ML17227A747
| ML17227A747 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 12/31/1992 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17227A746 | List: |
| References | |
| NUDOCS 9303030235 | |
| Download: ML17227A747 (49) | |
Text
FRG APPROVED 10 CFR 50 '9 SAFETY EVALUATIONS 9'303030235 930226 PDR ADOCK 05000335 R PDR
Title:
Chemical and Volume Control System and Waste Management System Boric Acid Heat Tracing Circuit De-Energization Safety Evaluation completed 12/20/91 Abstract:
The safety evaluation addresses the effect of de-energizing specific circuits of the heat tracing system associated with the Boric Acid Makeup System. Engineering packages (PC/M 336-189 Rev 2 and PC/M 094-188 Rev 0) reduced the boric acid concentration in the Boric Acid Makeup System. Portions of the Boric Acid Makeup System with boric acid concentrations of 3.5 weight percent or less do not require heat tracing. This evaluation focuses in on the Chemical and Volume Control System (CVCS) and Waste Management System (WMS) (or Boric Acid Makeup and Recovery Systems) for the purpose of de-energizing heat trace circuits that have been identified by St. Lucie Plant Maintenance as not being required.
Per Section 9.3.4.1 portions of the Boric Acid Makeup System are designed and built to meet the requirements of seismic class hence this safety evaluation is classified as nuclear safety related.
. This safety evaluation is in effect until June 30, 1993. An engineering package is scheduled to be completed by August 2,'992 to complete the permanent boric acid heat trace de-energization changes evaluated in this safety evaluation.
The de-energization of the boric acid heat trace circuits as described in this safety evaluation has been evaluated to have no adverse affect on the plant operation, safety, and safety-related systems. The change herein does not involve an unreviewed safety question and does not involve a change in the Technical Specifications or the FSAR.
Safety Evaluation:
De-energizing heat trace circuits do not affect the initiation of an accident evaluated in the FSAR nor increase the probability of occurrence. Desired boric acid concentration is maintained even with the de-energizing of heat tracing circuits; Therefore, the probability of an accident previously evaluated in the FSAR would not be increased.
The consequences of an accident previously evaluated in the FSAR are not increased by the de-energizing heat trace circuits. De-energizing heat trace circuits will not change, degrade, or prevent system functions described in, or assumed to occur in the
- Unit:
Title:
4 Chemical and Volume Control System and Waste Management System Boric Acid Heat Tracing Circuit De-Energizati.on Safety Evaluation completed 12/20/91
/
Safety Evaluation (Continued):
mitigation of any FSAR accident. This proposed activity has no impact on the LOCA analysis and the radiological consequences of an accident evaluated in the FSAR will not be increased. Therefore, the consequences of an accident previously evaluated in the FSAR would not be increased.
The probability of occurrence of a malfunction. of equipment important to safety previously evaluated in the FSAR has not been increased because the proposed activity will not result in new performance requirements being imposed on any system or components such= that any design criteria will be exceeded. The Boric Acid Makeup functional requirements are unchanged, therefore no new probability of malfunction has been imposed. Thus, the probability of a malfunction of equipment important to safety previously evaluated in the FSAR would not be increased.
De-energizing heat trace circuits do not change, degrade, or prevent actions described in, or assumed to occur in the mitigation of any FSAR accident. Therefore, de-energizing heat tracing circuits will not increase the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR.
The proposed activity to de-energize some heat trace circuits in the boric acid system at St. Lucie Unit 1 does not introduce failure modes of a different type than any previous analyzed in the FSAR. The system configuration and the design basis of the boric acid system has not been changed or affected.
The de-energizing of heat trace circuits has been evaluated and does not impact the structural integrity or performance capability of CVCS and Waste Management System. 'he possibility of, a malfunction of equipment important to safety of a different type than any previously evaluated in the FSAR is not increased by this proposed activity.
Removal of heat tracing in areas containing boric acid concentration of 3.5 weight percent or less does not reduce any margins of safety for boration sources and flow path requirements since the concentration of boric acid is within the requirements of the Technical Specification. De-energizing heat tracing does not affect the Technical Specification basis for borated water sources.
' Unit:
Title:
~
Chemical and Volume Control System and Waste Management System Boric Acid Heat Tracing Circuit De-Energization Safety Evaluation completed 12/2/92 Abstract:
This safety evaluation is a supplement of the preceeding safety evaluation; it tracing circuits.
was written to include a greater number of heat This safety evaluation addresses the effect'f de-energizing specific heat trace circuits within the Chemical Volume Control system (CVCS) and Waste Management System (WMS). Boric acid concentrations were reduced in portions of the following systems via two previous modification packages; (1) Boric " Acid Concentration Reduction Modifications, (2) Boric Acid Concentration Reduction.
This safety evaluation was written to determine the acceptability of de-energizing specific heat trace circuits within the CVCS and WMS. The PSL Unit 1 Boric Acid Concentration Reduction Report outlines the technical basis which permits a reduction in the PSL Unit 1 boric acid makeup tank concentration to a range of 2.5 to 3.5 weight percent. The boric acid concentration setpoint of less than or equal to 3.5 weight (wt.) percent, combined with the ambient temperatures that normally exist in the PSL-1 Auxiliary Building, provide conditions under which boric'cid will not
. precipitate. Therefore, for those parts of the CVCS and WMS where boric acid concentration is 3.5 wt. percent or less and the ambient temperature determined to be 50 degrees F or greater, the heat trace circuits at these locations can be de-energized since boric acid will not precipitate at these locations of the CVCS and WMS.
When boric acid concentrations are 3.5 wt. percent or greater, heat trace circuits shall remain energized.
De-energizing heat trace circuits has been determined not to adversely impact the functioning of equipment in the CVCS or WMS.
Where boric acid concentration is 3.5 wt. percent or less, heat tracing is no longer required to prevent boric acid precipitation because the systems and components affected are located in the Auxiliary Building where the minimum ambient temperature is 50 degrees F.
SAFETY EVALUATION The probability of an accident previously evaluated in the FSAR would not be increased because de-energizing heat trace circuits in the Boric Acid System does not affect equipment assumed to initiate accidents evaluated in the FSAR.
The consequences of an accident previously evaluated in the FSAR would not be increased by de-energizing heat trace circuits. The
Title:
Chemical and Volume Control System and Waste Management System Boric Acid Heat Tracing Circuit De-Energization Safety Evaluation completed 12/2/92 Safety Evaluation (Continued):
de-energizing of heat trace .circuits will not change, degrade or prevent system functions described in valued with the mitigation of any FSAR accident. This -proposed change does not impact the radiological consequences of an accident evaluated in the FSAR.
The probability of occurrence -of a malfunction of equipment important to safety previously evaluated in the FSAR has not been increased because the proposed change will not result in new performance requirements being imposed on any plant system or component.
The de-energizing of the heat trace circuits does not change, degrade, or prevent actions described in, or assumed to occur in the mitigation of any FSAR accident. Therefore de-energizing heat tracing circuits will not increase the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR.
The proposed change to de-energize certain heat trace circuits in the boric acid system at PSL Unit 1 does not introduce failure modes of a different type than any previously analyzed in the FSAR.
The system configuration and the design basis of the boric acid system has not been changed or affected. Therefore, the proposed activity does not create the possibility that an accident may be created that is different from any already evaluated in the FSAR.
The de-energization of heat trace circuits has been evaluated and does not impact the structural integrity or performance capability of CVCS and WMS. The possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR is not increased by this proposed change.
De-energization of heat tracing circuits in the boric acid system does not reduce any margins of safety for boration sources and flow path requirements since the concentration of boric acid is within the requirements of the Technical Specification. De-energizing heat tracing does not affect the Technical Specification basis for borated water sources.
Title:
The Use of a Temporary Spent Fuel Pool Cooling System Abstract:
-The purpose of this Safety Evaluation is to address the acceptability of the temporary Spent Fuel Pool Cooling System (SFPCS) piping and component installation and operation and to address the potential for adverse interaction with existing safety related structures, systems or components (SSC).
The temporary SFPCS is installed as a temporary substitute for the permanent SFPCS. The temporary system will remain installed and in operation for the duration of modifications to the SFPCS Heat Exchanger (HX) .
Although the permanent SFPCS is Quality Group C, Safety Related nor is it it is not Nuclear required to operate. during a seismic event. The temporary SFPCS also serves no Nuclear Safety related function, however, potential interaction with the stored spent fuel and Seismic Category l Spent Fuel Pool must be evaluated.
Therefore, this safety evaluation is classified as Quality Related.
Based upon the unreviewed safety question determination, the safety evaluation concludes that the location and operation of the.
temporary SFPCS components do not result in an unreviewed safety question and do not require a change to the Technical Specifcation.
