ML17229A207
| ML17229A207 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 07/27/1996 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17229A206 | List: |
| References | |
| NUDOCS 9701280370 | |
| Download: ML17229A207 (98) | |
Text
ST. LUCIE UNIT 1 DOCKET NUMBER 50-335 CHANGES, TESTS AND EXPERIMENTS MADE AS ALLOWED BY 10 CFR 50.59 FOR THE PERIOD OF DECEMBER 02, 1994 THROUGH JULY 27, 1996 0
970i280370 970%22 PDR ADOCK 05000335 R PDR
INTRODUCTION This report is submitted in accordance with 10 CFR 50.59 (b), which requires that:
I) changes in the facility as described in the SAR ii) changes in procedures as described in the SAR iii) tests and experiments not described in the SAR which are conducted without prior Commission approval be reported to the Commission in accordance with 10 CFR 50.59(b) and 50.71,(e)(4). This report is intended to meet this requirement for the period of December 02, 1994, through July 27, 1996.
This report is divided into three (3) sections; the first, changes to the facility as described in the Updated Final Safety Analysis Report (FSAR) performed by a Plant Change/Modification (PC/M); the second, changes to the facility or procedures as described in the Updated FSAR not performed by a PC/M and tests and experiments not described in the Updated FSAR; the third, a summary of any fuel reload safety evaluations.
TABLE OF CONTENTS 2RRLXQ~
84021 HIGH PRESSURE SODIUM VAPOR LIGHTS FOR RAB 86046 STATION SERVICE UPGRADE FOR SWZTCHYARD 10 93228 PERMANENT REACTOR CAVITY SEAL/SHIELD RING 94062 HEATING COIL REMOVAL FROM INLET SIDE OF HVS-6 & HVS-7 12 94119 DELETION/ABANDONMENT OF THE AUTOMATIC RECIRCULATION FUNCTION FOR CONDENSATE PUMPS 1A, 1B, 6 1C 13 94122 CONTAINMENT SPRAY NaOH INJECTION RECZRC LINE TRAIN SPLIT 14 94124 REACTOR VESSEL LEVEL INDICATING SYSTEM (HJTC)
REPAIRS 15 94145 CHANGE IN OPERATING POSITION OF HPSI HEADER CROSS CONNECT VALVE V3653 16 94146 CIS RADIATION MONITOR DETECTOR REPLACEMENT 94144 RCS FLOW TRANSMITTERS.DAMPENING BOARD ADDITION 18 95017 PRESSURIZER UPPER HEAD INSULATION REPLACEMENT 19 95082 HP TURBINE BLADE RING CHANGEOUT 20 95098 RCP VAPOR SEAL LEAK-OFF LINE MODIFICATIONS 21 95104 REPLACEMENT OF THERMAL RELIEF VALVE V3439 22 95108 MODIFY LETDOWN BACKPRESSURE CONTROL LOOP SETPOINTS AND REDUCE V2345 BLOWDOWN 23 95113 SAFETY RELIEF VALVE V3417 SETPOZNT BLOWDOWN MOD 24 95131 CHARGING PUMP SEAL LUBRICATION TANK HIGH LEVEL ALARM DELETION FROM LZS-02-1A, B, Ec C 25 95236 DISCONNECTION OF ICW/CCW PUMP LOCAL PUSHBUTTON STATION 26 96014 SHUTDOWN COOLING RELIEF VALVE V3483/V3468 SET PRESSURE CHANGE 27 96064 DELETION OF EDG AUTOMATIC START ON CIAS AND CSAS 28
RKXXQ~ (Continued) 96076 INSTALLATION OF PROPORTIONAL AXIAL REGION SIGNAL
, SEPARATION(PARSSEL)DEMONSTRATION INCORE DETECTORS 29 96084 REACTOR VESSEL 0-RING REPLACEMENT 30 96085 REFUELING WATER TANK VORTEX SUPPRESSOR 96027 REMOVAL OF THE POSITION INDICATORS FOR HPSI/LPSI FLOW CONTROL VALVES 32
SECJ-93-003 SAFETY EVALUATION FOR SPECIFICATION SPEC-C-014 GUIDLINES FOR iNSTALLATION AND USE OF RXGGING ATTACHMENTS 34 SECJ-93-011 SPECIFICATION SPEC-C-005, COMPONENT MOUNTING AND SUPPORTS 35 SECJ-93-012 SPECXFICATXON SPEC-C-019 TUBING AND TUBING SUPPORTS 36 SECP-96-053 INSPECTION PROCEDURES AND REPAIR CONTINGENCIES FOR THE RWT BOTTOM LINER 37 SEEP-96-049 480 V SWITCHGEAR CROSS-TZE 38 SEES-95-009 SAFETY EVALUATION FOR OPERATION WITH UP TO TWO CELLS OF THE 1B BATTERY JUMPERED OUT 39 SEES-95-011 SAFETY EVALUATION FOR OPERATION OF A SAFETY BATTERY WITH A SINGLE CELL BEING EQUALIZED BY A SINGLE CELL BATTERY CHARGER 40 SEFJ-94-022 CHANGE TO THE BASES OF TECHNICAL SPECIFICATION FOR STEAM GENERATOR LEVEL-LOW TRIP TO SPECIFY ACCEPTANCE CRITERIA FOR AFW REQUIREMENT 41 SEFJ-95-010 USE OF NEW FUEL HANDLING CRANE FOR MAINTENANCE AND TEST ACTIVITIES 42 SEFJ-95-023 SAFETY EVALUATION FOR 1995 BORAFLEX BLACKNESS TESTING RESULTS 43 SEFJ-96-017 SAFETY EVALUATION FOR NEW PSL1 CYCLE 13 CPC COEFFXCIENTS FOR THERMAL MARGIN/LOW PRESSURE TRIP 44 SEFJ-96-020 REFUELING EQUIPMENT UNDERLOAD AND OVERLOAD SETTINGS 45 SEIS-95-007 OPERATION WITH ONE WIDE RANGE NEUTRON FLUX DETECTOR FISSION CHAMBER DISCONNECTED 46 SEZS-96-028 WIDE RANGE NUCLEAR INSTRUMENTATION TEMPORARY SYSTEM ALTERATION 47 SEIS-96-050 ALTERNATE NIS CONTROL CHANNEL ARRANGEMENT 48 SEMP-92-051 TEMPORARY INSTALLATION OF STRAIN MEASURING DEVICES ON THE PRESSURIZER RELIEF VALVE DISCHARGE PIPING SEMP-94-083 SAFETY EVALUATION FOR REDUCED PRESSURIZER HEATER CAPACITY 50 SEMP-95-112 SAFETY EVALUATION OF INSTALLED WESTINGHOUSE STEAM GENERATOR TUBE PLUGS 51 SEMP-96-052 ZN-SITU HYDROSTATIC TESTING OF STEAM GENERATOR TUBE FLAWS 52 SEMS-95-022 FREEZE SEAL APPLICATION FOR LETDOWN DRAIN VALVE V1233 53 SEMS-96-007 SAFETY EVALUATION FOR THE ADDITION OF THREE MANUAL ISOLATION VALVES WITHIN THE RCGVS PROCESS VENT LINES 54
(Continued)
SEMS-96-020 PCV-18-5 ALTERNATE CONFIGURATION 55 SEMS-96-027 INSTALLATION OF TEMPORARY FIRE PENETRATION SEALS ZN FIRE BARRIER BW064 SEMS-96-031 FREEZE SEAL APPLICATION FOR V3651 & V3652 ON THE 1B SHUTDOWN COOLING RETURN LlNE 57 SEMS-96"034 FREEZE SEAL APPLICATION FOR 1A CONTAINMENT SPRAY HEADER (12-SI-406) UPSTREAM OF V07161 58 SEMS-96-045 INSTALLATION OF A HYDRO-TEST SEAL PLUG INTO ZCI PENETRATION FLANGE NO.7 AT CORE LOCATION R4 59 SENP-94-047 SIT DISCHARGE/LOOP CHECK VALVE STROKE TEST 60 SENP-95-022 REMOVAL OF THE UNIT 1 PRESSURIZER MISSILE SHIELD 61 SENP-95"049 ALTERNATE NIS EXCORE DETECTOR ARRANGEMENT 62 SENP-95-100 LETDOWN PRESSURE CONTROLLER PZC-2201 SET PRESSURE REDUCTION 63 SENP-95-106 JUMPER/LIFTED LEAD FOR PDIS-02-1 64 SENP-95-115 TEMPORARY MODIFICATION ON ANNUNCIATOR Y-15 REACTOR COOLANT VENT PRESSURE HIGH 65 SENP-96-007 POWER PLANT OPERATION AT REDUCED Tc 66 SENP-96-012 SAFETY EVALUATION FOR FSAR AND OPERATING PRACTICE INCONSISTENCIES 67 SENP-96-014 TEMPORARY RELOCATION OF CLASS BREAK ON ZCW 68 SENP-96-016 SAFETY EVALUATION FOR FSAR AND FUEL HANDLING INCONSISTENCIES 69 SENP-96-018 COMPENSATORY MEASURES TO MONITOR SPENT FUEL POOL TEMPERATURE DURING CORE OFF-LOAD WITH HIGH TEMPERATURE ANNUNCIATION OUT-OF-SERVICE 70 SENP-96-021 ELIMINATION OF THE PRESSURE RELIEF FUNCTION FOR RCD-1 Ec 2 71 SENP-96-065 ST. LUCZE UNIT 1 OPERATION WITH INCREASED STEAM GENERATOR TUBE PLUGGING 72 SENS-95-003 TEMPORARY INSTALLATION OF ACOUSTICAL MONITORING EQUIPMENT 73 SENS-95-016 ALTERNATE VALVE POSITION FOR SPRAY HEADER ISOLATION VALVE I-FCV-07-1A 74 SENS-95-027 SAFETY EVALUATION FOR APPLICATION OF VARIED EXTERNAL LOADS ON THE PRESSURIZER SAFETY RELIEF VALVE COMMON DISCHARGE PZPING HEADER 75 SENS-96-003 SAFETY EVALUATION FOR CHEMICAL AND VOLUME CONTROL SYSTEM OPERATION 76
(Continued)
SENS-96-010 USE OF A BLIND FLANGE ON CONTROL ROOM HVA-3A INTAKE DUCT 77 SENS-96-021 AC POWER SOURCE REQUIREMENTS FOR REDUCED INVENTORY CONDITIONS 78 SENS-96-033 SAFETY EVALUATION RELATED TO WASTE GAS SYSTEM OPERATION 79 SENS-96-035 SAFETY EVALUATION FOR UPDATING FSAR DESCRIPTION OF POST-LOCA HOT LEG INJECTION 80 SENS-96-046 SAFETY EVALUATION: USE OF THE STATION BLACKOUT FOR NON-LICENSED BLACKOUT EVENTS 81 SENS-96-060 REVZSZON OF STEAM GENERATOR BLOWDOWN SPENT RESIN TRANSFER METHODS 82 SENS-96-065 SAFETY EVALUATION TO SUPPORT THE ADDITION OF BREAKAWAY LOCKS ON UNIT 1 AND UNIT 2 HOT SHUTDOWN PANEL ROOM DOORS 83 112-295 RELOAD CORE DESIGN OF ST. LUCIE UNIT 2 CYCLE 9 85 7
SECTION 1 PLANT CHANGE / MODIFICATIONS
0 PLANT CHANGE/MODIFICATION 84021 This modification consisted of replacement and as building on a as-failed bases existing incandescent lights with high pressure sodium vapor lights for the Reactor Auxiliary Building. A reevaluation of the original lighting design criteria was performed and determined acceptable use of high pressure sodium vapor lights for the Reactor Auxiliary Building.
The original lighting criteria for St. Lucie Unit 1 prohibited Mercury illumination sources in plant areas where direct radiation exceeded a limit and where situations that might lead to contamination of the reactor coolant. As such, a reevaluation was performed and concludes that the lighting replacements will be in areas that 1) Do not contain primary cooling system or equipment,
- 2) Contain negligible amount of stainless steel subject to mercury contamination upon breakage, 3) Could not return mercury contamination to the primary cooling system upon breakage of e lamp. The lighting system is not a safety related system. The reevaluation determined that the replacement lighting does not constitute an unreviewed safety question. Therefore, prior NRC approval was not required for implementation of this, modification.
PLANT CHANGE/MODIFICATION 86046 ARD This modification provided for the replacement of existing DC load centers with higher capacity load centers equipped with fuse blocks and fuses. This change facilitates proper fuse coordination at the St. Lucie Switchyard. This included routing of a second 13.2kV feeder from Hutchinson 1sland Distribution Substation and the splitting of the station service load between the two normal sources through two new automatic transfer panels. Station step down transformers were replaced to provide a uniform 120/240 volt system. This system is Not Nuclear Safety, and does not adversely impact safety and safe plant operations.
The modifications to the power supplies and distribution systems are switchyard modifications. No safety related systems or equipment are affected by these changes. These modifications improve the reliability of internal switchyard power by providing two redundant sources of power, one from Hutchinson Island Distribution Substation and one from St. Lucie Unit 1 480V Switchgear 1A-1. This modification does not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
10
PLANT CHANGE/MODIFICATION 93228 This modification replaces the temporary reactor cavity seal ring and neutron shield water bags and support steel framing with a permanent reactor cavity seal and neutron shield ring. This modification significantly reduces the man-rem exposure for maintenance in the refueling cavity.
