ML17223A203
| ML17223A203 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 12/31/1988 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17223A202 | List: |
| References | |
| NUDOCS 8906200245 | |
| Download: ML17223A203 (66) | |
Text
7 m r r h n Via um er/Lift d L ad -1 FR 0.59 Eval nit 1 Request 0 8-035
, component / System Affected: PI-1181, 1B2 RCP MIDDLESEAL CAVITYPRES PIA-1182, 1B2 RCP UPPER SEAL CAVITYPRES Description of Change:
Installed a 100 ohm resistor in the pressure instrument current loop to develop the required voltage for a strip chart record~
hookup.- .
Safety Evaluation:
This temporary change is required to monitor RCP seal cavity pressure oscillation to determine if there is any correlation between the seal pressure oscillation and RCP pump vibration. PIA-1182 83 PIA-1183 indications are not affected by the installation of the 100 ohm resistor.
i) The proposed activity does not increase the consequences of malfunction of equipment important to safety previously evaluated in the FSAR.
ii) The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.
iii) The proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.
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m r r hn Vi m r/Lif dL - I FR .59 Eval Component / System Affected: LS 4420, FUEL POOL LEVEL ANNUNCIATOR TE 4420, FUEL POOL TEMP ANNUNCIATOR Description of Change:
Modify instrument mountings for spent fuel machine clearance Safety Evaluation:
Due to the loss of annunciator alarm capability, fuel pool level and heat exchanger outlet temp. willbe monitored every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. By following the alternate means of monitoring fuel pool level and temp., it can be demonstrated that the removal of LS 4420 85 TE 4420 does not have an adverse effect.
i) The proposed activity does not increase the consequences of malfunction of equipment important to safety previously evaluated in the FSAR.
ii) The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.
iii) The proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.
m r r h n Vi m r/Lif L -1 FR Eval it 1 Request 0 8-041 Component / System Affected: BQRQNQME'IZR Description of Change:
To install temporary high voltage power supply in order to restore Boronometer indication. The installed power supply unit failed while in service.
Safety Evaluation:
i) The proposed activity does not increase the consequences of malfunction of equipment important to safety previously evaluated in the FSAR.
ii) The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.
iii) The proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.
m rr hn Vi m Lif L Ev I it 1 Request ¹ 8-058 Component / System Affected: V6565 AND RE 26<2 Description of Change:
Allow operation of waste gas system to plant stack with Waste Gas Monitor (ch. 42) out of service.
Safety Evaluation:
This temporary jumper is necessary to allow waste gas to be aligned to the plant stack with RE-26-42, Waste Gas Radiation Monitor (ch. 42) out of service. RE-2642 initiates a closure signal to the waste gas header isolation valve, V-6565, on high radiation. Since RE-2642 is out of service, V-6565 failed closed and needs to'be opened via a jumper to align waste gas to the plant stack. The plant stack will be used to monitor the waste gas releases.
Additional information: Waste Gas Decay Tanks was also out of service i) The proposed activity does not increase the consequences of malfunction of equipment important to safety previously evaluated in the FSAR.
ii) The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.
iii) The proposed activity does not reduce the margin of safety as de6ned in the basis for any technical specification.
m rr hn Vi m r Lif L FR 9 Eval it 1 Request ¹ 8-061 Component / System Affected: PRESSURIZER PRESSURE (LOW RANGE) CH. 1103 Description of Change:
Plant Change and Modification (PCM) 033-188 relocate PR 1103/1 104 Safety Evaluation:
Removing PR 1103/1 104 from service does not involve an unreviewed safety question in that:
i) Jumpering out the pressure recorder does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. While the existence of the recorder is mentioned in section 7.6.2.1 of the FSAR as providing clear indication of system status, other redundant indication is also available on RTGB 104. Annunciation capability of the OMS (Overpressure Mitigation System) is not affected.
Therefore, while credit for the existance of PR 1103/1 104 is taken in the FSAR, other inputs are available which provide the same information.
ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR is not created because the instrumentation loops for PI'103 and PT 1104 will not be disabled. A jumper willbe installed between points Xl(+) and Xl(-) and another between X2(+) and X2(-) on terminal board 0 to ensure the circuit remains operationaL The signals from the pressure transmitters will still be operational and able to provide input to SDC interlock loop and the OMS loop.
iii) The margin of safety as deGned in the basis for any technical specification is not reduced as PR 1103/1 104 is not requited to be operable to meet any technical specification requirements.
m r r 1 n Vi m r Lif L Ev I it 1 Request 0 8-064 Component / System Affected: ANNUNCIATORK, WINDOWS K15,16,17,1831,2233,24,30 Description of Change: I Construction to make modifications to Annunciator K Windows K2132,23,24 per Plant Change or Modification (PCM) 005-188 Safety Evaluation:
PCM 005-188 replaces the existing CEA metrascope system with an upgraded CEA Position Display System which monitors and displays CEA position.
The removal of Annunciator K arming screw willaffect the annunciators K15,16,17,18,2l~g3~, and 30 as shown on the Controlled Wiring Drawings. These annunciators provide alarms for CEA positioning requirements during power operation. These alarms are not required during Modes 5 4 6. The removal of alarm capability has no adverse effects.
This work is related to PCM 005-188 and the Engineering Package provided the safety evaluation.
h n Vi m r Lif L Ev I it 1 Request 0 8-069 Component / System Affected: FEED TO "A" ANNUNCIATORFOR SS XFMER 1B2 CABLE "B"-SB Description of Change:
Replace Normal / Isolate switch in 1B2 bkr. cubicle as per Nonconformance Report 142-185-1974E Safety Evaluation:
During the performance of Plant Change and Modification 142-185, the Normal / Isolate switch located on the 1B2 bkr.
cubicle was found not able to operate properly in either Normal or Isolate position. This breaker is the 480 V Station Service Transformer 1B2 4160 V Feeder Breaker located on the 4160 SWGR Bus 1B3 Cubicle 2.
Engineering has evaluated the replacement of the switch in its disposition of the Non-Conformance Report and Field Change Notice. Also, Plant Change and Modification 142-185 has been previously approved by the FRG with its safety evaluation.
m r r h n Vi m r Lif L - I FR Ev I Component / System Affected: ANNUNCIATORL-2 "RX TRIP CEA BUSSES DEENERGIZED" Description of Change:
Plant Change and Modification (PCM) 007-188 construction requirements to relocate relay 27/RTS Safety Evaluation:
PCM 007-188 was to install new Bentley Nevada Vibration Monitoring equipment for the Reactor Coolant Pumps behind RTGB 104. This PCM calls for the relocation of the 27/RTS relay that is located behind RTGB 104 to a new location, This jumper is to facilitate removal of relay 27/RTS which has been identified by PCM 007-188 as one of the relays requiring relocation. The removal and relocation has been evaluated by Engineering via the PCM design verification and safety evaluation. The PCM 007-188 has been FRG approved for implementation.
Tm rr hn Vi m r Lif L -1 FR5 Ev I it 1 Request 0 8-073 ComPonent / System Affected: SPENT FUEL HDLG. MACHINEPOOL END, JIB BOOM (JUNCTION BOX)
Description of Change:
Temporary feed to Spent Fuel Handling Machine Pool End junction box Safety Evaluation:
i) The temporary feeds will be utilized to power non-safety related equipment, and willnot affect the function of equipment or systems important to safety as previously defined in the FSAR.
ii) The proposed temporary feeds willbe utilized to power non-safety related equipment. The operability of equipment or systems important to safety willnot be affected.
iii) The operability requirements of Technical Specification 3.8.2.2 can still be met with the temporary feeds installed.
