, RM-093. There L
are no applicable sections of the Technical Specifications associated with this change.
4 ECN-90-178 Tie-In to Existing Chemical Injection Point
-Description:
The intent of this ECN was to install a new injection point for Direct Chemical feed Injectinn. Concentrated Hydrazine will be cumped into this injection poir.t at the Steam Packing Exhauster location.
The new point is required at this time since the condensate system is depressurized. The work prescribed was to install a reducing sockolet, pipe stub, and an isolation vah ' with a cap.
Installation of this configuration would allow for an on-line installation of a new chemical feed system. The materials to be used are in accordance with the design application of concentrated hydrazine feed.
SJfety Analysis:
The USAR does not explicitly stcte that a breach in condensate is applicable. However, tne USAR does state that if condenser shutdown is required, tnat the Auxiliary Feedwater System be available tu remove decay heat.
- Thus, a breach in the Condensate System has been addressed as not being a safety issue.
Failure of a weld or valve is bound within Condensate System reliability.
The means of the availability of the Auxiliary feedwater system was not impacted by the change.
The tie-in is adjacent to the existing feed point.
No existing safety systems are affected by adding the tie-in point. The tie-in point has no interface with equipment required for safety as stated in USAR Section 10.2. Therefore, the t.hange does not increase probability of component failure.
No equipment, which is safety
- related, is located near the tie-in point.
Therefore, the consequer.ces of a failure of the valve or tie-in piping have not been altered.
Breach of the condensate piping is not a new accident type not previously evaluated in the UEAR. The Technical Specifications do not take credit for the Condensate Systcm in a margin of safety system.
ECN 90-189 Removal of Tubing from CW System
==
Description:==
The purpose of this ECN was to remove the unused copper tubing from the downstream sida of valve CW-120. This valve is located in the basement level of the intake structure on the inlet side of the screen wash strainer.
Safety Analysis:
The tubing is located in a non-CQE system in a non-CQE area.
The CW system is not required for any accident recovery.
The tubing to be removed is unused, has no purpo3e, will not impact any possible malfunction, and is not the basis for any Technical Specification.
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-ECN-90-266 Change to GE Drawing 14505719 and 791E611-
==
Description:==
-The bore on the upper guide bearing on the reactor coolant pump motors was opened by 5 mils via machining the soft babbit face on the bearing.
The bearing's surface finish and length remained at the manufacturer's original dimensions.
The manufacturer has been consulted on the tolerance change and agreed.
Safety Analysis-Historical performance data for this motor and the three other reactor coolant pump motors indicates that repeated starts and operation with bearings worn to tolerances exceeding those used in this case had no identifiable deleterious effects on the reactor-coolant pump. motor reliability or operation. Rotor seisure is less likely with increased tolerance. Consequences of an accident would be the same as those given in USAR Section 14.6-2, " Seized Rotor Event."
GE gave assurance that this machining is allowable.
Historical performance of worn bearings with greater tolerances have shown acceptable bearing vibration and temperature parameters, even with repeated cycling.
Motor reliability, function, and performance as a driver does not change therefore its safety function does not change. Flywheel inertia does not change and coastdown may increase due to less friction from guide bearing which would increase the margin of safety.
.a Special Procedures MW0 910625 Replacement of the High Vibration Trip Time Delay Relay K3001 While the Plant is On-Line
==
Description:==
The purpose of this change was to provide instructions for replacing the high vibration trip time delay relay K3001 in panel Al-55 while the plant is on-line.
1 Safety Analysis-1 During. the implementation of the procedure, the affected turbine trip circuits were disabled to prevent a sudden loss of load due to a turbine trip.
The trip was disabled 4 ways; by placing the Turbine Supervisory Instrument (TSI) power switch to,off', by placing the vibration trip switch to ' disable', lifting the trip circuit leads, and by removing the TSI power supply fuses. The potential severity of loss of load was not increased by the procedure change because the turbine power level was not affected by the procedure. None of the equipment affected by the procedure change was important to nuclear safety.
The TSI system including hi h vibration trips, is not safety related. All other turbi generator trips functioned normally before, duringleand a ter the implementation of this procedure. No credib new accident scenario could result from the procedure change. The ability of the turbine to trip as a result of a reactor scram or operator action would not be impaired.
The main turbine / generator and TSI systems are not important to nuclear safety. No Technical Specification margins of safety are based on or involve the TSI system.
SP-BURNUP-1 Burnup Determination for Storage of Spent Fuel p_escriotion:
SP-BURNUP-1 was performed prior to refueling operations.