The modification permanently installs a stainless steel seal (water tight during refueling) which spans the annulus of the reactor vessel seal ledge and the refueling cavity floor and includes a borated concrete shield ring. The cavity seal ring has hatches and neutron plugs that allow air flow for ventilation and access into the reactor cavity for removal and installation of the ex-core neutron detectors, loose parts monitoring system transducers, and temporary dosimetry. The design considers clearances for refueling equipment, head installation, and lift rigs for the upper guide structure and core support barrel. The permanent reactor cavity seal and neutron shield ring performs a quality related function, and is classified as a Quality Related seismic design component.
'This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications.
Therefore, prior NRC approval was not required for implementation of this modification.
PLANT CHANGE/MODIFICATION 94062 F
The modification removes a portion of steam supply piping and capping the remaining piping to cut-off the steam supply to HVS-6
& HVS-7. The heating coils for HVS-6 5 HVS-7 are degraded and have been removed. HVS-6 8 HVS-7 supplies outside air ventilation to the fuel handling building (FHB) and is desi.gned to reduce plant personnel exposure by preventing accumulation of airborne radioactivity in the FHB. 'The heating coils allowed the outside air to be heated to provide a60'F air for personnel comfort when the outside temperature is lower during the winter months. No temperature requirements exist for any safety related system, structure, component (SSC) withi.n the FHB to be maintained at a temperature a60'F. The only temperature requirements considered is the minimum temperature for borated water sources, however, since the duration of lower outside temperatures is limited in duration, and the fuel pool is a heat source, an adverse effect is not credible. The FHB ventilation system serviced by HVS-6 8 HVS-7 are not adversely impacted by the removal of the heating coils. The ventilation system used to minimize personnel exposure is not affected by this modification. No safety related function is affected by this modification.
The FHB ventilation system performs no safety related function.
The removal of the heating coils only impacts personnel comfort during periods of time when the outside air temperature is low (during the winter months). This modification has been classified as Not Nuclear Safety, and as such, does not adversely affect any safety related SSC within the FHB. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
12
PLANT CHANGE/MODIFICATION 94119 R AT This modification changes the condensate pump recirculation function from automatic to manual. The modification is required since the flow element for auto recirculation is inoperable and previous attempts to correct/repair the system have been unsuccessful. The inoperability of the automatic function for the recirculation valves has not adversely affected the operability or performance of the condensate system. The automatic function for the recirculation valves was originally designed and installed for operator convenience. Plant operating procedures currently require operators to manually open the recirculation valves prior to starting condensate pumps. The condensate system is classified as Not Nuclear Safety since it does not perform any safety related functions nor is it used to monitor any safety related parameters.
The design bases for the automatic recirculation function for the condensate system is to provide cooling flow through the pumps.
The condensate system preheats the feedwater prior to the feedpump suction. This modification changes how the recirculation valves are operated but does not change the original design function or performance of the condensate system. The condensate system is not nuclear safety and has no direct or indirect impact on any safety functions requir'ed for analyzed accidents and does not increase radiological hazards. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
PLANT CHANGE/MODZFICATXON 94122
'QKERXK This modification provides the independent routing of the containment spray discharge to each sodium. hydroxide eductor. An event occurred which .resulted in the lifting of a relief valve in the system. The source of pressurization for the event was identified and determined to be due to a common header. It was this header that was modified to provide for independent routing of piping to each sodium hydroxide eductor. This modification added a new Train A recirculation line. This modification ensures that a single passive failure upstream of either eductor will not prevent the operation of the NaOH system for both trains of containment spray. The addition and routing of the new piping was determined to have negligible impact on eductor performance, recirc flow, NaOH injection rate, thermal hydraulic or containment spray pump performance. Additionally, the system is classified safety related and as such, the modification was designed and installed to perform the same function with the original design configuration, including piping and supports which were constructed and implemented to the same quality and standards as original. This modification maintained the integrity of the original design and therefore plant safety was not adversely affected.
The design for this modification maintained the same quality and standards as the original design configuration. The independent routing of the containment spray discharge to each sodium hydroxide eductor restored train separation for the containment spray system.
Containment spray system performance and operation remains unchanged by the'odification, including the thermal/hydraulic parameters. Pipe routing, pipe supports, pipe penetrations, including Appendix R fire barriers remain unchanged by this modification. Based on the design, the installation of independent supply lines from the containment spray discharge to the eductors of the spray additive system did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modif ication.
14
PLANT CHANGE/MODIFICATION 94124 D
ggmmnxz This modification justifies replacement of failed HJTC with resistors and disconnects wiring for failed thermocouples. The failed heaters for HJTC are replaced with a resistor to maintain the electrical continuity of the entire probes'eater strings.
Technical Specifications define the acceptance criteria for operation o f the probes . Each channel requires a minimum o f 4 nodes operational. The design of the system far exceeds the minimum requirements to allow for occasional failures due to harsh environmental conditions. Redundant channels minimize the effect of losing indication at an individual node, allowing the operator to adequately make an assessment of the reactor vessel level. The modification is classified as safety related since the HJTCs input QSPDS, allowing the operator to monitor/mitigate the consequences of, losing cooling water in the reactor. These modifications do not render an entire string inoperable and thus replacement of the probe is not necessary. This package documented the failures and allowed for optimal operator information based on the probes current capabilities. These modifications do not adversely affect the system function which is intended for the purpose of emergency response data display (i.e. detect RCS inventory loss).
The design 'bases, original function, and the operation of the reactor vessel level indicating system remain unaffected by the modification to permanently disable inoperable nodes on HJTC probes. The QSPDS continues to provide the control room operator with reactor level data and the probes continue to provide their respective safety function in accordance with the Technical Specification requirements. There is no direct or indirect impact on the analysis of any accident, no increase of any radiological hazard, and the modification does not adversely affect safe plant operation. These modifications did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
PLANT CHANGE/MODIFICATION 94145 This modification changes the normal operating position of HPSI header cross connect valve V3653 from locked open to administratively maintained closed. This modification includes a change to the associated valve annunciator window. This modification is required since the 1C HPSI pump is no longer operationally used and has been abandoned, likewise, V3653 has been abandoned in place. The'losed position provides redundant isolation between the HPSI trains which is an enhancement to the HPSI sytem. The associated annunciator window, R-33, was changed along with the limit switch contacts to accommodate the window change/position change. The modification is classified as safety related since HPSI system is a safety related system and the valve is a Quality Group B component of the system.
Since valve V3653 is an isolation valve in the cross connect header between the A and B HPSI headers and serves to isolate the 1C HPSI pump from train B, and since the 1C HPSI pump is abandoned and not utilized operationally, the change in normal operating position of V3 653 to administratively closed "is determined acceptable. The position change provides redundant isolation between the HPSI trains which is an enhancement to the HPSI system. The change in the annunciator window is required due to the valve position change and addresses a nuisance alarm condition. The modification makes no physical changes to the HPSI system and maintains the quality, standards, and performs the same original design functions to inject borated water into the RCS pressure boundary.
if a break occurs in the reactor The change in normal valve position to administratively closed did not constitute an unreviewed safety question or require changes to the plant Technical:Specifications.
Prior NRC approval was not required for implementation.
16
PLANT CHANGE/MODIFICATION 94146 Qggm~
'his Engineering Package (EP). provides the design necessary to replace the existing high range area radiation detectors used for ESFAS containment isolation signal (CIS). The manufacturer, Victoreen is the supplier of a new model which replaces the old model detector. The replacement detector is a qualified replacement. The replacement detectors provide for the CIS on high containment radiation. This package is classified as safety related since the detector continuously monitors radiation levels and initiates a CIS on high containment radiation to mitigate consequences. Additionally, the radiation monitors are used post a'ccident to monitor radiation levels inside containment.
The new CIS radiation detector is supplied by the original equipment manufacturer. These detectors are certified and meet all the specified performance, accuracy, quality, and functional requirements, of the original design bases, including seismic and environmental qualifications. The technical specification requirements for ESFAS CIS are unaffected by the detector replacements The logi'c and actuation features are not impacted by the replacement detectors, including CIS diversity (containment pressure and SIAS). The detector replacements do not adversely affect the overall reliability, redundancy or diversity assumed in the technical specifications or bases and are determined to be equivalent to the original design. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
17
PLMT CHANGE/MODIFICATION 94144 D AD
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This modification adds dampening cards to the RCS flow transmitters to reduce undesired noise. The RCS low flow circuitry monitors RCS flow and provides a reactor trip if the total flow drops below the trip setpoint. The RCS low flow trip is designed to prevent exceeding DNB limits and is set at 95% of designed reactor coolant flow with four RCPs running. The noise seen by the RCS low flow trip circuitry is due to geometry of the RCS and not related to actual RCS flow. The engineering package is classified as safety related since the modification is, associated with the reactor protection system. The added dampening cards reduce the noise such that it minimizes any compensation for the noise which was required during past flow setpoint adjustments. The flow setpoint is adjusted following each refueling outage. These flow dampening cards are a standard feature for the flow transmitters and are supplied by Rosemount, the transmitter manufacturer. These dampening cards meet the design requirements, quality classifications, and code requirements of the original design. The design bases and function of the flow transmitters are not changed by the addition of the dampening cards.
The design bases and original function for the RCS low flow trip circuitry, part of the reactor protection system, as described in the FSAR remain unaffected by the addition of the dampening cards to the RCS flow transmitters. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
PLANT CHANGE/MODIFICATION 95017 P~gg~
This modification provides for the replacement of the pressurizer upper head area insulation with flexible blanket insulation. This modification improves thermal draft protection for relief valves and allows for easier removal and installation during refueling outages. Additionally, this material will be added to the code safety relief. valve discharge line for approximately three feet to reduce the heat sink. Industry experience with this material has demonstrated that it minimiz'es thermal draft effects at the relief valves. This modification provides a passive, not nuclear safety function to limit heat losses to the containment. However, the engineering package was classified as quality related since the modification evaluated ECCS sump screen blockage from insulation debris, interaction with adjacent safety related equipment during seismic events, and assured that the insulation materials do not compromise RCS boundary components.
The original insulation was a metal (reflective) type which was secured with latches. The material was heavy, did not flex, and required care when handling around safety related valves. Since the metal type insulation was not flexible, thermal drafts potentially affected relief valve reliability. The new material is flexible so as not to allow gaps and air spaces and thus limits thermal drafts. The original design bases, including form, and function are maintained by the flexible blanket insulation.
fit, The material, installation, and function have been reviewed for safety impact (ECCS sump blockage, etc. ) and determined acceptable.
This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications.
Therefore, prior NRC approval was not required for implementation of this modification.
19
PL2QFZ CHANGE/MODIFICATION 95082 This modification provides for the replacement of the HP turbine's 81 & $ 2 blade rings, outer gland seal casing, and associated hardware. These changes are vendor recommended to preclude premature failure of the components due to erosion or erosion-corrosion. The new turbine blades are stainless steel which contains chromium, molybdemun nickel and other trace elements demonstrated to slow or prevent flow assisted erosion. In addition, bolting material was changed to address galling based on the new stainless steel components. The new blade rings are functionally identical and maintain the same quality, standards, and requirements of the original design. Since the new blading design has superior strength -to the original components, the original internal missile analyses remains bounding and not impacted by the replacement turbine blades.
a 0 The replacement of the HP turbine's 01 E 52 blade rings, outer gland seal casing, and associated hardware are considered like for like replacements since the quality, standards, and requirements of the original design, including material compatibility are the same or exceed original design. The turbine blade replacements are
, manufacturer recommended and designed to preclude premature failure of the components due to erosion or erosion-corrosion. There is no direct or indirect impact on the. analysis of any design basis accident, nor an increased potential of any radiological hazards.
Additionally, the potential for an un-analyzed accident has not been increased by this modification. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
20
PLANT CHANGE/MODIFICATION 95098 This engineering package supports and justifies "the rerouting of RCP vapor seal leak-off lines to floor drains. The RCP vapor seal assemblies lacked proper drainage such that boric acid crystals on top of the RCP seals degraded seal flange bolting. The rerouting of RCP vapor seal drainage was covered under a jumper lifted lead.
This engineering package determined that the configuration is acceptable for the remaining fuel cycle until the modification can be made permanent during the extended refueling outage associated with S/G replacement. This engineering package is classified as quality related since the temporary leak-off lines could potentially interact with safety related equipment in the area of the floor drains. Rerouting of the seal leak-off floor drains defeated monitoring of RCP leak-off vialines to the the reactor drain tank, however, seal leakage is still monitored in a similar manner via existing measuring equipment in the reactor cavity sump area where the floor drains are routed. The RCP seal leakage is inconsequential and does not adversely affect the ability to detect RCS leakage from other sources. This configuration was determined acceptable and does not adversely affect safety or safe plant operation.
The RCP vapor seal leak-off lines are Quality Group D, non-seismic lines, classified as not nuclear safety, and were originally routed to the reactor drain tank. The temporary configuration reroutes the RCP vapor seal leak-off to floor drains. This configuration only slightly affects the reactor cavity sump flow indication, however, the requirements for RCS leakage detection systems per technical specifications are not impacted and does not interfere with the operation of the system. Additionally, there are no restrictions on plant operations as a result of this change. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
21
I I PLANT CHANGE/MODIFICATION 95104 343 This modification provides for the replacement and change in design criteria for thermal relief valve V3439. Valve V3439 provides protection against pressure developed due to fluid thermal expansion in an isolable section of the low pressure safety injection system (LPSI) header. The pressure set .point for V3439 was increased, valve blowdown reduced, and the discharge piping size increased. This modi'fication was performed to reduce the potential for (i.e.,
V3439 to lift during shutdown cooling operation LPSI pump starts, a relief which is not due to fluid thermal expansion) and reduce the time for the valve to reseat. The changes associated with V3439 are equivalent in quality, classification, and functionally the same as the original,design.
