ML20137K864
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{{#Wiki_filter:. . . . ._ _ _ _ . _ . _ _ . _ _._ __ ._ . REGION 11 ATLANTA, GEORGIA e PLANT STATUS REPORT I ST. LUCIE MARCH,1995
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r PLANT STATUS REPORT FOR ST. LUCIE (3/95) TABLE OF CONTENTS PART 1 - FACILITY DESCRIPTION 1.1 FACILITY / LICENSEE....................................Page 2 1.2 UTILITY SENIOR MANAGEMENT ...........................Page 2 1.3 1.4 NRC STAFF............................................Page 2 LICENSE INFORMATION... ..............................Page 3 ^ 1.5 PLANT CHARACTERISTICS................................Page 3 1.6 SIGNIFICANT DESIGN INFORMATION.......................Page 3 1.7 EMERGENCY RESPONSE FACILITIES / PREPAREDNESS...........Page 8 1.8 PRESENT OPERATIONAL STATUS (Past Six Months).........Page 9 1.9 OUTAGE SCHEDULE AND STATUS...........................Page 12 , PART 2 - PLANT PERSPECTIVE 2.1 GENERAL PLANT PERSPECTIVE.... .......................Page 13 2.2 SALP HISTORY (Past Two SALP Periods).................Page 12 ! 2.3 SELECTED SALP AREA DISCUSSIONS ................. ....Page 13 j PART 3 - SIGNIFICANT EVENTS i l 3.1 SIGNIFICANT EVENTS BRIEFINGS (Past 12 Months)........Page 18 3.2 ENFORCEMENT STATUS / HISTORY (Past 12 Months)..........Page 18 PART 4 - STAFFING AND TRAINING 4.1 OPERATIONS STAFF - 0VERALL...........................Page 18 4.2 WORK FORCE ..........................................Page 19 4.3 OPERATOR QUALIFICATION /REQUALIFICATION PROGRAM.......Page 19 4.4 P LANT S I MU LAT0 R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . P ag e 2 0 4.5 INP0 ACCREDITATION...................................Page 20 PART 5 - INSPECTION ACTIVITIES 5.1 OUTSTANDING ITEMS LIST
SUMMARY
..... .................Page 20 5.2 MAJOR INSPECTIONS....................................Page 20 5.3 PLANNED TEAM INSPECTIONS.............................Page 21 5.4 INFREQUENT INSPECTION PROCEDURE STATUS...............Page 21 5.5 SIMS STATUS (OPEN TMI ITEMS).........................Page 21 ATTACHMENTS
- 1. PERFORMANCE INDICATORS
- 2. ALLEGATION STATUS
- 3. NRR OPERATING REACTOR ASSESSMENT
2 PART l- FACILITY DESCRIPTION 1.1 FACILITY / LICENSEE FACILITY: St. Lucie Units 1 and 2 PLANT LOCATION: Hutchinson Island near Port St. Lucie, Florida LICENSEE: Florida Power and Light Co. (Corporate Office in Juno Beach, Florida) 1.2 UTILITY SENIOR MANAGEMENT CORPORATE: J. L. Broadhead (Jim), Chairman of the Board and CEO J. H. Goldberg (Jerry), President, Nuclear Division SITE: D. A. Sager (Dave) - St. Lucie Plant Vice President C. L. Burton (Chris) - Plant General Manager L. W. Bladow (Wes) - Nuclear Assurance Manager H. F. Buchanan (Hank) - Health Physics Supervisor g, R. et Dawson (Bob) - Licensing Manager D.fJ. Denver (Dan) - Site Engineering Manager H. L. Fagley (Herman) - Construction Services Manager P. L. Fincher (Pat) - Training Manager R. J. Frechette (Bob) - Chemistry Supervisor J. Marchese (Joe) - Maintenance Manager W. L. Parks (Bill) - Reactor Engineering Supervisor C. A. Pell (Ash) - Outage Manager J. Scarola (Jim) - Operations Manager J. A. West (Jeff) - Services Manager D. H. West (Dan) - Technical Manager C. H. Wood (Chuck) - Operations. Supervisor 1.3 NRC STAFF REGION II, Atlanta, GA: S. D. Ebneter (Stew), Regional Administrator, (404) 331-5500 L. A. Reyes (Luis), Deputy Regional Administrator (404) 331-5610 E. W. Merschoff, Director., DRP, (404) 331-5179 D. M. Verrelli (Dave), Branch Chief, (404) 331-5535 K. D. Landis (Kerry), Section Chief, (404) 331-5509 R. P. Schin (Bob), Project Engineer, (404) 331-5561 SITE: R. L. Prevatte (Dick), Senior Resident Inspector, (407) 464-7822 M. S. Miller (Mark), Resident Inspector. (407) 464-7822
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3 NRR: S. A. Varga (Steven), Director, Division of Reactor Projects-I/II, (301) 504-1403 J. A. Zwolinsky (John), Deputy Director, Division of Reactor Projects-I/II, (301) 504-1335 D. B. Matthews, Deputy Director, Project Directorate 11-2, (301) 415-1490 J. A. Norris (Jan), Senior Project Manager, Project Directorate 11-2, (301) 504-1483 AE0D: S. Israel (Sandy), Reactor Operations Analysis Branch, (301) 415-7573 1.4 LICENSE INFORMATION Unit 1 Unit 2 Docket Nos. 50-335 50-389 i License Nos. DPR-67 NPF-16 l Construction Permit Nos. CPPR-74 CPPR-144 Construction Permit Issued 7/1/70 5/2/77 Low Power License NA 4/83 Full Power. License 3/1/76 6/10/83 Initial Criticality 4/22/76 6/2/83 1st Online 5/17/76 6/13/83 Commercial Operation 12/21/76 8/8/83 1.5 PLANT CHARACTERISTICS Description Units 1 and 2 Reactor Type Combustion Engineering PWR, 2-loop i Containment Type Freestanding Steel w/ Shield Building l Power Level 830 MWe (2700 MWt) Architect / Engineer Ebasco NSSS Vendor Combustion Engineering Constructor Ebasco , Turbine Supplier Westinghouse Condenser Cooling Method Once Through Condenser Cooling Water _ Seawater 1.6 SIGNIFICANT DESIGN INFORMATION l l'.6.1 REACTOR INTEGRITY Reactor Pressure Vessel (RPV) With,the present fuel type and management policy, Unit 1 is expected to reach a 40-year RPV life. On this unit, tne fuel type and management policy have been modified to make that RPV life f d
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4 span possible. Presently, a proeram is evolving for RPV life extension beyond the projected 40 years, potentially to 60 years, 4 via a flux reduction program. A flux reduction program has I started with the addition of eight absorbers in core corner i positions, performance of vessel fluence calculations, and I determination of an optimum power profile for each core load. Calculations using current methodology and uncertainty predict a significant RPV life extension, but not to 60 years. Excore dosimetry installed for the current cycle (with planned removal in i October, 199/.; will be used to reduce calculation uncertainty. ; Due to different design and construction characteristics, Unit 2 RPV life expectancy exceeds 60 years. Low leakage core designs are now used for economic reasons, however the low leakage designu provide even greater life expectancy. Reactor Coolant Pressure Boundarv , l On this CE plant, ECCS-to-RCS injection points are isolated by at I least two check valves and one closed M0V. High pressure safety injection (HPSI), low pressure safety injection (LPSI), and containment spray (CS) pumps' common containment sump suctions are isolated from the containment sump by one closed M0V in conjunction with a closed seismic piping system. The CS headers , are isolated from containment by one closed M0V and a check valve in conjunction with a closed seismic piping system. CVCS has the normal complement of two automatic actuation isolation valves. 1.6.2 REACTOR SHUTDOWN Reactor Protection System The reactor protection system provides protection for the reactor fuel and its cladding by providing automatic reactor shutdowns (8 i trips) based on input from reactor power reactor coolant pressure, coolant temperature, coolant flow, steam generator pressure, and containment pressure. The RPS is a redundant four-channel system that operates on a two-out-of-four logic. ATWS Protection ATWS protection, outside the normal reactor protection system, is initiated via the ESF pressurizer pressure signal. It actuates by opening contactors in the output of the CEA MG sets, thereby interrupting control element assembly power at its source. This protection has been installed on both units per CE, the NSSS, tcommendations'. Remote Shutdown Facilities These facilities are located in the switchgear rooms beneath each unit's control room. 9
4 ! I i 5 i 1.6.3 CORE COOLING Feedwater System The main feedwater pumps are motor driven with each delivering 50 percent of the flow required for full power. Turt'ne Bvoass/ Steam Dump Capacity l l t Each unit has five steam bypass valves, providing 45 percent of l total capacity. Unit I has one atmospheric dump valve per train (two trains) and Unit 2 has two valves per train. Each unit has the capability of dumping nine percent steam flow to the atmosphere. l Auxiliary Feedwater System There are two motor-driven pumps on each unit with 100 percent l capacity per pump. There is one steam-driven pump on each unit with 200 percent capacity. Any of the three pumps can inject to l either steam generator. Automatic initiation and faulted steam j generator protection are provided by each unit's Auxiliary l Feedwater Actuation System provided by the NSSS. 1 Emeraency Core Coolino System In each unit, there are two HPS: pumps and two LPSI pumps with no j unit-to-unit cross-connections. One pump of each type per unit ' will handle a postulated c2CA. The LPSI pumps also provide decay heat removal as required when the unit is shut down. Decay Heat Removal As indicated above. the LPSI pumps also provide decay heat removal as required when the unit is shut down by taking suction from the RCS (hot legs), passing the fluid through the shutdown cooling heat exchangers, and returning it to the RCS (cold legs). The. heat removing medium is CCW - discussed in section 1.7.6 below. Shutdown cooling flow path overpressure protection is provided by i automatic isolation valve,s and various relief valves in the { system. l 1.6.4 CONTAINMENT Pressure Control / Heat Removal 1 There are two containment spray pumps and four containment fan
- coolers available per unit to suppress pressure spikes and cool ;
. the containment. One CS pump and two fan coolers will handle a l postulated LOCA. There are no unit-to-unit cross-connections.
! This engineered safety feature is automatically started sy ESFAS.
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- Hydroaen Control !
Containment hydrogen control post-LOCA is accomplished on each ) unit by two trains of hydrogen recombiners located on the ! i operating deck inside containment. By elevating, in a controlled ; manner, the temperature of containment atmosphere flowing through the recombiner, the recombiner units recombine hydrogen and oxygen to form water, thus preventing the buildup of hydrogen to potentially explosive levels. 1.6.5 ELECTRICAL POWER l Offsite AC. The station switchyard is connected to the transmission system by
- three independent 240 KV lines that share a right of way and interconnect with FPL's grid on the mainland approximately 10 miles West of the plant site. There are two independent offsite power feeds from the station switchyard to the Emergency busses, l
Onsite AC j Onsite AC power is provided by four EDGs (two per unit). EDGs are independent of other plant systems except vital DC power for control-of starting. A Station Blackout (SBO) cross connection is installed and tested. This cross-connection serves the emergency busses directly and reduces cross-connect time to less than 15 minutes. DC Power ; l Two trains of vital batteries per unit have been routinely tested for four-hour DC load profiles. Recently, due to cell replacement, they have been tested for three-hour battery capacity ! instead. The battery capacity test is harsher than the load profile test. There are four normal chargers per unit with swing chargers available for service. Non-safety batteries can be cross-connected to the safety-related swing bus if needed. Instrumentation Power Each unit has four inverters, two powered from each vital DC train, that provide four trains of instrumentation power. Station Blackout Resolution Status Unit 2 is a four-hour "DC coping" plant per the original license while Unit 1 is subject to the station blackout (SBO) rule of 10 CFR 50.63 requiring additional licensee action (unit-to-unit ; cross-connect of 4160V bus).
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1.6.6 SAFETY-RELATED COOLING WATER SYSTEMS Intake Coolina Water (Service Water) . Intake cooling water (ICW) for each unit originates in a common canal called the Intake Canal. The canal level varies with the tides since it is filled by a level difference between the Atlantic Ocean and the canal. One 16-foot and two 12-foot - diameter pipes pass under the beach to connect the ocean and canal. The intake pipe ends in the Atlantic are covered oy intake structures (rebuilt in 1991) intended to limit flow velocities, particularly vertical velocity, to reduce marine life entrapment. After use, ICW returns to the ocean through a Discharge Canal and under-beach pipes. Each unit has two trains of ICW plus a swing pump that can be
- 4. aligned to either train electrically and physically. The licensee has converted the deep draft ICW pumps from externally (water) lubricated to self-lubricated to increase reliability of the lubrication water source. The 100 percent (each) capacity pumps i take suction from the intake canal via a canal intake structure using traveling screen debris protection. The intake canal structures adjacent to the ICW pump suctions are continuously injected with a hypochlorite solution to reduce marine growth in the associated piping and heat exchangers. Commencing 3/92,
, periodic injection of a clamicide at the intake structures, primarily to control marine growth affecting the turbine condensers, has also somewhat reduced marine growth affecting the ICW system. ] l The ICW pumps move water through two trains of heat exchangers that cool component cooling water (CCW) and two trains of heat exchangers that cool main turbine cooling water. During a postulated accident. water flow isolates from the turbine cooling heat exchangers. The discharge from the heat exchangers returns via the discharge canal to the ocean. Increases in debris and silt in the heat exchangers during 1993 indicated that the intake canal needed dredging. As of September 1993, the utility was routinely cleaning main condenser waterboxes at reduced power and obtaining necessary dredging permits from the state and Corps of Engineers. The canal was dredged in December 1993 and January 1994 with ! immediate results of reduced waterbox fouling. l Closed Coolina Water Systems l Each unit has two trains of Component Cooling Water (CCW). The arrangement of two pamps and a swing pump mimics the ICW system. l l
i f i 8 . The swing pump can be aligned to either train. The 100 percent (each) capacity pumps drive water through the CCW/ICW heat exchangers and then on to the heat loads, mainly the containment l 4 fan coolers and the shutdown cooling (decay heat) heat exchangers (which also can operate as containment spray heat exchangers). Additionally, CCW cools a variety of bearings, seals, and oil coolers for the HPSI, LPSI, and CS pumps. A non-safety-related portion of the CCW system cools reactor coolant pump seals and the , spent fuel pool. This section isolates upon engineered safety features actuation. J 1.6.7 SPENT FUEL STORAGE Wet storage capability exists up to the year 2002 (Unit 2) and
- 2007 (Unit 1). '
] 1.6.8 INSTRUMENT AIR SYSTEM l 1 Instrument air compressors and driers, installed several years ago l
- on each unit, provide all instrument air for Unit 2 and all but ,
containment air for Unit 1. These have increased instrument air i reliability. Unit 1 also has instrument air compressors inside I containment, j i 1.6.9 STEAM GENERATORS ! i ! Each unit has two large steam generators (SGs) rather than the . j three or-four usually seen. The licensee has begun to focus on a 4 Unit 1 SG replacement in 1997. The SGs are under ccnstruction at
- the B&W Canada shops and a site organization is functioning.
7 1.7 EMERGENCY RESPONSE FACILITIES / PREPAREDNESS Emergency Operations Facility: 10 miles West of site,
- I-95/ Midway Rd. Exit Technical Support Center
- Onsite, Adjacent to 1
Unit 1 Control Room j Operational Support Center: Onsite, 2nd floor of , North Service Building The last annual emergency preparedness exercise was held February 9, 1994. Two followup items were identified; one involving the definition of containment failure and one involving the need to demonstraM a protected area evacuation. An evacuation drill on September 30, 1994, i satisfactorily demonstrated the accountability program. The next emergency preparedness exercise is scheduled for May, 1995. Since St. Lucie site has a high probability of hurricanes, communications facilities were improved following the Turkey Point experience with Hurricane Andrew in August, 1992. Improvements include: l
i. I a 9 l High Frequency Auto-link with other FPL sites and NRC. Enhanced 900 MHZ System for site and mobile communications, with radios also in the licensee's EOF and county emergency facility. l-Cellular phones with hardened antennas. Hardened Local Government Radio antenna ties. 1.8 PRESENT OPERATIONAL STATUS (3/9/95) Unit 1 is operating at 100% power and has been operating since a reactor startup on March 8 following a short notice outage for pressurizer code safety valve replacement. Unit 2-is operating at 100 % power and has been operating since a reactor startup on February 24 following a trip on February 21 due to an instrument failure. Availability Factors: Unit 1 Unit 2 1991 81.0 100.0 1902 96.5 75.2 1993 74.0 71.8 1994 86.8 79.6 1995 (through 1/95) 100.0 100.0 Cumulative (through 1/95) 77.3 83.0 ) 1.8.1 UNIT 1 OPERATING HISTORY (Past Twelve Months from 3/21/95)- Unit 1 operated continuously during the past 12 months with the following exceptions: On March 28. 1994, Unit 1 experienced an automatic reactor trip when a maintenance foreman opened the generator exciter breaker. The worker had been issued a clearance on the Unit 2 exciter breaker and mistakenly entered the wrong unit's exciter control cubicle. j On April 2, 1994, the unit was returned to power; however the unit automatically tripped on April'3 from 19% power while deenergizing a 4160 Volt non-vital bus to allow safe removal for maintenance of a failed startup transformer output breaker. The planned electrical lineup placed the A emergency bus on its EDG, which was running at a different frequency from the grid. The paralleled CEA MG sets, now with different frequency drivers, developed circulating currents, resulting in several tripped circuit breakers. A partial reactor trip tripped the turbine, which tripped the reactor. Unit I returned to power on April 4, j 1994. j l l 1
10 " On June 6. 1994, Unit 1 tripped during a severe thunderstorm. The main transformer locked out the generator, causing a reactor trip, due to a phase differential on main generator transformer lA. This occurred as a result of an approximately 8' length of flashing from an adjacent building which was blown across two phases of the 1A main transformer output. The licensee conducted inspections and tests of the 1A and IB main transferwrs and the main generator, and performed repairs to the 1A % ;6 % ..rformer. The reactor was taken critical on June 8; however, the licensee i elected to remain off-line until repairs were completed to the 1A i main transformer. Unit I was placed on line on June 11. Unit I reduced power and entered mode 2 on August 28 to repair a J DEH leak. The unit was returned to power approximately 18 hours j later on the same date. On October 26. the unit tripped from 100 percent power due to a { I loss of electrical load. This was the result of arc-over in a l potential transformer in the switchyard due to salt buildup. The i licensee then entered a unit refueling outage, which had been ' scheduled to begin four days later. The unit was returned to j service on November 29.