Safety Evaluation:
The temporary SFPCS performs no safety function and adverse interactions between the temporary SFPCS and any safety related SSC's are precluded by location of the temporary SFPCS piping and components. Also, there is no potential for fuel damage and the temporary piping sections do not constitute an initiating event for any accident. Therefore,'he probability of occurrence of an accident previously evaluated in the FSAR is not increased.
The temporary SFPCS piping and component locations have no impact on SSC's relied upon to mitigate the consequences of an accident previously evaluated in the FSAR. Equipment location will prevent any releases to the environment. Furthermore, any potential incident is bounded by previously analyzed conditions. Therefore, the consequences of an accident previously evaluated in the FSAR are not increased.
Handling of piping assemblies in proximity to the spent fuel pool, is controlled by approved administrative procedures and the
Unit: 1
Title:
The Use of a Temporary Spent Fuel Pool Cooling System Safety Evaluation (Continued):
inherent strength of the installed pipe prevents its failure and dropping into the pool. A loss of the temporary SFPCS capability will not cause any SSC important to safety to fail. Therefore, the probability of occurrence of'a malfunction of equipment important to safety previously analyzed in the FSAR is not increased.
The temporary SFPCS piping and component locations have no impact on any SSC's relied upon to mitigate the consequences of a malfunction of equipment important to safety. Also, adverse consequences from the potential to siphon water'rom the spent fuel pool is precluded by the primary pumps location being higher than the SFP water elevation,= and release of SFP effluent to the outdoor environment is prevented by the location of the SFPCS primary loop inside the Fuel Handling Building (FHB). Consequences of the temporary SFPCS failure are not increased since it is a substitute for the permanent SFPCS which does not cause adverse consequences to equipment important to safety. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the'FSAR are not increased.
The location or operation of the temporary SFPCS components does not change the operation, function or design basis of any SSC's important to safety as described'in the FSAR and no new hazards are created that can be postulated to cause an accident different from those previously analyzed in the FSAR. The SFPCS is not a basis for any FSAR postulated accidents. Therefore, there is no possibility that an accident may be created that is different from any already evaluated in the FSAR.
All temporary SFPCS piping and components are located in the FHB.
This precludes interaction with existing SSC s. Also, there are no new initiating features for equipment malfunction that have not been previously evaluated. Therefore, the possibility of a malfunction of equipment important to safety of a different type than previously analyzed in the FSAR is not created.
There are no Technical Specifications affected by location or operation of the temporary SFPCS components. Therefore; there is no reduction in the margin of safety as defined in the basis for any Technical Specification.
Title:
Temporary Installation of Strain Measuring Devices on the Pr'essurizer Relief Valve Discharge Piping Abstract:
This safety evaluation will allow the temporary installation of strain measuring devices on the three pressurizer relief valve discharge pipe lines.
St. Lucie Plant (PSL) informed FPL Nuclear Engineering that the Unit 1 Pressurizer Relief Valve V-1202 has been leaking and will be replaced with a spare relief valve during the short notice outage in September, 1992. The other two relief valves, V1200 and V1201, were not leaking and do not require maintenance. One of the root causes of the leakage could be due to asymmetric pipe loading on the relief valve discharge nozzle. Therefore, it is considered prudent to compare the loads being imposed on the piping adjacent the discharge flange of the relief valves V1200, V1201 and V1202.
This can be achieved by installing strain measuring devices on the ANSI B31.1 piping (Quality Group D) and monitoring the strain during the relief valve installation and plant heatup.
I The strain measuring devices, excluding the data acquisition equipment "Strain Indicator VISHAY P3500", are scheduled to remain in place for the remainder of Cycle 11 up through the beginning of Cycle 12 mode 2 (middle of 1993).
Safety Evaluation:
The proposed change does not affect any equipment whose malfunction is postulated in the FSAR to,'initiate an accident or prevent an accident from occurring. No physical modifications have been performed, to the RCS or connected "systems except to attach the strain gage shims and Omega clips to the pressurizer relief valve discharge piping using micro spot welding generating energy output less than 50 watt-seconds. The piping is 'not Safety Class 1, 2 or
- 3. Additionally, no failure modes have been identified that could initiate an accident previously evaluated in the FSAR. As such, the probability of occurrence of an accident previously evaluated in the FSAR has not increased.
The proposed change does not diminish in any way the ability of the pressurizer relief valve discharge piping or any other safety system to perform its intended function. There is no interaction with any safety related equipment. As such, the proposed change does not- increase the consequences of an accident previously evaluated in the FSAR.
The proposed change is to install strain measuring devices which do not interfere with the operation of any system. There is no
Title:
Temporary Xnstallation of Strain, Measuring Devices on the Pressurizer Relief Valve Discharge Piping Safety Evaluation (Continued)':
interaction with any safety related equipment. The strain gage an/or thermocouples (including temporary field routed cables) will not pass through the sump water screens loose and carried to the containment sump.
if the assembly becomes As such, the probability of occurrence of a malfunction of equipment important to safety-previously evaluated in the FSAR is not increased by the proposed change.
The installation, -operation or failure of the strain measuring devices will neither impact operation of any system nor cause any adverse affect to any safety related equipment. ,Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR are not increased by the proposed change.
The proposed change does not introduce any new failure modes to or impact safety related equipment in any way. The proposed change will not change or impact the requirements of the decision bases of the safety related systems as described in the FSAR. There are no postulated failure modes which could be considered accident initiating. Therefore, the possibility of an accident of a different type than any previously evaluated in the FSAR is not created by this modification.
The proposed change does not interact spatially or functionally with any SSC important to safety other than the attachment of strain gages to the pressurizer safety relief valve discharge piping. No new failure modes are created that can be postulated to interact with any equipment important to safety different than those previously evaluated in the FSAR. Therefore, the possibility of a malfunction of equipment important to safety which is of a different type than any previously evaluated in the FSAR is not created by the proposed change.
The Technical Specification requirements and Technical Specification Bases are not affected by the proposed change. The proposed change does not affect any plant Technical Specification requirement. The proposed change maintains the level of protection previously evaluated in the FSAR. Therefore', the margin of safety as defined in the bases for any Technical Specification is not reduced by this modification.
Title:
10, CFR 50.59 Evaluation for Excavation of Diesel Generator 1A Fuel Oil Transfer Line 1-2"-DO-13 Abstract:
This safety evaluation is prepared to document the acceptability of exposing the underground portion of pipe line I-2"-DO-13 from the floor penetration of the diesel fuel oil transfer pump 1A cubical to the inlet of the diesel generator building for leak inspection and repair. Line I-2"-DO-13 connects the diesel oil transfer pumps to the 1A diesel generator day tanks. This line is safety class three (3), seismic class I. Plant operation and safety are not affected by excavation of the fuel transfer line.
This evaluation concludes that the excavation of the line does not constitute an unreviewed safety question nor require a change to the Technical Specification. Therefore, prior NRC approval is not required.
The following restrictions apply:
Pipe is required to be supported at a maximum spacing of 17 feet between support points for straight horizontal runs and 13.5 feet for out-of-plane runs (ie. with elbows). Sandbags (85g or greater) or other equivalent restraining methods (each side of the line to the trench wall) shall be installed to provide adequate horizontal and vertical support for the pipe.
Safety Evaluation:
FSAR Section 15.2.9 evaluates station blackout. The probability of occurrence of a station blackout event has not been increased since excavating the underground portion of pipeline I-2"-DO-13 does not adversely affect the functionality of the line. No other evaluated accidents are applicable.
The consequences of an accident previously evaluated in the FSAR have not been increased since the proposed pipe excavation does not create a new path for uncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function. The excavation of the piping does,not adversely impact any equipment which is required to perform a safety related function or initiate actuation of any safety systems. The diesel generator system will continue to function as required.
The probability of= occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report has not been increased. No new failure modes for any equipment important to safety are introduced by excavating the
Title:
10 CFR 50.59 Evaluation for Excavation -of Diesel Generator 1A Fuel Oil Transfer Line 1-2"-DO-13 Safety Evaluation (Continued):
underground portion I-2"-DO-13 and no new components or equipment are introduced that could adversely interact with any equipment important to safety. The seismic integrity of the diesel oil transfer pipeline is not affected by the excavation provided the line is supported. Additionally, 'missile protection is maintained by virtue of the line's size and its location in the trench.
The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR have not been increased since the excavation of'the underground portion of I-2"DO-13 does not create a new path for uncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function. The proposed pipe excavation will not adversely impact any equipment which is required to perform an active safety related function or to initiate actuation of any safety systems. The diesel generator system will'ontinue to function as required.
The possibility of an accident of a different type than any evaluated previously in the safety analysis report has not been created since the proposed pipe excavation does not add or adversely affect any equipment capable of initiating an accident.