The LPSI portion of piping protected by V3439 is Quality Group B and ANSI B31.7 Class 2, and classified as safety related. This modification did not change the quality or classification of the system. There are no restrictions on plant operations as a result of this change.
The installation of the replacement valve, increase in relief setpoint, reduction in blowdown settings and increase in discharge piping size did not change any functional requirement for the relief and protection of the LPSI system piping. No accident analyses are adversely affected by this change. Plant safety and safe plant operation are not compromised by the modification. No new failure modes or system interactions are introduced by the replacement of V3439. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
22
PLANT CHANGE/MODIFICATION 95108 D 345 BL This modification provided a lower operating set pressure for the letdown intermediate leg pressure controller PIC-2201, modified the low alarm setpoint for PA-2201, and reduced the blowdown value for relief valve V2345. The purpose for the operating pressure change was to increase the margin between the system normal operating pressure and valve V2345 potential to lift lift set pressure. This minimizes the V2345 during normal plant evolutions. The blowdown value was reduced to increase the margin between the system operating pressure and the pressure at which V2345 will reseat following actuation. The reduced low alarm setpoint for PA-2201 was performed to maintain the original alarm margin to the operating pressure. Historically, the valve lifted during plant alignment changes and on occasion required letdown to be isolated to allow V2345 to reseat. These modifications have been evaluated and determined to maintain the design bases for subcooling within the letdown intermediate pressure piping system and letdown heat exchanger under maximum letdown and minimum charging conditions.
The valve blowdown adjustment has no adverse impact and the reduced low alarm setpoint for PA-2201 maintains the original design margin. The upper alarm setpoint for PA-2201 is maintained as the original design setpoint.
The operating set pressure change for V2345, PA-2201 low alarm setpoint change and the reduced blowdown value change for V2345 maintains the original design bases, quality, and requirements for CVCS letdown. The design basis function to provide over pressure protection by V2345 is maintained. The letdown backpressure controller still provides adequate backpressure to ensure that letdown fluid will not flash under design basis conditions. The setpoint value and the alarm setpoints are suitable for system backpressure control including instrument uncertainty. No new failure mechanisms are introduce by these modifications. Plant safety is not compromised and these changes do not adversely affect safe plant operation. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
23 I
PLANT CHANGE/MODIFICATION 95113 This modification provides an increased pressure set point and a reduced blowdown value for relief valve V3417. Valve V3417 provides protection against pressure developed due to charging pump discharge into an isolable section of the auxiliary high (HPSI system) . The modification was performed. to reduce the pressure'eader potential for V3417 to lift during system testing and off-normal operation and reduce the potential the valve will not reseat at system operating pressures. Charging pump alignment with the auxiliary high pressure header occurs during two different operating scenarios: 1) charging pumps are used to fill or adjust boric acid concentration of the SIT, and 2) should the charging lines be inoperative for any reason, the connection to the safety injection system is used as an alternate charging path. No piping modifications were performed for relief valve V3417 changes.
The pressure setpoint increase and reduction in the blowdown value for valve V3417 maintains the original design bases, quality, and requirements for the HPSI auxiliary high pressure header. The design basis function to provide over pressure protection utilizing V3417 is maintained as the valve lift the design pressure for the piping/header.
set pressure remains within This portion of the HPSI system, protected by V3417, is Quality Group B and ANSI B31.7 Class 2 piping. 'his modification. maintains the same quality and classification as original design bases. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
24 I
PLANT CHANGE/MODIFICATION 95131 I 1 8 This modification deletes the CVCS charging pump plunger packing lubrication system tank high level alarms from LIS-02-1A, B, S C.
This modification enhances plant operation by eliminating a control room nuisance alarm without adversely affecting the operation of the charging pumps (frequent high level alarms would come in with moderate, acceptable levels of leakage). The high and low level alarms which are annunciated in the control room served as a means of detecting primary or secondary plunger seal leakage. However, the tank level is monitored locally during the reactor operator walkdowns, and this level is used to trend charging pump plunger packing performance. The tank low level alarm was not affected by the modification. This evaluation determined that the shif tly operator rounds provided the information needed regarding charging pump packing condition.
Deleting the CVCS charging pump plunger packing lubrication system tank high level alarms from LIS-02-1A, B S C enhanced 'lant operation by eliminating a control room nuisance alarm without adversely affecting the operation of the charging pumps. The CVCS Charging Pump Plunger Packing Seal Lubrication high level alarm serves no safety function since it is not required for safe shutdown and not required for accident mitigation. An open overflow connection to the Waste Management System drain collection header provides a drainage path should the seal lubrication tank overflow. Excessive primary seal leakage can be indirectly detected by an indicated increase in the RCS leak rate or abnormal makeup to the VCT, thus maintaining the original design basis function for monitoring seal leakage. This change does not adversely affect any leakage detection system or impact technical specifications required for leakage detection. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
25 l
PLANT CHANGE/MODIFICATION 95236 This modification disconnects the local push-button control station from the 1A, 1B, and 1C intake cooling water (ICW) pump and the 1A, 1B, and 1C component cooling water (CCW) pump control circuits.
This modification was implemented to delete the operational requirements for manually resetting the RTGB control switch to "stop" and then back to "auto" which may be required to preserve the automatic SIAS start feature of these pumps. Resetting of the control switches was only required after a local push-button stop of a running pump which was originally started from the RTGB control switch. This modification has no impact on the safety related functions of the ICW and CCW control circuits and was implemented to prevent the operator work around associated with the manual reset required after a local push-button stop. There are no adverse affects on safety or plant operations since the local push-button stations are not required during normal, off-normal, emergency or maintenance operations. Response to a LOOP, SIAS, or LOOP/SIAS condition is not affected by this modification.
The design bases for the ICW and CCW systems are not compromised by this modification. There are no adverse affects to any protective function, accident mitigation function, or any function of safety related system, structures, or components. This design provides a procedural enhancement by reducing the manual actions associated with local push-button control station actuation and maintains automatic SIAS start feature. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
26
PLANT CHANGE/MODXFXCATXON 96014 This modification increases the design set pressure for relief valves V3483/V3468. The safety related function for relief valves V3483/V3468 is to protect the associated shutdown cooling (SDC) suction and discharge piping from . overpressure due to the simultaneous running of the charging pumps and the SDC system, with a water solid RCS. The portion of SDC/LPSI system piping protected by V3483/V3468 is Quality Group B and ANSI B31.7 Class 2, safety related. The SDC system/LPSI system original design pressure for suction and discharge piping remains unchanged. The new set pressure is greater than the expected peak pressure and the valve reseat pressure is designed to be greater than the maximum SDC operating pressure. The original design basis function, quality, and requirements are maintained by the set pressure changes. These set pressure changes were implemented to prevent recurring problems with V3483/V3468 lifting and not reseating.
The design basis function to provide over pressure protection by relief valves V3483/V3468 is maintained with the set pressure changes. The SDC system/LPSI system original design pressure for suction and discharge piping remains unchanged. This change only provides additional margin in the relief valve set pressure to prevent lifting during initiation of shutdown cooling and does not adversely affect plant safety or safe plant operation. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modif ication.
27
PLANT CHANGE/MODIFICATION 96064 Guam~
This modification deletes the CIAS and CSAS EDG start and non-essential EDG trip block signals. Testing of the EDGs require that the EDG under test be manually synchronized and loaded in parallel with offsite power. During this test, all.protective trips, such as reverse power, remains operable. Actuation of CIAS or CSAS without SIAS, either manually or by spurious actuation of the relay, results in blocking most of the protective trips and does not open the EDG circuit breaker. This results in the EDG running parallel with offsite power and with most of the protective trips blocked. Should an abnormal condition develop, the EDG would be unable to trip with the re'sultant possibility of damage. This modification eliminates the possibility of running in parallel with offsite power and the protective trips blocked. This change is classified as safety related since of EDGs.
it affects the starting circuits The SIAS automatic start and block of most of the protective trips of the EDGs are not changed by this modification.
The design analyses ensured that there are no adverse effects to any other component or system required for mitigation of a CIAS without SIAS event. The EDGs provide emergency AC power in the event of a loss of offsite power to those components and systems necessary for the safe shutdown of the. plant and accident mitigation.
The design basis for the EDG starting system is to automatically start and load the EDG for. a loss of offsite power by itself or coincident with an ESFAS signal and to automatically start in standby for an ESFAS signal with offsite power available.
will not actuate without a concurrent SIAS. The EDGs start on CSAS the SIAS and would not result in any time delay associated with removal of the CSAS EDG auto-start feature. Similarly, the CIAS actuates on high containment pressure, the same signal and setpoint which actuates SIAS. CIAS also actuates on receipt of a SIAS. The only CIAS which actuates and is not concurrent with a SIAS is "High containment radiation." However, no credit is taken in the analyses for an early start of the EDGs due to receipt of the CIAS prior to the SIAS (Modes 1-4, SBLOCA)'. CIAS is not required in Mode 5 or in Mode 6, CIAS functions only on high containment radiation for a fuel handling accident inside containment.
However, no isolation function occurs which requires EDG auto-start on CIAS. Removal of the CIAS EDG auto-start has no adverse affect on plant safety or safe plant operations. This modification did not'onstitute an unreviewed safety question or require changes to the p1ant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
28
PLANT CHANGE/MODXFXCATXON 96076 This modification provides justification and engineering for the installation of Proportional Axial Region Signal Separation Extended Life (PARSSEL) incore detectors. The PARSSEL fixed incore detector inputs DDPS and the core exit thermocouples input QSPDS used for accident monitoring. As such, this modification was classified safety related. This modification allows up to two PARSSEL assemblies be used to temporarily replace failed fixed incore detector assemblies. The PARSSEL detector is limited to two fuel cycles. The temporary installation is required in order to demonstrate that the mechanical stresses of installation do not result in shortened lifetime of the PARSSEL detectors, determine the power range over which usable signals can be obtained without additional signal processing, demonstrate that PARSSELs provide axial power shape resolution comparable to the fixed incore detectors, and certify operating and performance characteristics.
The PARSSEL detector is precluded from the scan used in power distribution measurement and the linear heat rate alarm function.
The PARSSEL detector is designed equivalent in form, fit, and function. The detector does not impact Technical Specifications or adversely affect safe plant operation.
This temporary modification is designed to demonstrate PARSSEL detector performance. The mechanical design ensures pressure boundary components are equivalent in form, fit, and function.
There were no modifications required for use of the PARSSEL detector. The PARSSEL is precluded from the scan used in power distribution measurement and the linear heat rate alarm function, and the associated core exit thermocouple is out of service.
However, adequate thermocouples exist to meet technical specifications. A PARSSEL detector failure could not adversely affect safety related equipment. The design basis for incore instrumentation system and core exit thermocouple remains unchanged by use of the PARSSEL detector. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
29
PLANT CHANGE/MODIFICATION 96084 This modification replaces the reactor vessel 0-rings with a new spring energized 0-ring design. The reactor vessel vendor has evaluated the replacement 0-ring seal and determined the reactor vessel boundary effects to be acceptable. This modification is classified as safety related. The replacement 0-rings are designed such that installation requires a decreased amount of honing and/or welding required due to any minor surface anomalies. This design therefore reduces the man-rem dose for this work. This design has no adverse impact on reactor vessel seal leakoff detection and
.maintains the design requirements for RCS pressure boundary integrity.
The reactor vessel 0-rings with spring energized 0-ring design have been evaluated and determined .to have acceptable reactor vessel boundary effects. No new failure modes are postulated. The spring energized 0-ring design is proven used successfully in over 75 nuclear installations. The design was evaluated and determined to result in no degradation, either directly or indirectly, to any safety functions required for analyzed accidents, and do not increase any radiological hazards. Implementation of this modification does not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
30
PLANT CHANGE/MODXFXCATXON 96085 This modification installs a vortex suppressor in the refueling water tank (RWT) to prevent possible air ingestion by the ECCS and containment spray (CS) pumps during a design basis safety injection scenario. The vortex suppressor design is a downward facing box elbow fabricated from plate aluminum, compatible with the RWT, and attached to the side of the RWT outlet.. The design increases the effective submergence of the tank outlet while also increasing the area of the effective outlet. This design reduces the vortex potential while ensuring that the top of the tank outlet is not uncovered during the high flow rates of full ECCS/CS actuation.
This. modification is classified as safety related.
The vortex suppressor design ensures that the RWT can be drawn down to the minimum credible level associated with a design basis accident and RAS actuation without entrainment of unacceptable air into the ECCS/CS suction flow stream. The design is passive, utilizes compatible materials, and is robust (considered conservatively high loadings for seismic, conservative operational and emergency hydro-dynamic forces, and a conservative deadweight
.load). This design enhances the margin of safety by preventing against ECCS and CS pump air ingestion. The borated water source, available water volume, and NPSH requirements, for safety related ECCS and CS pumps are not adversely. affected by this modification.
This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications.
Therefore, prior NRC approval was not required for implementation of this modification.