Since the existing Tech. Spec. requirements are not affected by the temporary feeds, no changes to the Technical Specifications are requiretL
m r r h n Vi m r Lif L Ev I it 1 Request ¹ 8-074 Component / System Affected: REACTOR BUILDINGCRANE Description of Change:
Temporary electrical feed to Reactor Building Crane, cable 0 10993C Safety Evaluation:
i) The temporary feeds will be utilized to power non-safety-related equipment, and will not affect the function of equipment or systems important to safety as previously defined in the FSAR.
ii) The proposed temporary feeds willbe utilized to power non-safety-related equipment. The operability of equipment or systems important to safety willnot be affected.
iii) The margin of safety as defined in the basis for any technical specification has not been reduced.
m r r b n Vi m r Lif L - I FR Eval it 1 Request 0 8-079 Component / System Affected: FCV-25-1 p j,4$ A6; HVE-8A; HVE-8B Description of Change:
Temporary jumper to open the containment purge valves without running the Containment Purge Fans. This jumper is to support M/M in repair of the purge valves.
Safety Evaluation:
i) The proposed activity does not increase the consequences of malfunction of equipment im portant to safety previously evaluated in the FSAR. The FSAR evaluates the purge assuming containment integrity. This jumper will be installed only while the maintenance hatch is open.
ii) The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.
iii) The proposed activity does not reduce the margin of safety as defined in the basis for. any technical specification.
r r h n Vi m r/Lif L -1 FR Ev t it 1 Request 0 8-099 Component / System Affected: BORONOMETER Description of Change:
To install temporary high voltage power supply to restore Boronometer indication. The installed power supply unit failed while in service.
Safety Evaluation:
i) The proposed activity does not increase the consequences of malfunction of equipment important to safety previously evaluated in the FSAR.
ii) The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different'ype than previously evaluated in the FSAR.
iii) The proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.
m r r h n Vi m r Lif L d FR 9 Eval it 1 Request ¹ 8-106 Component / System Affected: B FEED REGULATING CONTROL RACK Description of Change:
Suspect lead lag unit is causin'g 4 - 5% oscillation of feed flow signal Safety Evaluation:
i) The proposed activity does not increase the consequences of malfunction of equipment important to safety previously evaluated in the FSAR. Lead/ lag unit does not impact the ability to provide feedwater.
ii) The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR. Operability of feedwater regulating system is not altered; operation of safety-related system is not compromised or altered.
iii) The proposed activity does not reduce the margin of safety as defined in the basis for any technical speciflcation.
Feedwater regulating system is not required by Unit I Tech. Spec.
Ym r r h n Vi m r/Lift L d FR 0 Ev I it 1 Request ¹ 8-111 omponent / System Affected: PNUEMATICTUBINGONMAINFEEDWATERREG. VALVEACTUATORS (FCV-9011 4 FCV-9021)
Description of Change:
Installation of test ports on the instrument air supply line to facilitate testing of valve movement.
Safety Evaluation:
This temporary change was part of the ongoing troubleshooting activities to determine the cause of the erratic oscillations of steam generator levels.
i) The proposed activity does not increase the consequences of malfunction of equipment important to safety previously evaluated in the FSAR.
ii) The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.
iii) The proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.
m r r h n Vi um r/Lifted L ad -1 FR 50.59 Eval it 1 Request 0 8-122 Component / System Affected: CHEMNUCLEARSYS, INC.RESINTRANSFERFILLHEAD(NOTPLANT INSTALLEDEQUIP.)
Description of Change:
To bypass malfunctioning High-High Level alarm interlock in the disposal container.
Safety Evaluation:
This jumper bypasses an interlock which isohtes the resin transfer fillhead on the High-High level alarm. This interlock is still capable of isolating the resin transfer fillhead via the High level alarm which was demonstated to function properly. Furthermore, all resin transfers to the disposal container are monitored with video cameras to ensure proper transfer of resin and to monitor the container levels. This jumper does not interphase with any safety-related plant equipment; therefore i) The proposed activity does not increase the consequences of malfunction of equipment important to safety previously evaluated in the FSAR.
ii) The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.
iii) The proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.
Tm r r h n Vi um r Lif L d -1 FR Eval it 2 Request ¹ 8-027 Component I System Affected: PDIS.2216 Description of Change:
Jumper out PDIS-2216 because the switch failed while in service.
Safety Evaluation:
The purpose for PDIS-2216 is to isolate letdown upstream of the Regenerative Heat Exchanger in the event of a pipe rupture downstream of this heat exchanger. The pipe rupture would create a high differential pressure which would be sensed by PDIS-2216 and isolate letdown flow.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety evaluated in the safety evaluation report has not been increased since the pressure switch is not a consideration in the determination of the probability of accidents addressed by the FSAR. Further, the probability of an accident of a different type than previously evaluated in the safety evaluation report is not created because PDIS-2216 is not assumed to isolate letdown in the accident scenario outlined in Section IIIabove. Since PDIS-2216 is not reqtured to be in-service by the Technical Specifications, the margin of safety as defined in any basis for any Technical Specification has not been reduced. Technical Specifications are not affected by removing the subject pressure switch horn service.
m r r h n Vi m r/Lif L FR Eval it 2 Request ¹ 8-036 Component / System Affected: TR-22 Description of Change:
To re-route conduit to TE-22-8B, Impulse Chamber Steam (pt.N2 on TR-22-6)
Safety Evaluation:
This component has no control function 8t non-safety-related and is used for starting & loading of the turbine. The turbine is currently at base load.
i) The proposed activity does not increase the consequences of malfunction of equipment important to safety previously evaluated in the FSAR.
ii) The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.
iii) The proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.
I r r h n Vi I r Lif L Ev I it 2 Request ¹ 8-039 Component / System Affected: RCP 2Ali 1) UPPER OIL RESEVOIR, LIA-1156 2) LOWER OIL RESEVOIR, LIA-1157 3) CONTROLLED BLEEDOFF FLOW, FIA-1150 Description of Change:
To repair FIA-1150, 2A1 RCP Controlled Bleedoff flow indication Safety Evaluation:
This temporary jumper removes the power to FIA-1150 which also deenergizes LIA-1156 A LIA-1157 to facillitate the replacement of the Sigma.
i) The proposed change does not increase the probability of occurrence of an accident previously evaluated in the FUSAR because there are other available indications which provide information about seal condition and possible degradation.
Therefore, the ability of the RCOs to respond to a failure of the 2A1 RCP seal is not compromised. These indications are annunciation of high pressure in the upper seal cavity, and temp and pres. indication on the controlled bleedoff line.
ii) A possibility for an accident or malfunction of a different type than evaluated previously in the FUSAR is not created. Should all instiumentations on the 2A1 RCP fail, and the RCP seize, the Rx would trip due to low RCS flow trip on the RPS. The loss of coolant flow through the destmyed seal is enveloped by the small break LOCA accident analysis in Chapter 15 of the FUSAR.
iii) The margin of safety as defined in the basis for any technical specification is not reduced. Unit 2 T.S. 3.4.6.2, action statement (a) requires with RCS leakage, which is confirmed in a flow path with no flow indication, to be determined by the performance of an RCS inventory balance which is to be started within the hour. Should seal failure in the 2A1 RCP occur while the flow indication in the control bleedoff flow was out of service, the high pressure annunciator in the upper seal cavity would alert operators to the condition, which could be confirmed by the temp. and pres, indicator in the bleedoff line. Action statement 3.4.6.2.a would then be enteretL Therefore, the margin of safety defined by this specification is maintained.