There are no new failure modes associated with the pipe excavation.
The possibility of a malfunction of a different type than any evaluated previously in the safety analysis report has not been created since the excavation of diesel oil transfer pipe line I-2"-
DO-13 will not inhibit or otherwise adversely affect the operation of any equipment important to safety. There are no new failure modes for the excavated piping. In addition, this pipe excavation does not create any new modes of operation for any safety related equipment.
The proposed activity does not reduce the margin of safety as defined in the basis for any Technical Specification since the excavated portion of I-2"-DO-13 wi'll not affect system flow rates, set points or testing frequency. The seismic and missile protection are adequate for the excavated portion of the piping and will not adversely impact the operability of the diesel generator system as defined in Technical Specification 3.8.1.1 and 3.8.1.2.
Unit: 1
Title:
Safety Evaluation for Use of Betz Clam-Trol Abstract:
The proposed evaluation involves the use of an alternate biocide to control macrofouling in the Unit 1 saltwater cooling systems (circ water and ICW). The biocide, Betz Clam-Trol (CT-1), will be injected prior to the grizzly rakes into the circulating water stream. One circulating water stream will be treated at a time with CT-1 levels at 10mg/L or less for 12-18 hours. The treatments will be repeated every 2-4 months based on marine growth rate.
Clam-Trol has been analyzed for compatibility with the materials in our seawater systems and found to have no impact for concentrations less than 100 mg/L.
Safety Evaluation:
This evaluation does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety evaluation report since the levels of Clam-Trol will be maintained at less than one-tenth of the level required to assure no adverse impact to the system.
The possibility of an accident or malfunction of a different type than any evaluated previously has not been created because .the addition of CT-1 will not have a deleterious effect on plant equipment. In fact, if CT-1 performs as expected, the plant equipment will be in a better condition to perform its safety function because macrofouling will be reduced. Additionally, injection will not require any modification to installed plant equipment, and all temporary equipment will be placed outside the grizzly rakes.
The margin of safety as defined by the Technical Specifications has not been reduced because there will be no effect on safety related equipment.
Title:
Safety Evaluation for Operation with Diesel Oil Transfer Pump 1A Discharge Isolation Valve I-V17206 Closed Abstract:
This safety evaluation is to document the acceptability of plant operation with the Diesel Oil Transfer Pump (DOTP) 1A discharge isolation valve, valve I-V17206, in the CLOSED position.
Compensatory measures shall be established to open the valve upon operation of the 1A emergency Diesel Generator (EDG). I-V17206 is normally a LOCKED OPEN valve; however, due to a suspected leak in the underground piping downstream of the valve it is desired to isolate the piping until the leak is identified and repairs are made or the line has been replaced.. Isolation of the line will prevent the loss of an estimated 50 gallons per day of diesel fuel oil to the environment.
Valve I-V17206 is located on line I-2"-DO-13 which connects the DOTPs to the 1A EDG Day Tanks. This line is class'ified as safety class three (3), Seismic Class I; therefore, this evaluation is classified as Safety Related.
The purpose of this evaluation is to allow the plant to continue operation with the DOTP lA discharge isolation valve (I-V17206) in the CLOSED position, thus isolating the fuel oil leak to the environment.
Safety Evaluation:
The proposed activity affects only the "A" train EDG fuel oil system. There are no analyzed accidents for which the loss of an EDG is considered to be an initiating event; therefore, there is no increase in the probability of occurrence of an accident previously analyzed in the FSAR.
The. consequences of an accident previously evaluated in the FSAR have not been increased since the performance and operation of the 1A EDG will not be impacted by this change. Additionally,.this change will not create a new path for uncontrolled radioactive releases a'nd will not adversely affect any radiation monitoring equipment or equipment which is relied upon to . mitigate radiological consequences of an accident.
The proposed activity slightly alters the method for initiating fuel flow from the DOSTs to the EDG Day Tanks. Valve I-V17206 a LOCKED OPEN valve that does not require any actuation inis'ormally order to ensure a flow path from the DOSTs to the 1A EDG Day Tanks.
This evaluation allows I-V17206 to be placed in the CLOSED position provided the identified compensatory actions are implemented.
0 Unit: 1
Title:
S afety Evaluation for Operation with Diesel Oil Transfer Pump 1A Discharge Isolation Valve I-V17206 Closed Safety Evaluation (Continued):
These compensatory actions assure the reliability of the EDG fuel oil supply. Additionally, once I-V17206 is opened, the fuel oil transfer system functions as originally designed.
The failure of I-V17206 to open (due to either valve or operator failure) is possible. Such a,failure would result in the loss of the 1A EDG due to fuel starvation after approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of operation. A risk assessment was conducted by FPL's PRA group to determine the change in the reliability of the EDG System following implementation of the specified compensatory actions. Since the EDG System is only required .to perform its safety function following a loss of power to the safety electrical buses, failure of the system were taken in conjunction with a loss of offsite power.
In the proposed configuration, the change in frequency of a station blackout event (i.e. loss of offsite power and loss of both EDGs) was found to be neglible by use of PRA,techniques. The change is
'well within the accuracy of the analysis, considering that generic component failure rates and initiating event frequencies were used and that no recovery of the offsite power or use of the Unit 2 electrical crosstie was postulated. Based on the above, concluded that the probability of occurrence of a it can be malfunction of equipment important to safety previously evaluated in the FSAR has
.not been increased.
The consequences of a malfunction of equipment important to'afety previously evaluated in the FSAR have not been increased since the most limiting failure would result in the loss of a single EDG which is an analyzed event. No- other safety systems or equipment required for accident mitigation or radiation monitoring are impacted.
A failure modes and effects analysis has been performed fo'r the proposed activity. This analysis has identified two potential failures which would result in the failure of,I-V17206 to open.
Failure of the valve to open would, after approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, result in the loss of the 1A EDG due to fuel starvation. The loss of a single EDG is an analyzed event. No other failure modes have been identified for the proposed activity. Based on the above, the possibility of an accident of a different type than previously evaluated in the FSAR does not exist.
As stated in discussions above, a failure modes and effects analysis has been performed for the proposed activity. The analysis has identified two potential failures which would result
Title:
Safety Evaluation for Operation with Diesel Oil Transfer Pump 1A Discharge Isolation Valve I-V17206 Closed Safety Evaluation (Continued):
in the failure of I-V17206 to open. Such a failure would ultimately result in the loss of the 1A EDG, due to fuel starvation.. The loss of a single EDG is an analyzed event. No other failure modes have been identified for the proposed activity.
Additionally, except for the operator action to initially open I-V17206, the operation of the fuel oil transfer system is not impacted. No other systems are affected by the proposed activity.
Based on the above, the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR has not been created.
The proposed activity does not reduce the margin of safety as defined in the basis for, any Technical Specification since the proposed activity does not impact EDG operability (Techni'cal Specification 3.8.1.1 and 3.8.1.2). The compensatory actions required by this evaluation ensure a reliable supply of fuel oil to the day tanks of the 1A EDG and, upon the opening of I-V17206, the fuel oil transfer system will function as designed.
Title:
Safety Evaluation of the Effects of Increased Main Steam Safety Valve (MSSV) Blowdown Abstract:
Overpressure protection for the shell-side of the steam generators and the main steam lines is provided by the Main Steam Safety Valves (MSSVs). One of the design parameters for the MSSV is a blowdown of 44 (FSAR Table 5.5-2). Blowdown is defined as the difference between the set pressure and the resealing pressure of.
a relief valve, expressed as a percentage of the set pressure. IE Information Notice No. 86-05 "Main Steam Safety Valve (MSSV) Test Failures and Ring Setting Adjustments" describes a potential problem regarding MSSV ring settings. MSSVs manufactured by Crosby taken from the Seabrook plant and full flow tested at Wyle Laboratories were not able to achieve the full lift necessary to develop the required steam flow capacity. St. Lucie Units 1 & 2 utilize the Crosby MSSVs and therefore are impacted by the Seabrook/Wyle test results. The Wyle tests established that the original MSSV nozzle and guide rin'g setting on the "as-shipped" valves were significantly different from that required to achieve .
1004 relief capacity. The MSSV "as-shipped" ring (nozzle and quide) settings, which control steam flow capacity and blowdown, were factory set by using extrapolated values of the valve operational test results, conducted at lower prorated set pressures.
These settings could result in an inappropriate relief capacity.
FPL Nuclear Engineering has recommended -full flow testing of all MSSVs at Crosby to establish specific ring settings that will satisfy FSAR requirement of 100: relief capacity. It is expected that adjustments in ring settings could accompany some increase in MSSV blowdown. Increased blowdown changes the steam generator (SG) secondary side pressure and would slightly increase the Steam Generator Tube Rupture (SGTR) flow.