PLANT CHANGE/MODIFICATION 96027 This modification removes HPSI and LPSI flow control valve position indicators. The valve position indicators provided the percent open position for the valves but were not useful in the operation of the valve for flow control. Flow indicators are used to more accurately control flow verses the use of position indication for throttling a valve. These position indications were abandoned in place. Wiring modifications were performed to prevent adverse impact associated with control board equipment. These valve position indicators were not relied upon for safety related activities and did not satisfy Regulatory Guide 1.97 requirements for post-accident monitoring. Indicating lights are used to determine valve position if required and do not rely on any of the position indication circuits. This modification was classified as safety related since the HPSI/LPSI system is relied upon for accident mitigation.
~ ~ ~ ~
~
0 The modification removes HPSI and LPSI flow control valve position indicators. Since the valves essential to the operation of the safety injection system have there "open-closed" position indicated by status lights and/or are alarmed in the control room or locked in the required position, the removal from service of the position indication circuits does not adversely affect plant safety or safe plant operation. Valve position surveillance for each valve in the safety injection flow path that is not locked, sealed, or otherwise secured in position, is verified to be in the correct position.
This verification utilizes the control room indicating lights.
This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications.
Therefore, prior NRC approval was not required for implementation of this modif ication.
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SECTION 2 SAFETY EVALUATIONS 33
SAFETY EVALUATION iTPN-PSL-SECi3'-93-003 REVISION 1 P I This safety evaluation provides generic details for permanent lifting lugs to be used for small loads (e.g., pump motors, valves, pipe spools) and allows development of details for temporary rigging attachments on an as needed basis when plant procedures do not address the desired rigging configurat'ion. This includes fabrication and installation of lifting attachments developed to suit existing field conditions and lifting requirements. New lifting attachments are approved by engineering and temporary lifting attachments are for a one time installation only unless a specification clarification or change is made. This specification is designed to be used in the maintenance process.
This safety evaluation addressed the specification guidance and determined its acceptability to meet technical and licensing requirements. The safety evaluation concluded that the use of the specification does not adversely impact plant safety or safe plant operations. No safety or licensing bases are impacted by use of this specification. The specification provides scope and restrictions for use. The use of lifting attachments as specified in the specification and identified in this safety evaluation did not constitute an unreviewed safety. question or require changes to plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified in this safety evaluation.
34
SAFETY EVALUATION JPN-PSL-SECiT<<93-011 REVISION 1 This safety evaluation provides for generic component mounting instructions and support details. These mounting instructions and support details are used in conjunction with design output documents utilized in the procurement and maintenance process.
This specification is limited to the installation of replacement components weighing less than 50 pounds. New mounting considerations outside the scope of the specification requires engineering approval unless 'a specification clarification or change is made.
This safety evaluation addressed the specification guidance and determined its acceptability to meet technical and licensing requirements. The safety evaluation concluded that the use of the specification does not adversely impact plant safety or safe plant operations. No safety or licensing bases are impacted by use of the specification. The specification provides scope and restrictions for use. The use of mounting instructions provided in the specification and identified in this safety evaluation did not constitute an unreviewed safety question or require changes to plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified in this safety evaluation.
35
SAFETY EVALUATION JPN-PSL-SECaT-93-012 REVISION 1 This safety evaluation provides for generic installation and support details for safety related, quality related, and not nuclear safety tubing. These installation instructions and tubing support details are used in conjunction with design output documents. Additionally, maintenance and repair/replacement activities may be performed 'on tubing supports using the specified standard supports directly covered by the specification, otherwise, engineering approval is required unless a specification clarification is used or specification change is made. This specification also includes installation of test ports on air operated valves (AOVs) on an as-needed basis. These test ports may be installed on instrument air lines when routine maintenance or testing is performed on the valves identified in the maintenance AOV program.
This safety evaluation addressed the specification guidance and determined its acceptability to meet technical and licensing requirements. The safety evaluation concluded that the use of the specification does not adversely impact plant safety or safe plant operations. No safety or licensing bases are, impacted by the use of the specification. The specification provides scope and restrictions for its use. The use of tubing and tubing support instructions provided in the specification and identified in this safety evaluation did not constitute an unreviewed safety question or require changes to plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified in this safety evaluation.
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SAFETY EVALUATXON JPN-PSL-SECP-96-053 REVXSXON 0 D
A repair of a hole to the bottom liner of the refueling water tank (RWT) was required due to corrosion and pitting. The repair was performed under Relief Request 13A and used a reinforced vinyl ester liner in lieu of a Code repair. This repair was approved and remains in place until the Steam Generator replacement refueling outage. At that time, the 'RWT bottom would undergo an ASME Code replacement/repair. This evaluation covers the visual inspection of the coating performed during the refueling outage. To conduct the inspection, the RWT was drained for hands-on inspection of the material and included testing (comparison to controlled specimens) .
This evaluation included acceptance criteria for the inspection results and provided direction for minor repairs to be completed within the allotted time limit the RWT was out of service. The RWT is classified as safety related. This evaluation determined that the inspections and any repairs could not constitute new failure modes.
This safety evaluation evaluated the effects of the RWT liner temporary repair (reinforced vinyl ester liner) on nuclear safety, and determined that the inspections, including temporary scaffolding erected for the inspection, did not adversely impact plant safety or safe plant operation during the refueling outage.
The inspections and tests of the liner material, including any repairs, do not initiate accidents, create malfunctions, adversely affect accident mitigation or any equipment important to safety.
The activities associated with the RWT inspection identified in this safety evaluation did not constitute an unreviewed safety question or require changes to plant Technical Specifications.
Therefore, prior NRC approval was not required for implementation of the actions or changes identified in this safety evaluation.
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I SAFETY EVALUATION JPN-PSL-SEEP-96-049 REVISION 0
-TIE This safety evaluation provided an assessment and bases for operating the A and B trains of safety related 480V electrical system cross connected with one of the 4.16 kV busses (1A3 or 1B3) out of service for required maintenance during Modes 5 or 6. No electrical system and/or equipment capabilities or limits are exceeded in this configuration. The cross connect is accomplished by aligning load centers via the load center 1AB bus and jumpering out the 1AB load center interlocks that prevent cross connecting of the two independent electrical trains under normal operating conditions. This safety evaluation is classified as safety related. While in Modes 5 or 6, one train of electrical power at a time may be deenergized in order to perform maintenance on the busses or EDGs. The technical specification electrical requirements were maintained throughout the use of this temporary electrical configuration. This temporary cross tie was used to power loads required during refueling activities that would have been removed by bus maintenance.
This safety evaluation determined that the cross tie electrical configuration did not adversely affect the safety related equipment, plant'perations, or safety functions. The technical specification for electrical distribution, independent trains, and emergency power were determined acceptable and permitted under the technical specification. This temporary electrical configuration provided enhanced capabilities by powering the "swing" loads. No new failure modes were identified that could adversely affect plant safety or safe plant operations. The temporary electrical system cross connect did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions or conditions identified within this evaluation.
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SAFETY EVALUATXON O'PN-PSL-SEES-95-009 REVXSION 2 This safety evaluation demonstrated that plant operation with up to two of the 60 battery cells (1B battery) removed from service is acceptable and does not adversely affect plant safety or safe plant operation. This ~valuation was generated as a result of battery cell I43 having a voltage problem with voltage less than t'e minimum acceptable level. An equalizing charge was used to bring back the voltage, however, with continued potential voltage problems, the cell could'e jumpered out and/or replaced.
Acceptability of operating the battery with up to two cells removed from service had previously been evaluated and determined acceptable. Battery group 1B provides 125V DC service to group B safety related loads. This evaluation is classified safety related. In addition, this evaluation included the effects for the temporary use of the 1C DC battery bus on 1B and 1A, and provided alternate methods of monitoring the cross tie stability. The 1C battery and 1C DC bus are non-safety. This evaluation also determined acceptability of cross connecting 1C battery to the 1B DC bus (Note that the 1C battery charger is disconnected when in this configuration).
This safety evaluation considered the effect on plant operation with up to two of the 60 battery cells removed from service and /or cross tie of the 1C battery to the 1B DC bus ~ There are no failure modes associated with the jumpering of one or two battery cells.
The cross connect of the non-safety 1C battery to the safety related 1B DC bus worst case failure identified was the loss of one of two redundant DC buses and a plant trip which are previously analyzed events. The 1A bus is not affected by the cross connect configuration. The 1C to the 1B tie is accomplished through the 1AB 125V DC bus. The 1C tie to the 1AB bus is an existing plant design feature, which is implemented only under administrative controls (the breakers on each end of the non-safety to the safety bus tie are key operated). The ability of 1B battery to perform its safety function was not impaired with the cross connect configuration or up to two cells temporarily removed from service.
The temporary electrical system cross connect and/or jumpering of the battery cell did not constitute an unreviewed safety question or require changes to the plant 'Technical Specifications.
Therefore, prior NRC approval was not required for implementation of the temporary electrical configuration identified in this safety evaluation.
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1 J
I SAFETY EVALUATION JPN-PSL-SEES-95-011 REVISION 1 E
AF D BY This safety evaluation demonstrated that plant operation with a single battery cell being equalized by a dedicated battery charger is acceptable and does not adversely affect plant safety or operation. This evaluation was generated as a result of battery cell ¹43 (1B battery) having a voltage problem and voltage less than the minimum acceptable'evel. An equalizing charge was used to bring back the voltage, however, if voltage problems continue to occur upon return of the battery to float charge, a single cell dedicated battery charger was determined acceptable to equalize the cell. Acceptability of operating the battery with up to two cells removed from service had previously been evaluated and determined acceptable. This evaluation maintains the cell available until can be replaced during an outage.
it The ability of the 1B battery to perform its safety function will not be impaired by utilization of the single cell battery charger to recharge the cell. The battery charger configuration is such that it was electrically isolated by fuses and assured that charger failure would not adversely. affect operation of battery 1B. No new failure modes were created by use of the single cell battery charger.
This safety evaluation considered. the effect on plant operation with a single battery cell being equalized by a dedicated battery charger. There are no failure modes associated with the use of the dedicated battery charger. The 1B battery and the bus are not adversely affected by use of the battery charger to maintain cell
¹43 voltage. All equipment powered off this bus are capable of performing there intended safety functions. Safe plant operation is not affected as no equipment is impacted. The only impact is to the 1B battery itself. However, the battery is capable of providing the loads even with up to two cells jumpered out of service. The dedicated battery charger allows cell ¹43 to be maintained as part of the 1B battery. The temporary electrical system arrangement with the use of a dedicated battery charger did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the temporary electrical configuration identified in this safety evaluation.
40
l SAFETY EVALUATION JPN-PSL<<SEFZ-94-022 REVISION 1 E
This safety evaluation documented the bases for modifying the 10 minute margin description for the initiation of auxiliary feedwater specified in the bases of technical specificat.ion for SG low level trip. The change allows credit for the automatic AFW actuation logic present in both St'. Lucie Units. Recent re-analysis determined that specific acceptance criteria needed defining for use in the auxiliary feedwater system evaluation. The change made defines operator action time as based on the accident conditions evaluated and includes the acceptance criteria. SG dry out was not the appropriate criterion for these events. No plant modifications were involved with this change. The existing analyses described in the FSARs for these events meet the acceptance criteria specified in the evaluation. The margin of safety for both units are not reduced by the change.
The bases to the technical specifications for AFW state that there will be sufficient water inventory in the steam generator at the time of trip to provide a margin of at least 10 minutes before AFW is required. This margin is inappropriate for limiting accident conditions which provide for safety. related automatic actuation of AFW. The current FSAR analyses is bounded by the acceptance criteria and therefore do not reduce the margin of safety. No configuration changes were made, and no additional failure modes are created. Plant operating procedures remain unaffected with no new operator actions required. This change credits the safety related automatic feedwater actuation and therefore is bounded by existing analyses'here is no adverse impact on safe plant operation. The margin of safety defined in technical specifications was not reduced. This safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
SAFETY EVALUATION JPN-PSL-SEFJ-95-010 REVISION 0 gg~~
This safety evaluation documented the use of the new fuel handling crane for the purposes of maintenance activity when no new fuel is stored in the new fuel storage area. The use of this crane is considered for such activities as lifting the Blackness Testing calibration cells. The intent of regulatory position 3 of Safety Guide 13 is to prevent damage to any fuel in the fuel handling buildi'ng. This crane cannot physically travel over the spent fuel pool and no new fuel will be stored in the new fuel storage area during maintenance activities, thus no.operation of the crane can lead to fuel handling accidents.. This evaluation documents in the FSAR that the new fuel handling crane can be used for maintenance activities when no fuel is stored in the new fuel storage area.
A The fuel handling system and the associated new fuel handling crane electrical interlocks are not required for the safe operation of the plant or mitigation of consequences. A dropped spent fuel cask is an analyzed limiting fuel handling accident. The transfer of new fuel assemblies between the shipping containers, the new fuel storage racks and the new fuel elevator are performed using the new fuel handling crane. However, when used for. maintenance activities, no fuel can be stored or is stored in the new fuel storage'rea (new fuel handling crane travel is physically limited to the new fuel area and 'cannot pass over the spent fuel pool area). The FSAR states that the crane can be used for RCP seal assemblies under controlled procedural conditions. These limitations apply when there is a potential for damage of new fuel.