- On February 27, 1995, the unit was removed from service for the
]' replacement of pressurizer code safety valves which had been ; leaking by the seat since shortly after startup in November, 1994, p The unit was returned to service on March 8. l On March 4, 1995, the unit experienced a 14' minute loss of shutdown cooling. The apparent root cause was operator error by a reactor operator placing one loop of SDC in standby. The operator apparently closed the suction valve to the operating, vice standby, pump. The operator in question has denied the error. The licensee is considering disciplinary action and has relieved the operator of licensed activities. 1.8.2 UNIT 2 OPERATING HISTORY (Past Twelve Nnths from 3/21/95) Unit 2 operated continuously during the past 12 months with the : following exceptions: On March 16, 1994, during a unit refueling outage, the licensee identified boron deposits indicative of leakage from one of four pressurizer steam space instrument nozzles. Licensee investigations identified un. acceptable linear indications in three of four no7zle pressure boundary welds. The steam space nozzles were constructed of Inconel 690 and were installed in 1993 as replacements for existing Inconel 600 nozzles, which had been found to be cracked. The new nozzles were attached with Inconel 600-equivalent weld material, as 690-equivalent material was not , approved for use at the time. The licensee determined that ti e ** ,
I l i 11 l l indications were the result of Primary Water Stress Corrosion ! Cracking (PWSCC). i l The licensee's corrective actions involved repairing all four j l nozzles by creating new pressure boundary welds at the exterior l
- l. wall of the pressurizer. The new welds were of the inconel 690- l compatible material. During the repair efforts, region-based I inspectors found that the overall repair effort was well I controlled and that performance was good: however, one violation was identified involving incorrect bevel angles on two weld preps.
On Marcn 18. Unit 2 experienced a six minute cessation of shutdown l cooling when a misanalyzed clearance (tagout) resulted in automatic valve realignments that secured flow to one of two
- operating shutdown cooling trains. A second shutdown cooling loop l l was in operation at the time; however, operators stopped the '
j' operating pump as a precaution against damage after the unexpected valve realignments. Operators assessed the situation and restored , shutdown cooling in six minutes. ! Unit 2 completed the refueling outage and was returned to power on l April 19, 1994. l l On April 23, 1994, Unit 2 tripped due to a RPS cabinet manufacturer's wiring error which manifested itself during RPS troubleshooting. The wiring error existed since the original manufacture of the cabinet. Following Unit 2 trip stabilization of April 23. 1994, the steam
- bypass control system operated unexpectedly, resulting in a rapid j 7 degree cooldown and a resultant severe RCS shrink (pressurizer l heaters deenergized on low oressurizer level). Dromot operator ,
action was taken to secure the cooldown. Unit 2 was returned to I power en April 26. 1994. l On July 9. 1994, Unit 2 turoine was shut down and reactor power reducea to Mode 2 because the 2B1 RCP lower oil level indication showed a leak. The sump was not leaking and an unusual failure in the indication system was determined to be the reason for the ; indication. The reactor was returned to moae 1 and the turbine l i startec up on July 10, 1994. l On July 14,1994, Unit 2 was shut down to allow repair of a stuck-l closed trip circuit breaker. Operators did not follow Unit 2 Tech. i Spec LC0 time requirements regarding shut down on July 14 to allow i repair of a stuck-closed trip circuit breaker. The unit was restarted and olaced on line on July 15. 1994. ; l
; On February 21. 1995. the unit tr1oped as a result of low steam i generator water ievel. The conaition was the result of a ; feedwater regulating valve closure after a steam generator water ; }
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4 12 l level control level transmitter failed high. The transmitter was l replaced and the unit was returned to service on February 25. I 1.9 OUTAGE SCHEDULE AND STATUS Unit l's last refueling outage began on October 26, 1994, and ended on ] November 29, 1994. Major activities included: refueling; reactor vessel - . nozzle and flange weld ISI inspection; installation of a permanent i ] cavity seal ring; replacing reed switches for several CEAs; integrated l , safeguards test; steam generator tube inspection and plugging; steam generator sludge lancing; repair of refueling water storage tank: several instances of reduced inventory / mid-loop operations; replacement s of ICW/CCW LOOP logic [HFA latching relays) with pull-to-lock switches;
- removal (collection] of Rx vessel neutron flux dosimetry; modification of EDG skids to allow access underneath; inspection of ECCS sump area; replacement of a main transformer; modification of containment spray Na0H addition piping; ano mechanical, electrical. and I&C systems maintenance.
Unit 2's last refueling outage began on February 13, 1994, and ended April 17, 1994. Major outage activities included: refueling; steam , generator tube inspection and plugging; low pressure turbine blading ' replacement; emergency diesel generator inspection; replacement of two reactor coolant pump mechanical seals; and mechanical, electrical, and I&C systems maintenance. The next Unit 2 refueling outage is scheduled for October, 1995. 1 PART 2 - PLANT PERSPECTIVE j 2.1 GENERAL PLANT PERSPECTIVE A SALP presentation was conducted on February 15. 1994, covering the SALP period of May 3, 1992, through January 1, 1994. 'The facility was rated category 1 in all functional areas for the. second consecutive SALP period. In June 1994. St Lucie was dropped from the NRC management list of good performers after experiencing five unit reactor trips in the first half of 1994. 2.2 SALP HISTCRY (Past 2 SALP Periods) The last SALP period, SALP Cycle 10, ended on January 1, 1994. The current SALP period ends on July 1.1995. j 5
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13 ASSMT. OPS RAD MNT/SURV EP SEC ENG/ TECH SAQ'V PERIOD 5/1/89 - 1 1 2 1 1 1 1 10/31/90 11/1/90 - 1 1 1 1 1 1 1 , 5/2/92 PLANT OPS MAINTENANCE ENGINEERING PLANT SUPPORT 5/3/93 - 1 1 1 1 1/1/94 2.3 SELECTED SALP AREA DISCUSSIONS (3/21/95) Since the assessment of the SALP period ending in January, 1994, there have been no events that should significantly change the overall assessment of this facility. A new corrective action program, the St. Lucie Action Report Star System, was implemented in July, 1994. This program will be used to identify, review, analyze, resolve, track, and close out all plant discrepant conditions. It is intended to provide increased emphasis in this area. Plant Operations Summary of Previous Assessment Within the current SALP cycle, previous assersments have noted a potential decline in Operations' performance. Noted indicators included five reactor trips in the first six months of the cycle. No common root causes were identified. Operator actions with regard to the noted trips were generally good. Two entries into reduced inventory operations during the Unit .'. outage were noted as excellent. Procedural weaknesses which incicated a lack of rigor in the review process were noted, as was the fact that temporary changes to procedures were on the increase (indicating increasing attention to procedural adequacy). Management activities in response to the increase in operational events was determined to be strong, with an increase in overall focus directed at plant operations. The corrective actions program was enhanced, consolidating tens of programs into one which involves daily management reviews of all documented conditions. The previous assessments concluded that Operations remained strong in the current period, that management actions were aggressive in dealing with identified weaknesses, and that increaseo attention to procedural adeouacy may be warranted.
14 Last Six Months 1 The previous six months has shown a sharp decrease in the number l of reactor trips, with two having occurred; one due to salt buildup on switchyard components which led to a loss of electrical . load on Unit 1, and one due to a level transmitter failure which I led to a low steam generator level trip on Unit 2. l Operator Performance > l I Operator performance has been noted to be good to excellent ! in evolutions such as post-trip response, startups, mid-loop l operations (during the Unit 1 outage), and in the l performance of surveillance testing. Several findings have l indicated potential weaknesses in attention to detail on the l part of operators. They include: e Failure to sample a SIT within the TS-requirea time frame. Second occurrence in two years. e Failure to identify an inoperable LPSI header pressure control board indicator which had been failed in mid-range for approximately two weeks. e Minor refueling machine errors of omission e Failure to identify / notify in response to Unit I hot leg stratification. o Potential operator error in valve control manipulation which led to a loss of shutdown cooling on Unit 1. An issue of a lack of professionalism and of a casual attitude toward E0Ps on the part of one individual was identified during operator examinations. The licensee prepared and executed an aggrt.asive remediation plan prior to allowing the individual to return to watchstanding duties. Strengths Strengths have been identified in the ability of the Operations department to integrate plant activities. Aspects included multidisciplinary briefings prior to major evolutions (be they operations or maintenance evolutions), and an increased involvement by Operations in the prioritization of maintenance activities. Operations' management has played a major role in the identification and resolution of prominent problems, including: e 2B LPSI pump air-binding e Pressurizer relief valve seat leakage e Failure to sample SITS following volume additions e Minimum temperature for criticality discrepancies
i. 1 15 . e ; e CVCS letdown control anomalies j i In these and other cases, Operations has expressed an unwillingness to accept stopgap maintenance measures and has demanded that historically poorly performing systems be made i to operate correctly. l' Weaknesses Weaknesses have been identified in annunciator response procedures (several were discrepant either in setpoint ' !' reference or in content), corrective actions (a violation resulted from the licensee's failure to recognize the impact '
- l. of a failure to perform adequate surveillance testing on
, swing bus components), logkeeping (a violation resulted from l the modification by one snift, of another snift's log i entries), and aspects of the Facility Review Group s ,
- activities (some members were identified as " going through the motions" of review; however, the Plant General Manager i was consistently cited as a strong presence who articulated j expectations and kept members on track). I i
Conclusion I L i On. balance. Operations continues to be strong at St. Lucie. 1 The licensee continues to be aggressive in the I L identification and resolution of operational problems. In 3 the last six months, operations has shown a new involvement , in maintenance prioritization which has increased the safety i perspective applied to maintenance activities. Maintenance / Surveillance j Summary of Previous Assessment Maintenance was assessed as category 1 in the previous SALP. The previous assessments made during the current SALP cycle indicated that the performance level of maintenance activities had not abated. Strong performance had been
- '1oted in the support of the Unit 2 outage, and housekeeping l' ano plant preservation activities were aeemed 9000.
i last Six Months I Juring the past six months, twenty-seven , maintenance activities were observed in varying levels of depth. Two
<iolations were cited; one involving the misuse of an unreviewed vendor technical manual and one involving a-
#ailure to perform an adeouate independent verification of Landed leads. Workers were generally founo to De,well- .;
trained, conscientious, and skilleo. l l l m - _ _ . _ , _ . -
i I 16 ! i L Twenty-four surveillance activities were observed. No I l violations or major weaknesses were identified as a result. j t i I !
. Strengths j
l The Unit 1_ outage was considered a major strength. The , a outage was conducted in thirty-five days, three days ahead ' . of schedule, without a safety-significant reduction in work j scope. l The licensee developed and implemented a Critical l
- Maintenance Management program for the performance of I t
maintenance within TS ass. Maintenance activities performed i 5 under this program have been observed and found to be well- ' j planned and executed and have been performed after assessing . the increased risk associated with the activity. , 2 I ! The licensee's predictive maintenance program has continued ) j to provide early indications of impending failures. Most recently, the organization identified an unsatisfactory lug j . in cabling supplying one phase of a CEA MG. j l Weaknesses Isolated weaknesses have been identified both by the NRC and the licensee's QA organization. They include: e A failure to properly incorporate changes into VTMs. Required FRG reviews were not performed. o A failure to adequately define and perform independent verifications properly. 4 e A failure to properly implement elements of the site's welding program. This was identified by QA and , resulted in the Maintenance Manager issuing a stop ! work order, which was in place for one week, while concerns were addressed. Conclusion Maintenance continues to be performed at superior levels. , Planning and execution has improved through extensive l planning efforts. The addition of the Critical Maintenance Management program has enhanced on-line maintenance by requiring extensive review and planning prior to work. Surveillances continue to be performed well. Enaineerina Summary of Previous Assessment
17 The previous assessments for this SALP cycle concluded that engineering was generally strong. Good support of the Unit 2 outage was noted, as was. good QA with respect to fuel fabrication and receipt inspection. Potenti'al problems were noted in the area of vendor technical manuals. i k Last Six Months i The noted concerns with respect to VTMs were, in part, validated in the maintenance area, where a violation resulted; however, the violation was not reflective of a failure on the part of engineering. Good support to the + Unit 1 outage was noted, with engineering personnel assuming pivotal roles in the management of the outage. One NCV was ! identified, relating to the design of NaOH supply piping, however, the probl'm e had exi'sted since shortly after - construction and was appropriately addressed. Five plant modifications and several safety evaluations were reviewed and were generally found to be thorough and correct. The licensee's program for the control of containment coatings was reviewed and found to be i satisfactory. Engineering involvement has been evident in i each major plant challenge in the last six months, including: e Apparent air-binding of the 2B LPSI pump , e 1A LPSI pump relief valve lift e Unit 1 Pressurizer relief valve seat leakage e Post-event reviews of loss of Unit 1 SDC e Unit 1 RWT leak repairs Conclusion Engineering continues to perform well. No weaknesses have been identified in this functional area. Plant Support Radiolooical Controls Previous assessments this SALP cycle indicated an effective program. Inspections this period indicate good contrcl of . internal / external exposure and containment during outages. ALARA initiatives were noted; robotics, submersibles, and telemetry. The licensee was noted to be ahead of most of the region in the use of cameras, video and wireless communications. Emeroency Preparedness The licensee continues to maintain an effective EP program.
15 Security Security upgrades made prior to the last SALP were notable. The licensee continues to maintain a very effective security program. Fire Protection The licensee continues to maintain an effective fire protection program. Housekeepina Housekeeping has been generally very good. PART 3 - SIGNIFICANT EVENTS 3.1 SIGNIFICANT EVENTS BRIEFINGS (Past 12 Months) Unit 1: None this period Unit 2: Failure of a GE AK-25 Trip Circuit Breaker 3.2 ENFORCEMENT STATUS / HISTORY (Past 12 Months) Currently, there are no escalated enforcement actions pending at St. Lucie. PART 4 - STAFFING AND TRAINING 4.1 OPERATIONS STAFF '0VERALL (8/94) Above average performance of the operations staff has been noted. Control ' room demeanor of personnel is above average. Number of Shifts: (RC0, SR0) Six shift rotation. 8-hour shifts; (NPO, ANP0. SNPO) Five
. shift rotation. 8-hour shifts.
Number of SR0s: 22 active /21 inactive' / 43 total Number of R0s: 30 active /2 inactive / 32 total Total Licensed Operators: 52 active /23 inactive / 75 total 3 SR0s perform only R0 duties and maintain SR0 licenses active only for R0 duties. This practice is being reviewed by RII operator. licensing. 1
i ; i 49 l 4.2 WORK FORCE I8/94) -l i { FPL Contractor j 3 , i' Plant personnel (excluding 713 122 l disciplines below) l i Training 63 0 i Quality Assurance /ISEG/ SPEAK 0VT 49 0 1 } Materials Management 46 0 j Security 11 122 7 Site Engineering 42 0 4 4.3 OPERATOR OVALIFICATION/RE0VALIFICATION PROGRAM (Past Two fears) i j 4.3.1 REQUALIFICATION PROGRAM
, NRC-administered requalification exams were completed in October,
{ 1992. Results were good - 9 of 12 R0's passed and 12 of 12 SR0's
- passed. Three of the-R0's failed the written exam and one also i failed the JPMs. The program was rated satisfactory.
j ReyJalification exams are currently in progress (10/94). To date, 20 of 24 SR0's and 17 of 20 R0's have passed all portions of the ! exams Failures have incluoed 5 written exams, 1 JPM..and 1 l simulator failure. l $ 4.3.2 INITIAL EXAMS ! i Previous initial operator exams were conducted on April 29. 1991.
- Six SR0 upgrades were examined, and all six passed. Additional exams were completed October 25. 1991. Six operators 2 SR0 upgrades. and 1 instant SR0 were examined. All passeo. The last i i
initial exam was given April 27 through May 1. 1992, to 6 SRO l j upgrades and 2 R0s, and all passed. A hot license class of 15. 1 persons was started in late February, 1992 (14 still in class). Results of an initial exam conducted in October 1993 were that 10 j of 10 prosoective R0s passed. Resuits of initial exams in October l 1994 were that 9 of 9 pro'spective SR0s and 2 of 2 prospective R0s 1 passed. 4.3.3 GENERIC FUNDAMENTAL EXAM On an NRC administered Generic Funoamental Exam on June 6. 1990, 6 ; of the 10 St. Lucie operators who took the exam passed. On ! ( February 6. 1991. 3 of 3 coerators who took the exam passed. On June 6. '991,
. one cperator took the exam ano casseo. On February 10.1993, all 12 operators who took the exam passed. One person
l w ' $ took the exam on February 9, 1994. and passed. No further Generic Fundamental Exams have been taken. 4.4 PLANT SIMULATOR The simulator is on site and fully certified to meet ANSI /ANS 3.5, 1985. ! 4.5 INPO ACCREDITATION All training programs are maintaining INP0 accreditation. The site
- specific simulator has been used for training since 1988 and has been
. fully certified for approximately 4 years. Eight separate NRC , inspections in the form of operator examinations at the simulator have j found no serious problems. i 1 PART 5 - INSPECTION ACTIVITIES 1 5.1 INSPECTION FOLLOWUP OPEN ITEMS
SUMMARY
(UNITS 1 AND 2 COMBINED) (10/6/94) l Pre Change from I Division ,_9_3__ Total Last Report DRP 3 30 0 l DRS 0 7 -3 l DRSS 0 2 0 Totals 3 39 -3 Note: Each item that applies to both units is counted as one item. 5.2 MAJOR INSPECTIONS I R - N o_ . Date Type 89-02 1/89 RG-1.97 89-03 3/89 NDE 89-07 3/89 EQ 89-09 3/89 Design Control 89-?4 10/89 Maintenance Team Inspection 89-27 11/89 E0P Followup 90-09 4-5/90 OSTI 91-03 2-3/91 EDSFI 91-18 9/91 M0V (r;o negative findings) 91-201 9-10/91 Service Water Inspection 92-14 7/92 Emergency Preparedness Program 92-17 7/92 EDSFI Followup l 93-01 1/93 Check Valves 94-11 5/94 MOV Followup
.z a 5.3 PLANNED TEAM INSPECTIONS None 5.4 INFRE0 VENT INSPECTION PROCEDURE STATUS No core modules are overdue at this time.
5.5 SIMS STATUS - OPEN TMI ITEMS There are no open TMI items. 4 4 4 d i l 1 0
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PREDECI@ICNAL , ST. LUCIE 1 l PI EVENTS FOR 94-1 SSA 01/13/94 LER9 33594002 50.728: 26605 PWR HIST: POWER OPERATIONS AT 100% DESC : WHILE TESTING THE UNDERVOLTAGE RELAT, AN INADVERTENT UNDERVOLTAGE SIGNAL INITIATED AN EM RGENCY DIESEL GENERATOR START AND SusSEQUENT LOADING OF ITS ASSOCIATED VITAL BUS. SCRAM 03/28/94 LER# 33594003 50.728: 27006 PWR HIST: POWER OPERATIONS AT 68% DESC : AN AUTOMATIC REACTOR TRIP OCCURRED F 4 THE INADVERTENT OPEN NG OF THE GENERATOR EXCITER FIELD BREAKER. i i PI EVENTS FOR 94-2 SCRAM 04/03/94 lea # 33594004 50.728: 27035 PWR HIST: POWER OPERATIONS AT 19% l DESC
- A REACTOR SCRAM OCCURRED WHILE ATTEMPTING TO ISOLATE A 4KV BUS FOR REPAIR. DUE TO A PROCEDURAL i ERROR, AN ELECTRICAL PMASE MISMATCM OCCURRED BETWEEN THE CEDM MG TRAINS THAT RESULTED IN A LOS$ OF
- POWER 70 THE CEDM BUS.