Th).s Safety Evaluation is classified as Safety Related since the Main Steam headers to which the MSSV pipings are connected to are designed per ANSI B31.7, Class 2, 1969 Edition (Per FSAR Figure 10.1-la and Table 10.1-1).. This Safety Evaluation concludes that there are no unreviewed safety que'stions of Tech Specs changes required.
Safety Evaluation:
The probability of occurrence of an accident previously evaluated in the FSAR has not been increased since the excess blowdown will not effect the ability of the MSSVs to perform their intended function nor will it affect any equipment required to mitigate the
Title:
Safety Evaluation of the Effects of Increased Main Steam Safety Valve (MSSV) Blowdown Safety Evaluation (Continued):
H consequences of an accident.
~~
of an accident previously evaluated in the FSAR
~
The cons'equences are not increased by an increase of up to 124 in MSSV blowdown.
Increased MSSV blowdown has insignificant impact on the LOCA and SGTR analyses and radiological releases remain a small fraction of the 10 CFR 100 Limits.
The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR has not been increased because the proposed activity will not result in any hardware changes. The proposed activity does not alter the MSSVs capability to perform its intended safety function.
The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR have not been increased due to increased MSSV blowdown of up to 12% since the radiological releases from LOCA & SGTR events remain a small fraction of the 10 CFR 100 Limits, nor will it prevent the MSSV's from performing it their intended safety function and will not affect any equipment required to mitigate the consequences of an equipment malfunction.
The possibility of an accident of a different type than any previously evaluated in the FSAR has not been created. Increased MSSV blowdown has been evaluated against LOCA transients, SGTR and non-LOCA transients and was found to be acceptable. No new failures are created.
Increased MSSV blowdown does n'ot create the possibility of a malfunction of nuclear safety related equipment of a different type than any previously evaluated in the FSAR. No new failure modes are introduced via this modification since proposed activity did not add or remove any safety related equipment and will not prevent the MSSV's from performing their intended safety function.
Increased MSSV blowdown does not reduce -the margin of 'safety as defined in the basis for any Technical Specification at St. Lucie Unit 1. Effects of increased blowdown is minimal and within the limits of the FSAR Chapter 15 safety analysis.
Unit:
- O
Title:
Safety Evaluation for the Deletion of Submersible Service Design (Rev. 1)
ICW Isolation Valve Abstract:
This safety evaluation provides the basis and associated evaluation for 'emoving the submersible service design requirement (submersible requirements) for certain motor operated isolation valves in the Intake Cooling Water (ICW) system for St. Lucie Unit
- 1. It is consistent with the position described in Section 2.1.4a of the Service Water System operational Performance Inspection Report.
During worst case flooding the subject valves are postulated 'to become submerged. Since they are not designed for submersible service their continued operation can not be assured during such flooding. This evaluation demonstrates that a controlled safe shutdown of the plant, as described in the evaluation, is assure'd with the deletion of submersible requirements and that the required functions of the valves are maintained.
There are no 'changes in design or operating practices or 0 philosophies associated with this evaluation and no restrictions on plant operations are introduced. A FSAR change package is attached to this evaluation which will delete the submersible requirements for these valves. The FCP is scheduled to be incorporated in the Amendment 11 to the PSL Unit 1 FSAR.
Removal of submersible requirements for the subject valves does not involve an unreviewed safety question nor a change to the Plant Technical Specifications. Therefore, prior NRC approval is not required.
Safety Evaluation:
The proposed activity does not increase the probability of occurrence of an accident because deleting the submersible
. requirements from the subject valves does not introduce accident initiating devices. The isolation function of the valves is not required for Safe Shutdown of the unit as described in this evaluation. The valves remain capable of their isolation function during a SSE or Chapter 15 Class 3 Event. Concurrent flooding and a Chapter 15, Class 3 event or an SSE is beyond the design basis of the ICW system.
The proposed activity does not increase the consequences of an accident as the activity does not adversely affect the ability of the ICW system to respond to a postulated accident. This activity will not cause an increase in radiation dose levels during an O accident.
Title:
Safety Evaluation for the Deletion of ICH Isolation Valve Submersible Service Design tRev. 1)
Safety Evaluation (Continued):
The proposed activity does not, increase the probability of occurrence of a malfunction of equipment important to safety because the valves remain capable of the isolation function during an SSE or Chapter 15, Class 3 event. Concurrent flooding and a Chapter 15, Class 3 event or an SSE is beyond the design basis of the ICW system. The isolation function of the valves is not required for Safe Shutdown of the unit as described in this evaluation.
The proposed activity does not increase the consequences of a malfunction of equipment important to safety. The valves .remain capable of their isolation function during a SSE or Chapter 15, Class 3 event. Concurrent flooding and a Chapter 15, Class 3 event.
or an SSE is beyond the design basis of the ICW system. The isolation function of 'the valves is not required for Safe Shutdown of the unit as described in this evaluation.
The proposed activity does not create the possibility of an accident of a different, type than any previously evaluated because" the subject valves are not accident initiating devices. The valves remain capable of their isolation function during a SSE or Chapter 15, Class 3 event. Concurrent flooding and a Chapter 15, Class 3 event or an SSE is beyond the design basis of the ICW system.
The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated because the isolation function of the valves is maintained for a SSE or Chapter. 15, Class 3 event.
Concurrent flooding and a Chapter 15, Class 3 event or an SSE is beyond the design basis of the system. The isolation function of the valves is not required for Safe Shutdown of the Unit as described in this evaluation.
The proposed activity does not reduce the margin of safety as defined in the bases for any .Technical specification because the activity does not affect any safety margins as discussed in the bases of any Technical Specification. The associated Technical Specification basis is, in part, to ensure that the equipment and systems required for Flood Protection are maintained. No margin of safety is affected by this evaluation.
Unit: 2
Title:
AFAS Actuation of Main Feedwater Isolation Valves Abstract:
The St. Lucie Plant - Unit 2 FSAR states in Section 7.3.1.1.8 that the Main Feedwater Isolation Valves (MFIVs), HCV-09-1A, HCV-09-2A and HCV-09-2B are actuated by Auxiliary Feedwater Actuation Signal (AFAS) latching relay outputs. Similarly, the Safety Evaluation (NUREG-0843)'ssued- by the NRC in October 1981, Section 'eport 7.3.3.1, stated that the MFIVs are actuated through latching relays outputs. However, the current design utilizes AFAS cycling relay outputs to actuate closure of, the MFIVs.
This Safety Evaluation addresses the effects of the use of AFAS cycling relay outputs in lieu of latching relay outputs.
As long as the requirements of this safety evaluation are followed there are no adverse effects on plant operation or safety.
Safety Evaluation:
The proposed change in this instance has been defined as the use of the AFAS cycling relays to close the MFIVs in lieu of the AFAS latching relays and has been evaluated as follows:
The -failure modes have been analyzed and the probability of occurrence of an accident evaluated in the FSAR is not increased due to the use of the cycling relays in lieu of the latching relays.. Automatic reset of the AFAS output relays that close the MFIVs and inhibit its re-opening can in no way result in the occurrence of an accident previously analyzed in the FSAR.
The consequences of an accident previously analyzed in the FSAR are not increased due to the use of AFAS cycling relays in lieu of latching relays because the Auxiliary Feedwater System will continue to perform its safe shutdown functions as previously analyzed. Whenever the AFAS actuates, the MFIVs will rapidly close, isolate the non-safety related portion of the feedwater systems to assure that, the auxiliary feedwater is directed into the steam generators. Once the AFAS has reset, the MFIV position is not important 'unless a Main Steam Line Break (MSLB)" or Main Feedwater Line Break (MFLB) occurs. If a MSLB or MFLB, the Main Steam isolation (MSIS) will actuate on low steam generator pressure or high containment pressure to close the MFIVs and.MSIVs.
The probability of occurrence of a malfunction of equipment important to safety previously- evaluated in the FSAR is not increased due to the use of AFAS cycling relays in lieu of latching relays because the reliability of both types of relays is essentially identical, and they are subject to the same periodic
Title:
AFAS Actuation of Main Feedwater Isolation Valves Safety Evaluation (Continued):
testing.
The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR were not increased due to the use of AFAS cycling relays in lieu of latching relays.
The possibility of an accident of a different type than previously evaluated in the FSAR was not created by the use of AFAS cycling
. relays in lieu of latching relays. The consequences of the new failure modes introduced by the use of the cycling relays have been evaluated and cannot create an accident.
The possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR was not created by the use.of the AFAS cycling relays in lieu of latching relays. The failure modes of all equipment previously evaluated in the FSAR with the exception of the MFIVs remain unchanged.