However, for the maintenance activities described in this evaluation, no fuel will be in the storage area and therefore cannot cause a fuel damage type accident. The use of the new fuel handling crane for activities described in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
42
SAFETY EVALUATION aTPN-PSL-SEFiT-95-023 REVISION 3 This safety evaluation reviewed the results of blackness testing of fuel storage cells in the spent fuel pool. Blackness testing is used to obtain in-service performance data of Boraflex panels in the storage cells. This revision to the safety evaluation included a review of the fabrication records of the Unit 1 spent fuel pool racks. An inspection revealed irradiation induced Boraflex degradation as expected with the exception of approximately a 15 inch portion of a panel (143 inches long) missing. This evaluation looked at the cause for the missing section, looked at blackness testing results to define future testing requirements/criteria, looked at the effect of the missing boraflex on spent fuel pool operation and safety, and recommended actions that provides justification for spent fuel pool continued operation. The evaluation concluded that the missing panel was an isolated case of a hidden storage rack manufacturing defect not discovered by QA/QC inspections performed by FPL and the manufacturer. No additional plant actions were recommended and no new plant actions or impact on planned blackness testing frequency are required. No adverse safety implications regarding criticality could have resulted from the missing Boraflex panel since the missing portion is considered to be within end shrinkage of a Boraflex panel in a storage rack.
No changes or modification of the spent fuel pool configuration or structures are required. Safety analysis limits remain valid. The only action taken as a result of this evaluation was not to store fuel in the affect storage cells.
This evaluation determined there was no impact on the ability of the spent fuel pool system, structure, or components to meet the design bases function or adversely affect the spent fuel pool boron concentration. The spent fuel pool design features remain intact and are not affected by any of the evaluation recommendations. No equipment important to safe plant operation is adversely impacted and no change in reactivity is experienced since each storage cell is neutronically isolated (i.e., not using a storage cell on a permanent basis is no different from not using any other cell in the spent fuel pool). The plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications.
Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
43 I
SAFETY EVALUATION JPN-PSL-SEFQ'>>96-017 REVISION 0 This safety evaluation modified the core protection calculator (CPC) coefficient of the TM/LP trip function related to the drift allowance presently included in the equipment coefficient calculation. This was necessary due to occurrences of TM/LP pre-trip alarms while the plant was operating at a reduced operating pressure of less than 2250 psia. These hot leg temperature fluctuations which resulted in the pre-trip alarms had previously been observed and evaluated. The CPC coefficients used for TM/LP trip function result in the trip setpoint conservative relative to the technical specification requirement. This change removes the drift allowance from the trip setting and accounts for the corresponding allowance in the safety and setpoint analysis, thus the change of one of the equipment coefficients for CPC TM/LP trip function. With the proposed change, the TM/LP trip continues to remain in compliance with the technical specification. This evaluation was classified as safety related. This change is limited to the remainder of Cycle 13 operation.
The evaluation determined there was no impact on the ability of the reactor protection system TM/LP trip function. The removal of the drift allowance from the CPC coefficient setting required a higher uncertainty be applied to the safety analysis. The incremental uncertainty change, compared to the conservative estimate of available margin, was determined to not lead to any violation of safety acceptance criteria. No other trip function was affected by the change. The safety analysis determined that the margin of safety as defined in the technical specification was not reduced.
This change was limited to the remainder of Cycle 13 operation.
The plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that 'were identified within this evaluation.
SAFETY EVALUATION aTPN-PSL-SEFZ-96-020 REVISION 2 This safety evaluation documented refueling machine and spent fuel handling machine load cell settings. The overload and underload setpoints will provide adequate margin during the refueling operations. The margin will account for resistance while lifting or lowering fuel assemblies, without exceeding the fuel assembly and refueling equipment design limits. The methodology for the
. setpoint calculation was the same used for Unit 2 and per the Vendor technical manual. The changes did not affect any of the design considerations of fuel handling equipment or require physical hardware changes. The load setting changes are well within the loads analyzed in the FSAR. The analyses of the fuel handling accident and cask drop event continue to remain bounding.
The safety functions documented refueling machine and spent fuel handling machine load cell settings. The overload and underload setpoints provide adequate margin during the refueling operations.
The fuel handling equipment and refueling operations remain unaffected by the change. The new overload and underload limit s'ettings are based on the fuel assembly design. The interlocks associated with these limits are set at safe limits, well below fuel assembly damage considerations. The plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
SAFETY EVALUATION JPN-PSL-SEIS-95-007 REVISION 3 D
This safety evaluation documented acceptability of plant operation with a single wide range neutron flux fission chamber in the channel disconnected. The remaining fission chamber in the channel was operable and provided an adequate signal. The output signal from the wide range neutron flux monitors provide power level and rate of change of power information to the reactor protection system (RPS). RPS provides percent power indication, counts per second indication, audible count rate, zero power mode bypass removal, high rate of change of power pre-trip alarm, high rate of change of power trip, and enabling of the high rate of change of power trip. Two detectors are used for the channel, above 1000 cps, only one of the two is in service. The change made by this evaluation provided a design equivalent change for operation at all power levels.
This evaluation concluded that operation of the plant with a single wide range neutron flux fission chamber in the channel disconnected did not impact plant safety. The ability of the modified wide range channel to satisfy its design and regulatory functions are not affected. When operating in the extended range (s1000 cps),
the loss of one of the two channels results in the channel counts less than the remaining channels. When operating at power, (a1000 cps), there is no difference. The manufacturer has determined single detector operation acceptable and operation with one fission chamber is not uncommon. The wide range channel with the non-isolating fission chamber disabled is capable of performing its intended safety functions and considered operable for all Modes, therefore, no adverse impact on any technical specification. This plant configuration did not constitute an unreviewed safety question or require a change to the technical specifications.
Therefore, prior NRC approval was not'required for the conditions that were identified within this evaluation.
46
SAFETY EVALUATION JPN-PSL-SEIS-96-028 REVISION 1 This safety evaluation provided for installation of coaxial jumpers to allow for connection of Wide Range NI Detectors 02(MB) and 04(MD) to the preamplifier input for detector channels 51(MA) and 53(MC), respectively. This modification was necessary due to NI detector replacement during refueling operations. This arrangement meets technical specification for two operable wide range detectors and provides for initial fuel shuffle to be performed at locations adjacent to the detectors. These two detectors monitor sub-critical multiplication. This alternative arrangement was temporary until the other NI channels were installed.
This safety evaluation. provided an alternative wide range NI detector arrangement which substituted NI detectors to allow for continued fuel reload operation. These detectors are class 1E and provide safety related functions. The evaluation addressed train separation, channel independence and separation, and physical separation between redundant channels. The temporary azrangement required restoration of the NI detectors to original design prior to entry into Mode 5. This temporary NI detector arrangement meet technical specification requirements for two operable wide range detectors. This plant configuration did not constitute an unreviewed safety question or require a change to the technical specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
47
SAFETY EVALUATION JPN<<PSL-SEIS-96-050 REVISION 2 This safety evaluation provided for installation of an alternative arrangement for nuclear instrumentation system (NIS) control channels to connect one of the linear power range control channels to both linear power range control channel (CC1 8 CC2) outputs.
This change was necessary since the detector for CC1 was replaced during the outage and its c'ircuitry could not be fully calibrated until 25% power level. This NIS arrangement allows for a continued power level signal to the 1A low power feedwater regulating system.
This also includes signals to the reactor regulating system, the power ratio calculator and recorders JR-010 & 012. For each of these inputs, only one control 'channel was selected. Should a failure occur in the remaining single channel, the control systems could be placed in manual. Manual control is determined acceptable. This NIS detector arrangement is quality related and these control systems are not relied upon for safe shutdown of the plant or to mitigate accidents. No new failure modes are created by this configuration. The alternative, NIS control channel arrangement is limited until CC1 is fully calibrated at 25% power.
Once calibrated, the control channel will be realign to its original design (detector ¹9 to CC1).
This safety evaluation provided for. installation of an alternative arrangement for nuclear instrumentation system (NIS) control channels to connect one of the linear power range control channels to both linear power range control channels (CC1 8 CC2) outputs.
Technical specifications are not affected as axial shape index is not applicable less than 40 percent power. Prior to exceeding 25 percent power the control channel was restored to original design.
These detectors and the alternative NIS detector arrangement do not perform any safety related functions, and therefore can not adversely affect safety or safe plant operation. This plant configuration did not constitute an unreviewed safety question or require a change to the technical specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
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SAFETY EVALUATION JPN-PSL-SEMP-92-051 REVISION 3 This safety evaluation allows the use of temporary strain measuring devices and thermocouples on the pressurizer code safety valves for an additional fuel cycle until removal prior to Mode 2, Cycle 15.
The pressurizer code safety valves have seat leakage problems.
These strain measuring devices are used to monitor the time required for the valves to thermally stabilize during a plant startup (heatup) from cold'o hot zero power. Valve tail pipe loading was considered among potential root causes for safety valve seat leakage. The strain measuring devices provided input to data acquisition equipment. The strain measuring devices and thermocouples are not part of any plant system. This equipment is considered test equipment. This evaluation considered potential interaction with safety related equipment and determined acceptable use of these devices. No new failure modes were created by the configuration or use of the temporary test equipment. This test equipment is non-intrusive to the valves performance and does not adversely affect safety valve operability.
This safety evaluation considered the effects on RCS pressure boundary, piping, and the safety valve design. The test equipment installed is temporary and configured such .that it could not interact with any safety related function or adversely affect valve performance. The plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
49
SAFETY EVALUATION JPN-PSL-SEMP-94-083 REVISION 0 This safety evaluation documented acceptability of plant operation with reduced pressurizer heater capacity of 1375 kW. This permits up to 10 pressurizer heaters, a total of 125 kW, to be removed from service (2 proportional heaters, -25kW, and 8 backup heaters, 100 kW, with a maximum of 2 heaters in any one backup heater bank).
The pressurizer heaters have a total capacity of 1500 kW. Because of design margin, safe plant shutdown and the results of postulated events in the FSAR safety analyses are not adversely affected by operation with reduced heater capacity. This evaluation is classified safety related since the components are required for safe shutdown of the plant. Control of heater capacity will be based on normal plant operation and postulated plant events (FSAR) and will be determined from prior ABB/CE analysis and the heaters design basis. For any reduced pressurizer heater capacity or heater retirement, the facility change and configuration management will be docum'ented using a drawing change request (DCR) process.
The removal of heaters is controlled by technical specifications.
The retirement of heaters can be performed and still meet the minimum requirements for technical specification and the associated design bases for maintaining natural circulation sub-cooling.
This safety evaluation documented acceptability of plant operation with reduced pressurizer heater capacity of 1375 kW. The plant transients have been reviewed and determined acceptable design'ases with a reduced. pressurizer heater capacity. Sufficient margin exists for plant operation with reduced heater capacity. Technical Specifications provides the limits associated with meeting heater capacity requirements for plant conditions. The plant conditions identified in this safety evaluation for reduced pressurizer heater capacity did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
50
SAFETY EVALUATION JPN-PSL-SEMP-95-112 REVISION 1 This safety evaluation documented a time extension for remedial action of Westinghouse thermally treated alloy 600 mechanical steam generator tube plugs to allow continued plant operation until EOC 14 refueling outage. A WCAP identified a service life for the material with remedial actions for potentially leaking plugs. The concern for plug service li'fe was material cracking and potential tube perforation during plug top release. Based on previous tube plug remedial actions and replacement of plugs at the suspect tube locations, all of the remaining plugs are located in tubes not expected to experience perforation during plug top release. Per WCAP recommendation, remedial actions are scheduled over the next five year period beginning with Category 1 plugs during the next schedule outage. However, a technical justification was previously submitted which identified remedial action for only those plugs that potentially had leaked and provided an assessment for continued operation until the scheduled steam generator replacement outage 1997. This evaluation looked at tube and tube plug integrity and the replacement/remedial actions performed and determined the actions acceptable and meets the criteria for the recommended one time deferral of remedial actions.
This safety evaluation determined that a one time extension for remedial action of Westinghouse thermally treated alloy 600 mechanical steam generator tube plugs to allow continued plant operation until EOC 14 refueling outage was acceptable. The evaluation determined that the steam generator tube and tube plug integrity was maintained. The consequences of a postulated SGTR and MSLB are not affected by the tube plugs. A tube top release and tube perforation are bounded by the SGTR event. The actions taken were acceptable and do not adversely affect safety or safe plant operations. The criteria as recommended by Westinghouse for the one time deferral regarding plug remedial actions was met and determined acceptable. The plant conditions identified in this safety evaluation for continued power operation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
'I 51 I
SAFETY EVALUATION JPN-PSL-SEMP-96-052 REVISION 2 This safety evaluation demonstrates that in-situ hydrostatic tes ting to validate examination techniques for flawed s team generator tubes can be performed in accordance with the requirements of 10CFR50.59. Eddy current testing (ETC) is commonly used to detect defects in tubes and determines sizing capability for corrosion indications. However, recent industry experience and NRC criteria have resulted in tighter requirements for tube plugging. With these changes, -FPL has evaluated alternate inspection methods and this evaluation considers demonstrating via high pressure testing that defective tubes can sustain the pressure requirements of the draft Regulatory Guide 1.121 "Basis for Plugging degraded PWR Steam Generator Tubes" without bursting.
This technique demonstrates the adequacy of the existing ECT technique. This evaluation addresses the procedure for testing, contingency actions for a burst tube during testing and includes containment integrity and the actions for a subsequent loss of shutdown cooling. The process, procedures, and results with other plants have been compared and alternative inspection method.
it was determined to be an acceptable This safety evaluation provides justification and develops the test parameters for in-situ pressure testing of flawed steam generator tubes and validates the examination techniques for flawed tubes.