( SCRAM 06/06/94 LER8 33594005 50.728: 27359 PWR HIST: POWER OPERATIONS AT 8% DESC : DURING MIGH WINDS, A PIECE OF ALUMINUM FROM A STORAGE BUILDING WAS BLOWN INTO THE MAIN TRANSFORMER 9 CAUSING A CENERATOR/ TURBINE / REACTOR TRIP. PI EVENTS FOR 94-3 NONE 4 PI EVENTS FOR 94-4 SCRAM 10/26/94 m.# 33594007 50.72#: 27954 PWR MIST: POWER OPERATIONS AT 100% ( DESC A REACTOR TRIP WAS INITIATED WHEN THE MAIN TURtlNE TRIPPED AS A RESULT OF A MAIN GENERATOR LOCKOUT CAUSED BY A FAULTED POTENTIAL TRANSFORMER IN THE SWITCH TARD. SSA 11/22/94 LER# 33594009 50.728: 28060 PWR HIST: COLD SHUTDOWN DESC . SOTH EDCS STARTED ON AN $! AS. ONE EDG LOADED ITS BUS WHEN AN ABNORMAL ELECTRICAL LINEUP COMe!NED , WITH AD0!TIONAL LOADING FROM THE SIAS CAUSED THE BUS BREAKER TO OPEN. SSA 11/22/94 LER# 33594009 50.728: 28060 PWR MIST: COLD SHUTDOWN DESC : AN SIAS OCCURRED WHEN TWO OF FOUR PRESSURIZER PRESSURE TRANSMITTERS DRIFTED MIGH FOR UNKNOWN REASONS. THE CHARGING PUMPS WERE SECURED BEFORE THE LTOP SETPolNT WAS REACHED. SSA 11/24/94 LER8 33594010 50.728: PWR HIST: COLD SHUTDOWN , DESC : AN INADVERTENT SAFETY INJECTION ACTUAfl0N $1GNAL OCCURRED DURING REACTOR PROTECTIVE AND ENGINEE M , SAFEGUARDS STSTEM TESTING WHEN A PRESSURIZER PRESSURE BISTABLE TRIPPED.
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a PREDECISIONAL ST. LUCIE 2 PI EVENTS FOR 94-1 NONE PI EVENT 8 FOR 94-2 SCRAM 04/23/94 LERW 389%D03 50.728: 27150 PWR H! sis POWER OPERATIONS AT 29% DESC : DURING A NUCLEAR INSTRLME1TATION CALIBRATION, A REACTOR TRIP OCCURRED DUE TO AN !MPROPERLY WIRED BYPASS CIRCUIT. PI EVENTS FOR 94-3 NONE PI EVENT 8 FOR 94-4 NONE 1 i i 4 1 0
PAGE ST. LUCIE OPEN ALLEGATIONS AS OF 03/21/95 ALLEGATION: RIl-94-A-0105 FAClll1Y: ST LUCIE I DATE RCVD: 940520
SUBJECT:
I WELDING SUPERV RCVD NEGATIVE PERFORMANCE APPRAISAL AND THEN DUE DATE: 950228 REMOVED FROM A SUPERVISORY POSITION DAYS OPEN: 305 ACTION PNDG: AP:NO SAlETY ISSUE. 01 ACTION PNDG RSP DIV/lCA: EICS ACTION DUE DAIE: DUE: 12/30/94 : ALLEGATION: Ril-94-A-0119 IACilllY: SI LUCil 1 DATE RCVD: 940615
SUBJECT:
NOT EN00Gil GUARDS, LT HAS DRINGKING PROB, SECURITY OFFICER , DUE DATE: 950430 3MELLS OF ALC0HOL, SEC GUARD STRUCK BY LIGilTNING, WEAP0HS, DAYS OPEN: 279 ACTION PNDG: AP: LIC RSP DTD 2/3/95 RCVD. DRSS/NMSS PROVIDE CLOSURE MEM0 TO EICS j RSP DIV/lCA: DRSS/NMSS . ! ACTION DUE DATE: DUE: 03/17/95 , AllEGAll0N: Ril-94-A-0206 lAClllIY: ST LUCll 1 DATE RCVD: 941116
SUBJECT:
INADEQUAIE EVALUATION OF PIPE SUPPORT BOLTING DISCREPANCIES ; DUE DATE: 950330 AS REQUIRED BY GL 79-02 IN AN ATTEMPT TO COVER UP A GENERIC i DAYS OPEN: 125 ACTION PNDG: AP: DRS/ENG PROVIDE CLOSURE MLM0 TO EICS i RSP DIV/LCA: DRS/ENG;BLAKE ACTION DUE DATE: DUE: 03/15/95 ALLEGATION: Ril-94-A-0220 FACILITY: ST LUCIE 1 ! DATE RCVD: 941215
SUBJECT:
FITNESS FOR DUTY CONCERN. AN INDIVIDUAL Wil0 ilAS WORKED AT NU DUE DATE: 950528 CLEAR Pl. ANTS FOR MORE THAN 20 YRS CONSUMES SIGNIFICANT AMOUN DAYS OPEN: 96 ACTION PNDG: AP: DRP/PB2 AWAITING L!C RSP. IIC RSP DUE 3/20. DRSS/NMSS REVIEW LICENSEE'S RESPONSE WilLN RECEIVED RSP DIV/lCA: DRP/PB2 ACTION DUE DATE: DUE: 03/20/95 I i i
PAGE S1. IUCIL OPEN ALLEGATIONS AS OF 03/21/95 AILEGA110N: Ril-95-A-0001 FAClll1Y: ST LUCIE 1 DATE RCVD: 950103 SUBJECI: ARMS ROOM DOOR LEFT OPEN AND UNATTENDED DUE DATE: 950322 DAYS OPEN: 77 ACTION PNDG: AP: DRP/PB2 RESIDENTS PERFORM INITIAL CilECK FOR REQUIREMENTS OVERALL RESPONSIBILITY IS ASSIGNED TO DRSS/NMSS RSP DIV/LCA: DRP.PB2 DRSS/NMSS ACTION DUE DATE: DUE: ALLEGAT10N: Hil-95-A-0002 IACILIlY: SI LUCIE I DATE RCVD: 941222 SUBJECI: CI ALLEGES TilAT WELD REPAIRS CONDUCTED, A COUPLE OF YEARS AG DUE DATE: 950322 0, ON Tile TURBINE COOLING WATER AND COMPONENT COOLING WATER DAYS OPEN: 89 ACTION PNDG: AP: DRS/ENG INSPECTION COMPLETED. PROVIDE CLOSURE TO EICS RSP DIV/LCA: DRS/ENG ACTION DUE DATE: DUE: 03/15/95 ALLEGATION: RIl 95-A-0026 IACILI1Y: SI LUCIE I DATE RCVD: 950227
SUBJECT:
OlG REFERRRAL - ALLEGED DISCRIMINATION FOR REPORTING SAFETY DUE DATE: 950527 CONCERNS DAYS OPEN: 22 ACTION PNDG: AP: 01 OPEN CASE RSP DIV/lCA: ElCS 01 ACTION DUE DATE: DUE: ALLEGATION: Ril 95-A-0028 IACILITY: ST LUCIE I DATE RCVD: 950308
SUBJECT:
ANONYMOUS - WORK BEING CONDUCIED UNDER PWO 69/4185 BY DUE DATE: 950608 CONTRACTOR IS NOT IN ACCORDANCE WITil SITE'S WEDLING PROGRAM DAYS OPEN: 13 ACTION PNDG: AP: DRP/PB2 RESIDENT INSPECT AND DOCUMENT RLSULTS IN IR RSP DIV/LCA: DRP/PB2 ACTION DUE DATE: DUE: m v e -. e-.-n
. 79 PAGE
~
SI. tutIL 0 PEN ALLEGATIONS AS Of 03/21/95 ALLEGATION: Ril-95-A-0033 FACillIY: ST. LUCIE & TURKEY POINT DATE RCVD: 950312
SUBJECT:
ANONYMOUS - THE CONTRAC10R Wil0 WON Tile SECURITY AND flRE DUE DATE: 950612 WATCil SERVICES IS NOT COMPETENT DUE TO QUESTIONABLE SECURITY DAYS OPEN: 9 ACTION PNDG: AP: DRSS/NMSS (TOBIN) INSPECT RSP DIV/lCA: DRSS/NMSS ACTION DUE DATE: DUE: i 6 - - - - - ,w- .------ --. - _ - -- --. - __.- -.- ------ - -----.--.-- -- - .- -- - _ - - -__
? o ATTACHMENT 3 NRR OPERATING REACTOR ASSESSMENT NRR ASSESSMENT FOR ST. LUCIE October 1994 l CURRENT ISSUES l 1
-Seismic qualification of electrical and mechanical equipment (GL 87-02, USI A-46) issue on Unit 1 is still not resolved. The staff issued a letter in early 1994 providing a general framework of criteria which would resolve this issue. FPL responded in May 1994 r'estating their previous' position and stating that they believe that further NRC reouests for work, evaluations, or ciant changes would l provide no additional safety benefit to their nuclear facilities. The staff is l considering performing a backfit analysis to determine the possibility of ordering i FPL to implement additional . actions or accept the licensees position. A third )
alternative being evaluated is performance of a site inspection'to determine if any ; safety-significant issues exist in the areas of disagreement. l 1
-Unit I will be replacing steam generators in 1997. The licensee is well into planning for the event. '
.-An alternative approach to the resolution of the Thermo-Lag issue was proposed by i FPL, however, the staff did not pursue review of this performance based approach i based on Commission direction of this issue. The licensee.is scheduled to submit to {
the staf ~ by early November 1994 a schedule and method for resolution of the Thermo- < Lag issue. l 1
.The plant continues to perform well. The latest SALP evaluation had ratings of 1 i in all categories. )
Contact:
Jan A. Norris 504-1483
1 4 1 i Marcn 15. 1995 ST LUCIE Integrated Plant Performance Review 2 Current Plant Status Unit I has been operating at power since Mar.a 8. 1995. ~he unit was returned to power follow ;g a short notice outage for pressurizer code safety valve replacement. The next refueling outage is scneduled for April 4, 1996. Unit 2hasbeenoperatinoatpowersinceFebruary25.j995. he unit was returned to power rollowing a reactor tr1p aue to sow steam generator water level.resulting from a level transmitter failure. The next refueling outage is scneduled for September 25. 1995.
- 3. Manaaement The current organization nas been effective since Septemoer 1. 1994:
C. Plant Performance
- The units experience an unusually high number of reactor trips (4) between March 94 and June 94. However, during the past nine months, the units have experienced two reactor trips, two unit snutdowns, one unit j
off-line for maintenance and one loss of shut down cooling event. . Operator performance was good during the trips and shutdowns. July 8. 1994 - Unit 2 was taken off line (Mode 2) when 231 RCP lower oil level indication failed, incorrectly showing leak. l July 14. 1994 - Unit 2 was shut down to allow repair of a-stuck-closed reactor trip circuit breaker. October 26.1994 - Automatic trip-of Unit I due to arc-over from a potential transformer in the switchyara which resulted in a loss of load. February 21.1995- Automatic trip of Unit 2 when a SG water level transmitter failed high. The failure resulted in a i closure of a FRV, starving the SG for water and leaaing to a trip on SGWL. February 27,1995- Unit I was shut down to replace pressurizer code safety valves, which had been leaking by the seat since restarting from the Unit 1 outage in November ; Marcn 4. 1995 - Unit 1 experienced a loss of snutdown cooling for approximately 14 minutes when a hot leg suction valve to the operating shutdown cooling train closed. Root cause has not been established: however, operator errorinmanipulatingvalvesisthemostlikelycause,j
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ST. LUCIE I " Operational Period April 5,1994 through March 2,1995 1 2 3 4 100 l' i E go - , o, a..,e t. . ,,,s. unit . ......a.t .o... ,,,,,.d thunderstosa as a result of aluminum flashing from en adjecent d......,. R building being blown across two pleases transformer output. The reactos aften of the oss taken la main critiest on June 0 main transf ormer repaare C and returned on line on June 11 wer- <<-eieted. E a. "- t i ww-- -<> >> - i> >'- 2>
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2 lates on the same date. N leek The manit wee returned to newer Ore Octoter 26 1994 Unit 3 aut tendt itally t r ipped due f utraildup ar c-over of a switthyard potenitial transtoneer due to salt on insulators. The unit then ses:sined shut down for a: 4h - refueling outage. Unit I was tentarte.1 on Pto.emier 29 and j) placed on line on Decend>er I. e use telaussy 27 theit 1 was sinnia.e l l y ut.ut down tu sepala leaking
,we unit u.. re.ta ted oo pre..u. ire. ...ety e.i.e..
ggi 20 - It R i 0 mmmmlimumimm ritimillmnmimmmimmmimmmmmmmmmilmillmmitimmmmummmimmmim11m N D .I F M M .I .) A S O A PERIOD OF OPERATION Graph does not include power reductions . for routine repairs, waterbox cicaning, or required repairs.
I
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ST LJCIE
# of ISSUES vs TIME 25 - - - - -- -
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ST. LJCIE ISSUES BY ROOT CAUSE AND SALP AREA 20 18 m 15
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2 2 UOPERATlONS
- k. A Id I$ b b o/ / s ***~ *'
0 0 0 0 PLANT SUPPORT 1 6 4 10 1 0 MAINTENANCE 8 1 1 0 OPERATlONS 1 3/94 THRU 2/95
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3/16/94 Engineering jNRC Regional inspector haa two Unit 2 SL4 corrective i violations: 1) corrective action for action an 11/24/92 water hammer event was done without cocumented instructions or procedures. resulting in operating until 3/94 with five snubbers on the SRV and PORV tailpipes inoperable. 2) Failure
, to write a nonconformance reoort for a i damaged pipe suoport in Maren 1994. (IR 94-08) l 3/28/94 Personnei Licensee Unit 1 auto reactor. trip. Maintenance l error foreman opened generator exciter breaker
- on wrong unit. Operators had
, clearance on Unit 2. ',IR 94-12. LER U1
, 94-03) 4/3/94 Operations ! Unit 1 auto reactor trip from 19% power proceaure l Licensee while deenergizing the 4160 Volt non-error (Lack vital bus for SU Tx output breaker of maintenance. With the A emergency bus sufficient on the EDG. which was running at a depth in different frequency from the grid, the review of paralleled CEA MG sets developed procedure circulating currents and tripped sevetci change) reactor trip circuit breakers. A partial reactor trip tripped the turDine, which trippea the reactor. (IR 94-12. LER U1 94-04) 4/3/94 Surveill. Licensee During testing for Unit 2 moaifications error '
, the licensee discovereo that the 4160 V
; [AB Bus) swing aus components [C ICW Pumo ana C CCW Pumpj would not strip from the Dus upon unoervoitage if the
] bus were alignea to the B bus. A missing w1re in the swittngear (from initial construction) was the cause.
(SL4, Inacequate Corrective Action for 1992 NRC VIO for inaceouate surveillance I test - IR 94-12) 4/7/94 Maintenance lNRC Contractor personnel maae and contractor l error QC accepted pressurizer nozzle weld prep that did not meet procedural requirements for bevel angle. Licensee .
. engineering had specified overly tight l l tolerances. (SL4 94-10-01) j 4/21/94 Op ator Licensee ' Unit 2 reactor ::ower "ncreasea from 26 !
inatlentive to 31% due to positive MTC ano. operator inattentiveness. t ness (IR 94-12) ,
[ .
- a l j ,>
i a : I 7/14/94 Equipment I Licensee During surveillance test. TCB 5 failed failure- to open. It had stuck shut. A broken ! piece of bakelite had fallen into the j trip mecnanism. The unit was restarted and placed on line on July 15. 1994. (IR !' 94-15. LER U2 94-06) 7/14/94 Operations NRC Operators did not follow Unit 2 Tech , error Spec LCO time reauirements regarding i shut down on July 14 to allow repair of
- l a stuck-closed trip circuit breaker. i (SL4 94-15-01) ;
l 8/12/94 Operations / NRC The licensee was unloading new fuel for i Maintenance Unit I with a hoist graople that was ! error missing the safety latch sleeve locating i pin. The safety sleeve functioned by , friction only. (IR 94-18) ' 8/28/94 Equipment Licensee Unit I was taken off line (Mooe 2) to failure repair a DEH leak. The un1t was
- returned on line later the same day. !
(IR 94-20) 8/29/94 Operation.; NRC Operators placed 1A EDG in an electrical ; errors lineup for which TS-required . surveillance tests had not been ; performed (with the safety-related swing l bus powered from it). Also. related control room log entries appeared to be inaccurate. Inadequate corrective i action to NRC violation regarding Surveillance Testing. (SL4 94-22-01) Imorocer modification of controi room loos. ' SL4 94-22-02) l 9/30/94 Operations. NRC Plant personnel not trained on IPE and maintenance not using it for work planning and 1 deficiency scheduling. (IR 94-20) l 9/30/94 Operations NRC During requal exam, a licensed operator weakness exhibited an apparent disregard for
,, E0Ps. (IR 94-19) 9/30/94 Operations Licensee Failure to notify the NRC of changes in error status of licensed operators' medical conditions. (NCV 94-19-01) 10/26/ Storm Licensee Unit 1 automatically triopeo due to arc- !
94 over from a potential transformer due to salt buildup on switchyard insulators. l (IR 94-22)
1
.* l -4 e
6 1 2/16/95 ' Maintenance Licensee Loaa snea of the IA3 lE '160 cus cue to I
! Error / inaavertent jumper contact.while 1 ; Proceaural replacing a aegraaea voltage relay. (IR , ! Wea.. ness I 95-04) l t
2/20/95 Eauipment Licensee 2B LPSI pumo founa air-couna during Anomaly surveillance testing. The licensee has theorized that the n1grat1on of air in the system resuitea in the concition as i a result of orevious surveillance
! testing. "he pumos are not seif- ,
l I i ventino. (95-04) ! l 3 2/21/95 Eauipment Licensee Unit 2 trip due to failure of a SGWL l control level transmitter. Transmitter i l Failure l' faii.ed high resulting in closure of the ) FRV ano a suosecuent tr1 on iow SGWL. 1 I I(95-04) 2/27/95 Eau 1pment Licensee Unit I was shut down for the replacement Failure of 3 pressur1zer code safety valves.