m r r hn Vi um r/Lif dL d -1 FR 0 Evl i 2 Request 0 8-041 Component / System Affected: 2A2 RCP: FS-1166, FS-1167, PS-1160 Description of Change:
Remove DC power to components in containment due to electrical ground Safety Evaluation:
REVERSE ROTATION INDICATIONSWITCH:
Removing the annunciator capabilities from FS-1166 and FS-1167 would not constitute an unreviewed safety question based upon the following:
i) The probability of occurrence or the consequences of an accident or malfunction previously evaluated in the FSAR is not increased. Both flow switches monitor oil flow in the lube oil system at the main thrust bearing bracket. In order for a reverse rotation condition to exist, the RCP would have to be stopped and restarted, and the Anti-reverse Rotation device would have to malfunction in its entirety. If the RCP were to be stopped while the unit was at power, the unit would trip due to low RCS flow (RPS trip unit). Should the pump be restarted, and the motor somehow reverse rotation, one pin in the anti-reverse rotation device is capable of holding the pump stationary against the torque produced by the application of 100% voltage in such a reverse phase rotation. Therefore, even though the control room operators were unaware of a reverse rotation condition existing, the anti-reverse rotation device would prevent the pump from turning in the wrong direction.
ii) The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR has not been created. When the RCP shaft stops rotating when the motor is stopped, the pins in the anti-reverse rotation device prevent the RCP from rotating in the reverse direction, even against 100% voltage applied to the motor in the reverse phase. With annunciator capabilities lost to the operators, it is conceivable that the pump may be restarted; while damage to the motor may result, and the anti-reverse rotation device would prevent backflow through the RCP and the 2A steam generator, thus corresponding to a cooldown using less than 4 RCPs for RCS circulation. As the unit is designed to accomodate a natural circulation cooldown, no new accidents or malfunctions are assumed to be created.
iii) The margin of safety asdefinedin thebasis for the TechnicalSpeciflicationsisnotreduced. Reverse rotation indication is not required by Technical Specifications.
GASKET LEAKAGEINDICATION/ANNUNCIATION Removing the annunciation capabilities on PS-1160 does not constitute an unreviewed safety question based upon the following:
i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. PS-1160 is discussed in section 5.2.5, "Detection of Leakage Through RCS Pressure Boundary," of the FSAR. The primary indications of RCS leakage are given as the containment sump level and containment radioactivity alarms. As the loss of annunciator in the 2A2 RCP gasket pressure switch does not affect any of these primary indications, no existing analysis are affected.
ii) A possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. With no indication of gasket leakage in the control room, the operators would still be made aware of any RCS pressure boundary leakage detectable by the RCS Pressure Boundary leakage detection system; this system is capable of detecting unidentified leakage of 1.0 gpm or less within one hour. This redundant capability ensures that RCS inventory is maintained within analyzed limits.
iii) The margin of safety as defined in the basis for any technical speciTication is not reduced. Unit 2 T.S. 3.4.6.1 and 3.4.6.2 state the operability requirements for the RCS Leakage detection systems and RCS leakage. The inoperability of the RCP gasket pressure switch is not included in these specifications and is bounded by the capabilities of the leakage detection system. Therefore, the margin of safety defined by the basis for the technical specifications is not affected.
m r r h n Vi m r/Lif L Ev I it 2 Request ¹ 8-052 Component / System Affected: TURBINE RUNBACK CIRCUIT Description of Change:
PIS-22-36 failed at pressure for 100% power (runback would not terminate), the unit was at 859o when the failed pressure switch was discovered.
Safety Evaluation:
i) The proposed activity does not increase the consequences of malfunction of equipment important to safety previously evaluated in the FSAR. Turbine runback feature is not important to safety per FSAR section 7.7.1.1.10.3.
ii) The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR. Equipment associated with the turbine runback is not important to safety.
iii) The proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.
Although turbine runback produces an energy imbalance between the primary and secondary, FSAR section 15.2.1.2 analyzes for the worst case condition of energy imbalance which is isolation of turbine at 102% power.
ST. LUCIE UNIT 2 RCS INSTRUMENT NOZZLE CRACKING PURPOSE Safety evaluation written to conclude that there are no significant nuclear safety issues surrounding the concerns of potential intergrannular stress corrosion cracking ( IGSCC) in the seven instrument nozzles which have been determined to be susceptible to IGSCC at St. Lucie Unit 2.
SAFETY EVALUATION The issues surrounding the RCS instrument nozzle IGSCC susceptibility do not involve an unreviewed safety question for the following reasons:
The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not increased. The RCS instrument nozzles have been shown not to create a small break LOCA if they should develop through-wall cracks. Normal shutdown can be accomplished since the maximum leakage would be within the normal charging capacity. The pressurizer lower head level nozzle has been shown to be the most limiting case. The RTO hot leg nozzles and the reactor flange leak detection nozzle are not a concern for Cycle 4 operation. The pressurizer can operate for up to 604 days with a nozzle leak of app'roximately
- 0. 1 gpm before unacceptable corrosion results. Modifications implemented by PC/M 137-287 to insulation near the subject nozzles ensures leakage would be directed to the coated concrete floor. Therefore, with respect to the consequences of through-wall cracks, the possibility of corrosion of the pressurizer and other materials in the vicinity is not a concern.
The possibility of an accident or malfunction of a different type than evaluated previously in the Safety Analysis Report is not created since no new equipment or operating procedures are added as a result of this evaluation. In addition, the preceding engineering evaluation has shown that no new types of accidents or malfunctions are created.
The margin of safety as defined in the basis for any technical specification is not reduced. Technical Specification 3/4.4.6 will require shutdown should pressure boundary leakage occur.
ST. LUCIE UNIT 2 RCS INSTRUNENT NOZZLE CRACKING Page Two CONCLUSIONS Based on the foregoing analysis, there are no significant nuclear safety issues surrounding the concern of potential IGSCC in the seven RCS instrument nozzles described herein. The analysis demonstrates that in the limiting case, the pressurizer lower head level nozzle, the structural safety margin of the pressurizer is not compromised for assumed leakage of one operating cycle duration or less.
EXTENDED BURNUP ANALYSIS FOR ST. LUCIE UNIT I PURPOSE Safety evaluation to address St. Lucie Unit 1 operations with higher burnup limits. This evaluation supports the oper ation of the unit with a core average burnup to 40,000 MWD/MTU and a maximum assembly burnup up to 52,000 MWD/MTU.
SAFETY EVALUATION Since all the analyses affected by the extended burnup operation have been reviewed and demonstrated to meet the currently approved safety criteria, it can be stated that:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased.
The increased assembly burnup limits do not change the overall configuration of the Plant. The increase in burnup does not require a physical change to the fuel or fuel handling equipment at the Plant. Since the mode of operation remains unchanged, the probability of occurrence of an accident or malfunction of equipment important to safety is not impacted.
The . safety analyses performed in support of increasing the assembly burnup beyond 47,000 MWD/MTU demonstrates that the consequences of an accident or malfunction have not been increased beyond those evaluated previously. This is demonstrated in the radiological release calculations where conservative release calculations result in site boundary doses which are a small fraction of the 10CFR100 limits.
The possibility of an accident or malfunction of a different type than any previously analyzed in the safety analysis is not created.
As discussed earlier, the increase in burnup does not require a change in the overall configuration of the Plant. The mode of operation remains unchanged since the increase in burnup does not require changes in fuel design.
EXTENDED BURNUP ANALYSIS FOR ST. LUCIE UNIT I PAGE TWO 111 ) The margin of safety as defined in the basis for every technical specification is not reduced.
The re-analyses of the radiological assessment and rod bow effects to support extension of the maximum end of life (EOL) peak assembly average exposures, have shown that the results are well within the design basis. The effects of rod bow, which affect the MDNBR criteria, have been shown to be bounded by the existing 1.22 safety limit for assembly exposures up to 52,500 MWD/MTU. The analysis shows a small rod bow penalty (<0.5X) above the 1.22 safety limit for assemblies with burnups in the range of 50,900
- 52,500 MWD/MTU. However, due to the depressed power levels associated with these high burnup assemblies, the calculated DNBR (for these extended burnup assemblies) is bounded by the DNBR associated with higher powered fresh fuel assemblies. Based on these findings, it is determined that the increase in assembly exposures does "not result in a reduction in the margin of safety relative to the Technical Specification basis for St. Lucie Unit 1.