The associated testing is limited to Modes 5 or 6 when the steam generators are out of service. Contingency actions are defined for a tube burst and include containment integrity to isolate the RCS from the secondary steam generator to the containment atmosphere.
Steam generator tube integrity is maintained by this type of testing. This testing is performed prior to tube plugging operation. The plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical'pecifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
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SAFETY EVALUATION JPN-PSL-SEMS-95-022 REVISION 1 33 This safety evaluation demonstrates that a freeze seal is acceptable on the letdown line to isolate V1233 while in Mode 5 with the pressurizer vented to make repairs on a body to bonnet leak. This activity rendered the letdown line out of service.
V1233 is a normally closed drain valve for RCS letdown between the 1B1 RCS loop and letdown isolation valve V2519. valve is the first of two RCS pressure isolation boundary andThisa blind flange downstream of V1233 forms the second isolation boundary. Since this valve is unisolable from the RCS, the freeze seal establishes a temporary isolation of the RCS. The freeze seal is a not nuclear safety device, as such, freeze 'seal integrity was evaluated for this configuration and determined acceptable. The freeze seal process is proceduralized with adequate measures to ensure low probability for freeze seal failure and addresses the unlikely event of leakage/failure. No RCS leakage technical specification Limiting Condition for Operation exist for Mode 5 operation.
Additionally, no credit is assumed in any, accident analyses for Mode 5 which considers the letdown path, and safe shutdown ability does not require letdown flow.
This safety evaluation demonstrates that a freeze seal is acceptable on the letdown line to, isolate V1233 while in Mode 5 with the pressurizer vented to make repairs on a body to bonnet leak. Safe shutdown capabilities, accident mitigation, and RCS integrity is maintained while in Mode 5 with the pressurizer vented and the temporary freeze seal installed. No new failure modes are created by this configuration. Contingency actions are defined for a leak or failure to maintain RCS inventory. The plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
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SAFETY EVALUATION JPN-PSL-SEMS-96-007 REVISION 0 This safety evaluation justifies the installation of manual isolation valves within the reactor coolant gas vent system (RCGVS). These valves were added to enhance maintenance and/or test efforts associated with the primary solenoid-actuated vent valves and secondary solenoid-actuated block valves in the RCGVS.
This evaluation covers both units. The valves are Quality Group B, Seismic Category I and designed to ASME Section III, Class 2. The evaluation was classified as safety related. These valves are maintained as locked-open manual isolation valves on the RCGVS process vent lines. Since these valves do not affect the original design basis and do not adversely impact safe plant operation, the new configuration was determined acceptable. These new valves are backups to the existing line valves and offer additional isolation capability. These valves provide more flexibility, especially at elevated RCS temperatures and pressures should the original valves need replacement/repair. The primary post-accident function to vent the RCS was not adversely affected by the addition of the new manual isolation valves.
This safety evaluation justified the installation of new manual isolation valves within the reactor coolant gas vent system (RCGVS). The original vent valves could be required to function during post-accident conditions. This function is not impacted by the addition of the three new normally locked-open manual isolation valves. New failure modes are not created by this new valve configuration. The plant configuration with the new manual isolation valve maintained in the locked-open position identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications.
Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
SAFETY EVALUATION JPN-PSL-SEMS-96-020 REVISION 0 This safety evaluation justifies the alternate valve configuration by maintaining the valve position for containment instrument air supply, PCV-18-5, full open during conditions of high demand. This was necessary since the valve fluctuations to maintain supply pressure during high usage resulted in large air pressure fluctuations that were undesirable. The highest demand for instrument .air usage is" during outage conditions. This configuration, therefore, applies during outage conditions or when maintenance activities require high demand use of instrument air.
PCV-18-5 was opened to prevent pressure fluctuations and continuous adjustment by the valve. By maintaining the valve open, the air supply, although at a higher pressure than normal, was determined acceptable. This configuration provides air at a higher operating pressure but does not have exhibit large fluctuation in the operating pressure.
PCV-18-5 is a self regulated valve to maintain the containment air supply at the appropriate pressure. If the containment air compressors are unable to maintain pressure, PCV-18-5 opens to supply air to the inside containment instrument air system. As the line pressure drops during high demand, the, valve continuously adjusts to compensate and results in undesirable air pressure fluctuations. Thus, the valve is maintained open during outages.
PCV-18-5 is not nuclear safety and not required for safe shutdown, accident mitigation, or serves any safety function. This evaluation has determined that the new higher pressures maintained by PCV-18-5 open does not have a detrimental affect on the air system, air system operation or performance of any of the end use components, and maintains the requirements of the original design bases. The plant configuration with PCV-18-5 open during outages and maintenance activities which require high demand use on containment instrument air identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
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SAFETY EVALUATION JPN-PSL-SEMS-96-027 REVISION 0 g gym~~
This safety evaluation justifies the installation of temporary fire penetration seals on piping located between the pipe penetration room and the containment purge room, fire barrier BW064. The temporary penetration seals is KBS Sealbags. This product meets a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rating per ASTM E814 and U.L. 1479. A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> fire watch was posted at these locations until the temporary penetration seal was installed. The pipe openings had been installed for modification to the Steam Generator Blowdown System. However, the exhaust system for the pipe penetration room has maintained the required slightly negative pressure and has met the technical specification requirements. When installed, the seals do not represent an operability concern or restrict any Mode of operation.
This installation is passive and installed on passive portions of the steam generator blowdown system. The penetration seals are required by the FSAR since the wall with the piping penetrations is an Appendix R wall.
The installation of temporary fire penetration seals on the identified piping of fire barrier wall BW064 is required to meet Appendix R requirements. The design basis for. the temporary seal is to stop possible air flow between the two fire areas. However, the pipe penetration room has always maintained the slightly negative air pressure and met technical specification requirements.
The temporary configuration meets the three hour fire barrier.
This configuration will remain in place until the next scheduled refueling outage and replaced with a permanent fire penetration seal. The plant configuration with the temporary fire penetration seal identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
56
SAFETY EVALUATION iTPN-PSL-SEMS-96-031 REVXSION 0 This evaluation justifies the temporary application of a freeze seal to the RCS 1B hot leg shutd'own cooling return line while refueling operations are in progress. This activity rendered the "B" shutdown cooling loop out of service. The freeze seal is not nuclear safety and as such, freeze seal integrity was evaluated for this configuration and determined acceptable. The freeze seal process is proceduralized with adequate measures to ensure low probability for freeze seal failure and addresses the unlikely event of leakage/failure. V3651 and V3652 are the primary motor operated valves for the loop'1B shutdown cooling return line which isolate the RCS from the low pressure lines of shutdown cooling.
The purpose of the freeze seal is to maintain the pressure boundary when the RCS is open to the refueling pool during refueling operations. The use of a freeze seal to maintain the pressure boundary during refueling operations is consistent with the use of other non-permanent equipment used during outages in similar operating conditions (e.g. the steam generator nozzle dams, and the RCS hot and cold leg plugs).
This evaluation justifies the temporary application of a freeze seal to the RCS 1B hot leg shutdown cooling return line while refueling operations are in progress. Potential loss of the freeze seal will not result in the loss of inventory below the centerline of the hot legs and still allow cooling water to circulate. Safe shutdown capabilities, accident mitigation, and RCS integrity is maintained while in Mode 6 with the temporary freeze seal installed. No new failure modes are created by this configuration.
Contingency actions are defined for a leak or failure to maintain RCS inventory. The plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
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SAFETY EVALUATION JPN-PSL-SEMS-96-034 REVISION 1 This safety evaluation evaluates the safety significance of applying a freeze seal to the 1A containment spray header with the plant in Mode 5 or Mode 6 with no core alterations in progress.
The freeze seal is not nuclear safety and as such, freeze seal integrity was evaluated for this configuration and determined acceptable. The freeze seal process is proceduralized with adequate measures to ensure low probability for freeze seal failure and addresses the unlikely event of leakage/failure. V07161 is a normally locked open manual isolation valve upstream of FCV-07-1A in the 1A containment spray header. This valve is unisolable from the shutdown cooling loop that isolates the "A" containment spray header from the "A" shutdown cooling header. The purpose of the freeze seal is to maintain the pressure boundary function during refueling operations so that repairs can be made with shutdown cooling in service and operable configuration. This activity is consistent with the use of other non-permanent equipment used during outages in similar operating conditions (e.g. the steam generator nozzle dams, and the RCS hot and cold leg plugs).
This evaluation justifies the temporary application of a freeze seal to maintain the pressure boundary function during refueling operations so that repairs can be made with shutdown cooling in service and in an operable configuration. The containment spray header is not required during refueling outage Modes of operation.
Safe shutdown capabilities, accident mitigation, and RCS integrity is maintained with the temporary freeze seal installed. No new failure modes are created by this 'onfiguration. Contingency actions are defined for a leak or failure to maintain RCS inventory. The plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical 'Specifications'herefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
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. SAFETY EVALUATION aTPN-PSL-SEMS-96-045 REVISION 0 RQH&KD.
This evaluation addresses plant modification 96076 which installed proportional axial region signal separation extended life (PARSSEL) incore detectors'ne detector location, incore instrumentation system guide tube location R4, was not installedwith a detector.
As .such, this safety evaluation justifies the acceptability of installing a hydro-test seal plug for the R4 incore detector location for the duration of Cycle 14. The seal plug will only be installed for the duration of Cycle 14 as the plug will be replaced with a functional incore detector following the cycle. This plug performs the pressure boundary function identical to the existing detectors. RCS integrity is the same and the plug meets all the design requirements associated with RCS integrity. Without the detector in this location the technical specification requirements for incore detectors are met and maintain the 75% available with the desired symmetry. The loss also results in the loss of a core exit thermocouple however, it does not reduce the number below that required by technical specifications. No new failure modes are created by the use of the hydro plug since equivalent to the normal seal plug.
it is designed This safety evaluation justifies the acceptability of installing a hydro-test seal plug for the R4 incore detector location for the duration of Cycle 14. This plug has been designed in accordance with ASME Class 1 requirements and maintains the integrity of the RCS. The available incore detectors or core exit thermocouples are not reduced in numbers, requirements, or symmetry as required by technical specifications and do not adversely affected safe shutdown, accident mitigation, or safe plant operations. The plant conditions identified in this safety evaluation for use of the hydro-test seal plug for the duration of Cycle 14 did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
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SAFETY EVALUATION JPN-PSL-SENP-94-047 REVISION 1 The purpose of this safety evaluation was to demonstrate the acceptability of performing a partial flow, full stroke test of the Safety Injection Tank (SIT) discharge/loop check valves.
Successful performance of the test was used to satisfy NRC requirements for SIT check valves as delineated in Generic Letter 89-04. There are four SITs with two check valves associated with each tank. The check valves tested were V-3215, V-3225, V-3235, V-3245, V-3217, V-3227, V-3237, and V-3247. Partial flow, full stroke testing supersedes disassembly and inspection as a method to demonstrate acceptability. Plant restrictions applicable to the test were identified and listed:
This safety evaluation addressed the effect of the test on safety related components including the fuel, steam generator nozzle dams, refueling water cavity clarity, and induced cyclical thermal stresses in the SIT. The actions or plant changes in procedures identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified within this evaluation.
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SAFETY EVALUATION iTPN-PSL<<SENP-95-022 REVXSION 0 E E This safety evaluation documented acceptability of permanently removing the St. Lucie Unit 1 pressurizer missile shield roof. The purpose of the pressurizer missile shield roof is to prevent credible pressurizer missiles from penetrating and puncturing the containment vessel. The containment vessel is a 2" thick steel right hand circular cylinder with a 1" thick steel hemispherical dome. The evaluation concluded that there does not exist any credible missile which require the pressurizer missile shield roof to protect the containment vessel. The plant benefits identified with removal of the pressurizer missile shield roof include the reduction in risk from dropping a heavy load, reduction in thermal aging effects on components located in the pressurizer cubicle and it facilitates inspection activities while the unit is at normal operating pressure and temperature.
The safety evaluation concluded that the removal of the pressurizer missile shield roof does not adversely affect safe operation of the plant and does not constitute an unreviewed safety question and does not require a change to the Technical Specifications.
Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.
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J SAFETY EVALUATXON 8PN-PSL-SENP-95-049 REVXSXON 1 This safety evaluation provides an assessment of an alternative NIS excore detector arrangement which uses linear range control channel CC2 in place of a failed linear range detector. This restores channel D to operable status and the reactor trip logic from 1/3 to 2/4 coincidence. The detectors are the same design and use the same cables. The change associated with this evaluation considers the geometry/symmetry effects associated with the detector locations. These changes were determined acceptable for the Modes of operation. The loss of a single control channel was evaluated and determined acceptable and bound by technical specifications and the FSAR. The options presented in this evaluation did not adversely affect safe plant operations or accident mitigation functions'his safety evaluation provides an assessment of an alternative NIS excore detector arrangement which uses linear range control channel CC2 in place of a failed linear range detector. This restored linear range detector channel D to operable status and the reactor trip logic from 1/3 to the normal 2/4 coincidence. The plant systems affected by the loss of the control channel are not nuclear safety systems. This evaluation applied until the NIS detectors were replaced during the refueling outage. Based on the design of the alternative NIS detector arrangement, the use of linear range control channel CC2 in place of failed linear range channel D did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.