'The valves were leaking oy the seat. (IR 95-04) 3/4/95 Apparent Licensee Unit 1 experienced an approximate 14 Personnel minute loss of shutdown cooling while Error shifting from one shutdown cooling loop-to the other. The apparent root cause i is the closing of the wrong SDC suction l isolation valve (the valve for the j operating, vice 1dle, pump) on the part ,
of the operator. The enerator continues ) to aeny nis involvement. althougn SOER l ana other aata strongly support the 1
.I I I conclusion. (IR 95-07 ?enaino) l ! !
i l l I I I I NRC CONCLUSION: he mispositioned valve and water hammer occurred over a year ago. Personnel errors generally reflect inattention to detail. While events in the last year inaicated declining performance, the numoer of events has i While it is too early to concluae that the
~
decreased in the last quarter. Jecline nas abatea. recent management actions appear to be having positive affects on corrective actions, maintenance, ano operations. Further very
-iose inspection ano assessment is reautred.
l l l
b'= s . h. i ROI 2210, REV. 3 ENCLOSURE 5
- INSPECTION RESULTS AND SALP INPUT l Summary of inspection Results
{ Plant Inspected : St. Lucie Dates: February 7-11,1994 4 Inspectors - Lead: Fred Wrioht
- larry Cohen i Bill Sartor i
Glen Salvers Safety Objective: (Provide a brief statement of the safety objective of this inspection) This routine, announced inspection involved the observation and 4 evaluation of the annual emergency preparedness exercise. Emergency organization activation and response were selectively observed in the licensee's Emergency Response Facilities including: the Simulator Control
- Room; Technical Support Center; Operational Support Center; Joint Information 3
Center and Emergency Operations Facility. The inspection also included a review of the exercise scenario and observation of the licensee's post exercise critique. The exercise was a full participation exercise involving State and County participation. Offsite activities were evaluated by FEMA. The exercise was conducted on February 9, 1994, between the hours of 7:00 a.m.
- and 3:00 p.m. '
! Results: In the areas inspected, violations or deviations were not I i identified. Two Inspector Fol_ low-up Item was identified to: 1.) Review _ j licensee emergency plan implementing procedures and training concerning the
- definition of containment failure as it relates to emergency classifications j and protective action recommendations (Paragraph 10); and 2.) Observe a site
! accountability drill in 1994 (Paragraph 10). Overall, the licensee j demonstrated the capability to implement the Emergency Plan in the event of a i radiological emergency. I 4 I l 4 a
I i ROI 2210, REV. 3. ENCLOSURE 5 I SALP EVALUATION i If more than one SALP function area is involved in this inspection. then ! , complete this form for each functional area. l SALP Functional Area Inspected: Emeraency Preparedness-
)
Key: 0 - Not Inspected; 1 - Above average; 2 - Average; 3 - Below average EVALUATION CRITERIA ASSESSMENT , 1 l a. Assurance of quality, including management involvement . and control; 0 1-V 3
- b. Approach to the identification and resolution of technical issues from a safety standpoint; O1y3 )
- c. Enforcement history; 01J3 !
- d. Operational and construction e'ents (including response to, j analysis of, reporting of, and corrective actions for); V123
- e. Staffing (including Management); and 01V3 ,
- f. Effectiveness of training and qualifications program. 01J3 '
COMMENTS: Provide a brief (2-3 sentence) summary of the licensee's performance as observed in this inspection, that can be used in the next SALP Report. Overall, the exercise demonstrated an effective capability to implement the ' Emergency Plan and implementing procedures. In the areas inspected, violations, deviations or exercise weaknesses were not identified. See Attachment. LEAD INSPECTOR: F,N. Wriaht DATE: Noveiber 22, 1993* DISTRIBUTION: - Inspector's Branch Chief W.E. Cline Inspector's Section Chief K.P. Barr Project Section Chief (for reactor inspections) K. Landis DRS Division Director A.F. Gibson Original (In folder with cover letter) 4 a - , , . . ~ , - - . , . , , , - . ,. .- -
4 . ! i e
- .ll l St. Lucie Annual EP Exercise 1994
- ' Exercise Strengths
- i
)
o T Players worked hard and well as a team throughout the Emergency Response ; Organization, during the exercise, i o Command and Control in the SCR, TSC and OSC as evidenced by frequent ! briefings, establishing priorities and noise and attention discipline of staff. . l o Licensee improved exercise controls with use of a Scenario control \ center with radio communications to key exercise controllers. ! o Licensee has improved the response capabilities of the licensee's staff i reporting to the Emergency Operations facility. ) Areas For Improvement: 4 l 50-335, 389/94-04-01, IFI: Review licensee emeroency plan implementing procedures and training concerning the definition of containment failure as it relates'to emergency classifications and protective' action recommendations. 1 (Paragraph 10). 50-335, 389/94-04-02, IFI: Observe a site accountability drill in 1994. (Paragraph 10) Licensee's program for personnel accountability appears to.be l i very weak and the Licensee has not tested it since 1989. ' Other areas for improvement: i o Licensee does not have forms for NRC notifications out of the E0F. o Licensee State Notification Form does not include a review / approval. ' space to indicate message approval by Emergency Coordinator or Recovery Manager. O Licensee pre-stages licensee media personnel in order to respond to local new agencies during drills. Licensee should attempt in an exercise to provide media information during period when Emergency News Center is responding to E0F, c Licensee does not have minimum expectaticns defined for exercise objectives. O Licensee does not define minimum time requirements for activating ERFs. o Do'se assessment procedures have been waiting for Crystal River implementation of EPA 400, scheduled for March 31, 1994. Licensee procedures should be reviewed and should clearly define all methods and procedures used to determine radiological releases (Manual and Computer) for ERO use on and off-site, o Licensee Dose Assessors, Emergency Coordinator, and Recovery Manager
~ - _
l
- 9 1 3 procedures should address need to consider extended offsite dose i
projections for periods longer than next 2 hours when controls for radiological releases have not been maintained.
]
o l 3 Licensee needs to improve methods for prompt identification and tracking ' of Emergency Response Team members. Action on Previous Inspection Findings (92701)
- a. (Closed) IFI 50-335, 389/92-01-02: Calculatina radioloaical release dose oro.iections. (Paragraph 9)
In the 1992 annual Emergency Preparedness exercise inspectors determined that the TSC dose assessment team did not provide useful or timely dose projection estimates to the EC. During the 92 exercise both the TSC and EOF dose assessors had responsibili' ties to performed offsite dose assessments. The TSC dose assessors were responsible for making all offsite dose assessments until the EOF was activated. Once the EOF was operational it acquired the responsibilities for calculating dose l projections for any release and the TSC dose projection staff's ' responsibility was to perform parallel calculations to validate the EOF staff's calculations. A simulated radiological release began after the TSC and EOF facilities were both activated. The TSC dose assessors completed calculations about 22 minutes after ! the EOF dose assessors had completed their calculations. ' The inspector learned that the TSC dose assessor had keyed in a wrong value into a computer program that had cause a different dose projection estimate than that determined by the E0F. The dose assessor decided to recalculate the dose calculations which caused a delay. Following the exercise the inspector discussed the dose assessment delay with licensee representatives. The licensee agreed to review the event to determine the need for additional training or procedural guidance for the TSC dose assessors and the item was identified as an IFI. To address the problem the licensee developed a training brief to all Chemistry Department (dose assessors) personnel for review TB 9208167-S-01, "Off-Site Dose Calculations-Computer Method Training Brief". The brief provided procedures and guidance for using the computer bases process to derive calculated off-site doses in a manner similar to the procedure for manual calculation. During the 1993 annual EP exercise the inspector identified additional concerns with dose assessment activities. During the
- 1993 EP exercise dose assessors did not calculate release rate for a simulated release from containment. The dose assessors utilized a LOCA cose assessment program which used a design bases leakage rate for calculating offsite dose rates. The leak rate was much larger than a design base leak rate. Failure to calculate the release rate for the simulated conditions and factor it into the offsite dose programs resulted in an underestimate of offsite dose rates by about two to three orders of magnitude. To correct the
l l
/ :
problem identified in 1993 the licensee developed several methods i to calculate a burp release and provided " Containment Leakage and i Emergency Dose" training on the procedures to staff dose assessors. The inspector reviewed the licensee's procedures and subject training records concerning the issue. The licensee's activates appeared appropriate for improving dose assessment l capabilities. The inspector stated that the item would be closed. '
- b. (Closed) IFI 50-335, 389/93-16-01: Ensurina consistency of ;
scenario technical data.. This issue concerned inconsistencies in scenario technical data during the 1993 EP exercise and licensee efforts to ensure the congruity of scenario data. In the 1993 exercise the offsite dose projections differed from the associated field monitoring values by two or three orders of magnitude. To prevent a similar occurrence the licensee assembled a qualified scenario development team which spent more time verifying exercise data than in previous years. i Inspectors observed that the offsite dose rates measured by offsite field monitoring teams generally agreed with the offsite ^ dose rates calculated by the dose assessment staff during the 1994 1 exercise. No other significant technical data issues were I identified by the inspectors. The inspector stated that the item would be closed.
- c. (Closed) IFI 50-335, 389/93-16-02: Reviewing REP and EPIP quidance and reouirements related to provision of periodic olant status updated to offsite authorities durina an Alert or hiaher emeroency classification. This item concerned licensee's procedure i and requirements for making periodic updates of emergency conditions to State and local agencies during an event.
Licensee EPIP-3100031E, " Duties and Respansibilities of the i Emergency Coordinator" specified that offsite authorities would be ) updated regarding any significant changes in plant conditions and ' that appropriate notification forms would be used for all updates. However. during the 1993 exercise the licensee failed to make a notification to offsite agencies that a plant shutdown was made following the Alert classification. State of Florida representatives in De EOF expressed concern that no updates had been provided to tht State between the period of 4:12 a.m. and 6:25 a.m. In accordance with NRC and industry guidance the licensee's REP 1 required periodic updates. However the EPIPs did not establish l any fixed or variable time interval that would implement the commitment. Licensee management committed to review the applicable guidance and requirements, and to discuss the matter with State representatives, in order to ensure at a minimum that all parties clearly understand what was meant by "significant changes" in plant conditions. The licensee discussed the periodic notification needs o'f the State with state representatives and revised EPIP-3100021E to
,, . . . . - - ~. . . -
- n O
i p I require updates be provided, by notification forms, every 60 i' minutes or upon any "significant changes" in plant status. Several examples of "significant changes" were also included in the text of the procedure. The procedure allowed for changes in i the frequency of updates following agreements by offsite authorities. The insp?ctor stated that the item would be closed. !
- d. (Closed) IFI 50-335, 389/93-16-03: Evaluatino orocedural recuirements with respect to conveyina information to offsite authorities reaardina classifiable incidents of lesser severity occurrina durina an extant emergency classification. The IFI concerned a statement in EPIP- 3100021E:
"If one unit is in a classification event and the same or the other unit enters into an event where the same or lesser emergency class would apply, a new classification should NOT be declared. The event should be issued as a update at the earliest practical time. No regulatory time limits would apply to the update."
The inspector reported that the last sentence in the procedure guidance referenced above did not appear to be appropriate i guidance based upon any specific commitment or specification in the REP. Licensee management committed to evaluate the procedural , guidance. The licensee removed the last sentence from the ' referenced guidance in revision 30 to EPIP-3100021E, dated September 28, 1993. The inspector stated that the item would be closed. '
- e. (Closed) EW 50-335, 389/93-16-04: Failure to activate the EOF in a timely manner. In the absence of any licensee criteria for timely activation of the EOF the inspector used the guidance in Supplement 1 to NUREG-0737 to formulate acceptance criteria ;
relative to the timeliness of E0F activation. The cuidance l specified that the EOF is to be staffed by a des'gnated i senior I licensee manager and appropriate technical staff within one hour (item 8.4.1.1). The EOF was declared operational 109 minutes after the Alert declaration ano an exercise weakness identified. In a letter to the NRC dated September 1,1993 the licensee reported the following corrective actions:
- The current Recovery manager operations Advisors at St.
Lucie have been trai.ned to perform as the Interim Recovery Manager until one arrives at the E0F. 3 The auto dialer will no longer be used in the " notify" mode for drills. l l
- An unannounced activation drill was conducted after normal !
working hours for the initial E0F responders on August 26, ! 1993. :I I The licensee's corrective action letter reported that, following the implementation of the above corrective actions, the EOF was
1 J- ~0 staffed in 37 minutes of the decision to staff the facility. In the 1994 exercise, the decision to activate the E0F was made at about 9:lf a.m. - following the declaration of an Alert emergency classification at 9:05 a.m. The E0F was minimally staffed by 9:56 a.m., approximately 41 minutes following decision to-staff the EOF and 51 minutes following the Alert declaration. The EOF was activated following a turnover, from the EC in the TSC, at 10:28 a.m., approximately 73 minutes following the decision to activate the EOF with initial EOF responders. The inspector reported the observed response during the 1994 exercise was a significant emergency response program improvement and the item would be closed. 4
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p - . 1 l*,J' ST. LUCIE MAINTENANCE EVALUATION i A. l
. Assessment: Adverse trends were noted in maintenance. The EDG problems indicate q a weakness in EDG Maintenance. 'Ihe number of problems related to personnel '
errors and procedure problems may indicate possibly attitude problems. Also, Safety j Equipment performance has failed to meet the industry average on all safety system. 1 Maintenance performance has declined. B. Basis:
- 1. Adverse trends were noted on the Site Integration Matrix. ;
1 Recent Equipment Failure: 01/23/96 - Elect. arc during maint. caused loss of 25% of control room annunciators 12/20/95 - Pitting of reactor flange o' ring groove : 11/06/95 - Failure of EDG 2A relay sockets i 10/05/95 - EDG IB fuel oil leak at threaded connection ' l 09/20/95 - EDG 1 A/1B governor control problems resulted in load oscillations l Recent Personnel Errors: 01/01/96 - ICI wiring error during Rx head installation 08/31/95 - Damaged cy. head on IB EDG due to loose lash adjustment 08/09/95 - Inoper.ble Unit 1 PORV due to maintenance error / testing inadequacy Recent Procedure Problem: 12/09/95 - 2A2 RCP seal destaged due to inadequate or weak procedure 09/15/95 - Failure to have clearance for work on cond. water box 08/25/95 - Failure to sign off procedure steps as work completed
- 2. Safety Equipment Performance (Availability %) has been below industry a average.
Actual Industry Averages Unit 1 Unit 2 AFW 99.5 97.8 99.6 - 99,I HPSI 96.0 98.8 99.7 - 99.2 EDGs 95.6 99.8 99.1 - 98.6
- AFW wiring error on 2B AFW
- 2C AFW steam admission valve did not open and mechanical trip linkage problem
- HPIC 2B Breaker Failure
- 3. Numerous equipment failures have Caused power reductions during the last 6 h
_. , , - ~
.j a U e months. 10/95 Unit 1 Heater level Control
- 10/95 Unit 1 IB Heater Drain Pump 11/95 Unit 1 FW Reg. Viv. Control 11/95 Unit 1 1A Main Transformer
- 01/96 Unit 1 FW Viv.13V-24-A2 08/95 Unit 2 Heater Drain Valve
- 08/95 Unit 2 Heater Drain Pump
- 08/95 Unit 2 Cire. Water Valve 08/95 Unit 2 Htr. Drn. Pmp. PCM 01/96 Unit 2 Hydrogen Sys. Prob.
01/96 Unit 2 MFW Pump C. Future Inspections: . Assist the Resident inspector with the routine Resident Maintenance / Surveillance Inspections program. Conduct Regional Initiative : inspection, focus on outage activities BOP and EDG maintenance, procedure adequacy { and safety system performance. Perform maintenance rule inspection in September 1996. Conduct the ISI inspection. I i i e
ST. LUCIE MAINTENANCE EVALUATION - February 1996 l
,1 A. Assessment:The following adverse trends related to mamtenance were noted for 1995: (1) 9 cases of equipment failure 3 of the 9 for i
EDO equipment, (2) with 10 cases of personnel error, ar.d (3) 13 cases failure to follow procedures, inadequate procedures, or weak procedures The EDG problems may mdicate a weakness in EIX) hiaintenance The number of problems related to personnel errors and procedure problems mdicate weaknesses, possibly attriude problems. in the hiaintenance Program. Also, Safety Equipment performance ;
~
has failed to meet hcensee goals for Umt I (Numbers provided are averages for 1995). B. Basis:
- 1. Last SALP Rating: Category 2 (Vs94 I'5 96) Nest Period End: 04 15 97 l
- 2. Staintenance Trend Charts:(Site integration hiatnx)(Jan.1995 - Dec.1995) fVI m i n te n a n c e' I a s LJ e a i i
- . . a o. a . .
. j
-em ce r IE ra.
~
l a .s ,- - -. .- l
- 3. Equipment Failure Examples 01 '23 96 Elect. arc durmg maint. caused loss of 25*. of control room annunciators 12'2095 - Pitting of reactor flange o'rtng groove i 19695 - Failure of EDO 2 A relay sockets 10'0595 - EDO 1 B fuel oil leak at threaded connection 092095 - EDO INI D governor control problems resulted in load oscill.
Personnel Error Esamples 01'01195 - ICI wiring error durmg Rx head installation 08'3195 - Damaged cy. head on 1!! EIXI due to loose lash adjustment 08'0995 -Inoperable Umt 1 PORY due to mamtenance errortestmg inadeq Procedure Problem Examples 12 09 95 - 2A2 RCP seal destaged due to inadequate or weak procedure 091595 Failure to have clearance for work on cond. w ater box 08/25 95 - Failure to sign off procedure steps as work completed
- 4. Saf* Equipment Perfonnance(Asallability %)(1995 Astrage):
Actual Goal t' nit 1 Unit 2 AFW 99.5 97.8 97.9 IIPSI 96.0 98 8 97.7 EDGs 95.6 99 8 98.4 AFW Wiring error on 2BAFW, *2C'AFW steam admission valve did not open and mechanical trip hnkage problem. IIPlc - Ilreaker Failure 211.
- 5. S/G hiatus: SG "l A" 13.3*e plugged,50 "l!!" . 9. I'e plugged,50 "2A" 3.2*. plugged. SG "20" 2 49. plugged - A S 'O replacement for Unit 1 is planned in the Spring 1998.
- 6. Equipment Failures That Caused Power Reduction in the Last 6 %1onths:
Umt 1 10 plus 14 equipment failures during startup after hurncane ; Umt 2 - 10 l Date l'ait Equipment Date Unit Equipment 10 1395 I Excore Nuc. Instr. 10'1495 I litr. level Control 10'1595 i Excore Nuc. Instr. 10'1595 I Safety Valves 10:1795 1 IB litr. Dm Pmp 10 23 95 I ASI Control Il'1695 i FW Reg. Viv. Cont. 11tS95 1 Water flox Cleaning 11/2195 l 1 A %1am Trans. 01/1996 I FW Viv. LSV 24-A2
. 0823 95 2 litr. Drn. Valve 082395 2 Water Dox Cleaning !