CONCLUSION As per Federal Regulation 10CFR50.59(b), the above Safety Evaluation provides the basis to conclude that operation of the St. Lucie Unit I with a core average burnup of up to 40,000 MWD/MTU and/or with peak assembly burnup up to 52,500 MWD/MTU does not involve any changes which introduce an unreviewed safety question. Therefore, implementation of this change is permissible without prior NRC approval.
I REVIEM OF CHAPTER 15 EVENTS FOR ST. LUCIE UNIT 1 MITH 15$ STEAN GENERATOR TUBE PLUGGING - 10CFR50.59 DETERMINATION PURPOSE A review of the FSAR Chapter 16 non-LOCA events for St. Lucie Unit 1 was performed by Advanced Nuclear Fuels (ANF) to support operation with up to 15% steam generator tube plugging level. In addition, a LOCA-ECCS accident re-analysis for up to 15K average steam generator tube plugging was completed and has been approved by the NRC.
SAFETY EVALUATION Since all the events have been reviewed and proved acceptable, it can be stated that:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, previously evaluated in the Safety Analysis Report, is not increased.
The increase in tube plugging level does not change the overall configuration of the Plant.. The mode of operation of the Plarit remains unchanged. Therefore, the probability of occurrence of an accident or malfunction, previously evaluated in the Safety Analysis Report, is not increased.
The ANF report demonstrates that the consequences of an accident or malfunction have not been increased beyond these evaluated in the previous analyses since all transients meet current criteria.
A possibility of an accident or malfunction of a different type than any previously analyzed in the safety analysis is not created.
There are no hardware or procedure changes for the increase in tube plugging level. The mode of operation of the Plant remains unchanged. Therefore, a possibility of a new accident or equipment malfunction has not been created.
REVIEW OF CHAPTER 15 EVENTS FOR ST. LUCIE UNIT I WITH 15$ STEAM GENERATOR TUBE PLUGGING - 10CFR50.59 DETERMINATION PAGE TWO 111) The margin of safety, as defined in the basis for every Technical Specification, is not reduced.
'he event by event evaluation presented in the ANF report has shown that all events are bounded by the results of previous cycle analysis except for the Loss of External Load event. The results of the re-analysis presented in the ANF report demonstrate that the maximum system pressure was calculated to be 2725 psia, which is below the vessel pressuri zati on cri teri on of 2750 psi a (110K of design pressure 1 imi t) . Therefore, there i s no reduction in the margin of safety relative to the Technical Specification basis.
CONCLUSION As per Federal Regulation 10CFR50.59(b), the above Safety Evaluation provides the basis to conclude that there is no unreviewed safety issue associated with the increase of the steam generator tube plugging level to 15%. This review within Fuel Resources meets gI-FRN-07 requirements.
ST. LUCIE UNIT I HYDROGEN REGULATOR TO THE VOLUME CONTROL TANK 10CFR50.59 EVALUATION DESCRIPTION OF CHANGE Setpoint change for the hydrogen regulator (V-6758) to the volume control tank (VCT) from 15 psig to 50 psig. The listed operating pressure given in the FUSAR for the VCT is 15 psig. This pressure corresponds to the minimum setpoint for required hydrogen concentration. The reason for change is to allow greater flexibility for hydrogen injection into the RCS in order to maintain specified hydrogen concentration requirements.
SAFETY EVALUATION The probability of an accident or malfunction previously evaluated has not been increased. Also, the possibility of creating an accident or malfunction of a different type than any previously evaluated has not been created. This proposed change enhances the operability of the charging pumps by providing a higher suction pressure and enhancing the net positive suction head. This change does not affect the basis of any technical specification since the component or parameter is not referenced in the Technical Specification and will allow maintenance of chemistry requirements referenced therein.
Based on the above, it is concluded that the change in question does not constitute an unreviewed safety question and prior Commission approval is not needed.
SAFETY EVALUATION FOR REMOVING THE PAGE CAPABILITY FROM THE BELL PHONES DESCRIPTION OF CHANGE This change removed the paging capability from the Bell phones because of unrestricted access of high volume paging. The Control Rooms still have the capability for high volume paging via the Gaitronic phones.
SAFETY EVALUATION The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FUSAR are not increased because this communication system is not considered in the evaluation.
The possibility of an accident or malfunction of a different type than any evaluated previously in the FUSAR will not be created.
111) The margin of safety as defined in the basis for any Technical Specification is not reduced.
INCREASE IN RCS pH AT ST. LUCIE UNIT 2 DURING FUEL CYCLE 4 DESCRIPTION OF CHANGE This evaluation is concerned with changing the Reactor Coolant System (RCS) Lithium/Boron Control Program at St. Lucie Unit No. 2. The proposed change involves increasing the RCS pH at St. Lucie Unit No. 2 for the remainder of Fuel Cycle 4, from 6.9 to a 7. 1 - 7.4 band by increasing the RCS lithium concentration to a maximum of 2.4 ppm. The lithium will then be held constant (7. 1 - 7.4) for the remaining period of core life. This change is applicable only to St. Lucie Unit No. 2 and only for the remainder of the current fuel cycle (Fuel Cycle No. 4).
The purpose for raising the pH flow from 6.9 to 7. 1 - 7.4 is because recent data indicates that slightly high pH will lower the transport of corrosion products from the core surfaces; thus lower the Plant radiation levels.
SAFETY EVALUATION The change in RCS pH is not an unreviewed safety question based upon the following:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FUSAR are not increased.
The bases for Conclusion I are as follows:
a) The proposed operation with a 7.1 - 7.4 RCS pH does not affect any safety features of the Plant.
b) Operation with a pH of 7.4 is within the normal range of operation allowed by the fuel vendor' longstanding traditional recommendations for RCS lithium concentrations (1.0 - 2.0 ppm).
c) The basis for this change is consistent with the FUSAR stated objectives for chemistry and purity control, Section 9.3.
d) Lithium Hydroxide is the design basis pH control chemical for the RCS.
e) The maximum lithium (2.4 ppm) for this operation is within the range of the vendor's current recommendations for RCS lithium concentration.
INCREASE IN RCS pH AT ST. LUCIE UNIT 2 DURING FUEL CYCLE 4 PAGE TWO With respect to any concern for Zircaloy hydriding and fuel failures, the threshold for accelerated attack on Zircaloy is approximately 35 ppm. The maximum lithium concentration for this operation is 2.4 ppm, providing a significant margin from an area of concern.
Consequently, the probability of fuel cladding failures is not significantly altered by the proposed change.
g) The St. Lucie Unit No. 2 steam generators do not contain tube material considered highly susceptible to primary side cracking. However, with regard to steam generator tubing failure concerns, it has been found that within the typical range of PWR chemistry control, primary side cracking of susceptible Inconnel 600 steam generator tubing is relatively insensitive to pH. Recent tests conducted by EPRI indicate that there is no significant effect of lithium or pH over the range of practical interest on highly stressed specimens of susceptible tube material. Similar indications have been observed for less highly stressed specimens and less susceptible materials.
h) With regard to other primary system boundary failures, no negative effects are postulated for the proposed change.
Existing Technical Specification chemistry limits and LCO's are unaffected by this change in operation.
- 2. The possibility of an accident or malfunction of a different type than any evaluated previously in the FUSAR will not be created.
The basis for Conclusion 2 is as follows:
a) The proposed operation does not involve any changes to the Plant or its design basis other than a short term increase in the RCS lithi.um concentration and a new constant pH band being maintained.
- 3. The margin of safety as defined in the basis for any Technical Specification is not reduced.
The basis for Conclusion 3 is as follows:
a) No technical specification changes are necessary to implement this operation. All technical specification chemistry limits and LCO's remain in effect.
INCREASE IN RCS pH AT ST. LUCIE UNIT 2 DURING FUEL CYCLE 4 PAGE THREE CONCLUSIONS Based on the above evaluation, it is concluded that the mid-cycle pH change is safe to perform and the change can be implemented without prior NRC approval pursuant to 10CFR50.59 because it does not involve a change in Plant technical specifications and because it is not an unreviewed safety question.