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SAFETY EVALUATION JPN-PSL-SENP-95-100 REVISION 0
-2 1 E P D This safety evaluation demonstrated the acceptability of operating with a reduced set pressure as low as 360 psig for Letdown Backpressure Controller PIC-2201 under controlled conditions going from Mode 3 to cold shutdown. The normal operating set pressure is 450 psig with the controller in the automatic mode and the unit at 100% power. The reduced operating set pressure was necessary to provide additional margin between operation and V-2345 relief valve lift set pressure. V-2345 provides over pressure protection for the piping section between LCV-2110 PSQ and PCV-2201 PEQ. PIC-2201 would have been restored to the normal operating set pressure (450 psig) should a plant transient occur.
This safety evaluation~considered the effect on the operation of the Chemical Volume and Control System with a reduced set pressure on PIC-2201 under controlled conditions. It was determined that there were no adverse effects on plant safety under the conditions described in the safety evaluation. The design basis requirements of the CVCS were not impacted by the reduced set pressure operation. Over pressure protection was maintained. The temporary reduced set pressure provided a significant assurance that no flashing in the Letdown line would occur. Flashing was evaluated however, and determined to have inconsequential impact due to the operating restrictions at the reduced set pressure. Therefore, the conditions or changes in procedures and design documents addressed under this evaluation did not involve an unreviewed safety question or require a change in plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified in this safety evaluation.
SAFETY EVALUATION JPN-PSL-SENP-95-106 REVISION 0 D
This safety evaluation demonstrated the acceptability of applying a jumper to the pressure differential switch (PDIS-02-1) in the chemical & volume control system (CVCS) letdown line. This switch measures the differential pressure across the regenerative heat exchanger as a means of detecting high fluid flow that would occur from a downstream line break. The jumper of this switch defeats closure of a letdown isolation valve on high differential pressure.
Without the differential pressure switch, letdown isolation occurs from a temperature element (TE-2221) located immediately still downstream of the regenerative heat exchanger. TE-2221 senses high temperature (470'F) downstream of th'e regenerative heat exchanger which is indicative of a line break and is credited in the FSAR for isolating letdown by closing V2515.
This safety evaluation addressed the technical and licensing requirements for the jumpering of pressure switch PDIS-02-1 and concluded that the proposed plant configuration and mode of operation was bounded by the Technical Specifications and did not change the analysis of accidents addressed in the FSAR or the results and conclusions of any previous safety evaluations. The actions or changes identified and evaluated in this safety evaluation did not have any adverse. effect on plant safety or plant operations. The actions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified in this safety evaluation.
SAFETY EVALUATION JPN-PSL-SENP-95-115 REVISION 0 B~iRRX~
This safety evaluation justifies changing the indication logic of annunciator Y-15, reactor coolant vent pressure high. The current logic causes an alarm on high pressure, indicating leakage past valves V-1441, V-1442, V-1443, or V-1444. These valves are part of the primary reactor coolant gas vent system (RCGVS). Changing the logic as proposed, results in an alarm on low pressure if leakage past the second set of valves downstream (V-1445, V-1446, or V-1449) is greater that the leakage past valves V-1441, V-1442, V-1443, or V-1444. This new logic ensures an alarm as an indicator of leakage past one of the second set of valves downstream.
Leakage detection of the RCGVS is required. This method enhances leakage detection in combination with quench tank level and containment sump level. The features of this method ensure that RCGVS leakage past the second set of valves will continue to be identified and is unchanged by modification of the annunciator logic. No plant restrictions were identified with the new annunciator logic. Safe plant operation, accident mitigation, and plant safety are not adversely impacted by this temporary annunciator logic change. This temporary change remains in effect through Cycle 13 until leaking valves are repaired or replaced.
The logic for the annunciator would be restored to the original design following the outage.
evaluation justifies changing the indication logic of changers This safety annunciator Y-15, reactor coolant vent pressure high, from high pressure to a low pressure alarm. This logic change was determined acceptable since all of the features and requirements for ensuring RCGVS leakage is detected. No new failure modes were created by the annunciator logic change. No plant restrictions are associated with the Therefore, the conditions or changes under this evaluation did not involve an unreviewed safety question or require a change in plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SAFETY EVALUATION JPN-PSL-SENP-96-007 REVISION 0 E
This safety evaluation demonstrated acceptability of full power operation at a reduced nominal T~ range of between 548'F to 549'F.
With T~ at a reduced nominal value of 548'F, increased operational margin is provided for the limit on the technical specification DNB related parameter., This evaluation considered the operational and design considerations associated with reduced T~. The effects of the approximate 1'F droop in Tgyg on other plant systems have been evaluated and determined to be acceptable. No protective functions require change. Plant protection and control were operated in the same manner. Operation at reduced T~ is limited until the end of each unit's current fuel cycle.
This safety evaluation demonstrated that safe plant operations is maintained with a reduced T~. This action results in increased margin to the DNB technical specification limit for T~. This evaluation determined that no safety limits will be violated as 'a result of operation with a 1'F droop in T>>~. No fuel parameter limits are exceeded and FSAR accident analyses results remain bounding. This change results in a benefit for DNB related events.
Additionally, plant operation at a reduced T, is bounded by the current setpoint analysis. Therefore, the conditions for safe plant opertion at a reduced nominal T, range of between 548 F to 549'F did not involve an unreviewed safety question or require a change in plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
66
SAFETY EVALUATION iTPN-PSL-SENP-96-012 REVISION 0 This safety evaluation documents FSAR changes required as a result of review of operating procedures for plant heatup, cooldown, and startup procedures. This evaluation provided the bases for revisions to correct the FSAR. This evaluation was classified as safety related since safety related systems were involved.
Operating procedure changes were not required as a result of Each of the FSAR inconsistencies identified were this'valuation.
documented in a Condition Report and this evaluation provided resolution. This evaluation determined safety condition existed and if if an adverse operating or changes to plant procedures were warranted.
This evaluation determined that the identified FSAR inconsistencies do not impact safe operation of the plant, warrant changes to plant procedures, or adversely impact accident mitigation. The conditions identified in the evaluation and the changes to the FSAR identified in this evaluation did not involve an unreviewed safety question or require a change in plant Technical Specifications.
Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SAFETY EVALUATION JPN-PSL-SENP-96-014 REVISION 0 This safety evaluation demonstrated the acceptability for a temporary system alignment for the intake cooling water (ICW) system. During a walk down, a flow restricting orifice, S0-21-5A, was not installed per plant modification PCM 341-192. The flow restricting orifice was intended to be the quality group boundary between Class C and D piping. As a temporary corrective measure, the downstream valve, SH-21001, was closed to protect the integrity of the ICW flow paths and represented the quality group boundary.
This configuration was temporary until the flow restricting orifice SO-21-5A was installed and the piping quality group boundary returned to its original design configuration/location.
0 This evaluation justified the temporary system alteration related to the ICW quality group boundary. Valve SH-21001, located approximately five feet downstream of .the intended boundary, was closed to reestablish the quality group boundary, and protect the integrity of ICW flow paths. The flow orifice functioned to provide lubricating water to the non-safety related circulating water pumps. With the valve closed, lubricating water was available from the "B" side ICW and the alternate city water supply via PCV-21-26. This evaluation determined that the safety related function of the temporary quality group boundary met all code and regulatory requirements. The non-safety related function for lubricating water was also met in this temporary configuration.
Safe plant operation was maintained and the temporary configuration did not adversely affect safety or accident mitigation. Therefore, the condition for safe plant operation with the temporary system alteration related to ICW quality group boundary did not involve an unreviewed safety question or require a change in plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SAFETY EVALUATION aTPN-PSL-SENP-96-016 REVISION 0 This safety evaluation documents FSAR changes required as a result of review of operating procedures for refueling operations. This evaluation provided the bases for resolving the inconsistencies and correcting the FSAR. This evaluation was classified as safety related since safety related systems were involved. Operating procedure changes were not required as a result of this evaluation.
Each of the FSAR inconsistencies identified were documented in a Condition Report and this evaluation provided resolution. This evaluation determined existed and if if an adverse operating or safety condition changes to plant procedures were warranted.
This safety evaluation demonstrated that the identified FSAR inconsistencies associated with refueling operations do not impact safe operation of the plant, warrant changes to plant procedures, or adversely impact accident mitigation. The conditions identified in the evaluation and the changes to the FSAR identified in this evaluation did not involve an unreviewed safety question or require a change in plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SAFETY EVALUATION JPN-PSL-SENP-96-018 Qe REVISION 1 B~QHlXK!.
This safety evaluation justified acceptability for providing contingency measures for an out-of-service spent fuel pool high temperature alarm. Refueling operations were in progress and required monitoring of the spent fuel pool to determine adequate cooling. Guidance provided in the reload plant change modification PCM 054-196 contained guidelines for ensuring adequate spent fuel pool cooling. Although the PCM required this alarm circuit to be operable, it also states that deviations from the guidance provided will be acceptable if appropriate compensatory measures were taken.
This evaluation documented and permits periodic monitoring by operations personnel in the event that circuit providing control room annunciationthewas temperature alarm determined to be out-of-service. The operator function to monitor was determined to fulfill the system design bases requirements. Spent fuel pool cooling is not safety related. However, the spent fuel pool is designed seismic category I, but the instrumentation is not seismic required. Sufficient time to establish cooling or a source of makeup is credited in accident analysis.
This safety evaluation justified acceptablity for providing contingency measures for an out-of-service spent fuel pool high temperature alarm. Refueling operations were in progress. This evaluation provided the basis for the frequency requirements for monitoring of the spent fuel pool and was based on the results from a conservatively performed analysis. Significant margin existed between the actual spent fuel pool temperatures and the analysis.
No changes were made to the fuel pool cooling system and no new failure modes were created by the actions taken. The conditions identified in this evaluation did not involve an unreviewed safety question or require a change in plant Technical Specifications.
Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SAFETY EVALUATION JPN-PSL-SENP-96-021 REVXSION 0 D-This safety evaluation demonstrated the acceptability of eliminating the pressure relief function of reactor cavity dampers RCD-1 6 2. The pressure relief dampers were originally designed to open du'ring LOCA events to vent mass and energy released to the cavity area from RCS pipe breaks in the annular space between the reactor vessel and the primary shield wall. The safety related vent function, however, is eliminated by this evaluation from the damper design basis by taking credit for the leak-be fore-break qualities of the RCS main loop piping. During normal operation, the reactor cavity pressure relief dampers are maintained closed to aid in maintaining proper reactor cavity ventilation flow evaluation justified changes to the FSAR. paths'his This safety evaluation demonstrated the acceptability of eliminating the pressure relief function of reactor cavity dampers RCD-1 8 2. The original design function to relieve the pressure associated with subcompartment pressurization following a LOCA is no longer required to be part of the plant design basis. The vent function of the reactor cavity pressure relief dampers is eliminated and determined to be acceptable. This change did not adversely affect plant 'safety, accident mitigation, or safe shutdown. No new failure modes are created by maintianing the dampers in the closed position which is the normal operating position for the purpose of proper reactor cavity ventilation. The conditions identified in this evaluation related to the reactor cavity dampers, RCD-1 R 2, did not involve an unreviewed safety question or require a change in plant Technical Specifications.
Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SAFETY EVALUATION iXPN-PSL-SENP-96-065 REVISION 2 This safety evaluation provided an assessment for plant operation with increased steam generator tube plugs. Increasing the steam generator tube plugging level from 11.2% of cycle 13 to a cycle 14 value of approximately 23.4% resulted in certain operational impacts including a reduction in steam generator pressure and an increased core aT. Plant operation and plant response to off-normal conditions were reviewed and determined to be acceptable.
The increased level of tube plugging results in no adverse impact to the safety of Unit 1 operation or its ability to respond to transient or accident conditions. The NRC approved the license amendment for a reduction in the required value of the RCS flowrate from 355,000 gpm to 345,000 gpm and a reduction in the RCS low flow trip setpoint. These changes provided additional operational margin. Plant operation under these conditions will remain in effect until the steam generator replacement outage.
This safety evaluation provided an assessment for plant operation with increased steam generator tube plugs and determined that power operation is acceptable. The operational impact of reduced steam generator pressure and increased core aT did not adversely impact plant safety, accident analysis, accident mitigation, or safe shutdown. The NRC approved the license amendment for reduced RCS flow and a reduced RCS low flow setpoint which provided additional operational margin. The conditions identified in this evaluation did not involve an unreviewed safety question or require a change in plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SAFETY EVALUATION JPN-PSL-SENS-95-003 REVISION 1
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This safety evaluation demonstrated the acceptability of applying temporary acoustical monitoring equipment used in an effort to determine the source of increasing leakage into the quench tank.
This evaluation provides justification for mounting these instruments to the pressurizer upper head and various safety related and non-'safety valves and adjacent piping during power operations. These acoustical monitoring devices can remain in place following leak repair until the end of the current operating fuel cycle. This action prevents having the equipment removed during "Hot" at power conditions. The acoustical devices have been determined not to interact with the host components. These devices
, do not adversely impact the operation, integrity, or qualification of any of the host components. The installation and wiring are not mounted external to any permanent electrical conduit and do not interfere with safety related equipment.