The margin of safety as defined in the basis for any Technical Specification is not reduced as the result of using AFAS cycling relays in lieu of latching relays. The design, requirements and operation of the MFIVs is described in the FSAR Sections 10.4.7.3, 7.3.1.1.5, 7.3.1.1.8, 6.2.1.4.2, 6.2.1.4.1.1 d) and 6.2.4.4 a) and the Limiting Conditions for Operations and Surveillance requirements are specified in Section .3/4.7.1.5 of the Plant Technical Specifications. The capability of the MFIVs to perform their safety function as described in the FSAR and the valve capabilities to satisfy Technical Specifications .will not be diminished.
Title:
Safety Evaluation for the Deletion of ICH Isolation Valve Submersible Service Design (Rev. 0)
Abstract:
This safety evaluation provides the basis and associated evaluation for removing the submersible service design requirement (submersible requirements) for certain motor operated isolation valves in 'the Intake Cooling Water (ICW) system for St. Lucie Unit
- 2. It is consistent with the position described in Section 2.1.4a of the Service Water System operational Performance Inspection Report.
During worst case flooding the subject valves are postulated to become submerged. Since they are not designed for submersible service their continued operation can not be assured during such flooding. This evaluation demonstrates that a controlled safe shutdown of the plant, as described in the evaluation, is assured with the deletion of submersible requirements and that the required functions of the valves are maintained.
There are no changes in design or operating practices or philosophies associated with this evaluation and no restrictions on plant operations are introduced. A FSAR change package is attached.
to this evaluation which will delete the submersible requirements for these valves. The FCP is scheduled to be incorporated in the Amendment 7 to the PSL Unit 2 FSAR.
Removal of submersible requirements for the subject valves does not involve an unreviewed safety question nor a change to the Plant Technical Specifications. Therefore, prior NRC approval is not required.
Safety Evaluation:
The proposed activity does not increase the probability of occurrence of an accident because deleting the submersible requirements from the subject valves does not introduce accident initiating devices. The 'isolation function of the valves is not required for Safe Shutdown of the unit as described in this evaluation. The valves remain capable of their isolation function during a Safe Shutdown Earthquake (SSE) or Chapter 15 Class 3 Event. Concurrent flooding and a Chapter 15, Class 3 event or an SSE is beyond the design basis of the ICW system.
The proposed activity does not increase the consequences of an accident as the activity does not adversely affect the ability of the ICW system to respond to a postulated accident. These activities will not increase any radiation dose levels during an accident.
Unit:
Title:
Safety Evaluation for the Deletion of ICH Isolation Valve Submersible Service Design (Rev. 0)-
Safety Evaluation (Continued):
The proposed activity does not, increase the probability of a malfunction of equipment important to safety of'ccurrence because the valves remain capable of the isolation function during an SSE or Chapter 15, Class 3 event. Concurrent flooding and a Chapter 15, Class 3 event or an SSE is beyond the design basis of the ICW system. The isolation function of the valves is not required for Safe Shutdown of the unit as described in the evaluation.
The proposed activity does not increase the c'onsequences of a malfunction of equipment important to safety. The valves remain capable of their isolation function during a SSE or Chapter 15, Class 3 event. Concurrent flooding and a Chapter 15, Class 3 event or an SSE is beyond the design basis of the ICW system. The isolation function of the valves is not required for Safe Shutdown of the unit as described in this evaluation.
The proposed activity does not create the possibility of an accident of a different type than any previously evaluated because the subject valves are not accident initiating devices. The valves remain capable of their isolation function during a SSE or Chapter 15, Class 3 event. Concurrent flooding and a Chapter 15, Class 3 event or an SSE is beyond the design basis of the ICW system.
The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated because the isolation function of the valves is maintained for a SSE or Chapter 15, Class 3 event.
Concurrent flooding and a Chapter 15, Class 3 event, or an SSE is beyond the design basis of the system. The -isolation function
'alves is not required for Safe Shutdown of the Unit as of'he described in this evaluation.
The proposed activity does not reduce the margin of safety as defined in the bases for any Technical Specification because the activity does not affect any safety margins as discussed in the bases of any Technical Specification. The associated Technical Specification basis is, in part, to ensure that the equipment and systems required for Flood Protection are maintained. No margin of safety is affected by this evaluation.
Title:
, Safety Evaluation for the Installation of Simulated Incore Detector Assembly in Location L13 Abstract:
This safety evaluation is prepared to document the acceptability of the installation of a simulated fixed incore detector assembly (supplied by Combustion Engineering) in location L13 during,St.
Lucie Unit 2 operating cycle 7; The detector which was to be installed in this location was damaged during insertion, and there are no spare incore detector assemblies available. .The simulated detector assembly consists of a seal plug which is materially and dimensionally equivalent to the normal detector, except 'that the seal plug is solid (similar to the plug used for hydrostatic testing), and has a solid rod, which simulates the outer sheath tube, calibration tube, and the instrumentation (thermocouple, rhodium detectors, and background detector). The ICI (Incore Instrumentation) rhodium detectors themselves do not perform a safety related function as demonstrated in FSAR Section 7.7. The ICI detector assembly does perform safety related pressure boundary function, and houses a Core Exit. Thermocouple (CET), which performs a safety related function per FSAR Section 7.5.4.2. The installation of the simulated ICI detector assembly will functionally abandon detector location L13. Technical Specification 3.3.3.2 requires that at least 75~ of all incore detector locations be operable and a minimum of two quadrant symmetric incore detector locations per core quadrant be operable.
Although the installation of the simulated detector will reduce the number of available incore detectors, it will not reduce their numbers below that required by the Technical specifications. The simulated incore detector incore assembly will perform a pressure boundary function identical to the existing detectors and is, although it is not necessary, designed in accordance with ASME Code requirements to provide added assurance of reliability, and as such will not adversely affect the integrity of the Reactorresult Coolant System. The installation of the simulated detector will in the loss of one CET from the B-channel in core quadrant 1, however, the available number of CET's and,inputs to the Qualified Safety
-Parameter Display System (QSPDS) will not be reduced below that which is required by the Technical Specifications. This evaluation concludes that the proposed configuration described herein does not represent an unreviewed safety question and has no impact on plant safety or operations. A review of the plant Technical Specifications and the FSAR has shown that there are no Technical Specification changes involved. This evaluation is valid through the end of Cycle 7 operation.
Safety Evaluation:
The probability of occurrence of an accident previously evaluated
Title:
Safety Evaluation for the Installation of Simulated Incore Detector Assembly in Location L13 Safety -Evaluation (Continued):
in the FSAR has not been increased since the proposed replacement does not adversely affect any accident initiating components. The ICI detectors do not perform any active functions necessary for the safe shutdown of the plant and the proposed replacement does not create any new unmitigated failure modes for any equipment or.
systems capable of initiating an accident. The simulated incore detector assembly will perform a pressure boundary function identical to the existing detectors and is designed in accordance with ASME Code requirements to provide added assurance of reliability. In addition, the seismic integrity of the Incore Instrumentation'system is not affected, by the installation of a simulated detector.'he removal of the ICI detector and subsequent introduction of water moderator will not significantly affect the local power distribution within the fuel assembly residing in Location L13. The removal of this ICI detector will not affect the ability of the ICI system to perform its intended function, i.e.,
measurement of the core power distribution.
The consequences of an accident previously evaluated in the FSAR have not been increased since the proposed replacement does not create a new path for uncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function. The ICI detectors are not required to perform any active safety related functions and the proposed replacement does not adversely impact any equipment which is required to perform a safety related function or initiate actuation of any safety systems. The simulated incore detector assembly is dimensionally equivalent and seals in the same manner as the normal detector assembly. The simulated incore detector assembly will perform a pressure boundary function identical to the existing detectors and is designed in accordance with ASME requirements to provide added assurance of reliability. The installation of the simulated detector will result in the loss of one core exit thermocouple from the B-channel in core quadrant 1, however, the available number of CET's will not be reduced below that which is required by the Technical Specifications.
The probability of 'occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR has not been increased. No new unmitigated failure modes for any equipment important, to safety are introduced by the proposed replacement and no new components-.or equipment are introduced that could adversely interact with any equipment important to safety. In addition, the seismic integrity of the Incore Instrumentation system is not affected by the installation of a simulated detector and the simulated incore detector assembly will perform a pressure boundary
Title:
Safety Evaluation for the Installation of Simulated Incore Detector Assembly in Location L13 Safety Evaluation (Continued):
function identical to the existing detectors and is designed in accordance with ASME Code requirements to provide added assurance of reliability.