This procedure verifies the acceptability of fuel storage in region 2 of the spent fuel pool in accordance with requirements in the Technical Specifications.
Safety Analysis:
This procedure uses a CECOR printout and compares the values with the limitations of Technical Specification Figure 2-10 and does not constitute an unreviewed safety question.
SP-CP-08-DEVAR-1A3, Calibration of the DEVAR Relay and Associated Timers lA4, TIAl, TIA2, TIA3, TIA4 pescriptior.:
l The purpose of this procedure change was to change the j
subject relay trip setpoints to reflect Design Engineering l
Calculation EA-FC-91-017.
Safety Analysis:
The resetting of the relay has the conservative effect of l'
providing the safeguards loads with reliable power earlier in any accident. This ensures that all 480V loads continuc to have sufficient voltage to prevent damage to the motor.
l
.e l
The safeo.Jards loads will continue to perform all accident functions as described in the USAR analysis for each of those accidents. With all equipment functioning according to design, no increase in ceasequences will occur.
This procedure change speeded-up the response time of the electrical system to a degraded system voltage. The Offsite Power Lc,w Signal (0PLS) setpoints are detailed in the Technical Specifications.
The setpoints installed in the relay moved conservatively from the described setpoints.
This ensures proper voltage is maintained at the safeguards motors which is the basis for the OPLS setpoints. Thus, the margin of safety is increased.
SP-DEN 'P-0001 Fire Door Inspection SP-DEN-FP-0002 Fire Damper Inspection SP-DEN-FP-0003 Fire Barrier Penetration Seals Inspection
==
Description:==
These procedures are new procedures to document fire barrier wal kdown.
Safety Analysis:
These completed procedures do not constitute an Unreviewed Sa fety Question because they only provided for the inspection of fire barriers and no fire barriers were physically degraded due to these inspections.
SP-FAUD-1 fuel Assembly Uplift Condition Detection Descriction:
The purpose of this procedure is co prevent fuel assembly wear and fretting by detection of a fuel assembly uplift condition.
Safety Analysin The Monthly RCS Flowrate Test is a Technical Specification requirement,
- however, the Fuel Upli f t Test is act.
Performance of this test is simply a manual calculation using data obtained from the RCS Flowrate Determination. It does not constitute an unreviewed safety question.
SP-ISI-SURFPREP-1 Surface Preparation without any Detectable Change in Material Thickness for Class 1, 2, and 3 Components and Welds
==
Description:==
The RCS pressurizer was inspected at certain weld locations.
To perform that inspection, the insulation was remosed and the surfaces of the welds were cleaned.
SP-ISI-SURFPREP-1 was used to control this process.
Safety Analysis:
No unreviewed safety question was involved in this process; it was only an activity to clean up a weld area.
The procedure followed the guidance of ASME Cocle for surface preparation.
l
i SP-RP-1 Cold Shutdown Initial Radiological Survey Procedure
==
Description:==
The purpose of this procedure is to provide a general procedure for performing the initial radiological surveys of containment and the balance of the plant after cold shutdown.
Safety Analysis:
The procedure is guidance for completion of the initial radiological survey of the plant after reactor shutdown, special Procedure SP-RP-1 does not constitute an unreviewed safety question.
SP-SITFILL-1 Injection of Boric Acid into Safety Injection Tanks SI-6A, SI-68, SI-6C and SI-6D
==
Description:==
The purpose of this procedure is to provide written instructions for the in,)ection of concentrated boric acid into the safety injection tanks in order to increase the concentration of boric acid for nitrogen sparging to ensure proper boric acid mixing.
Safety Analysis:
The perf ormance of this procedure did not represent an unreviewed safety question because no safety-related equipment was removed from service and the tanks were maintained operable in accordance with Technical Specification 2.3(1)c.
SP-VA-TPACER Tracer Gas Characterization Study of Spent Fuel Pool Area Ventilation - VA-66 Carbon Filter
==
Description:==
The purpose of this procedure change was to incorporate electrical load considerations for the 480 volt systcms necessary to complete tracer gas testing of VA-66.
Safety Analysin Operation of the additional 480 volt loads does not increase the probability of occurrence of any accident.
Since the motors can continue to operate and perform their design function there is no increase in the consequences
' any accident.
No tew accidents have been introduced.
No additional component failures have been created.
All equipment and systems will function as designed.
4..
O BLmporary Modifications TM 91-012 Remeval from Service of the Welding Receptacle in Room 81 ILeSJ;riution:
TWir temporary modification isolated the welding receptacle in Room 81 from the remaining welding receptacles powered from 480 Volt Breaker MCC-4Al-E06. The receptacle in Room 81 was isolated by disconnecting the cable (384H) at JB 448A which goes to the welding receptacle.