ST. LUCIE UNIT I FUEL HANDLING AND CASK DROP RADIOLOGICAL ASSESSMENT PURPOSE Safety evaluation to address fuel handling and cask drop radiological releases to a maximum assembly burnup of 47,000 MWD/MTU. This evaluation supports fuel handling operation of the Batch H assemblies, currently in the core, whose burnups will be exceeding the previously analyzed radiological assessment burnup limit of 44,600 MWD/MTU.
SAFETY EVALUATION Based on the technical evaluation performed and the acceptable results shown, it can be stated that:
The probabi1 i ty of occurrence or the consequences of an accident or mal function of equipment important to safety previously evaluated is not increased.
The increase in assembly burnup limit does not change the overall .configuration of the Plant. The increase in burnup does not require a physical change to the fuel or fuel handling equipment at the Plant. Since the mode of operation remains unchanged, the probability of occurrence of an accident or malfunction of equipment important to safety is not impacted.
The radiological consequences of an accident or malfunction of equipment important to safety has been evaluated. As demonstrated in the safety evaluation, the radiological consequences of the Fuel Handling and Cask Drop accidents result in site boundary does which are a small fraction of the 10CFR100 limits.
The probability of an accident or malfunction of a different type than any previously analyzed in the safety analysis is not created.
As mentioned earlier, the increase in burnup does not require a change in the overall configuration of the Plant. The mode of operation remains unchanged since the increase in burnup does not require changes in fuel design.
ST. LUCIE UNIT 1 FUEL HANDLING AND CASK DROP RADIOLOGICAL ASSESSMENT PAGE TWO The margin of safety as described in the basis for every technical specification is not reduced.
The radiological assessment of increasing the fuel assembly discharge burnup from 44,600 MWD/MTU to 47,000 MWD/MTU has shown that the results are well within the design basis.
Based on this finding, it is determined that the increase in assembly exposure does not result in a reduction in the margin of safety relative to the technical specification basis for St. Lucie Unit 1.
CONCLUSION As per Federal Regulation 10CFR50.59(b), the above Safety Evaluation provides the basis to conclude that Fuel Handling and single assembly cask operations involving fuel'ssemblies with burnups up to 47,000 MWD/MTU do not involve any changes which introduce an unreviewed safety question. Therefore, implementation of this change is permissible without prior NRC approval.
REMOVAL OF GUIDE TUBE PLUGGING DEVICES ON ST. LUCIE UNIT I PURPOSE St. Lucie Unit 1 was originally designed with part length control element assemblies (PLCEAs). The PLCEAs were intended to help control the effect of xenon oscillations on axial power distributions during normal Plant operation. Prior to the startup of Cycle 2, a study was completed that determined the PLCEAs were not necessary. In order to preserve the dynamic operating characteristics of the reactor, the PLCEAs were replaced by Guide Tube Plugging Devices (GTPDs).
Each GTPD was designed identical to the PLCEAs; i.e., five firigers per plugging device. A total of six GTPDs were installed. As a result, during each fuel reload, these plugs must be removed from the core and re-installed in their proper core locations prior to startup.
Removal of these GTPDs can potentially result in a savings of three or four hours of critical path fuel movement time each refueling outage.
SAFETY EVALUATION Based on the techni ca 1 eva1 uati on performed, i t can be concluded that the removal of the GTPDs, during the Cycle 9 refueling, does not result in a violation of any design criteria, is bounded by the reference analyses and can be implemented with no changes required to the existing St. Lucie Unit 1 Technical Specifications. Therefore:
i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased.
Removal of the St. Lucie Unit I GTPDs does not result in a change to the overall operation and performance of the Plant. Therefore, the probability of occurrence of an accident or malfunction of equipment important to safety is not impacted. The impact of the change in flow characteristics was determined, and the effect on all transient analyses and setpoints was evaluated. The consequences of the analyses evaluated do not exceed their respective acceptance criteria.
The possibility of an accident or malfunction'f a different type than any previously analyzed in the safety analysis is not created.
RggVAL OF GUIDE TUBE PLUGGING DEVICES ON ST. LUCIE UNIT 1 PAGE TWO 111) The margin of safety as defined in the basis for every technical specification is not reduced.
The removal of the St. Lucie Unit 1 GTPDs does not impact the neutronics input to the safety analysis assumed for Cycle 9. The safety analyses have been reviewed, and in all cases the results are well within the acceptance criteria of the design basis. Based on an FPL safety evaluation, it can be determined that the removal of the St. Lucie Unit I GTPDs during the Cycle 9 refueling outage does not result in a reduction to the margin of safety relative to the Technical Specification basis for St. Lucie Unit 1.
CONCLUSION As per 'federal Regulation 10CFR50.50(b), the above Safety Evaluation provides the basis to conclude that the removal of the St. Lucie Unit 1 guide tube plugging devices (GTPDs) during the Cycle 9 refueling outage does not involve any changes which introduce an unreviewed safety question. Therefore, implementation of this change is permissible without prior NRC approval.
ST. LUCIE UNIT I CYCLE 9 FUEL RELOAD ANALYSIS PURPOSE The St. Lucie Unit 1 Cycle 9 Safety Analysis Report presents the evaluation of the reload core characteristics with respect to the safety analyses presented in the St. Lucie Unit I Cycle 8 Safety Analysis Report.
SAFETY EVALUATION Based on the technical evaluation performed and the results of the re-analyses included in this SAR report, it can be concluded that the St. Lucie Unit I Cycle 9 reload design meets all the design criteria, is bounded by the reference analyses and can be implemented with no changes required to the existing St. Lucie Unit 1 Technical Specifications. Therefore:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased.
The St. Lucie Unit I Cycle 9 reload does not result in any changes to the overall configuration of the Plant.
The mode of operation remains unchanged. Therefore, the probability of occurrence of an accident or malfunction of equipment important to safety is not impacted. The physics data input to safety calculations incorporated in Cycle 9 were found to be .conservatively calculated. The fuel design modifications incorporated for Cycle 9 were evaluated.
The CEA ejection event and the Mode 5 and 6 Boron Dilution events were the only transients re-analyzed for Cycle 9.
As discussed in the safety evaluation of this document, the results of both the CEA ejection and the Boron Dilution events were well within the acceptance criteria. Based on the conclusions of the independent FPL review and the attached safety evaluation, the consequences of accidents or the malfunction of equipment important to safety previously evaluated is not increased.
The possibility of an accident or malfunction of a different type than any previously analyzed in the safety analysis is not created.
The St. Lucie Unit 1 Cycle 9 reload design does not result in any changes to. the overall configuration of the Plant, or in the actual . Plant operation. The possibility of an accident or malfunction of a different type than previously analyzed in the safety analysis is not created.
ST. LUCIE UNIT I CYCLE 9 FUEL RELOAD ANALYSIS PAGE TWO iii) The margin of safety as defined in the basis
'technical s'pecification is not reduced.
for every The St. Lucie Unit 1 Cycle 9 reload design neutronics input and the resulting safety analysis have been reviewed, and in all cases, the results are well within the acceptance criteria of the design basis. Based on FPL independent technical reviews of the SAR report, it can be determined that the St. Lucie Unit I Cycle 9 reload design does not result in a reduction to the margin of safety relative to the Technical Specification basis for St. Lucie Unit 1.
CONCLUSION As per Federal Regulation 10CFR50.59(b), the above Safety Evaluation provides the basis to conclude that the implementation of the St.
Lucie Unit I Cycle 9 reload does not involve any changes which introduce an unreviewed safety question. Therefore, implementation of this change is permissible without prior NRC approval.