08 2595 2 litt Drn. Pmp. 082795 2 Circ Water Valve I 08'2895 2 litr. Drn. linp. PChi 09'l496 2 Water llox Cleaning 092595 2 Water llox Cleanmg 01/0$'96 2 Ilydrogen Sys. Prob. 01'08 96 2 hlFW Pump 01/1196 2 Excore Nuc. instr. C. Future Inspections: Routine Resident hiaintenance survedlancs inspections, assist the resident core mspections with one inspection per quarter (2-man weeks) Add 4-man ueeks ofinspection m the maintenance aren (Regional Initiative), focus on outage actmties HOP and f) EDO mamtenance (2-man weeks). procedure adequacy an.d safety system performance (2-man weeks) Perfonn manaance rule y ' inspectmn (September 9-13.1996) Conduct (1 man week)ISI(core mspectmn). i L
ST. LUCIE PERFORMANCE REVIEW D 11. Function *f Are Assrssm nts
- 1. Plant Support
/
A. The last SALP cycle ended 1/6/96. Plant Support was Category 1. The licensee continues to maintain a satisfactory level of performance in the area of Plant Support. Some decline in Radiation Protection has been noted due to the loss of control of contaminated tools and exceeding dose goals. Insufficient information is available at this time to assess the Chemistry. Effluent, Waste. Transportation and Emergency Preparedness programs. Hurricane preparations for hurricane Bertha were conservative Overall, site secunty has been adequate. Training and qualification noted as a strength and management observed to be aggressive en pursuing issues, but not aggressive in doing indepth review of events. Implementation of the fire protection program continued to be satisfactory. Fire protection activities were primanly monitored by the resident inspectors dunng the PPR period. B. Basis Radiation Protection NCV for failure to control contaminated tools used in RCA (96-04, p 45) Violation (repeat of above NCV) for numerous examples of failure to control contaminated tools. (E09, p 25) Internal and external exposures below 10 CFR Part 20 limits. (E04, p 45 and 96-04, p 23) (1996 dose levels????????) 1995 dose was 412 person-rem. Unplanned maintenance and rework caused 1995 dose goal of 283 person-rem to be exceeded by 129 person-rem. (96-04, p 50) Rad Techs decreased from 32 to 30 and 2 supervisors lost (E04, p 48) a Decon staff reduced from 22 to 12 persons. Levels of contaminated equipment and materials increasing (96-04, p 46) Good radiologica! housekeeping and controls. (96-09, p 28) The total area contaminated was at 250 ft'. (96-04, p 47) Licensee accreditation of the FP&L DADS a good examp;e of Radiation Protection staffs technical capabilities. (96-04, p 44) Emeroency Preoaredness Conservative actions taken to prepare for Hurricane Bertha (E11, p 3) Secunty Failure to report a confirmed tampering event within one hour, which resulted in a violation. Two events in prior to the above tampering event were documented as tampered or unauthorized work, but management failed to notify secunty of these events. Numerous problems discovered by a QA audit determined the FFD program to be weak. 1 Fire Protection
!1 A backup fire pump was installed to replace an out of service larger fire pump without an engineering evaluation.
C. Recommended Inspection Effort insoections Rationale Health Physics (SALP 1 decline maintain; watch) Operational HP(83750) 2-inspections with focus on procedure compliance; rework doses Effl/RadWast(84/86750) 3-inspections with focus on accident / process monitor installation & maintenance Tl 133 Rad Waste Combine with 86750 Emergency Preparedness EP Prog (82701) 1 inspection with focus on Self A,ssessment results 1-Reg Init. inspection on allegation followup (2 weeks) , Secunty . l Secunty Prog (817U0) Core insp to review secunty audits, corrective actions, management support and effectiveness, and review protected area detection equipment. Sec. Prg/FFD (81700/81502) One regional initiative to followup on tampenng and FFD issues. t VBS (Tl 2515/132) Inspection of Vehicle Barner System TI Fire Protection None
S; int Luci) in:pectiJn Activiti.s During The A ctsament P;riod 02/01/96 09/30/96
% IR 9b-01: period 01/07/96 - 02/17/96: Issued 03/18/96: IP 71750 1
R:sident Report, Radiological Protection topics (71750) were reviewed by residents. IR 96-03: period 01/26/96 - 01/30/962 i ssued 02/22/96 Special inspection of Over Dilution Event of January 22,1996 IR 96-04: period 02/18/96 - 03/30/96: issued 04/29/96: IPs 71750 and 83750 R:sident Report, Emergency Preparedness topics (71750) were reviewed by residents. PSB Inspection, Radiological Protection (83750) topics were reviewed by Region it inspector IR 96-05: period 04/30/96 - 05/03/96; issued 05/31/96 Security inspection by Region 11 inspector .* JR 96-06: period 03/31/96 - 05/11/96: issued 06/07/96: IPs 93702 R;sident Report, Emergency Preparedness topics (93702) were reviewed by residents. IR 96-03: period 05/12/96 - 06/08/96: Issued 07/08/96: IPs 71750 Rrident Report, Radiological Protection topics (71750) were reviewed by residents. IR 96-09: period 06/09/96 07/06/96: Issued 08/05/96: IPs 71750 and 83750 Rxident Report, Fire Protection topics (71750) were reviewed by residents. PS3 Inspection, Radiological Protection (83750) topics were reviewed by Region 11 inspector, IR 96-11: period 07/07/96 - 08/03/96: Issued 08/05/96: IP 71707 Resid nt Report, No Plant Support Section in report, preparations for hurricane Bertha (71707) were reviewed by residents. IR 96-12: 07/12/96: issued 08/26/96: Speciil 50.59 Inspection St. Lucie Violations 02/01/96 09/30/96 Dunng the assessment period three violations were identified. VIO 96-04-01, Failure to Follow Procedures lead to Unit 1 Containment Particulate-lodine-Noble Gas Monitor in-operabihty. This violation resulted in a Technical Specification violation during start-up. (IR 96-04; pages 14-16; paragraph 04.2) NCV 96-04-05, improper Health Physics Practices, Failure to follow licensee radiation protection procedures. The poor health physics practices identified as an URI in IR 96-01 later resulted in a NCV 96-04-05 (IR 96-04, pages 39 and 40, paragraph MB.1) NCV 96-04-06, Failure to Follow Contamination Control Procedures for the Control and Use of Contaminated tools in the RCA. (IR 96-04, pages 45 & 46, paragraph R3.1) l
h , ATTACHMENT RG II FIRE BARRIER PENETRATION SEALS (IN 88-04 AND 88-56) STATUS SITE INSP FOR IN'S INSP REMARKS-SATISFACTORY? i
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BROWNS FERRY YES YES LATEST 1992 l BRUNSWICK YES YES LATEST 1993 , t CATAWBA NO NO INF0 l AVAILABLE CRYSTAL RIVER NO PENE SEALS WERE INSPECTED IN 1985 t FARLEY YES YES AN URI WAS I ISSUED, PROGRAM ! WAS FOUND SAT 1 GRAND GULF NO TESTING WAS i REVIEWED IN 1985 HARRIS NO INSP WERE DONE 1 IN 93 & 94 BUT ! NOT SPECIFIC TO PENE SEALS HATCH YES NO LACK OF X-REF BETWEEN SEAL & TEST DATA- NO FOLLOWUP WAS CONDUCTED MCGUIRE YES YES LATEST INSP 1993 NORTH ANNA YES NO LICENSEE COMPLETING PROGRAM IN '94 NEEDS FOLLOWUP OCONEE YES NO TEST DOCUMENTS WERE INADEQ REPAIRS NEEDED - SCHEDULED TO BE COMPLETE IN '88 NO FOLLOWUP DONE ROBINSON YES YES FINAL REPAIRS WERE NOT INSPECTED ST LUCIE YES NO TESTED CONFIG OF PENE WERE NOT REVIEWED
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2 ; SEQUOYAH YES NO PROGRAM TO BE I COMPLETED 12/94 : NO FOLLOW DONE i SUMMER YES NO IN '87 & '88 INSP OF A SAMPLING 0F : SEALS l SURRY NO LICENSEE WALKDOWNS FOUND 1 SOME PROBLEMS ( NRC HAS INSP(4) i TURKEY POINT YES YES LATEST INSP 1992 l V0GTLE YES YES 1993 VERIFIED CORRECTIVE
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P k . p INSPECTION RESULTS AND SALP INPUT Summary of Inspection Results P1 ant Inspected: 8T, 4 "d-r i Dates: / > -/ f Jut y 9[ i Inspector (s) - Lead: [. [ 'tC4rd/d/od Other(s):
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Safety Objective: This routine, announced inspection was conducted in the areas of confirmatory measurements and the release of potentially radioactively-contaminated materials offsite. Results: The licensee demonstrated that an effective Count Room radiochemical analysis program was in place. The licensee had an effective program in place to control the release of material from the RCA.
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R01 2210, REY. 3, ENCLOSURE 5 SALP EVALUATION If more than one SALP functional area is involved in this inspection, then complete this form for each functional area. SALP Functional Area Inspected: eM T Iv#/e'4r ,eu revd 82crec r<o. Key: 0 - Not inspected; I - Above average; 2 - Average 3 - Below average l EVALUATION CRITERIA ASSESSMENT
- a. Assurance of quality, including management involvement and control; Ofh 3
- b. Approach to the identification and resolution of technical ,,
issues from a safety standpoint; ' 0;l 2 3 M
- c. Enforcement history 123 l d.
l Operational analysis of, reporting and construction of, and corrective events actions(including for); response to, h12 3
- e. Staffing (including management); and 0(1)23
- f. Effectiveness of training and qualifications program.
0}.23 COMMENTS: Provide a brief f 2-3 sentence) sunnary of the licensee's performance as observed in this insoect90n. that can be used in the next 5 ALP Report. rna t , . < ~s r c um) m ,w rn c v si. -c.e, a r, i nu c. .w., .e
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NOTE: Ple.ase attach additional sheets if needed for coments or for other SALP functional areas. i , LEAD INSPECTOR: / / "/' : r-DATE: - DISTRIBUTION: Inspector's Branch Chief Inspector's Section Chief Project Section Chief (for reactor inspections) DRS Division O'1 rector Original (In folder with cover letter)
( 4o i INSPECTION RESULTS AND SALP INPUT Summary of Inspection Results Plant Inspected: 6 T, A W/( Dates:_ft-Io 6c r M Inspector (s) - Lead: [ [O4#e od Other(s): Safety Objective: ' l l This routine, announced inspection was conducted in the areas of the l organization of the Chemistry Department and Radwaste Group, plant water chemistry, the Radiological Environmental Monitoring Program (REMP), the Post Accident Sampling System (PASS), the Control Room Emergency Ventilation System, .'adioactive naterials handling and transportation documentation, volume reduction of solid radwaste, the status of the Refueling Water Tank (RWT)-leak migration, contaminated sludge disposal, and the leaking Unit 2 Spent Fuel Pool (SFP) Ion Exchanger. Results: The licensee's organization of its Chemistry Department and Radwaste Grnup satisfied Technical Specification (TS) requirements. The licensee's plant water chemistry program continued to be effectively implemented. d The licensee had an effective program in place to analyze radiological effluents, direct radiation, etc. due to plant operations, as evidenced by the Radiological Environmental Operating Report. The licensee's PASS was capable of fulfilling its intended sampling function and the technicians had maintained their capability to operate the system. The Control Room Emergency Ventilation System was adequate for its intended function and was being maintained in compliance with the applicable TSs. The licensee's radwaste processing and shipping was conducted in a competent, professional manner and the radwaste shipping documentation was thorough and-in compliance with the applicable regulations. The licensee continued to make good progress in the reduction of its solid radwaste. The licensee continued to monitor isotope migration due to the RWT leak. v
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1 k i 2 The licensee had acted prudently on the issue of contaminated. sewage sludge disposal. The licensee had taken a proactive position in the resolution of the leaking Unit 2 Spent Fuel Pool (SFP) lon Exchanger. i l
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ROI 2210, REY. 3, ENCLOSURE 5 1 SALP EVALUATION If more than one SALP functional area is involved in this inspection, then complete this fors for each functional area. SALP Functional Area Inspected: d 4s - 8 Ne*# ' i Key: 0 - Not inspected; I - Above average; 2 - Average; 3 - Below aversge EVALUATION CRITERIA ASSESSMENT
- a. Assurance of quality, including management involvement ,s and control; 0(1,'23
- b. Approach to the identification and resolution of technical issues from a safety standpoint; 0123
- c. Enforcement history h123
- d. Operational and construction events (including response to, analysis of, reporting of, and corrective actions for); Oh23
- e. Staffing (includingmanagement);and Ofl[)2 3
- f. Effectiveness of training and qualifications program. 0(l',23 COMMENTS: Provide a brief (2-3 sentence) summary of the licensee's performance as observed in this inspection, that can be uses in the next SALP Report.
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O!STRIBUTION: Inspector's Branch Chief Inspector's Section Chief Project Section Chief (for reactor inspections) DRS Division D'1 rector Original (In folder with cover letter)
b 1' INSPECTION RESULTS AND SALP INPUT Summary of Inspection Results Plant Inspected: $7, 4eei Dates: 3 - 7 Ma n M
!nspector(s) - Lead: [.[64W#mJ Other(s):
Safety Objective: This routine, announced inspection was conducted in the areas of the - organization of the Chemistry Department and Radwaste Group, audits, primary water chemistry, the Annual Radioactive Effluent Release Report, Information Notice (IN) 94-81, the leaking Unit 2 Spent Fuel Pool (SFP) Ion Exchanger, monitoring of the Refueling Water Tank (RWT) leak migration, the status of the Unit I nitrogen system check valve installation, the Meteorological Monitoring Program, radioactive waste handling and transportation documentation, and contaminated sludge disposal. Results: The licensee's organization of its Chemistry Department and Radwaste Group satisfied Technical Specification (TS) requirements. The licensee's audit process was capable of identifying programmatic ! weaknesses and making recommendations for corrective action. ! The licensee's primary water chemistry continued to be effectively !
-implemented.
I The Annual Radioactive Effluent Release Report indicated that the licensee had met the requirements of the TSs and that doses to the public were minimal. l l The licensee had taken a very proactive position concerning the issues raised l by IN 94-81. l The licensee had taken a proactive position in the resolution of the leaking l Unit 2 Spent Fuel Pool (SFP) Ion Exchanger. The licensee continued to monitor isotope migration due to the RWT leak. The licensee had completed the installation of the double check valves and completed the upgrade of the Unit I nitrogen system to prevent future valve leakby potential. The licensee's Meteorological Measurement System was capable of fulfilling its equired functions. l
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} 2 The licensee's radwaste processing and shipping was conducted in a compe, tent, professional manner and the radwaste shipping documentation was thorough and in compliance with the applicable regulations. The licensee had succeeded in its efforts on the issue of contaminated sewage sludge disposal. I 1 j i l I O
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A ! l ,V i !' R01 2210. REY. 3. ENCLOSURE 5 i SALP EVALUATION If more than one SALP functional area is involved in this inspection, then 4 complete this form for each functional area. SALP Functional Area Inspected: #sr 3 uu,<w A 43 / A rec rio J ! Key: 0 - Not inspected; I - Above average; 2 - Average; 3 - Below average i EVALUATIONCRITERE ASSESSMENT ! a. Assurance of quality, including management involvement and control; {I 0([23 ! b. Approach to the identification and resolution of technical issues from a safety standpoint;
- 0[23
- c. Enforcement history h123 4
- d. Operational and construction events (including response to, l analysis of, reporting of, and corrective actions for); I
{. 0@23 I e. Staffing (including management); and @23 1
- f. Effectiveness of training and qualifications program. Od 2 3
- COMMENTS
- Provide a brief f 2-3 sentence) summary of the licensee's performance
! as observed in this insoectson. that can be used in the nest 5 ALP Report, j m in r w + ru c em , > .me r aa a nd rrpr' ro *'e1 ~- m +,? rw
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Inspector (s) - Lead: T.#, G Isr / / ~ Other(s): M T4 , Nu r/o NAN /d Safety Objective: insoectinn) (Provide a brief statement of %. sa fety obiective of +81s l This routine, announced inspection ofprogram the Radia
, external l-and contamination, surveys ard monitoring, and ma exposures As Low As Reasonably Achievable (ALARA).
Results: . The RP technician staff appeared knowledgeable and well trained. The licensee continued all exposurestolessimplement than 10 CFR effective internal and external exposure Part 20 limits. programs Program was considered to be adequate in identifying potential issuesThe Au Contamination control and overall housekeeping practices were also con adequate. . . Licensee efforts in the performance of ALARA initiatives in the area ofinrespirator strength meeting ALARA reduction goals. was determined by the inspector to be a prog Based on records review departments, the inspector found the radiation protect adequate in protecting the health and safety of plant employees. 4
j. l 401 2210. EEV. 3. ENCLOSURE 5 l { IALP EVALUATION domplete this form for eacn functional area.;f e, ore than one then sal.P func SALP Functional Area Inspected: h////*e < /o A, d' oe Ma dedon [MN J/c.5) y Key: a 0 - Not inspected; 1 - Above average; 2 - Average; 3 - Below average E
- EVAL.UATION CRITERIA,
- ASSESSMENT i a.
Assurance of quality, including management involvement ! and control; b. Oh23 j Approach to the identification and resolution of technical issues from a safety standpoint; 0@23 l c. Enforcement history ' 1 0@23 i d. Operational and construction events (including respons~e to, analysis of, reporting of, and corrective actions for); e.
@l23 Staffing (including management); and f.
012@ Effectiveness of training and qualifications program. 0@)23 COMMENTS: Provide a brief (2-3 sentence) sumary of the licensee's performance as coserved in tnis insoection. :nat can ce useo in the next SALP Report. TJe limn <r e /)/rktru s uM // in can fnnsins +/> s cris &rs / aa re r1d eos hol n /C n it hers e ra din aversor -lo hial a i ?;u t /u Plas h o& cs e nNins ti/u VJ:rors JKr Apenare sS A/ol/does e /n m/h / 3/r m
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IawfVular/( son hrs //4. , rad /D1e f/u!*,f'a a n d pre / Urs e A 4kJ/p HF %rm / rn +,% Pz/r+1 VL is w A Afp wnsber oP // ft < f ?> ) in R e (nuitfru f/S tKp / hose e f Hr a v 2 usoh/4 . RAefd/N r:.umno ii< As4/se/, NOTE: functional Please areas.attach additional sheets if needed for coments or for other LEAD INSPECTOR: 1) . /3 , For / e 5 OATE: 3//#/ N _ A DISTRIBUTION: . Inspector's Branch Chief I Inspector's Section Chief .' Project Section Chief (for reactor inspections) DRS Division Director l Original (In folder with cover letter) ) 4
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Operations Department St. Lucie Nuclear Power Plant ! Night Order DISTRIBUTION: Unit 1 Control Room Unit 2 Control Room OPS Support Work Control Group i System Specialists Training l Simulator i From: Operations Supervisor's Office , Date: August 12,1996 To: All Operations Personnel l l
- 1. NPS, please review with your crew the attached Crew l
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Self Assessment concerning the U.E. on 8/9/96. Take special note on the lessons learned. 1
- 2. Unit 1 ia currently scheduled to come off-line at 2000 hrs on 8/23/9.6 for a balance shot. This should reduce the vibration on # 9 bearing.
- 3. The PM on the 2C AFWP will be performed tonight on Mid-night shift. The surveil 1ance procedure has not been chacg=d for Human Factors as of this date. Therefore, plaase purform the 2C AFW pump surveillance following the PM.