Safety Analysis:
Removing a potential load from M"C-4A1 would reduce the probability of an accident by reducing the probability that a non-EEQ load (Llding Receptacle in Room 81) would result in failbre of CQE equipment.
Removing the welding receptacle could reduce the consequences of accidents by ensuring the breaker to MCC-4A1 does not trip due to fault on the welding receptacle. No technical specifications are associated with the welding receptacle. No safety margin of the Technical Specifications is affected.
TM 91-024 Jumpering of Cells 15 and 16 in Battery No.1
==
Description:==
Cell 15 in Battery No.1 (EE-8A) had developed a crack which initially ailowed a significant amount of electrolyte ta leak from the cell. The cell was temporarily repaired, but needed to be replaced. Had the temporary repair failed or the replacement process required that the battery been removed from service fer greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, it would have been desirable to jumper out cells 15 and 16 and therefore minimize the amount of time the battery was out of service.
Both of the cells required jumpering because cell 15 and 16 are contained within a single jar which cannot be separated.
In performing the evaluation for jumpering out cells 15 and 16, it had been deter.nined that:
1.
Battery I had sufficient capacity to allow the removal of these two cells.
This hhd been evaluated in memo PED-SYE-90-984J using the results of previous discharge tests.
2.
The overall t,attery terminal FLOAT voltage remained at 130VDC and the EQUALIZE voltage at 135VDC. This was the recommendation of DEN-Electrical. The individual cell voltages would rise but not beyond manufacturer's established voltage limits.
3.
The jumpers had sufficient capacity to ccnduct and carry the maximum load.
Safety Analysis:
l The station batteries are designed to respond to postulated accidents.
The battery was able to perform its required function in the altered configuration.
All previous assumptions based upon battery capacit) were still valid.
The changed battery configuration did not provide another
y.
5 y
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m means of ' creating = any Section 14 accidents.
- Accident L
mitigation was not affected.
~ No new failure modes were introduced.
Battery-current conductors installed were aopropriately sized and utilized the-same method of mechanical-connection - or. better than the existing configuration.
- Even though = the battery's ca)acity was-reduced - it was-not below design-basis levels. "he battery was stiIl able to perform its required safety functions.
The only credible failure associated with this temporary modification was the loss of either the battery or he-DC bus.
The loss of a DC bus, due to a battery failure ar a different equipment mal function, had been previously 4
addressed.
The Technical Specification basis states that the system is designed such that no single failure-could cause-enough engineered safeguards to become inoperable to prevent safe shutdown of - the plant.
This temporary modification does not introduce any failure mode which
-,;.cce# the single failure criteria of the original-design.
TM 91-037 Temporary! Battery to Support Battery Replacement Descriotion:
Battery No. 1--(EE-8A) was reloacted to the Turbine Deck to serve as a backup DC nower source for DC Bus 1 EE-8F and DC-Bus 2 -(FE-GG)- curing replacement of Battery N(o.1 a)nd 2.
The battery was connected to Disconnect Switches EE-8M and EE-8N which were to be open.
Safety Analysis:-
installed only while RCS The-temporary battery was
.temperaturc was-less than 300*F.
A permanent battery was always in service as a first-line of defense. The temporary battery was only to be used in the event of an emergency and would not have performed safety functions.
The tem)orary battery was normally isolated from the 125VDC distri)ution system.
No new malfunctions other than those previously/ analyzed were introduced since-the - temporary
. battery was a station battery. Since this is a backup power source for the-125VDC system, its use was a-last resort in the event of' failure of the preferred safety related source.
No new accidents were -created by adding the temporary battery.
The temporary battery was isolated by two disconnect switches from the DC distribution system.
1
-TM 91-048 Leads to ~63x-1/PIC-103 AC/DC-2 Aux. Relay Back-Up lieaters Control Cir.cuit, CB-1/2/3, 736
==
Description:==
TM 91-048 lifted leads to burned out HFA relay coil so that the other non-safety related instrumentation powered from the same circuit could be restored for the 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> that it took to replace the burned out relay.
The relay provided automatic operation of pressurizer heater banks 3 and 4.
Manual operation is unaffected.
Safety Analysin The loss of thi automatic backup capability of the f --
.n
+
Os
-em
.o i
Pressurizer Heater Bank' 3 and 4 makes no difference in the probability of any accident previously evaluated.
The subject HfA relay is non-safety related and it is wired out of the circuit by lifting the eods.
This circuit does not affect the margin of safety reqaired or defined in the basis of the Technical Specifications.
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