The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR have not been increased since the proposed replacement does not create a new path for uncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function. The proposed replacement will not adversely impact any equipment which is required to perform an active safety related function or to initiate actuation of any safety systems.
The installation of the simulated detector will result in the loss of one core exit thermocouple from the B-channel in core quadrant 1g however, the available number of CET's will not be reduced below that which is required by the Technical Specifi.cations. In addition, the simulated incore detector assembly will perform a pressure boundary function identical to the existing detectors and is designed in accordance with ASME requirements to provide added assurance of reliability.
The possibility of an accident of a different type than any evaluated previously in the FSAR has not been created since the proposed replacement does not add or adversely affect any equipment capable of initiating an accident. The proposed replacement does not present any new paths for the loss of reactor coolant system inventory since the simulated detector is dimensionally equivalent to the normal detector and seals on the reactor head in an identical manner. The simulated incore detector assembly will perform a pressure boundary function identical to the existing detectors and is designed in accordance with ASME Code requirements to provide added assurance of reliability. There are no new unmitigated failure modes for the simulated detector assembly. The seismic integrity of the Incore Instrumentation system is not affected by the installation of a simulated detector.
In addition, the detector assemblies are pas'sive measurement devices only, and their failure would not result in the initiation of an accident of a different type. The removal of the ICI detector and subsequent introduction of water moderator will not significantly affect the local power distribution with the -fuel assembly residing in Location L13. The removal of this ICI detector will not affect the ability of the ICI system to perform its intended function, i.e., measurement of the core power distribution.
The possibility of a malfunction of a different type than any evaluated previously in the FSAR has not been created since the
Title:
Safety Evaluation for the Installation of Simulated Incore Detector Assembly in Location L13
. Safety Evaluation (Continued):.
proposed replacement will not inhibit or otherwise adversely affect the=- operation of any equipment important to safety. The ICI detectois are passive measurement devices only, and are not required to perform an active safety related function or activate any safety related systems. The physical interfaces of the detector assembly are not affected by the proposed configuration, and the seismic integrity of the Incore Instrumentation system is not affected by the installation .of a. simulated detector. The simulated 'incore detector assembly will perform a pressure boundary function identical to the existing detectors and is designed in accordance with ASME Code requirements to provide added assurance of reliability. There are no new unmitigated failure modes for the simulated detector assembly. The installation of the simulated detector will result in the loss of one core exit thermocouple from the B-channel in core quadrant 1, however, the available number of CET's will not be reduced below that which is required by Technical Specification 3.3.3.6. In addition, this simulated detector does not create. any new modes of operation for any safety related equipment.
The proposed activity does not reduce the margin of safety as defined in the basis for any Technical Specification since the proposed replacement, of incore detector location L13 will not adversely impact the minimum number of incore detectors required for operation as defined in Technical Specification 3.3.3.2. The simulated incore detector assembly will perform a pressure boundary function identical to the existing detectors and is designed in accordance with ASME Code requirements to provide added assurance of reliability, and as such will not adversely affect the integrity of the Reactor Coolant System. The installation of the simulated detector will result in the loss of one core exit thermocouple from the B-channel in core quadrant 1,. however, the available number of CET's will not be reduced below that which is required by Technical Specification 3.3.3.6.
0 Unit:
Title:
2 Safety Evaluation for 2B CCW Heat Exchanger Flange Gasket Abstract:
The purpose of this evaluation is to address the potentially degraded sealing capability of the tubesheet/channel flange gasket in the inlet and outlet water boxes of the 2B Component Cooling Water (CCW) Heat Exchanger and to justify the use of an Arcor TS-RB joint coating system to enhance the sealing of the joint for Cycle 7 operation.
Repair welding performed to resolve two Non-Conformance Reports
,(NCR's) slightly warped the tubesheet flanges at the 12 0'lock position, which may impact the ability of the existing gasket to seal at the subject flange joints. A Plant/Change Modification (PC/M) provided for coating the tubesheet/channel flange joint with an Arcor S-16/Arc-thane joint coating system to allow for thermal movements and prevent seawater from contacting the carbon steel flange. The Arcor S-16/Arc-Thane joint coating system was replaced during the current 1992 refueling outage with TS-RB per the manufacturer's recommendation. This will enhance the capability in the area of the detected warpage.
This condition will have no impact on plant safety or operation.
A review of the plant Technical Specifications and the FSAR has shown that there are no unreviewed safety questions or Technical Specification changes involved since this condition does not hinder or change the operation of any components or systems.
This revision is provided to address the joint coating material change from Arcor S-16/Arc-Thane to Arcor S-16/TS-RB and to justify this current configuration through Cycle 7 operation. During this revision, an interfacing calculation was identified which was used as the new justification for this evaluation. The conclusions of the original evaluation remain valid.
Safety Evaluation:
The probability of occurrence of an accident previously evaluated in the FSAR has not been increased since this condition does not affect any accident initiating components. The 2B CCW heat exchanger is not an accident initiating component.
The consequences of an accident previously evaluated in the FSAR have not been increased by this condition since this condition does not have a detrimental affect on any equipment required to mitigate the effects of an accident. Failure of the gasket in the warped area on the 2B CCW heat exchanger flange would allow only minor leakage compared to the total ZCW flow rate. The leakage would
Title:
Safety Evaluation for 2B CCW Heat Exchanger Flange Gasket Safety Evaluation (Continued):
still allow sufficient margin for the heat safety related function. Failed'oating exchanger to perform its pieces would be carried through the system and be released into the discharge canal without affecting the operation of the system. Any failed coatings blocking the tubesheet would be identified through tubesheet differential pressure monitoring and pieces becoming lodged in the tubes would be addressed in a manner similar to that currently used for shells.
The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR has not been increased. This condition does not alter the function of any existing components, and thus does not increase the possibility of their failure. The addition of the Arcor system in the 2B CCW heat exchanger tubesheet/channel flange joint provides an enhancement to the existing configuration to ensure the joint is sealed.
The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR have not been increased since this condition does not nave a detrimental effect. on any safety equipment or components.
The possibility of an accident of a different type than evaluated previously in the FSAR has not been created since this condition does not add or affect any equipment capable of initiating an accident. This condition only affects the 2B CCW heat exchanger flange joint.
The possibility of a malfunction of a different type than any evaluated previously in the FSAR has not been created since this condition will not inhibit or otherwise- 'adversely affect the operation of the CCW or ICW systems. The components of the condition are in compliance with the FSAR requirements for the system elements.
This condition does not reduce the margin of safety as defined in the basis for any Technical Specification since this condition is only to a joint configuration on the 2B CCW heat exchanger. No changes are being made to the system design, modes of operation or assumptions in the bases for the Technical Specifications or the FSAR.
Title:
Evaluation of Operation with Motor Operated Valve V-3540 Locked Open to it's throttled position Abstract:
On June 17, 1992, St. Lucie Plant informed Nuclear Engineering (FPL) of the degraded condition of V-3540, a containment isolation
. valve on the Hot Leg Injection line of the High Pressure Safety Injection System in St. Lucie Unit 2. The Hot Leg Injection piping is used for long term cooling for an accident (per Section 6.3.2.2.3 of the FSAR). The High Pressure Safety Injection (HPSI) pumps are manually re-aligned for simultaneous Hot an Cold Leg Injection. This ensures flushing and ultimate sub-cooling of the core coolant independent of break location. The piping configuration is provided in the safety evaluation (Figure 1). V-3540 was undergoing diagnostic testing under the requirements of the NRC Generic Letter 89-10 Program. The diagnostic information showed the torque to thrust conversion was deteriorating, i.e. the resultant stem thrust was successively less for the same torque switch setting. Continued operation of the valve at this time may cause further deterioration. Spare parts are not available to repair the valve at this time, therefore, the change proposed by this safety evaluation is to "lock open to it's throttled position" V-3540 and use the second, redundant valve in the line, V-3550 (Figure 1), as both the containment isolation valve and the Hot Leg isolation valve. The change will place V-3540 in its "safe position" for Alternate Hot Leg and Cold Leg Injection. In addition, the safety function of V-3540 for containment isolation and Hot Leg isolation is transferred to V-3550. Since the quality and level of protection of the original design basis has been maintained, this change has no effect on plant operation or safety.
The proposed change would require a change to plant procedures which normally electrically key lock close the valves. This change has no effect on power operation and does not place any restriction on plant operation. In addition, valve V-3550 is de-energized to provide single failure protection under specific plant scenarios.
Therefore, the re-energization of valve V-3550 requires adequate direction be provided to plant Operating personnel prior to the need for simultaneous Cold Leg-Hot Leg Injection. The proposed change is an adequate measure until the next outage of sufficient duration at which time the valve (V-3540) will be repaired. The proposed change has been reviewed to determine whether an unreviewed safety question exists. Based on the evaluation herein, it has been determined an unreviewed safety question does not exist. The plant Technical Specifications have been reviewed as part of this evaluation and it has been determined the proposed change does not affect the plant Technical Specifications. Based on the above, prior notification of the NRC is not required.