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1 r' CREW SELF ASSESSMENT i EVENT TIMELINE I 8-9-96, 0205 Completed fission gas purging of volume control tank. This had begun on peakshift to increase VCT hydrogen
. concentration. I i
8-9-96, 0220 ' Began routine RCS inventory balance calculation. The ! starting of this had been delayed due to the VCT fission gas ' purging and associated VCT level changes. I 8-9-96, Sometime ddring inventory balance. Board RCO notices that VCT level drop seems more than usual and notifies ANPS. 8-9-96, 0420 ' Completed RCS inventory balance. Calculated leak rate was
.56 gallons per minute. Board RCO notified NPS of results for morning report. Board RCO and NPS note that the higher than usual leak rate may be due to the recent large VCT level changes. Board RCO's VCT level drop observation not mentioned. NPS directs board RCO to begin a second RCS inventory balance using the stop data from the first one as the start data for a second one.
8-9-96, 0530 SNPO completes fuel pool fill and is directed to check charging pumps for leakage. 8-9-96, 0535 SNPO tells ANPS of observed 1A charging pump center plunger leakage. 8-9-96, 0550 I RCS inventory balance calculates to 1.5 gallons per minute I over a 90 minute duration. Offgoing and oncoming NPS's l
' notified of RCS leak rate results and that it appears that I most of the leakage is from the 1A charging pump center ,
i plunger leak. The situation is discussed and the leakage I was considered identified leakage of less than 10 gallons per minute. ANPS directed to start the 1B charging pump, l l
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l then stop and isolate the 1A charging pump. This was done ! to be able to perform another inventory balance with the 1A i charging pump isolated. 8-9-96, 0610 Further discussions take place between the affgoing NPS, oncoming NPS and acting Operations Supe"~isor concerning the need to make an E-plan declaration. Things considered were:
- 1. The possibility that the RCS leakage may be from an unknown and unidentified source other than the 1A charging pump.
- 2. It would be better to error conservatively than non-conservatively.
8-9-96, 0615
- Offgoing NPS goes to unit 1 control room to declare an unusual event. On the way there he stops to consult with STA. Both the offgoing and oncoming STA disagree with declaration of unusual event. They state their belief that the RCS leakage is identified leakage from the 1A charging-pump.
1 8-9-96, 0622 Offgoing NPS notifies Acting Operations Manager that he is declaring an Unusual Event. Event declared. CREW ASSESSMENT Strengths:
- 1. Delaying of initial RCS inventory balance until VCT purging completed.
- 2. Immediate initiation of a confirmatory RCS inventory balance due to higher than expected value.
- 3. Rapid identification of leakage by SNPO after directed to do so by ANPS.
Weaknesses:
- 1. Failure to begin an immediate search for source of leakage after the higher than expected RCS inventory balance ,
results.
- 2. Failure to notify NPS of RCO's observed decrease in VCT level during initial RCS leakrate calculation.
I i
- 3. Failure of offgoing NPS to classify event in a timely l manner.
Contributin'J factors: 1. Belief that large VCT level changes during or immediately preceding RCS leakrate calculation can affect results. This delayed the initial leak search. 2. SNPO was busy performing 1A CCW pump code run and filling the spent fuel pool in response to level alarms during the confirmatory leak rate calculation. This may have delayed l dispatching SNPO for leak search. ' 3. Notification to NPS of confirmatory RCS inventory balance , results and that it seemed to be identified as from the 1A ! charging pump occurred during NPS shift turnover.
- 4. There was agreement among many experienced crew members that the leakage was identified based on the observed, but still unquantified, leakage from the 1A charging pump.
5. Recent change in our consideration of charging pump leakage . as reactor coolant system leakage. In the past we did NOT ! enter an Unusual Event every time we experienced charging ! pump leakage. The fact that 5 SRO's and 2 STA's struggled with this determination is evidence that it is not yet i clearly understood. PMAI PM96-04-157 discusses this issue.
- 6. Our most common reason for a greater than usual RCS leakrate at St. Lucie is from charging pump leakage. Due to the configuration of the plunger area drain piping it is not possible to quantify charging pump seal leakage without isolating (and disabling) a charging pump.
)
Lessons learned:
- 1. Do NOT delay search for suspected RCS leakage when there is a higher than expected inventory balance calculation.
- 2. Keep all crew members informed of relevant information (VCT level drop). Earlier identification and isolation of 1A charging pump leakage may have prevented meeting the ,
criteria for E-plan initiation.
) 3. Do NOT delay a classification of an Emergency when it meets the classification criteria.
i
.- 1 } '
Actions taken: i 1. NPS performed crew / event assessment and briefed crew about lessons learned and areas for improved crew performance. l 2. Initiated condition report to make charging pump leakage more easily quantifiable. l I summary: l 1. The operating procedures crew took to mitigate the the event. correct actions as required by 3 q) 2. Eliminating the self identified weakness har incorporating the lessons learned will improve future crew performance. l l 1 I i l ! l ! l l
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Resoonse to Ouestions Concernino Calc. PSL-1FJM-93-016 Run "75dp10" (Att. 2 to Calc. PSL-1FJM-93-016) provides the accident flow rate through the ICW system based on a 25% degraded . ICW pump and having 10 psid with a minimum flow of 8000 gpm across the CCW Heat Exchanger during normal operation. The pump - flow is 11104 gpm with a developed head of 112.4 ft of water. Run "221" (Att. 10 to Calc. PSL-1FJM-93-016) was performed through iterations to determine what "K" for the CCW heat exchanger would produce a flow of 9000 gpm through the heat exchanger at accident conditions. This "K" value was then converted to a pressure differential in psi under normal operating conditions with a flow of 8000 gpm to generate the , curve on page 4 of Attachment 10. The pump flow is 9292 gpm with a developed head of 119.6 ft of water. ; Since run "221" has a lower flow rate and higher developed-head than run "75dp10", 'the hydraulic grade'at node D is higher in run "221". f, . r
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PAGE 1 0F 19 PURPOSE / SCOPE.j' g gf.Gf7 . 0.( , 1 ' p m,.~ ;... , { ca cu.3{p foritisq';s:. Q f,'ormed. t J 3 liCW } Sis. tem'.Pei r ance'C fo,'m%,er.e;7 : ;!to create a .St. Lucia Un
'ecep)V
" is 'sfacross liCCWH A(Com]ponent Cooling
} ERENCES - 3 ,iM' n$ Eld'E. . ~ i Itn;6.Eifff%!? . ' . T^I *' . r 'u. ' cu tloh' #132-73.6000; - Re'v.1[. 'ICW System - Pressure Drop asco$r. 1 Di t%#s".k : -F ~ '- V' 4 1 9 4: f & R Q;;. k r J.'- p
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$}!Q.uit %.W.' .7.irie Techriical': i.'-
' Paper 410,-Twenty Second Printi rawing.if 2998-G.-644, .Rev. 6 . . . >
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.ygqp 's . .r'u ; ~~ fi- : .
rawing #87.70AG-125,2 Sheet.CW-F-4, Rev. 7. i -V: $ i .. & L .. * . .... i Drswfng #8770-G-125, Sheet CW-F-10, Rev. O.
.?! y'$4.'lSt ;G-633., . Sheet 501, Rev. 5.
rawingif2998-W 2.W :~.l. h.' ii.y 'nif'#2998G-633, : SheetRev. S05O. I Dra - We, ! EMDRACf0r@awi'ngsf U .' rf, .' 3,Mp#8770-6979, cRev.0; 877 { " Jest' Curve. &, Data - ICW Pumps".. 4 g,$Q@.$Yt?'?lM.:.6=h '. :' i f;9 .cCalculation'f PSL-1FJM-92-056, Rev.1. ~ l G4 b:8::.':'- . : & i 2?10 )):l2 Nonon.%* C fori iance' Report:(NCR)' 1-723*(9/8/92). 1 ed.A .- . :::@. # '. . . PSG11 FSARJ Amendmentfil. ;, l (,g S.t 'r t.' ;N.y . ' j 2 PSIF1 Off-Normal. Ope ating P'rocedure,; ICW System, #1-01i40030, 'Rev.15. i ' Oj.MQ L *: i 1 Instruction Manual; #8770-3599, Rev. 5. i i-7* .'t .' 51((D'rawin Q:y~ff'it .
- t#8770;G-125 Sheet.CW-F-16, Rev. 2.
i eq' ..g:. . - 53ifMiir[ s aA w , . , , tgt;. f Y4 2%* ' M.:gYf875hdil25,.
' T
.: W SheetNW[F-5, Rev. 3.
!!C.'Eng'ineeredl Software,,InE.1- FLO-SERIES, " PIPE-FLO", Version 4.02.
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. METHODOLOGY 1, d!': r g.c :" : ,,,_ pg'ajydraulie-medel of the ICW Sys.em deleloped for use by the PIPC-F!.0 4
1 9 ... com) uter: program, with values obtained from the Referente 2.1 calculation and l ' J.. 1.dt.her values developed within this calculation, a enrve is duveloped showing j L ;,. p.erformance of the ICW System. This curve is created by varying the i i l g erentiah pressures (AP) across the CCK'HX at various ICW flowrates.
) .l
?ASSUNPTIONS/ BASES ff .
.:*?
.. O 1
@, ..lBased ,on the test data provided by PSL site (Reference 2.10), an ICW Pump
; i y 'f head.has'been! degraded to 84% of its hydraalic performance. An ICW pump is 4 , 3. 3; degraded' to,75% of- its design performance throughout the pumps entire l l ? ,
; range, to provide some conservatism in this calculation.
\ W u ;
i ay .2- The ICW temperature control valve (TCV) position is 90* (full open), based
- i lr .i f. ,on Reference 2.1, page 13.
pp - w 4 l, The ICW flowrate used to determine the CCW strainer K-value and the CCWHX l s: M" . 4.34.cj K-values is 8,000 gpm, based on the degraded performance of the p i ; .- - D.- based on the lowest expected flow that would be available for the
'I j[#k 3 $ [, p* .H. ! Table 9.2-3).(essential (safety-related) header during normal operation g- s.L w.<-
l i ff p;f,g h 4j4: Based on the results of the verification performed as BASE 1 (Attachment o g tl #8) being essentially the same as results of the Reference 2.9 (Addendum
'*f .
9
'; 1) 341192 run, this version of PIPE-FLO is acceptable for use as a tool to I
)1 r perform this calculation. Slight differences between BASE 1 and 341192 are because Version 4.02 of PIPE-FLO uses:
j V e-iy ...
- ., 4 slightly different equations / algorithms that result in slightly lower (more conservative) system flows; l
$ ff [ 7:J. i9 l l If 0h points, which
- l j'g .
.h. :.;.creates . + ,a pump a more curve realisticdeveloped pump curve. with 5 points inster.d of
'IE s'ed o'n the similarity of the design performance curves of the IA, the IB
- I ' and thi IC ICW pumps (Reference 2.8), this calculation is applicable to i j ' nd' bounds any one of these 3 pumps when operating with system conditions j@.m!@.% Q
' i. N : e.7, lgiven and with a pump degraded to 75% of its design hydraulic performanc Q:!:*f-g : 4To provide a worst case operating condition scenario, the following items
- g . , .y " 'are being included in the system model:
i . , , . .- . m..
- ;di 1 [' W!" + The non-safety related CW (Circulating Water) pump lube water I f3 y .'
@.c ', supply line ruptures, allowing ICW to be lost from the 3" diameter J. .i. i piPei - ! 4 F :f f .c .
j's[1 (r. p.;., + )A non-safety related instrument line ruptures, allowing ICW to be
. .yp lost from the %" diameter pipe; j 'y ., ; ,. 4 ;
l d //, + For conservatism, an additional 100 gpm is lost from the ICW
- bN 1. system by an opening in the ICL' system.
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-W p '.j4;0 ASSUMPTIONS / BASES (C0iiT'D) .
ik.7 The CCMHX strair.er dd t-ferential pr, .aure of 1.5 ps- u e.! . h w c:.:his y ;" E,,
' 3. . . ff being the maximum allowed wi thou; c ' uniin !'y btch- P. ::h 't.g.
(Reference 2.12) h 1 .. g [y fr. t' j4J8 To provide realistic, yet conswvative foiicwing: v:.iu ct, tim "*1 " uses the
;J-
},u ., !a.
- t. ,' Elevations (page 12 for P.ciarcnce Dimmsiorc):
Q( (( .g.), T$: ePump intake ~
. 10. 5 f t '. i t',t . cr enc e P . 'y
. .s f p.-
g.
'b. 7 4 Intake well water ie. val c Pump discharge
= - Ir ft. (ho mr a.ce ?.3)
= +21.08 f L 0:cIm.u:2 2.4j M;[
M 9, 4CW discharge
* =+11.51 f tO'.cforem:e 2.14) i 3, CInstrument tap discharge =+21 ft. (lie f:'ren:.e 2. . .: 5)
. 3y +Dircharge canal water 'levelm 46 ft. (9:fr. u 12.n
"; +Discharije pipe s.5 ft (kcta .;3 2.3) t.* i Pipe Roughnesses:
i + Absolute roughness (E) = ') . 01 f t , c r 0. . :n. j.: {.3 ..,i' (concrete lined pipe] (Referev.. _'.2)
' + Absolute roughness (E)
Oc 04 = 0.00015 it. er 0.0t)18 in, f{t'
.' i [i.e (stainless steel pipe) (Reference 2.2)
To enable a performance curve to be created, varirvis v: lues of d
- f, j E it
-(differential pressure) across the CC',.'HX are used in the nrdo'. Tha rar.ge of the AP values is fro:n a AP of I psi to a AP of 12 psi, t'u.ed or, a
' ji [ l N' normal" AP of 3.8 psi. (Reference 2.13)
, .r ,
t-
! I'All piping between the ICV pump and the discharge :.ami ; (eith the r j ,i @y g, ,' exception of the points where ficw is being lost froni t!m sy:,t w.) is at
'+ o .J the same elevation. This is acce;: table, based on the syrtein beino open-
'M E j' ended and ary changes in elevation between the pump and 150 dischar'ge v:ill fj ! '
pcancel out. Also, inputting the actual elevation of the flow loss prnnts E- e. , Jwill be realistic and conservative. 41 ,
; i 9r117 The specific gravity (SG) used by the computer in 1.02 (see paans 11 t
- .'1b 12). The SG used in developing the K-values fer the CD.' 'P.rciaer a.:d the j 7'( d. h"'..' ,: CCWHX (PIPE 009 and PIPE 010, pg 5) ,is 1.03 (fren Ref errm e 2.1). The
. . v: -
^! ' } , correct SG is 1.03, but due to this small differm.o. ine results are
'$ n Insignificant and negligible,
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n 1
.M'. [ values n.ry,ste py;r. I','i- } >! :. PIPE 010 IB CCWHX 6,0.9)
(,.t'i-., I. E ~ Y' PIPE 011 FE-21-9B 9.80' 12.7; i'/a. l A
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y4' PIPE 013 I-50-21-1B N/A ( .p 4 ,, t;/A
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PUMP DISC. (39.6 FT.) n ' M f* 1 : N/A
.t f: '
PIPE 015 DISC. CANAL N/A fi/A i 2,4 i. r v7 FT.) PIPE 016 STAP. TING PO!?;T lI/A N/A N/A 4 f[ :1. j' . ' ' PUMP INTAKE
,s '.h ! . (0.0 FT.)
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i f.' P PIPE 017 %" OIA.INSTRL'MEitT N/A 1.572 f.'/, . LINE FLOW LOSS
! b' N/A 3'i l'A
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- l lQ e
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WATER SUPPLY FLOW LOSS l h t U 'E r.d y:gf 4.- l 5 'j 'M i*J h.@ALUES FROM REFERENCE 2.1, based on.a flow of 14,500j l k 'b 5 etermined within this calculation) ' f'h j- ? %yp $y@ l M.. '. , . . . . 3id t?.w ' b[- M@i
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. yL ELIST: 75dp12 SPECT Ft.CI.TIv; . > " :. ' . '
- at Steel nrsIcti ccNDT. r to'.' '.r + 1 x it y
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siidule 40 ~ min: o bsL, roughness: 0.0018 in 4 j l ; r.w : 12 h-
- Y' ' ! .'res vec min: C pai g l ~ q _
i'ze: 0.75 in :aa>:: 0
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- (.h- pgth: 0.5 6 ft p;. e c . O uni o
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- f. . - de::: 63.64 lb/fts l
1 A'h b ~ iey.ation. in: 0 ft a.- out: 0 ft sr;: 1.02
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vsc: 0.9242 cpois kh I . ',' N! l l' [ PIPELINE RESISTANCE CUF9; - .- jf i Msi.g FFP, , VEL HL DP ft non y, L :, ', ft/sec ft psi w ,,j } !) v w : ! 1F6 7.0372 1 0.029 0.013 ' ~ i'0323 2 0.112 0.050 _.,,, f'4.
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f; . VALVES and FIT ^.'INGS -- - - - - - - - - - - - - -
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K-VALU! c DESCRIPTION K-VALUE COUNT DECCPI'<TIr.:1 0.07:0: 1 EF noe' Sharp-Fdged O.5 i Ball Valve ~ 1 gEagg.;,snarp-Edged - - -
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4 TOTAL K: 1.572 Avg Percent of Tocal Las.s: 88 't
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' 'I out: 0 ft x,: 1.02 6
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, I ,t 'Q; l 6.0 RESUl.TS _(CONT'D) $
3 e
. . ,. 9.. ,
-g .. ,
Reference 2.17 require: an ICW flo. c f 12.000 e; m .it h 'i to .woport t',e L ? g.G- , , I be ,- .-. results of the containment 2nalysis. From tht- I Ci.' Pc r 7e .'..v.e. c.urve on [ M '; ' Page 16, it can be seen that the maxbum combime) aiior #.e -/> icr ,c , the ,_ s '.gf g ;.{.%'.CCWHX and the CCW strainer ~7.2 psici at a fla.<rhu or d 600 som during
] p g.
~
;g n,ormal operation. Based on AP(psi) being proportional to Q't,ppr.), a curve ,
e
' fp P, .tJcan be prepared to reflect the allowable combined 6P scros the CCVH) and 1
' 54 e* ;~
'l e th CCW strainer at various flovis during norcal ont.rster..
; $8 !j ; e
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- !3 h ! 75, 4* , 6.'l - A validation of the PIPE-FLO (versh 14.02) cennuter proarm wa pedon.v'd p' FL i. . by inputting the same data as used in the PSI-! pertier af the Refer ence t; n 2.9 calculation (341192, Addendun 41). The resdi;s, B.MH, are u
- .sentially
- i j. y ' the same and are attached to this calculat;on as .ttacncs,t #8 ( 5.w
- 18. i}.';.v
..i- , Assumption 4.4).
'd ?-ft (f' .] :
i l .(: $ a 6.'2 Following are the flowrates of the ICW fluic ther. ugh the C'I H/ drring . ) d pp.,iaccident conditions based on the v rious differential- prwn cm across the QU(f
- .h V " j!iCCW HX (to be combined with 2.5 psid across the strainer' d; ring norr.al e ;.$
- ",ih*Toperating conditions, 8,000 gpm flow and an ICW Pur.o at 7M'. t.ydraulic i
7" ! j. 41" performance: l
; a " .s a
, 8 A '5 l .