This safety evaluation demonstrated the acceptability of installation and operation of temporary acoustical monitoring equipment. ~
These, devices are used to monitor and determine the sources of increased leakage into the quench tank. This evaluation justified the mounting of these instruments to the pressurizer upper head and various safety related and non-safety valves and adjacent piping during power operations. The use of these temporary acoustical monitoring devices does not adversely affect safe plant operation, plant safety, accident mitigation, or safe shutdown. The use of temporary acoustical monitoring equipment identified in this evaluation did not involve an unreviewed safety question or require a change in plant Technical Specifications.
Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SAFETY EVALUATION JPN-PSL-SENS-95-016 REVISION 0 A
This safety evaluation demonstrated acceptability of positioning the normally closed containment spray header isolation valve I-FCV-07-.1A to the open position. This action was necessary since the valve failed to stroke to the open position within the required stroke time. I-FCV-07-1A design basis function is to open on a Containment Spray Actuation signal (CSAS) for accident mitigation.
The air operated valve, normally closed, fails open to its safety related position on either a loss of air or power. Since the valve failed to stroke open within the required time, the valve was placed in the open position and therefore, capable of performing it's design basis function. This evaluation assessed the effects on plant safety related to the open position. This evaluation determined that the open position did not adversely affect safe plant operation, or the engineered safety feature containment spray function since the valve was already aligned, for the safety related function. No new failure modes were created by this action since contingency actions adequately addressed inadvertent spraying of containment.
This safety evaluation concluded that operation with I-FCV-07-1A maintained in the open position during power operation was acceptable. Maintaining the valve open did not adversely affect the ability of the containment spray system to perform its safety related containment spray function. Likewise, in Mode 4, when containment spray" is not required, the valve could be repositioned to close to provide isolation and shutdown cooling integrity. The reposition of I-FCV-07-1A to be maintained opened during power operation identified in this evaluation did not involve an unreviewed safety question or require a change in plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SAFETY EVALUATION aTPN-PSL-SENS-95-027 REVISION 2 D
EL This safety evaluation demonstrated acceptability of varied, external loads on the safety relief valve (SRV)applying common discharge piping header in order to quantify the effects of the load and load distribution among the discharge nozzles of the three SRVs. The SRVs had experienced seat leakage with known experience that nozzle loading could adversely effect the leakage rates.
Strain gages were utilized to collect the data and the loads applied were limited to be within the ASME code allowable. The PORVs remained in service and were not adversely affected by this testing. The testing was performed with the plant in Mode 5 and the SRVs out of service. The evaluation determined that plant safety, accident mitigation, safe shutdown, or shutdown cooling was not adversely affected by the testing or SRV configuration.
This safety evaluation demonstrated acceptability of varied external loads on the safety relief valve (SRV)applyingcommon discharge piping header. The stress analysis for the affected piping and valves had been determined acceptable and within ASME code allowable. The original design configuration was restored prior to any mode changes and meets technical specification requirements. The external loads applied and stress analysis testing identified in this evaluation did not involve an unreviewed safety question or require a change in plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions/testing identified in this safety evaluation.
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SAFETY EVALUATION a7PN-PSL-SENS-96-003 REVISION 0 gym~a~
This safety evaluation documents FSAR changes required as a result of review of operating procedures and plant operation for RCS boron concentration changes: This evaluation provided the bases for resolving the inconsistencies and clarifying options for conducting boron concentration changes. This evaluation provided appropriate FSAR wording changes. This evaluation was classified as safety related since safety related systems were involved. This evaluation determined existed and if if an adverse operating or safety condition changes to plant procedures were warranted. This evaluation provided operations flexibility in how to operate the system, including the selection of flow paths and mode of operation.
This safety evaluation demonstrated that the identified FSAR clarifications associated with RCS boron concentration changes did not impact safe operation of the plant or adversely impact accident mitigation. The conditions identified in the evaluation and the changes to the FSAR identified in this evaluation did not involve an unreviewed safety question ar require a change in plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
SAFETY EVALUATION O'PN-PSL-SENS-96-010 REVISION 1 This safety evaluation demonstrated the acceptability of a temporary system alteration to the control room air conditioning system. Xt allows motor-operated damper D-20, D-21, or D-22 to be removed from the system for'aintenance and replaced with a blind flange. The blind flange is a temporary proxy for the damper and creates part of the control room A/C boundary. Only one damper from the system at a time was removed to ensure technical specification requirements. The system design bases was maintained by this temporary configuration during maintenance and was restored to the original design configuration following maintenance. This evaluation considered system operability, seismic design of the blind flange, pressure boundary, and failure modes. The temporary system alteration was determined acceptable and did not impact plant safety, safe plant operation, or accident mitigation.
This safety evaluation demonstrated the acceptability of a temporary system alteration to the control room air conditioning system and allowed motor-operated damper D-20,,D-21, or D-22 to be removed from the system for maintenance. The dampers were replaced with a blind flange. This activity was acceptable and did not adversely affect plant safety or safe plant operation. System integrity was maintained by, use of the temporary blind flange. The temporary system alteration to the control room air conditioning system identified'in this evaluation did not involve an unreviewed safety question or require a change in plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SAFETY EVALUATION aTPN-PSL-SENS-96-021 REVISION 0 D
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This safety evaluation demonstrated the acceptability of providing back-up AC power via the station blackout cross-tie and an emergency diesel generator (EDG) from the opposite unit in lieu of the local unit's (local means the shutdown unit ) second EDG.
Plant procedures had required two EDGs be operable whenever the RCS level was in a reduced inventory condition. This requirement had potential to extend outage durations. This evaluation reviewed NRC and industry guidance and plant licensing commitments and determined acceptability for the use of the station blackout cross-tie as a contingency. The requirements were met and the capability to supply loads were determined acceptable with no adverse affect on plant safety or safe plant operation. Technical specifications requirements are meet by the use of the station blackout cross-tie and an EDG from the opposite unit.
This safety evaluation demonstrated acceptability of back-up AC power via the station blackout cross-tie and an emergency diesel generator (EDG) from the opposite unit during operations with reduced RCS inventory conditions. Failure of an opposite unit EDG is no different than a failure of the second EDG. No new failure modes are created by the station blackout cross-tie and opposite unit's EDG. There are no plant restrictions on the normal operation of both the local and opposite units as a result of this evaluation. Providing back-up AC power via the station blackout cross-tie and an EDG from the opposite unit during operations with reduced RCS inventory as identified in this evaluation did not involve an unreviewed safety question or require a change in plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SAFETY EVALUATION JPN-PSL-SENS-96-033 REVISION 0 BQKIERXK This safety evaluation documents FSAR changes required to clarify permissible modes of operation for the waste gas holdup system. In particular, this evaluation identifies acceptable waste gas system vent paths for periods of high RCS fission product inventory. This evaluation provided the bases for .resolution and clarifying the FSAR wording. Operating procedure changes were not required as a result of this evaluation. This evaluation determined adverse operating or safety condition existed and if if an changes to plant procedures were warranted.
0 This safety evaluation demonstrated that the identified FSAR
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clarification associated with permissible modes of operation for
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the waste gas holdup system to maintain ALARA, do not impact safe operation of the plant, warrant changes to plant procedures, or adversely impact accident mitigation. The conditions identified in the evaluation and the changes to the FSAR identified in this evaluation did not involve an unreviewed safety question or require a change in plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SAFETY EVALUATION JPN-PSL-SENS-96-035 REVISION 0 A H gmmn~~
This safety evaluation demonstrated acceptability of updating the FSAR associated with long term cooling portion of the LOCA recove~
efforts. These updates are based on the guidance provided in CEN-152, Emergency Procedure Guidelines for Combustion Engineering Owners Group plants. This evaluation also provided the basis for the required flow rate to the core when using alternate injection methods. Hot leg injection method is the method utilized to ensure that excessive concentrations of- boric acid do not develop in the core and result in boron precipitation and flow blockage. This evaluation adds a secondary alternate method for hot leg injection as approved in CEN-152 and updates the FSAR description. No physical changes are made to the plant. No plant operational restrictions are imposed by this evaluation.
This safety evaluation demonstrated acceptability of updating the FSAR associated with long term cooling portion of the LOCA recovery efforts related to hot leg injection. The methods described in CEN-152 are being used to update the FSAR. This evaluation does not impact safe operation of the plant, warrant changes to plant procedures, or adversely impact accident mitigation. The conditions identified in the evaluation and the changes to the FSAR identified in this evaluation did not involve an unreviewed safety question or require a change in plant'echnical Specifications.
Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SAFETY EVALUATION JPN-PSL-SENS-96-046 REVISION 0 This safety evaluation demonstrated acceptability of utilizing the station blackout cross-tie for those events in Modes other than Mode 1 operation and revised the FSAR accordingly. The individual plant examination of external events (IPEEE) analysis concluded there were several fire scenarios where a fire related failure of one electrical train and subsequent independent failure in the other train (same unit) would result in a unit blackout condition.
Such scenarios account for a significant portion of the total core damage frequency (CDF). It was determined that a recovery action utilizing the SBO cross-tie could provide a significant reduction in the CDF for those scenarios. This evaluation recognizes the use of the SBO cross-tie to provide offsite power from the non-black-out unit to the blacked-out unit. Plant procedures already address using the SBO cross-tie under blackout conditions initiating with plant operation in any mode, however, the FSAR only discussed use from a Mode 1 condition and to maintain the unit in hot standby (Mode 3). For plant conditions other than the Mode 1 assumption, use of the SBO cross-tie does not invalidate or otherwise impact the licensing commitments for SBO. This change was determined acceptable and did not adversely impact plant safety or safe plant operation.
This safety evaluation demonstrated acceptability of utilizing the station blackout cross-tie for those events in Modes other than Mode 1 operation. No licensing commitments are violated by use of the station blackout cross-tie as identified in this evaluation.
No new failure modes are created by the SBO cross-tie configuration identified in this evaluation. The conditions identified in the evaluation and the changes to the FSAR did not involve an unreviewed safety question or require a change in plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
SAFETY EVALUATXON aTPN-PSL-SENS-96-060 REVZSXON 0 This safety evaluation demonstrated acceptability of using alternate methods to transfer spent resins from the steam generator blowdown system and its facility. The FSAR provides one method of transfer, but other methods are determined to be acceptable and can include reuse of the resins in other plant systems. This evaluation documents those acceptable methods in the FSAR. This evaluation is classified as not nuclear safety since it does not involve a safety related system. There are no plant restrictions associated with the methods for transfer of spent resins identified in this evaluation.
This safety evaluation demonstrated acceptability and documented in the FSAR alternate methods for the transfer of spent resins from the steam generator blowdown system and its facility. The alternate methods identified in the evaluation enhance resin transfer by improving ALARA and reducing plant operating costs.
The activities describe do not adversely affect plant safety, safe plant operation, accident mitigation, or safe shutdown. No new failure modes are created by the alternate methods identified in this evaluation. The conditions identified in the evaluation and the changes to the FSAR did not involve an unreviewed safety question or require a change in plant Technical Specifications.
Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SAFETY EVALUATION O'PN-PSL-SENS-96-065 REVISION 0 This safety evaluation supports the installation of breakaway locks on the Unit 1 and Unit 2 Hot Shutdown Panel (HSP) doors. Review of the FSARs, fire protection reports, and NRC safety evaluation reports were performed and determined that the breakaway locks are acceptable for use. This activity allows a passive increase in security of the HSP rooms while maintaining the ability of control room operators unrestrictive entry. This evaluation demonstrates that no new failure modes are introduced and that an increase in security can be achieved which benefits plant safety. The FSAR was updated as a result of this evaluation.
This safety evaluation demonstrated acceptability for the installation of breakaway locks on the Unit 1 and Unit 2 Hot Shutdown Panel doors. There is no licensing requirement for a door lock on the HSP. The installation of breakaway locks on the Unit 1 and Unit 2 HSP doors identified in the evaluation and the changes to the FSAR did not involve an unreviewed safety question or require a change'n plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
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SECTXON 3 RELOAD SAFETY EVALUATXONS 84
PLANT CHANGE/MODIFICATION 96054 This engineering package provided the reload core design of the St.
Lucie Unit 1 Cycle 14. The Cycle 14 core is designed'for cycle length of approximately 14,100 EFPH,. depending on a cycle 13 length of 10,018 EFPH. The cycle lengths for Cycle 14 included an end of cycle inlet temperature coast down with a maximum reduction in primary coolant temperature of 26'F at full power.
The primary design change to the core for Cycle 14 is the replacement of 88 irradiated fuel assemblies with fresh Batch T fuel assemblies. The fuel mechanical design has been evaluated for longer fuel cycle operating strategy. The margins to fuel mechanical design limits will not be degraded as a result of longer cycle operation. The analysis to support the reload were performed assuming 30%27% steam generator tube plugging and were consistent with an amended RCS flow rate of 345, 000 gpm (Amendment No. 145) .
The engineering package includes a change to the Low Flow Trip setpoint. The cycle 14 reload are of the debris resistant long end cap design. The mechanical design of Batch T fuel is the same as the Batch S and Batch R.
v u The safety analysis of this design was performed by Seimens Power Corporation (SPC) and independently reviewed by Florida Power and Light Co. It has been de'termined that the operation of the Cycle 14 reload core does not pose an unreviewed safety question and can be implemented with no changes to the St. Lucie Unit 1 Technical Specifications. Note that based on the amended RCS flow rate (Amendment No. 145), a restriction in power is imposed such that after 7000 EFPH power is limited to 90% RTP until a revised SBLOCA is performed at the reduced RCS flow and 30% steam generator tube plugging conditions. Prior NRC approval was not required for implementation of this modification.
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