Title:
Evaluation of Operation with Motor Operated Valve V-3540 Locked Open to it's throttled position 8afety Evaluation:
The proposed change does not affect any equipment whose malfunction is postulated in the FSAR to initiate an accident or prevent an accident from occurring. The proposed change maintains the Emergency Core Cooling. Systems ability to perform its intended functions. No physical modifications have been performed to the RCS or connected systems. The only change is a= change'o operational valve line-ups. Appropriate isolation capability is maintained between the RCS and HPSI system such that the -
probability of an Inter-system LOCA event has not increased. As such, the probability of occurrence of'n accident previously evaluated in the FSAR has not been increased.
The proposed change does not diminish in any way the ability of the Safety Injection System and Containment Isolation System to perform their intended function to mitigate the consequences of an accident previously evaluated in the FSAR. The containment isolation function of V-3540 is performed by valve V-3550 which is adequately designed and procured for this function.
The Safety Injection System Alternate Hot Leg-Cold-Leg injection is accomplished through operation of V-3550. In addition, proper isolation between the RCS and the HPSI system has been established, through plant restrictions and/or redundant plant equipment. As such, the proposed change does not increase the consequences of an accident previously evaluated in the FSAR.
The proposed change maintains the quality level and the level of protection previously established for the Safety Injection System.
No physical plant modifications have been performed and no new-equipment is added. The only change is a change to operational valve line-ups. The hot-leg injection path isolation function remains protected against a single credible failure. As'such, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR is not increased by this proposed change.
Safety Injection System operation is not affected by the proposed change. The system will respond in an accident as previously evaluated in the FSAR. The containment isolation function of V-3540 is performed by valve V-3550 which is adequately designed and procured for this function.
The Safety Injection System Alternate Hot Leg-Cold Leg injection is accomplished through operation of V-3550. In addition, proper isolation between the RCS and the HPSI system has been established through plant restrictions and/or redundant plant equipment.
Title:
Evaluation of Operation with Motor Operated Valve V-3540 Locked Open to it's throttled position Safety Evaluation (Continued):
Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR is not increased by the proposed change.
'The proposed change does not introduce any new failure modes. The proposed change is intended to maintain the requirements of the design bases of the Safety Injection System and Containment Isolation System as described.. in .the FSAR. Therefore, the possibility of an accident of a different type than any previously evaluated in the FSAR is not created by this modification.
The proposed change does not interact spatially or functionally with. any structure, system or component important to safety other than the valves and valve operators themselves. No new failure modes are created for the subject MOV's that can be postulated to cause a malfunction of equipment important to safety different than those previously analyzed in the FSAR. The FSAR postulates a single failure to one of the hot-leg injection path MOV's. This configuration will ensure that hot-leg injection isolation remains protected against a single credible failure. Therefore, the possibility aof malfunction of equipment important to safety which is of a different type than any previously evaluated in the FSAR is not created by the proposed change.
The Technical Specification requirements and Technical Specification Bases are not affect'ed by the proposed change. The proposed change does not affect any plant Technical Specification requirement. The proposed. change maintains the quality and level of protection previously evaluated in the FSAR. Therefore, 'the margin of safety as defined in the bases for any Technical Specification is not reduced by this modification.
Title:
Safety Evaluation for the Use of Sealant Injection on Valve I-V09252 Abstract:
The gasket sealing surface on the valve body hinge pin bore of valve V09252 (feedwater supply check valve to 2A steam generator) was weld, repaired in accordance with NCR-2-307. As a contingency in the event, the, valve 'continues to leak at this joint during startup, NCR-2-308 was initiated to request installation of Leak Repairs, Inc. capnuts and wire wrap. The valve is located such that installation of the 'capnuts would be difficult while the leak-is in progress. If leakage is detected during startup, the sealant would be injected through the capnuts into the gasket area.
The purpose, of this evaluation is to evaluate installation of the capnuts and wire wrap, and should leakage occur during startup, provide a method for temporarily repairing the body to hinge pin leak on valve I-V09252. The method of repair will be sealant injection. The valves are scheduled to be permanently repaired during the 1993 refueling outage or outage of sufficient duration.
The use of this. method to repair the subject valves will have no impact on plant safety or operation. A review of the plant Technical specifications and the FSAR has shown that there are no unreviewed safety questions or Technical Specification changes involved.
Safety Evaluation:
The probability of occurrence of an accident previously evaluated in the FSAR has not been increased. Failure of the injection seal is comparable to a gasket failure and is therefore encompassed by the original design bases.
The consequences of an accident previously evaluated in the FSAR have not been increased by this repair. FSAR Section 15.2.5 discusses the large feedwater line break (18" line downstream of
. the check valve). Total .failure of this gasket/sealant would in no way approach this scenario.
The probability of- occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR has not been increased. The valve is required to maintain pressure boundary and to preclude backflow from the steam generators. Its ability to do so will not be affected by this repair since the capnuts perform an identical function as the nuts, -and bolt loadings are not affected by the injection of sealant. The gasket and/or sealant does not perform a safety function. The sealant will be limited to the volume of the gasket area void and therefore, will not adv'ersely
Unit: 2
Title:
Safety Evaluation for the Use of Sealant Injection on Valve I-V09252 Safety Evaluation (Continued):
affect operation of the valve.
The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR have not been increased. The consequences of the failure of the injection seal is the same as the failure of the gasket, which would result in a loss of system fluid into the containment.
The possibility of an accident of a different type than any evaluated previously in the FSAR 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. Thus, no new accident initiators are introduced through this repair.
The possibility of a malfunction of a different type than any evaluated previously in the FSAR has not been created. By limiting the injection of sealant to just interacting outside the valve pressure boundary precludes a malfunction of a different type.
Leakage of sealant into the feedwater system is precluded by limiting sealant injection volume to the volume of the gasket area void, and the method of injection.
Chemistry limits are not altered and no other change is proposed to the plant design, modes of operation or assumptions in the basis for the Technical Specifications or Safety Analysis. Therefore, this repair does not reduce. the margin of safety as defined in the basis for any Technical Specification.
JUMPER/LIFTED LEADS DETERMINED TO BE 10 CFR 50 '9 CANDIDATES
Jumper/Lifted Lead g 15 Unit.: 1 Component and System Affected:
Unit 1 Intake Hose Station Reason for the Request:
This Jumper/Lifted Lead request, is to install equipment necessary for biocide injections in the Unit 1 Intake Cooling Water (ICW) and Circulating Water Systems. The involves the use of an alternate biocide to control proposed'valuation macrofouling in the Unit 1 saltwater cooling systems (Circulating water and ICW). The biocide, Betz Clam-Trol (CT-1), will be injected prior to the grizzly rakes into the circulating water stream. One circulating water stream will be treated at a time with CT-1 levels at 10mg/L or less for 12-18 hours. The treatments will be repeated every 2-4 months based on marine growth rate.
I 'I Clam-Trol has been analyzed for compatibility with the materials in our seawater systems and found to have no impact for concentrations less than 100 mg/L.
Safety Analysis:
This evaluation does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety evaluation report since the levels of Clam-Trol will be maintained at less than one-tenth of the levels required to assure no adverse impact to the system.
The possibility of an accident or malfunction of a different type than any evaluated previously has not been created because the addition of CT-1 will not have a deleterious effect on plant equipment.
expected, the plant equipment will In fact, be in if CT-1 performs as better condition to perform its safety function because macrofouling will be reduced. Additionally, injection will not require any modification to install plant equipment, and all temporary equipment will be placed outside the grizzly rakes.
The margin of safety as defined by the Technical Specifications has not been reduced because there will be no effect on safety related equipment.
Jumper/Lifted Lead g 18 Unit: 1 Component and System Affected:
Hydrogen Dryer Automatic Drain Valve for the Turbine Generator Reason for the Request:
This jumper and temporary piping/valve assembly will be installed to help troubleshoot the reason for the inefficient operation of the Unit 1 Hydrogen Drying System. By installing this jumper and temporary piping assembly, manual control of the Separator Drain Valve will be established, which will identify whether the level control system or the NOV is not operating properly.
Safety Analysis:
The proposed activity does not increase the probability of occurrence of an accident previously evaluated in the FSAR.
The proposed activity does not increase the consequences of an accident previously evaluated in the FSAR.
The proposed activity does not increase the probability of occurrence of a malfunction of equipment important to safety evaluated in the FSAR.
The proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR. The generator is mentioned in the FSAR, but does not describe the Hydrogen Drying System.
The proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the FSAR.
The proposed activity does not. create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated. The generator is mentioned in the FSAR, but does not describe the Hydrogen Drying System.
The proposed activity does'ot reduce the margin of safety as defined in the basis for any Technical Specification. The generator is mentioned in the FSAR, but does not describe the Hydrogen Drying System.
Jumper/Lifted Lead g 21 Unit: 2 Component and System Affected:
The 2A5 and 2B5 Load Centers Reason for the Request:
To maintain power to the Load Centers during the Outage. This Jumper/Lifted Lead request is to allow the installation of a temporary bus tie between 480 Volt Load Centers 2A5 & 2B5 while maintenance is being performed on the 2A3 & 2B3 4160 Volt Switchgear thereby assuring power is available to the refueling equipment.
Safety Analysis:
The PSL Unit 2 FSAR (Sections 8.3.1.1.2, 8.3.1.2.1, 8.3.1.2.2 and 8.3.1.4) discusses the independence of redundant systems.
Per PSL Unit 2 Technical Specification 3/4.8.3 (Onsite Power Distribution) two independent electrical trains are required to be operable in modes 1,2,3, & 4 and only one train is required in modes 5 and 6.
The proposed temporary tie between 480 Volt Load Centers 2A5 and 2B5 will only be installed in modes 5 and 6 when only one train onsite AC power is required. All equipment required to be operable in these modes will be powered from their normal sources.
The temporary double ended tie will only connect one A and one B train 480 Volt Load Center (2A5 and 2B5) together for the purpose of providing power to the refueling equipment while a 4160 Volt bus is de-energized. Safety related breakers will be installed to isolate the bus fed via the temporary tie in the event of a bus or cable fault.
The probability of occurrence or the consequences of a design basis accident or malfunction of equipment important to safety as previously evaluated in the FSAR will not be increased and the possibility for an accident or malfunction of a'different type than any previously evaluated in the FSAR will not be created.
The basis for Technical Specification 3/4.8.3 states that the minimum required power source for shutdown and refueling ensures the facility can be maintained in the shutdown and refueling condition for extended time periods and sufficient instrumentation and control capability is available for monitoring and maintaining the unit status. This temporary
Jumper/Lifted Lead g 21 Safety Analysis (Continued):
change will not,affect the availability of required equipment and Technical Specification 3/4.8.3 would still apply. The margin of safety as defined in the basis for a Technical Specification will not be reduced.
Jumper/Lifted Lead g 41 Unit: 2 Component and System Affected:
Fire System and Screen Wash System Reason for the Request:
The 2B2 Traveling Screen is damaged and the 2B2 Intake Well will have to be dewatered to affect repairs. This dewatering will remove the suction supply to the Screen Nash Pumps and jeopardize the continued operation of Unit 2 differential pressure across the screens increases in the if the other three wells. The jumper will be installed fiom the Fire Water System at Fire Hydrant f25 to= the Screen Nash System, and will be used under the observation of the Assistant Nuclear Plant Operator (ANPO) necessary.
if the screen washing becomes Safety Analysis:
The proposed activity does not increase the probability 'of occurrence of an accident previously evaluated in the FSAR.
The proposed activity does not increase the consequences of an accident previously evaluated in the FSAR.
The proposed activity does not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR.
The, proposed activity does not increase the consequences of malfunction of equipment important to safety previously evaluated in the FSAR. Use of the jumper will reduce the Fire Water System pressure causing the starting of the Fire Pumps, which is the purpose of the Fire Pumps, to maintain system pressure.
The proposed activity does not create the possibility of an accident, of a different type than any previously evaluated in the FSAR.
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. Use of the jumper will reduce the Fire Water System pressure causing the starting of. the Fire Pumps, which is the purpose of the Fire Pumps, to maintain system pressure.
The proposed activity does not reduce the margin of safety as defined in the basis for any Technical Specification because
Jumper/Lifted=Lead g 41 Safety Analysis (Continued):
Fire Pumps and the Fire Water System are not in the Technical Specifications.
ST+ LUCIE UNITS 1 AND 2 MANGROVE PHOTOGRAPHIC SURVEY I
PURSUANT TO TECHNICAL SPECIF CATION 4 ~ 7 ~ 6 ~ 1 ~ 2
Based on the evaluation of the false-color infrared photograph on July 3, 1993, including field evaluations, the condition of the mangroves situated between the intake and discharge canals (Impoundment SE) show indications of an increase in plant growth in several areas especially in the northern half of the impoundment. There is approximately a 224 increase from a year ago increasing the mangrove coverage to 404. The coverage in SE is still below the 1975 baseline condition, however; the mangroves have shown a significant improvement.
This trend of increasing coverage as well as the improvement in the health and vitality of the trees, is expected to continue.
ST LUCIE UNITS 1 AND 2 1992 MAN-REM REPORT PURSUANT TO TECHNICAL SPECIFICATION 6 ~ 9 ~ 1 ~ 5
DATF. 01/I PLANT-TINE- 15:24. REPORT NRRBC2 PAGE 0001 F L 0 R I D A P 0 N E R e L I GHT RADIATION EXPOSURE NONI TORING E ACCESS CON TROL SYSTEN RENACS 1992 tlAN-REN REPORT STANDARD FORNAT FOR REPORTING NUNBER OF PERSONNEL C HAN-REH FOR MORK C JOB FUNCTIONS NUttBER OF PERSONNEL 100 HREti TOTAL NAN / HREH ttORK ANO JOB F UNCT ION STATION UTILITY CONTRACT STATION UTILITY CONTRACT REACTOR OPERATIONS SURV ENGINEERING C
0 0 000 F 000 000 ~ 105 000 '25 HEALTH PHYSICS Z5 13 008 '40 000 F 000 003 '3?
NA INTENANCE 2 000 '20 000 '40 001 F 450 OPERATIONS 35 2
6 015 '16 000 '85 002 '70 SUPERVISOR 0 0 000 F 100 000 F 000 000 F 000 ROUTINE H AINT ENANCE ENG INEERING 0 3 000+000 000 '75 001 F 190 HEALTH PHYSICS 16 49 006 '30 000 F 000 016 '50 HA INTENANCE 116 203 o5o.4es 002 '20 071 ~ 247 OPERATIONS 15 004 '20 001 655 002 ~ 081 SUPERVISOR 1 0 000 '65 000 F 000 000 F 000 INSERVICE INSPECTION ENGINEERING 0 000 F 000 000 '15 000 F 000 HEALTH PHYS ICS 0 000 '20 000 F 000 000 F 000 NA INTENANCE 1 ooo.4e5 000 F 010 000 '70 OPERATIONS 1 000 '90 000 F 000 OOO O95 SUPERVISOR 0 000 F 000 000 F 000 000 F 000 SPECIAL NAINTENANCE ENGINEERING 0 000 F 000 000 '60 ooo.seZ HEALTH PHYSICS 11 3 003 '55 000 F 000 001 '50 NA INTENANCF. 80 121 025 '12 001 '00 048 '27 OPERATIONS 0 42 000 '32 000 '40 012 ~ 545 SUPERV I SOR 0 0 000 '50 000 F 000 000 F 000 HASTE PROCESS ING ENGINEERING 0 000 F 000 000 F 000 000 F 000 HE ALTH PHYS I C5 33 000 '10 000 F 000 009 '80 ttAINTENANCE 0 001 ~ 030 000 F 000 000 '40 OP ERAT IONS 5 000 '25 000 ~ 000 001 ~ 489 000 F 000 SUPERVISOR 0 000 F 000 000 F 000 REFUELING ENGINEERING 0 000 F 000 000 '35 000 '35 HE ALTH PHYS I C5 0 000 '55 000 F 000 000 F 000 ttAINTENANCE 2 001 F 020 000 '25 000 ~ 485-OPERATIONS 2 Dole 225 000 ~ 000 000 '00 SUPERVISOR 0 000 F 010 000 F 000 000 F 000 TOTALS ENGINEERING 0 3 7 000 F 000 001 '90 002 ~ 312 HEALTH PHYSICS 35 0 94 018 '10 000 F 000 031 '17 163 11 317 079 '15 003 '95 122 '19 019 '80 MAINTENANCE OPERATIONS 45 e 60 02'08 002 '80 SUPERVI)OR 1 0 0 000 '25 000 F 000 000 F 000 GRAND TOTALS .
Zo 478 1ZO ~ 558 007 '65 175 '28
CHEMISTRY RESULTS PURSUANT. TO TECHNICAL SPECIFICATION 6 '. 1 5 AND 3 4 '
In accordance with Technical Specification 6.9.1.5 the primary coolant . specific activity did not exceed the limits of Technical Specification 3.4.8.