ST. LUCIE UNIT I SAFETY EVALUATION FOR GAGGING V-3483 SHUTDON COOLING SUCTION RELIEF PURPOSE This safety evaluation addresses gagging V-3483 to allow testing the St. Lucie Unit I Integrated Safeguards Test and to continue with the refueling outage activities. The Shutdown Cooling Return Relief Valves V-3468 and V-3483 provide redundant overpressure protection for the Shutdown Cooling System during solid RCS operation with all charging pumps running. V-3483 had unsatisfactory performance with unrepeatable lift setpoint; therefore, it was gagged until repaired, protection for the Shutdown and V-3468 provided the overpressure Cooling System.
SAFETY EVALUATION The proposed activity does not increase the probability or consequences of an accident, nor cause an increase in the probability of malfunction or consequences of malfunction of equipment important to safety.
Gagging of V-3483 does not create the possibility of an accident previously unanalyzed in the SAR or create the possibility of a malfunction of eguipment important to safety of a different type than previously evaluated in the SAR.
ST. LUCIE UNIT 1 EVALUATION OF STEAM GENERATOR SLUDGE LANCING CONCURRENT WITH REFUELING OPERATIONS PURPOSE A safety evaluation was written to support steam generator sludge operations using Penetration 854, ILRT Pressurization Station, and Penetration 856, Hydrogen Purge Makeup. This evaluation is applied to the sludge lance activities performed on St. Lucie Unit I steam generators during the Summer 1988 refueling outage.
SAFETY EVALUATION The steam generator sludge lancing arrangement concurrent with refueling operations does not:
increase the probability of occurrence or the consequences of an accident or mal function of equipment important to safety because these penetrations used for sludge lancing activitiqs having direct access from the containment atmosphere to the outside atmosphere are capable of being closed by a manual isolation valve, or the direct pathway has been fitted with a leak-tight stuffing box to maintain an air tight seal at the containment penetration; 4
create the possibility of an accident or mal function of a different type than previously analyzed; iii) reduce the margin of safety as defined in the basis for any Technical Specifications.
ST. LUCIE UNIT 1 POTENTIAL LEAKING STEAM GENERATOR TUBE PLUGS SAFETY EVALUATION PURPOSE Ouring the July 1988 refueling outage at St. Lucie Unit 1, visual inspection of the primary side of the steam generator tube sheet indicated possible leakage of Westinghouse mechanical plugs. These Westinghouse plugs were installed during the Oecember 1985 refueling outage. Westinghouse has written a safety evaluation to conservatively address subsequent Plant operation with steam generator mechanical plugs. which may be leaking as a result of degradation such as primary water stress corrosion cracking or plugs which may be leaking at the tube-to-plug interface.
SAFETY EVALUATION 10CFR50.59 allows a change to a nuclear facility without prior NRC approval if an unreviewed safety question does not exist and if changes to Technical Specifications are not involved. Based on the evaluation of failure modes, it is concluded that an unreviewed safety question does not exist relative to the potentially leaking plugs since, The probability of occurrence of a. design basis accident is not increased since the leakage through a crack in the plug or between the tube and plug interface would not be expected to produce any greater leakage than that caused by small cracks in a steam generator tube, and loss of a steam generator tube plug would not be expected since the structural integrity of the plug is not in question.
The consequences of a previously postulated design basis accident are not made more severe for the same reason given in (i) and since no accident-mitigating equipment or systems have been altered.
The possibility of an accident of a different type than previously addressed in the FSAR does not exist since a steam generator tube rupture concurrent with tube plug failure is not considered a possible failure mode. Only tube plug leakage (within Tech Spec limits) is expected for the limiting case of a tube fish-mouth failure.
iv) The margin of safety as defined in the basis for any Technical Specification is not reduced since no change is being made with respect to the number of plugged steam generator tubes, . and any primary-to-secondary leakage indicative of a degraded tube plug could be identified by existing radiation monitoring equipment and secondary chemistry sampl.ing procedures. This would permit an orderly Plant shutdown.
ST. LUCIE UNIT 1 POTENTIAL LEAKING STEAN GENERATOR TUBE PLUGS SAFETY EVALUATION Page Two CONCLUSION The assessment has concluded that continued operation of the unit is acceptable since the evaluation has shown that no accidents or safety concerns outside of those analyzed in the FSAR are generated, and operation within the current Tech Specs will continue to provide appropriate detection and control parameters.
Finally, the margin of safety of the Plant, as defined in the basis of any technical specification has been evaluated and is not reduced.
Even if a plugged n'on-active tube experienced through-waIl degradation (including a fish-mouth opening as a result of the flow diode effect),
a large primary-to-secondary leakage event would not be expected and orderly Plant shutdown could be achieved.
Therefore, subsequent Plant operation of St. Lucie Unit 1 does not represent an unreviewed safety question pursuant to 10CFR50.50 criteria.
ST. LUCIE UNIT 2 SAFETY EVALUATION FOR GAGGING SR-14350 2A SHUTDOWN COOLING HEAT EXCHANGER PURPOSE The change involved the gagging of a safety relief valve SR-14350 that provides overpressure protection of a protected boundary which includes the 2A Shutdown Cooling Heat Exchanger. The safety relief valve was gagged to terminated the premature lifting of the safety relief.
SAFETY EVALUATION The self-actuated spring loaded relief valve is described in paragraph 9.2.2;2 of the FUSAR. The relief valve is described as a device used for overpressure protection of the shutdown cooling heat exchanger.
The overpressure protection is required on the SDC heat exchanger to prevent an overpressurization event caused by thermal expansion of fluid should the protected boundary be isolated while filled with water. However, the normal lineup of the component cooling water system requires that the protected boundary isolation valves (SB-14348
& SB-l4365) to be placed in the locked open position except for maintenance purposes. The position of these valves is verified to be locked open on a quarterly basis per Administrative Procedure 1-0010123.
Therefore, based on the above:
The change does not increase the probability of an accident previously evaluated sin'ce the protected boundary is not isolated, thus an overpressurization event would not occur.
The proposed change does not create conditions for a different type of accident than was previously evaluated since a redundant shutdown cooling heat exchanger is available and there is no single failure that could prevent the shutdown cooling system from performing its safety function.
The margin of safety as defined in the Technical Specifications is not reduced since the safety relief valve is not referenced therein and the gagged relief valve wi 11 not render the shutdown cooling system inoperable.
Furthermore, based on the previous statement, a change to the Technical Specifications is not required.
ST. LUCIE UNIT 2 SAFETY EVALUATION FOR GAGGING SR-14350 2A SHUTDOMN COOLING HEAT EXCHANGER PAGE TWO CONCLUSION Based on the above, it is concluded that the change in question does not constitute an unreviewed safety question nor a change to the Technical Specifications; therefore, prior approval by the NRC is not required.
ST. LUCIE UNIT I SAFETY EVALUATION TO PERMIT INCREASING REACTOR COOLANT SYSTEM LITHIUM CONCENTRATION TO REDUCE RCS RADIATION LEVELS BACKGROUND Lithium hydroxide is used to control RCS pH to maintain a zero coefficient of solubility for dissolved corrosion products, i.e.,
crud. This results in crud going into solution in hotter regions of the RCS (the core) and crud deposition occur ring in cooler regions of the RCS (the steam generators). The overall core crud load would be reduced by preventing crud from depositing on the fuel surfaces.
Although the crud load in the steam generators would increase, the net effect is to minimize the activation of corrosion products by reducing their residence time in the core. Currently, RCS pH control is maintained at constant pH 6.9 by a coordinated lithium-boron control program. This program is based on crud being comprised primarily of magnetite, and the zero coefficient of solubility of magnetite occurs at pH 6.9. Recent investigations of crud solubility, however, suggest that crud is made up mostly of cobalt and nickel substituted ferrites whose zero solubility coefficient occurs at pH 7.4. It is surmised that significant reductions in RCS radiation levels are attainable by operating at this higher pH.