1 PSID at 13,127 gpm
'lI If d, 2 PSID.at 12,797 gpm 4 PSID at 12,199 gpm
) . g n' ,
- i rC . ,.G 6 PSID at 11,682 gpm j
#d.
i.' +
--i f ' 8 PSID at 11,229 gpm 10 PSID at 10,828 gpm l , Yf 4M.7[.. dhi'i 12 PSID at 10,470 gpm
- ; e - d a it w, e.
. < s (r [ 4 L' An "lCW System Performance Curve" (Page 16) h:s been creat W by plotting e above points (by adding 2.5 psid to each 6P value).
. ! (%b l y 9 l j S J L
'n
~
i' d ,I, ' and [AMitionally, a CCW Strainer"an "ICW curveAllowable (page 18) isDifferential Pressure provided to show a range Acros:. of both
/ Y flowrates. luring potential
;p j . -[: -
l g;possible 6P's across a range of possible njrmaioperatingconditions. i 4 . e A.n i
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. ro act:;PsL c..,,.
',.4 *','.' }}by_: :;'TMECH / BOP .' j l ! I l
h 'i i /lW - G [. . .; . p LINELIST REPCRT t ef : .s.
. i .;
',{ ' N .
reate'd: 05/14/93 r inu. : t,c!as: 19
*O Des gn file:
e .P Specs: 3 r .:N o m' u : 2 ( ;: 7p - l
~ , .? 12PSID, ICW PUMP 75%, 2 OPEUINGO, 100 Grt: ..350
- ) . *;.%. ..
[ . , ' ,1,E. PROJECT SPECIFICATIouri ) , b IN' ,, PIPE MATERIAL _ FLUID VALVE TABLE l 1 .. . ., Sch / Roughness Temp / Pres { *;;;v: , in *r psi l sl4 4, . i I
, i f CONC LINED scawtcr S'candard ;
i j ', ,.,, a6*edj]F.05/20/ 93 STD / 0.12 95 / 0 i
- t. r 45qh .
i l. p asign, 36 ft/sec Limits: 0 - 12 ft/sec 0- 100 poi (5 l. 4 .- j 'y C' .; .- Steel sec.? iuct Stinuhrd W
$ O'3 2 \ (-
'ed ;05/24/9,3 . s 40 / 0.0018 95 / 0
$l N ' x'*6[
s f t/ s'ec Limits: 0 - 12 ft/sec j
~
. 0- 100 psi jf*4 rt
**N-i $E ,, , ,
seawater Standard TEEL f ' i Steel i I r a'ted:. 05/24/93
- 40 / 0.0018 95 / 0 s :M
# Design:.6 ft/sec Limits: 0 - 12 ft/sec I ; I'? O- 100 psi
%'h' l
= !~. * *:'f Ej g-
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~
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-. . . . . . _ ~ _ . - . . - - ~ -- - _ . . .--. --... -. .. - -- . _ - - - . ..
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CIPING LINELIST s; . .
. ,' ,r<_i. . ^
75dp12
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I;' LJ ' . SPEC MATERIAL . LEN?TH VALv:-;s i k. ' Sizo / Sch " / Pros _ Totzti-r p ,j
,l? in *'- '
ft p ya .-
- . .t 01 CONC LINED 469.6 c e a ve. t.<
- r 0 Ih,)., , 30 / STD
[{,< 4 95 .' o
~
) ,
;y;lf!':,,, 01 CONC LINED 172.4 cLawa;.cr 0 j j ),'. T 36 / STD 9 c, / o i ; 01 CONC LINED 110.2 cavater
'l
]
{y j hj'[' ,< ,- 30 / STD 95 / 0
- Q ,' ,[ , 0 1 - CONC LINED 1447 seawater 0 l
( 30 / STD 95 / 0 f1 '
- v. j ? 01 CONC LINED 765.6 seawater 0 l 1 '1 ,**- 30 / STD 95 / 0 >
l I 01 CONC LINED 235.G sc.awa ter 9
'(
j {. ,, [' 30 / STD 95 y 0 i ,3 i 07 i <03> SS 65.1 crarcac.< c o
'TN 30 / SS j { ) 95 / c l $ 'I 08 01 CONC LINED 700.7 cc-a'<ca tor o i
I f 1; g$- :-. . 30 / STD 95 / e i (: - ~ ,
- ., PI 09, 01 CONC LINED 1 seawater 24 l '{ 'lif Tif .
- 30 / STD 95 / 0 l' %
- d. Ye s-01 CONC LINED
~
I t 110 Vd fr- 1 seawater 115.2 l ,
>- t f 30 / STD 95 / 0 i i 3 Olb ,~ 01 CONC LINED 1 ccavater 12.74
*1 fi)h d.f., 30 / STD l t -
95 / 0 f 01'23 01 CONC LINED 1 ceavater 0.3 c 1 [j (I 'i 30 / STD 95 / 0
- y s- t :.7 4 i I :. PI .13 j. 01 CONC LINED 1 seawater 45.5 I
[ ,U t c4 7 30 / STD 95 / 0 3..i9. ..; CONC LINED r PI 14.- 01 1 coava tc-r 0
.) U @I.If-- 30 / STD 95 / 0 I '
R M ' 01 CONC LINED 1 scawater 0 1 W 5 ] 30 / STD 95 / 0 e q r g <
)- ab; p'a 01 CONC LINED 1 seawater 0 hN g
.'. 30 / STD
, -d i k 90 / 0 i
9 PI 1 <D2> Stcol 0.5 s ta t'r ':er 1.572
, it
'- Q+ -.
,,4 0.75 / 40 95 / 0 h I <02> Steel 0.23 senwater 345
( $ M~ $f.'- 3 / 40 95 / 0 ( , y) P W H .J g >: C y
~
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gb:, .
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^
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l; : MATERIA 13 REPORT ... ,, . ' l 'a
,y . r ..=_ ; 'y b . Created: 05/14/93 ,s j pes: ,,. g 4 ~ .$ 9N4filC: No:lus : 19 1- P8' Specs: 3 'congc,ents: 2 i ? ..
g.) "W i 19 9 .. 12PSID, ICW PUMP 75%, 2 OPENINGS, 100 CFM LDSS i ! 'c '; d.4 .. l 1 . $ '9 # 7 PIPE MATERI?.LS LIST I'
' ' j'E -$
.5? .
1 [PKP NE-9.'.Q MATERIAL LENGTH VALVES S FITT EGS j . Si::e / Sch ft
' PIPE 014 r ~ ?- CONC LINED 469.6 i
7.{ ,.! jSpec 01 30 / STD ta t. } (l PIPE 02 CONC LINED 172.4 4 U ; & .. eSpec 01 36 / STD T' 41 -PIP
.1 :
03 ' - CONC LINED 110.2 l 4 i - E ' .- ~ hdc '., 01 30 / STD 1 = '
' L l f. IP T.' CONC LINED 1447 01: 30 / STD
)gk 4. }ii .
Q0 . CONC LINED 765.6 l 1 4M p .s.
.30 / STD 1 l
$ i CONC LINED 235.6
.' l 30 / STD b
'$ . .11 g-e l
ITIP I d i r SS 65.1 l 4 d2 .,3>
.co 30 / SS l M ~
A .y i j UPIP " CONC LINED 706.7 3 , {# -
',g O *. 30 / STD l i . . . l
I h. . .k ' CONC LINED 1 1-Fixed K 24 '
e,,.c. . 01' 30 / STD ! ! ~>e ' . l : ,
-i- CONC LINED 1 1-Fix er.' K 115. .'
.t ec 01 30 / STD l
! '%;Sp?
i l
*PI 1i CONC LINED 1 1-Fixm K '.2./4
* . Spec 01 30 / STD
, , c.
Y IP 1-Fi):cd K 0. 3 1 1 12 7 CONC LINED 1 , ; y' ' Spec 01 30 / STD
- 4 h .[yPIP '3f[l' CONC LINED 1 1-Fixed F 45.5 f ( ,.', -
[SPe6 01 30 / STD i .. . . .
; . PIP 0.14 . .
CONC LINED 1 4
!g
-i @ Spec 01 30 / STD
. ?
I l . I h, 4 t, PIPE-FLO rev 4.02 9" i
_ _ - _ _ .- -. . . - - - - - - - - - - - - - - - - ' ~ ~ ~ ^ - ~ ~ ^ ~ ~ ^
'Y W U $ 'id5i5N O S.;iOt;.. '* % . "; .s..: a ~ . i. . . % . :.i~., n . ' b52 s .'. .'
'; y j!!e -
I; f. 4 [ PIPE ?!ATERI".LC L ' '"'
, .'. '/ . ' . f
-' % d p 12
( ,
' > [ (. 05/24/93 P E NE . ), MATERIAL L211GTH - VALVEE a 'r'. .'
} .- .
Size / Sch ft - -
. PIPEC15 ., . CONC LINED 1 Pe5 ' .01 30 / STD PIPE 016 CONC LINED 1
;0 7% #( H..
. { ,lppec 01 30 / STD
- h PIPEC17 .; Steel 0.5 1-Entrance SI:arp-rdcg.d h :.U,:= ~ pec ,<02>
0.75 / 40 1-Ba11 i'alve
.h rc t>
hr(' 1-Exit Sl.srp-Fdgr.d j f '
'. I
]T5.lp,;
- l ,- & h 1
! T P 'f i f -
Steel 0.23 ~1-Fixed 1: 344.5 l h f S . 02S 3 / 40 1-Entrance Sharp-Edgcd N ; d i N-l 0 , .4 1 i
- i. . g; 1 :. ,j.
' { .gw.-,
j f 1 s 9 t. . a9 3 l t. -3 M,.s. .
. sa-i' q .4 >
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" ;* h e*. '
PIPE MATERIAL SCHECULE - c a ,,,, e :1
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4
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1 tit^L. sQ 1 Steel 40 0.75 c 4 j I- [ e,.t,f.,, 3 t,-
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l f i;, b~ . f, ' d r f,1t . g.h:s i '- VALVE & FITTING SL"ril'.ARY i i F r) L #: SCHEDULE VALVES & FIT'! D CS I. 1 . SPEC ON %: MATERIAL J '1 L. *; ,n ([
! D E CONC LINED STD l'e )
l
- g c-d, i1 jb. '*
l
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! l 6' ' D .7 Size: 30 in 1-Fixed K 24 t. I
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< f b f (d 4 - 1-Fixed K 115.2 El
$; f' s c !. W , - 1-Fixed K 12. '/4
; >g
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. 14f' 'M i g s :.
I [,i1M. y - Size: 0.75 in 1-Entrance Sharp-Edged 1 4
;i / i[ ,,. j. -* h. . / #' . 1-Ball Valve
-l l1 .
4,-. l 1-Exit Sharp-Edged [l 3 ' 4 I. 4,h. , 1-Fixed K 344.5 i 1
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Size: 3 in
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- .7 ..
' !p 4 et dt.'05/14/93 y inc3: le j u l k - Ifh88. cs 3 ,3 c...npenonts : 2 NoU .'s : 19 j PSID, ICW PUMP 75%, 2 OPENINGS,
.-l12 10C GPn LOSS -
[ d ' SYSTEM NODELIST f 'y . .b..M"(. : j, *; ODE ELEVATION PIPELINES IN PIPJLINiG OUT l l j {, dj.p ft . 6 2 , p ' I "" PIPE 014 39.6 PIPE 001 l e <>
'f 4 .h,
. r'* ,
j
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39.6 PIPECC1 P.! P7.0 0 2
'1
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1
. j.k ,., 39.6 PIPE 002 PIPE 003 j
(
;. 0,1 . 'fgCW F , ,'.
9.6 PIPC018 : i {=.,ggl;t?
- r D~ 39.6 PIPE 003 ?IPE004
,< G.; .h ,.-) PIPE 013 :
2 ISC 17 PIPE 015
- hlb'd a; M. 39.6 PIPE 004 PIPE 005
'{? i .
ll C 1 ; 39.6 PIPE 005 PIPE 006 s .s l 39.6 PIPE 006 PIPE 007
/ $N' 4 y 39.6 PIPE 007 PIPE 008
.; i39.6 PIPE 008 PIPE 009
) p; Ef O PIPE 016 w.
) .-
* .- S 39.6 PIPE 009 PIPE 010 PIPC017 f 5 If[h' .
6 fk- N :I- 39.6 PIPE 010 PIPE 011
& m '?
l y[] $1' D, 39.6 PIPE 011 Plf'E0 3 2 p 1.; n. <!' (\ .[ ' O, ;- fiO Y 39.6 PIPE 012 P!PE0 3 :' PIPE 015 5 .a r
. f EI
.y . c 17 PIPE 013
[ ' t' ! '.' ? ~ 0 PIPE 016 PIPE 014 Phjlk fO
" '
- AP 39.5 PIPE 017 h ' .
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. _ _ _ . _ _ . . _ _ _ _ . - - - --. - - - - - -- - - - - - - - ^ - ~ ~ ^ - ^ ^ ~ ~ ~ ~ ~ ~ ' " ~ ~ " ~ ~ - ~
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$ *g' 'y , (I . ... ' '. '
, 75dp12 l
b : tc / - 7 ~ -c / (, 0 5 / .' 4 / ? *,
/ PIPELINE SPEC FRCM- NODE Td . CO.P'ON NT 4
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p@ PIPE 003 01 C D Pb'O 4 01 D k
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! ; p ,J.'.'
a-: PIPE 005
.,.3 PIPE 006 01 F G j j' Lfi I
' , PIPE 007 ,
<03> G 11 .
- j.d > yy, ,;. $ PIPE 008 01 H I
.. .u: :
l i 9 . i: 9-s t I ., PIPE 009 01 I J i r .. . PIPE 010 01 J K I, . f. '
- . n > ,
i' PE011 01 K L
}l' h)( ? pe> >
- t -
PE012 01 L M
- i W
[- Wr: I $ #l@ p PE013.) 01 M N
<r
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-l
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01 N DISC t i F V5 IPE016 01 INTAKE O j < ' *: .or. . ibE PE017 vrt i r
<02> J TAP
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;,4 '.NE.NT PERFORMA!!CE lU.TI. [..
. IC7 I/
4 I" ~
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~
gpm; O 7000 10000 14 2 "(. 15000 93 I ft: 154.5 129 11 '. . 5 202.27 s { [ kl. , s - s~~~~ M i 14000 16000 [ gpm: 0 - 7000 10000 h d ?
- ..U h - ft: 164.8 137.6 173.2 108 99 '
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~
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1.8 .)[- NELIST: ted s,. 7 5dp12 m:;n m ora: 9 . 7 a c _ c.c'c. ; g '.'d e 05/24/93
- .2fter
- 9 iteraticns s . .s
- g. ,9:.43 i
s atric!flod Eates require constant fluid propertler, in all pipalines. j {.iFluidpropertiesinthefirstpiporpecil'icationwereused. qa. ~ $ i DEMAND DoDE 3 a.i.3..t c NODb . S j';p gpm , gpe. 1 i- !$3 L , ' I ,','{. .j >>> 100 f J 1.~k t
%. f .' . .t.
l, s W. N ET FLO'<" OUT: 100 gpr.. i f .t. % I l'* t .* f a
, f.;t
.l
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. y.o.! .
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J p., l 0. f. n' ~. .. W.F k I 6 i'lPRESSURE CONNECTIONS PLOW PREssi. M j
,t [y M i 'gj oda , Pipelino gpm psi 9 l l 'r.'y ;. h,,, f '..;@ c W i .
<<< PIPE 018 104,6 <<< 0
- i .g - [ [CW.
7 F' ?f, .j; ; . ' *. 3
! .* j, ' . :"DISC <<< PIPE 015 10470 <<< 3.3 fy
'o t i- . INTAKE >>> PIPE 016 10749 >>> 4.2
-:-f -+ i'.' .
ir F"n ,'.
<<< PIPE 017 74.59 <<< 0
! ',~ bl.i. TAP i , <e D ti?:..9 ,' n : l
't ..j NET FLOWS IU: 300 gpm
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75dp12 !' h v.&,, ;? ' ' - o . c ?': :n ..aP4w((i 2 i n
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- o i efi. ' ' p.u;3.c[,
(i' ciut.r j f gy.A ., . ELEVATION DEMAND ~ - l ft gpm pgi g ft t) li,"{ 'j. , l I3) If ..! .i: 'j i ,1 'J .
" s 'i 39.6 36.98 123.3
.120.9
'1 J 39.6 35.93 4
- r. . ,.'S....,. . . ay
- ij .
?;'e 39.6 35.78 120.6
- 4. k: b
- Q7 1
4 .-Oc a' 9.6 p 0 9.6 Ibec .J'
~Q- 120
-; b,,-
$." .I .}f: y 39.6-
.y, ,4 f .:5.54 p .3 24.47 j
<v,.,@w. 17 i
4 i.s. D SC',4'( ~:{..P . . i Yi, o N *39.G 32.37 112.9 )
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, t F. .F,. ? :- ' ' . 39.6 30.7 109.1
- w. f n.
107.9
.i 5
i G. i. 4
- 39.6 N. 1D
> 39.6 20.12 107.8
- 4,_' {..*'j.,;- .;
1 39.6 28.5n 104.3
,3 s.
? . .'. p 4.2 9.504 e
'. c{ ~ w, h.h,pO s y6 I
', e. 39.6 > 100 ; . . l'i 94.29
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39.6 3.715 48.01
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' s. o c
) 39.6 1.451 42.30 l f.- - 39.6 1.395 42.76 h,
- l. .)
I.' %^ ~ 17 ~.302
$ 24.47 h'.ko 0 4.198 9.499
'4 <f hi39.5 p o 39.5
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)g LINEUP P." LIU2d 76dp1';
,. 3 ' '77. ; / /. Op/24/C' i v ' . h[ * .
s ' / o .:, . il !!1 t ,-:
- e i ll i ,,3 :P.19ELINE FROM o 71cu- v: ,
opn oci. .y. er Q f.r . i t ,. it ! APIPE001 A D 10749 5.225 1.048 2.072
- h y,IB002' B C .0749 2.587 0.143 0.323 b IPI ' 03i,' C D 10749 5.225 c.c40 0.S57 1 1 $f) J.d' G . h, O E 10644 5.174 3.167 7.160 l
! r 0 fl .-([ ,
f(. g F 10644 5.174 1.676 3.792
' . l
% y 10644 4 '}
G 5.174 0.516 1.167
- 4 l ik kl I -
G H 10644 5 0.057 0.129 FB l Dr,ci 10644 5.174 1.547 3.501 1 H I
,,, [p g I J 10644 5.374 4.406 ,9.971 i
)".I [ ,
!- J X 10470 5.089 20.45 46.29 j - i,.}C,
~ g L 10470 5.0d9 ?.264 5.123 i p ((j . L M 10470 0.089 0.;55 0.125 "h '
'hi y. "l M N 10470 5.039 (1.907 ) 18.28 j .kik - o A 10749 5 . "' 2 5 (32.78 ) (113.8) lP.ft.iCW75 --- dP: (50.28) --- H1: (113.8) f eQ- N DISC 10470 E.039 0.002 0.005 INTAKE o 10749 5.225 0.002 0.005 f ,
foi -h.L.