SCOPE This safety evaluation proposes ta increase the lithium concentration in the RCS, as a test, from the current limit of 2.2 ppm to 3.5 ppm for the duration of St. Lucie Unit 1 Cycle 9. This corresponds to an initial pH 6.9 at 2000 ppm boron. Lithium concentration would be held constant at 3.5 ppm from BOC until pH 7.4 is reached at approximately 650 ppm boron. A coordinated lithium-boron program would then be instituted to maintain pH 7.4 until the end of cycle.
SAFETY EVALUATION Fuel Rod Performance No design changes are being made to the fuel assembly dimensions, nor will any occur as a result of the higher Li/pH control. Given that fuel rod performance is within limits of previous analysis, rod growth, and assembly irradiation induced dimensional changes will also remain within analyzed limits.
The effect of operating at the higher lithium and pH levels proposed does not result in an increase in the probability of fuel failure either directly or indirectly. Corrosion failure is not probable as neither sufficient lithium concentration nor a high enough pH exist for accelerated corrosion to occur, nor is there any secondary failure mechanisms that might increase the probability of fuel cladding failure. Corrosion levels will not be greater than previously experienced, and no accelerated hydrogen pickup due to lithium has been demonstrated; consequently no weakening of the cladding or failure due to hydriding will occur.
ST. LUCIE UNIT 1 SAFETY EVALUATION TO PERMIT INCREASING REACTOR COOLANT SYSTEM LITHIUM CONCENTRATION TO REDUCE RCS RADIATION LEVELS Page Two Fuel Rod Performance (cont'd)
Neither the probability nor the consequences of any accident previously analyzed are increased because operation at a higher lithium concentration and pH does not result in fuel corrosion or mechanical behavior either greater than, or different from, previously considered in the input to any safety analysis.
The possibility of a different accident than any already evaluated is not increased because neither the fuel nor its modes of operation will be changed by operating at the proposed lithium and pH limits.
No changes to the fuel or to the fuel operating environment, other than increasing lithium and pH, are being proposed. Hecause these changes have no impact on the design of the fuel or its operation, there is no increase in the possibility of creating a new or different type of accident than previously analyzed.
The margin of safety as described in the basis of any technical specification is not reduced because no changes in any fuel analysis input or assumptions are required as a result of the proposed changes; nor are any changes to analysis methodology necessary to describe fuel rod behavior. As no inputs, assumptions, or methods have changed, the results of previous safety analysis remain unchanged.
In conclusion, increasing the St. Lucie Unit 1 RCS pH to 7.4 and the maximum lithium concentration to 3.5 ppm, poses no threat to fuel rod integrity. Experiments to determine the corrosive effects of lithium on zircaloy demonstrated that lithium attack only occurs at high pH and high Li concentrations, neither of which is present at St. Lucie Unit 1. No change in fuel rod performance is anticipated due to the elevated pH program. As such, no safety analysis is impacted nor is any new analysis required. Programs in place at Millstone and Calvert Cliffs will be examined for important conclusions applicable to FPL at St. Lucie.
BALANCE-OF-PLANT The proposed increase in lithium concentration does not increase the corrosion rate nor increase the incidence of stress corrosion of the components wetted by primary coolant or the letdown to other systems. The resultant increase to a slightly basic primary water solution will have no effect on the design life or performance of equipment important to safety, since corrosion rates will not be adversely affected. Therefore, the proposed increase in lithium levels does not increase the probability of accidents or malfunction to equipment previously evaluated.
ST. LUCIE UNIT 1 SAFETY EVALUATION TO PERMIT INCREASING REACTOR COOLANT SYSTEM LITHUM CONCENTRATION TO REDUCE RCS RADIATION LEVELS Page Three BALANCE-OF-PLANT (cont ')
The operability of the spray addition system ensures that sufficient amounts of sodium hydroxide is added to the containment spray in the event of a loss of coolant accident (LOCA) in order to maintain a pH value of 8.5 to 11.0 (T.S. 3/4.6.2.2). The increase in lithium concentration is negligible compared to the sodium hydroxide concentration in the containment sump solution and does not alter the resultant pH for the analyzed events. Therefore, the basis for the spray additive system is not affected, and the consequences of an accident previously analyzed in the FSAR are not affected.
The proposed increase in lithium level wi 11 not create a malfunction or a different failure mechanism than previously evaluated, since the corrosion rates will not increase for the Plant components which contact the primary coolant or its letdown.
The limitations -on the RCS chemistry (T.S. 3/4.4.7) ensure that corrosion of the components wetted by primary water is minimized and reduce the potential for RCS leakage or failure due to stress corrosion. Increasing the lithium concentration as proposed will not reduce the corrosion protection nor increase the potential for stress corrosion, and the structural integrity of the wetted components will not be adversely affected. The limitations for RCS pH levels and lithium concentration are not provided for in the Technical Specifications, but are specified in the updated final Safety Analysis Report and Plant chemistry operating procedure. Therefore, the proposed change in lithium levels will not reduce the margin of safety as defined for'ny Technical Specification basis.
In conclusion, the wetted materials within those systems contacting primary fluid were determined to be insensitive to the prov .sed increase in pH level, since this results only in a slight increase in pH from 6.8 - 6.9, to 6.9 - 7.4. Corrosion rates or incidence of stress corrosion cracking for the wetted materials at elevated lithium level and pH level will not increase. The performance of systems credited in previously analyzed events in the updated final Safety Analysis Report would not be adversely affected.
CONCLUSION Title 10 of the Code of Federal Regulations, Section 50.59, permits tests to the Plant as described in the Safety Analysis Report without prior Nuclear Regulatory Commission approval, provided certain criteria are met. It must be shown that the Plant test does not involve an Unreviewed Safety guestion or require a change to the Technical Specifications.
ST. LUCIE UNIT I SAFETY EVALUATION TO PERMIT INCREASING REACTOR COOLANT SYSTEM LITHIUM CONCENTRATION TO REDUCE RCS RADIATION LEVELS Page Four CONCLUSION (cont'd)
This safety evaluation concludes that increasing the reactor coolant system lithium concentration as described, will not pose an Unreviewed Safety guestion, require a change to the Technical Specifications, or require prior Nuclear Regulatory Commission approval.
ST. LUCIE UNIT I TEMPORARY TEE CONNECTION ON THE INSTRUMENT AIR LINES TO THE FEEDMATER REGULATING VALVES, FCV-9011 Si FCV-9021 DESCRIPTION OF CHANGE A temporary mechanical jumper connection installed in the instrument air supply line to the feedwater regulating valve actuator. The purpose of the jumper is to measure the air signal supplied to the valves actuator, which was part of the ongoing trouble-shooting activities related to erratic oscillation of steam generator levels.
SAFETY EVALUATION This temporary modification does not involve an unreviewed safety question because:
The probability of occurrence or the consequences of an accident or malfunction evaluated in the FSAR is not increased because the modification meets all of the applicable design requirements of the existing tubing.
Additionally, the isolation valves will be normally closed and capped to preclude leakage. Furthermore, the failure of the tie-ins will result in transients already addressed in the FSAR.
The possibility of an accident or malfunction of a different type than evaluated in the FSAR is not created. As discussed above, all failure mechanisms of. the tie-ins result in transients and scenarios that are already evaluated in the FSAR. Protective equipment exists to mitigate these transients.
This modification does not reduce the margin of safety as defined in the basis of the Technical Specifications as this portion of the feedwater system is non-safety, non-seismic and not a portion of the Technical Specifications.
As per 10CFR 50.59, this change does not involve an unreviewed safety question, and prior NRC approval for implementation is not required.
No Technical Specification changes are required for implementation.
ST. LUCIE UNIT I CYCLE 8 EXTENDED BURNUP ASSEMBLIES PURPOSE Safety Analysis Report to support continued power operations on St.
Lucie Unit I, Cycle 8 beyond 9450 EFPH.
SAFETY EVALUATION Since all the analyses affected by the extended burnup assembly operation have been reviewed and demonstrated to meet the currently approved safety criteria, it can be stated that:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased.
The increased assembly burnup limit does not change the overall configuration of the Plant. The increase in burnup does not require a physical change to the fuel or fuel handling. equipment at the Plant. Since the mode of operation remains unchanged, the probability of occurrence of an accident or malfunction of equipment important to safety is not impacted.
The safety analyses performed in support of increasing the assembly burnup beyond 44,500 N(0/NTU demonstrates that the consequences of an accident or malfunction have not been increased beyond those evaluated previously. This is demonstrated in the radiological release calculations where overly conservative release calculations result in site boundary doses which are a small fraction of the 10C FR100.
The possibility for an accident or malfunction of a different type than any previously analyzed in the Safety Analysis is not created.
As mentioned earlier, the increase in burnup does not require a change in the overall configuration of the Plant. The mode of operation remains unchanged since the increase in burnup.does not require changes in fuel design.
ST. LUCIE UNIT I CYCLE 8 EXTENDED BURNUP ASSEMBLIES PAGE TWO The margin of safety as defined in the basis for every technical specification is not reduced.
The re-analyses of the mechanical design, radiological assessment, LOCA/ECCS and rod bow effects, to support extension of the maximum end of life (EOL) peak assembly average exposures, have shown that the results are well within the design basis. The effects of rod bow, which affect the MDNBR criteria, have been shown to be bounded by the existing 1.22 safety limit for assembly exposures up to 52,500 MWO/MTU. The analysis shows a small rod bow penalty (<0.5%) above the 1.22 safety limit for assemblies with burnups in the range of 50,900 - 52,500 MWD/MTU.
However, due to the depressed power levels associated with these high burnup assemblies, the calculated DNBR (for these extended burnup assemblies) is bounded by the ONBR associated with higher powered fresh fuel assemblies. Based on these findings, it is determined. that the increase in assembly exposures does not result in a reduction in the margin of safety relative to the Technical Specification basis for St. Lucie Unit l.
CONCLUSION As per Federal Regulation 10CFR50.59(b), the above Safety Evaluation provides the basis to conclude that operation of Cycle 8 beyond 9450 EFPH does not involve any changes which introduce an unreviewed safety question. Therefore, implementation of this change is permissible without prior NRC approval.
ST. LUCIE UNIT 2 SAFETY EVALUATION FOR USE OF SEALING COMPOUND ON VALVE FCV-23-6 STEAM GENERATOR BLOMDOMN CONTAINMENT ISOLATION VALVE PURPOSE St. Lucie Unit 2 Blowdown .Containment Isolation Valve FCV-23-6 was reported to be leaking. This valve was previously repaired per Nonconformance Report 2-113 at the bonnet connection. The purpose of this evaluation is to provide a method for a temporary leak repair by injecting a sealing compound for the body to bonnet connection.
This valve is non-isolated and normally open.
SAFETY EVALUATION The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety previously evaluatd in the Safety Analysis Report has not been increased.
This repair does not adversely affect the integrity of the blawdown boundary other than to aid in sealing a leaking valve's bonnet and lower flanged joint. The probability of a pipe rupture is not increased since the stru'ctural integrity of the blowdown system components is not impacted.
The impact of the added weight of the sealant, fittings, and cap nuts on the stress analysis and supports has been evaluated and determined to be satisfactory.
The possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report has not been created.
The proposed repair does not provide a new mode of normal or emergency Plant operation. The valve is required for maintaining containment isolation, and its ability to do so will be enhanced with the injection of the sealant into the valve. In addition, new Plant hardware other than the cap nuts and fittings previously described, are added by this repair.
The margin of safety as defined in the basis for any Technical Specification has not been reduced.
Chemistry limits are not altered and no other change is proposed to the Plant design, modes of operation or assumptions in <he Technical Specification or safety analysis.
0 ST. LUCIE UNIT 2 SAFETY EVALUATION FOR USE OF SEALING COMPOUND ON VALVE FCV-23-6 STEAM GENERATOR BLOWDON CONTAINMENT ISOLATION VALVE PAGE TMO CONCLUSION 10CFR50.59 allows changes/repairs to a facility as described in the FSAR if an unreviewed safety question does not exist and if a change to the Technical Specifications is not required. As shown in the preceding sections, the change proposed does not involve an unreviewed safety question or a Technical Specification change because each concern posed by 10CFR50.59 can be positively answered.
ST. LUCIE UNIT I SAFETY EVALUATION FOR RAISING LOW PRESSURE ANNUNCIATION SETPOINTS IN THE INSTRUMENT AIR SYSTEM DESCRIPTION OF CHANGE Raise the low pressure annunciation setpoints in pressure switches PS-18-7 and PS-18-4. The current low pressure setpoints do not provide adequate warning time for a problem with the Instrument Air System because of the higher operating pressure band. PS-18-7 will be raised from 75 psig to 95 psig and PS-18-4 will be raised from 80 psig to 100 psig.
SAFETY EVALUATION The Instrument Air System performs no safety related function and is not required for Plant safe shutdown nor is it required to mitigate the consequences of an accident. Changing the low pressure setpoint for the subject pressure switches does not involve an unreviewed safety question, and the following are bases for this justification:
The prpbabi1 i ty of occur rence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not increased. The Instrument Air System is not used in any safety analysis for accidents or malfunctions of equipment and therefore changing the subject setpoints will not have any adverse effect on equipment important to safety.
The possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report is not created. Pressure switches PS-18-4 and PS-18-7 perform no safety related function and will not, therefore, introduce any new failure modes to safety related equipment.
This change has no effect on the operational design of the system.
1 The margin of safety as defined in the basis for any Technical Specification is not affected by this change.
PS-18-4 and PS-18-7 are not part of the bases for any Technical Specification.
This change requires no change to the Unit 1 Technical Specifications.
ST. LUCIE UNIT 1 SAFETY EVALUATION FOR RAISING LOM PRESSURE ANNUNCIATION SETPOINTS IN THE INSTRUMENT AIR SYSTEM PAGE TMO CONCLUSION The foregoing constitutes, per 10CFR50.59, the safety evaluation which provides the conclusion that this change does not involve an unreviewed safety question or a change to the Technical Specifications, and may be made on a temporary basis until Supplement 2 to PC/M 050-186 is issued. Prior NRC approval is not required for implementation of this change.
ST. LUCIE UNIT I SAFETY EVALUATION FOR USE OF SEALING COMPOUND ON VALVE FCV-23-6 STEAM GENERATOR BLONON CONTAINMENT ISOLATION VALVE PURPOSE St. Lucie Unit 1 Blowdown Containment Isolation Valve FCV-23-6 was reported to be leaking per Nonconformance Report 1-283. The purpose of this evaluation is to provide a method for a temporary leak repair by injecting a sealing compound for the body to bonnet connection.
This valve is non-isolated and normally open.
SAFETY EVALUATION The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report has
. not been increased.
The valve is required for maintaining containment isolation, and its ability to do so will not be affected because the leak seal will be capable of preventing leakage at 985 psig under normal operating conditions vs psig post-LOCA.
This repair does not adversely affect the integrity of the blowdown boundary other than to aid in sealing a leaking valve body to bonnet joint. The probabi.lity of a pipe rupture is not increased since the structural integrity of the blowdown system components are not impacted. The impact of the added weight of the sealant, fittings, and capnuts on the stress analysis and supports has been evaluated and determined to be satisfactory.
The possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report has not been created. The proposed repair does not provide a new mode of normal or emergency Plant operation.
In addition, no new Plant hardware other than the capnuts previously described are added by this repair.
The margin of safety as defined in the basis for any Technical Specification has not been reduced.
Chemistry limits are not altered and no other change is proposed to the Plant design, modes of operation or assumptions in the Technical Specifications or Safety Analysis.
As shown in the preceding sections, the change proposed does not
, involve an Unreviewed Safety. guestion or a Technical Specification change.