- I 01 J TAP 74.59
- 44.91 24.17 54.79
'O d 8',c D CW 104.6 4.542 35.54 110.4 l { ,
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. }-. -' 75dp10 f f .'('2l'. '
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MATERIAL
/ / // 9' ..
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i p' . SPEC . LENc;; - vf,ny g;
'.j ; y Size / Sch ,_, :. 1.r e.s Tota; y i.
5 , gd . *g . in ft ar pm , j 3' 1
-; M 'e' 6Q dQ!, ;
E, i
,' PIPE 01 CONC LINED 409.6 s e a ., a t. c r 0
,, , g/
^
g.. 30 / STD 95 / 0 ( iPIPE 01 CONC LINED 172.4 conwatar o ;' p
'j ,
36 / STD 95 / O p P g:y ' 01 CONC LINED 30 / STD 110.2 soawater 95 / 0 0 k 1 ' CONC LINED p .. I, 01 1447 seawater o i ! -l J4
- p 30 / STD 95 / 0 i
1 1
. , , .. N' 01 CONC LINED 765.6 acawater o [ ll t.
., .i i 30 / STD 95 / 0 [<
i 1 i c - j P 0 6' s. 01 CONC LINED 235.6 r.enwater 0 [l [ ' .j,. g ,j ' 30 / STD - : ( 95 / 0 [l IO7.;.y <03> SS 65.1 ccavater 0 j p'J ; i .,N T. 30 / SS 95 / 0 , i j 4.. i ' 01 CONC LINED 706.7 seawater 0
- PIPE 008 .
j
.{,.; 30 / STD 95 / 0
~
01 CONC LINED 1 ceawater 24 gi 30 / STD 5 f. (PE009 P
~.f ,(.(f -
01 CONC LINED 1 95 / 0 seawater 96 q 4
- g 10.;
4 l,j j'c. ' 30 / STD 95 / 0 Il
' PIPE 011 .-
01 CONC LINED 1 seawater 12.74'- " 1
]
' ;'1 !- ,,3 30 / STD 95 /O 01 CONC LINED 1 seawater 0.3 j 2 j 30 / STD
@ {c. .
'jj h, 95 / 0
) -[ 01 CONC LINED 1 seawater 45.5 3h" .(.h
.l
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e - 30 / STD 95 / 0
' CONC LINED 1 seawater 0 f 5 01 30 / STD 95 / 0 .
s
!, , 4
. ? 01 CONC LINED 1 seawater o {;
h .I '? 95 / 0 i l
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P 3 30 / STD ; 41 . scawater 0 FPIP 'i .9 01 CONC LINED 1 ,
~* ;- 30 / STD 95 / 0
{. 4 i , 1.572 ?
<02> Stcol 0.s scavater i . EIP -'
H
- h.f'jj ,
W)'O; .) 0.75 / 40 96 / 0 0.23 seawater 345 f 1 b:PI N <02> Steel ! I y 51
- k. } 3 / 40 95 / 0 ,'
- WF,@.
P g !
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! PIPE-FLO rev 4.02 10 PSID .
j N M g s.t2 W ,- : - - ~.s.. . . . , , . . .... . .--..m............m....... , , . , , , , , .. ..,,. j . .u.. ..uT, .. .. .. ,s. . i.wu. O t.j * . . 4 R .y 3 e',cy.'.- ! -
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J' ' d l 't' p m aq .':, - LINEUP REPORT r ..v/ L ' .
- g rev
- 05/25/93 't q ;D , . . . '.
i
;WIl IST: 75dp10 DEVIAT1oN: 0.000101 1 ated: 05/24/93 l
!.[
i: wh h .-' - after: .9 iteration ! '. .1V5 metric flow rates require constant fluid prose: ties !.n all pipelincc. i{.',F.luid properties in the first p.ip< specification were used.
^
A' . i ~ - ! ; .4 h ' NddE . , DE!'.AND NCOE D EM.t.N D l l f i^ . Ok;'1' :- 9pm gpm 1 .
~ ; *. Y m: . \
- "4 L.3
; f t.
' J.- r b. -
>>> 100 l
'. m b g - .e ,
i t.
.1 g, I '. .]. NET FLOWS O'JT: 100 gpn
.7
[ . h .; ' E .4 t
. :t
" J -
PRESSURE CONNECTIONS FLOW PRESSURE
,' 4 Noda.J: Pipeline gpm psi 9
.V. . ,
;t-I PIPE 018 103.8 <<< 0 p- .e. ;<<<
. PIPE 015 10828 <<< 3.3
*i 1 .
p] Eg'$ >>> PIPE 016 11104 >>> 4.2
- t. .
'.i S* <<< PIPE 017 72.24 <<< 0
'f .l f
, j. ' ' -
NET FLOWS IN: 100 gpm i
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ELEVATION _ DE!!AND --
- P.7. c '" ' '< / .H C!U,OI:
.g.-pE T i..
p. ft 7.0;". ::i o ft 1 ts ~m) < 4 -
.; A N 39.6 35.:S 121.9 - ..I. (, s 1 m l 4 e. -
39.6 3 5 . .' 3 119.3 } gj i.h -
, 39.6 3 5. t? S 119
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39.6 29.03
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r P..". 105.3 l i. 39.6 27.38 101.G f.$.:. G.T.Y
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O p 4.2 9.504 i C ' #': 2 d t? - 39.6 > 100 22.67 90.91 l [ ., r '- j 39.6
. ,,, : ;,.7 ,
4.439 49.55
. 4s"..- 39.6 2.018 44.17 TT- - 39.6 1.953 44.03 4
I.#}I2 17 3.302 24.47 {..f;7i- 0 4.198 9.499
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j 'd IP le 1h %~e A 3 1 104 5.390 L.119 3.531 j i ti, si 7 Mr R y; . B C 11104 3 . ~/ c 5 0.152 o,344 ! . . ! PIPY = . -. >. .,
. :.I'IP
! 3 C D 11:04 6.303 0.2t2 2.594 i
~ {r'IP e ? $3, 1
a 2 ( ,.1 4, ,.:>
?
D, E 11000 5 . 3 4 ." 3.333 7.655 p ! 'IP $ 3 M'. E' F 11000 S.3'e 1.79 4.05
- g ,9 .
d Q .. . FI
- i' LI 6 ;. . G 11000- 5.243 0.551 1.246 i
3
?
y I 7 V {..: t 1 '.
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'.G f H 11000 5.168 0.061 0.138
> h i
8* N r}- H- I 11000 5.348 1. 652 3.738 , ?
,, b p' .-- '
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'IP '9) .j I J 11000 3 . 3 e. F 4.706 10.65 l 9- J K 10828 5.264 .23 4t.26 h e*IP v. 0 .'o.1 vt JPIP 10, : K L 10S28 5.264 2. :: 5.48 a g.g ra' ..
PIP 012' L M 10328 5 . .M / 0.' '
. 0.134
! : y :.T y 3 4 4.'- M N 10828 5.264 (1.344 ) 19.56 l
. CPI.PE013 4 . 2865s : .-
1 PIPE 014
~
O A 11104 5.398 (32.15 ) (112.4) . l cjf ' - ,ICW75 --- dP: (49.65) --- H1: (112.4) cN i. . e . . . 0.002 0.005 1 .. PIP 5;
- N DISC 10828 5.264
- e 4- og g ....} '- .
i RtIPE016'< F INTAKE O 11104 5.398 0.002 0.005 i j# .g. . , peg e s i PIPE 017 J TAP 72.24
- 43.5 22.67 51.41 i e.g 4 -
j zTIPE018[( D CW 103.8 4.509 34.82 108.8
- w. s .
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/;.p e - ;. / (. OS/25/M j , .p4.- ,,.
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yAtygg j 'kLINEf SPEC MhTERIAL - L .NaT![ . :: l j ';* Size / Sch .,p To t:,i l -i l / rres - i I'*- in ft - :- pai 4 {N,? PE001- 01 CONC LINED 4 6 9 . t. sa n;;.r.cr 0 l
,-{
'" 30 / STD 9c / o
] - - PE002 01 CONC LINED J72.4 r.t v.un te r 0 l k' .g 3d / STD 99 j 0 0 CONC LINED 110.2 c::avater 0 l - i IPE0037-
- . f.4 y;4y 30 / STD ,5 / o
! b PE004p. 01 ' CONC LINED 1447 neawater 0 l }l $ % 60 'Si#* (' IPI'kD.p f 30 / STD 95 / 0 CONC LINED 765.G seawater 0 l 0 E '- ' 01 4 ' , 30 / STD 95 / 0 01 CONC LINED ? 3:> . 6 ;e.e.;,.t 3 r 0 ! '4 . PIPE 006 30 / STD 90 ! 0 l .
,,h) i PEC07 <03> SS 65.1 s c r. er 0
) 'g . , 30 / SS 95 j c s 0 i
- PE0 0 8,.. 01 CONC LINED 70b.7 nocwater
{ ;. y., 30 / STD 95 / 0 01 CONC LINED 1 .s c a :n.t c r 24 30 / STD j 2 009'h.{j ,
'15 / 0 CONC LINED seawater 76.S-j
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- h. 01 30 / STD 1
95 / 0 j p 01 CONC LINED 1 acawater 12.74 j 7 .' 30 / STD 95 / 0 - ?. 01 CONC LINED 1 scawater 0.3 e !- j l '. 30 / STD 95 / 0
' ~
01 CONC LINED 1 seawater 45.5 ! 01 .-
*
- 4,[
- 30 / STD 95 / 0 l 0 01 CONC LINED 1 crawater j E0iAf .
- / ,
30./ STD 95 / 0 0 l E01 [g.. ! '
- 01 CONC LINED 1 se.svater
'#I? 30 / STD 95 / G i
i ht. . .: seawater 0 CONC LINED 1 PE016$'- 01 i
.e 30 / STD 95 / 0 I.4 f '
O.; seavster 1.57: l IPE017 <02> Steel
;- 0.7 S' / 40 BC / 0
- m.
- 0.23 seawater 345 J
IPE018. <02> Steel 95 / 0 j ,g
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3 / 40 i 4
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, , p ,g y'4 ;
s op ..ht*t-1. LINEUP REPORT .ev: or / 2., j c , e,g,;Ji . ! $ d- LINELIST: 75dpo otvinTIou: e .ocola;, -
'. 1' dated: 05/24/93 after: 9 iterat. m olumstric flow ratos require constant ftuid prciperneo in a]) p:pelinoe.
}
]Q Fluid proporties in the first pipe spccii'icatic'a win c ured.
5
*.} DEMAND HODE D .!1/JO j c'4 NODE gpm U rr.
j. { j i >>> 100
~
NE"' FLO'.:S CUT: 3.00 gpm
.[
t - l . t ,i , f .tM i FLOW PRESSURE ! 'd :. > PRESSURE CONNECTIONS
; Node, Pipoline gpm psi g l
'ft g ,' 102.9 <<< 0 I s . . CW . ' <<< PIPE 018
! f.' 4 .
<<< 3.3 1 J DISC'
<<< PIPE 015 11229 f >>> 4.2 i i f.db-[t INTAKE >>> PIPE 016 11502 l .
69.47 <<< 0 i l'.' TAP
? <<< PIPE 017
! 7. ' l 'I- NET FLC'.1S IN: 100 gpm l
; C. T*- ,
l 0 L : l A;}. ,
.k.M . ! ; .4 . _l.t 4.
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l
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'5dpa
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r$' 41 ]O 05/25/93
'p*O ELEVATION
- 1. cy .' ..u! t 4 [ ' ' .- DEMA!!D -
PMtSSU- H GP16DC
!J 'f .ft gpm pig f,,t fh.Oh -{lg
, .I e '.p< '. i. t n :q, ,.
4 .. i.:.. f N 'i$' I#'~.39.6
.k 11 D '
30.G5 120.3 i 0 ' I1 of s 9 l! f '
~
- 1 3.i . ', 5 117.6 c.1 ' U L 1 4 .t:.39.6 1 ,
4
.) , . -
I F '.4 39.6 M.28 117.2 1
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- 9.6 p 0 9.6 d' '( 3.':
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1 54 34 316.5 1 i , ..
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;- . s- . ,.
s P 33 24 47
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30.37 10 8 . ~a
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/ ,4l & ,3-h':'
l [ jJ.;}' 39.6 .39 .6 27.79 102.5 i 5;3 e 1.i. ' ~ 39.6 26.02 90.48 [ [ o p 4.2 9.504 d 4Y 39.6 > 100 20,37 87.05 1 f a L'.
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) (f ,I . vs 39.6 5.279 51.55
, , '{ i' JI .;; :
; 39.6 2.674 45.65
. 'j f(
f .' 39.6 2.611 45.51 ,
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,17 3.302 24.47 i' ' IL k '
g .o 'L ., -o 4.197 9.499
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; a P Ol' A B 11E02 5. 5 :.1 1.2 2.715 c.- 4
'2 ' B C 11502 2 . r.' 3 0 0.16: 0.369
- f. ' P ,
0.637 11502 5.691 0.28?
* ) fIVE003 C D nts . t :, -
3 PIPE 004:#'-
~
D E 11395 5.541 3.631 8 218 1 ,.r. ' '
. a. s?ft; c PIPB005,f. E F 11399 5.541 1.921 4.343 j = L. i :'F 1. - -
.: PIPp 3006'fi .* F G 11399 5.601 0.591 1.330
, v N '* - -'
1 '."$ PIP: g&:l007{f-P' i G H 11399 5.255 0.065 0.147 N d8 H I 11399 5.541 1.774 4.013 g 1 r , *+N
. PIPE 009 I J 11399 5.541 5.063 11.44 w ai 5 . ', 5 9 16 . t. 3 35.5 f PIPE 0'10 J K 11229 i s !. sps L 11229 5.459 2 . i.N 5.394 PIP 011' K S-Ip g -
L M 11229 5.459 0.uv4 0.144 4.,.
- ,, 'I 13 9 -- M N 11229 5.459 (0.692 ) 21.03 j
4 .O
,p O A 11502 5.591 (31.45 ) (110.8)
.( I 1-
;. L ~
CW7.5 --- dP: (48.95) --- H1: (110.3) j .h h'
- N DISC 11229 5.459 0.002 0.006 )
. ' e O 11502 5.591 0.003 0.006 E IP , INTAKE j b$ 75,i,h(. J TAP 69.47
- 41.83 20.97 47.55 106.9
. :. j' [hil ' D CW 102.9 4.47 34 n
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'f PIPELINE SPEC MATERI.%L - L g g s ;;; ,,
'.' yn gj;;3 T' D' Si':c / Och - p/ r t s T.:,t .i l - L j
f in rt - p.a 4 4 <v.. ... 2
- ,,, ,E001- 01 CONC LINED 469.0 a c.3va t.r,; o
.g. j . . .l. 30 / STD ;5 / o
.. ':P E002 01 CONC LINED 172..: 'ex.!wccr
. 0 l
;.J '[
i, y 36 / STD es / o
, , 1 . * -
01 ]i i r} ... l , .f, CONC LINED 30 / STD 1 10 . .' .;cawa te r qS / o 0 I
) .
1 d: + 01 CONC LINED 1447 ncawater 0 ! j f [; fj 30 / STD 95 / 0
't ,t;
' \ fi 01 CONC LINED 765.6 .cawatcr o I } t * 's , . . , , ['; ' . 30 / STD 95 / 0 ) i h[ i
P
,1 ly'. ..
01 CONC LINED 30 / STD 2 3 b . r. ueawate-95 / e 0 00 gf. l g . <03> SS 65.1 acawa: .; o y 1
')
.r.'>{ , 30 / SS 95 / 0
,, ) 8* v. 01 CONC LINED 706.7 s e awa te.- 0
, [ '.I ;
- 30 / STD 95 / 0 I
l ' h 009 ' 01 CONC LINED - cavator 24
, k . f. .' -
30 / STD 'iS / 0 010 :# 01 CONC LINED 1 seawater 57.6 T f. ' . 30 / STD 95 / 0 4 0* . -
. 011' 01 CONC LINED 1 ceawater 12.74 y
m .
/ e 7.
, 30 / STD 95 / 0
$ ,., ,012 01 CONC LINED 1 scawater 0.3
' l e. 4 30 / STD 95 / 0 PIPE 013- ' C 01 CONC LINED 1 seawater 45.5 4 i fl'
#i 30 / STD 95 / 0
- 1 }y PIPE 014 01 CONC LINED 1 seawater 0 W
4 ..>,y { .! 30 / STD 95 / 0 3 . F'P PE015 01 CONC LINED 1 seawater 0 ! f 30 / STD M / 0 i i P'IPE016 01 CONC LINED 1 ceawater 0
.i N 30 / STD 95 j 0 i ! .2 dt
- f PIPE 017 <02> Steel 0.5 se
- .wa ter 1.572 i e d-e ..
1 ; 0.75 / 40 05 / 0 q .i PP 18. . <02) Steel 0.23 scawater 345
- f kdj.- 3 / 40 95 / 0
{ h r i; jO
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_ _- __ ---- ~~~-
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1
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- 7' ..;
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LINEUP REPORT rev: 5/25/93
'g. [. g[ .i f4,. c;x.x;Ien: o, ocolos 1,
'LI IST);,.7 5dp6 after: 9 iterations
( t
'h p kt4dV'05/24/93 n .w . .
t > :. i lines. i
. .g. :.a 5 .
*Wo etzi'c'. flow rates require in the constant first fluid prcporticIii pipe specification wm:cinused. all p pe
,{ .
,j,R1uid proporties s
{: l: Fe b .. DEMAND NODE DE!!.it:D g e;.-
.2
0d . NODE .' gpm
^
- r. f
' i '. ~ {+
t 2 iP ! I
$y >>> 100 lo . ,
+ t
.q : ..
s a sf ,, (Q;..tf. t, r NET FLO'AS QUT: 100 gpm
.? ..
l f -(' [d
',g .
1 f{ $, 4-Floh' PRCS SUP.E psi g E ^..' PRESSURE D CONNECTIONS Pipeline gpm E ? Node <<< 0 I . ,k v;. . PIPE 018 101.9
~CW' <<<
T 22 <<< 0.3
> .; 11682
.K' isa <<< PIPE 015 j p~, DISC >>> 4.2
,N, Yd 6 11950
>>> PIPE 016
.'1 (y9'.)- '
l L' INTAKE
<<< PIPE 017 66.13 <<< 0 e n . TAP I, .
4 $ ', 1 NET FLOWS IN: 100 gpn
, 3-I '.'. L' f s e .g. ,
!! 3.p" e
.1 s_ ;,et 4
5 > ad f.-. o p . a ,$i.a:; e. f A L 49 2 F F: . : q w :-
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a r- ---
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, ..q u. ; LINEUP 'I'DE'i ..' 75dp0 l ,. , t'. l- - ' ' '; I G cs j 5; e, , 2; 1. : g , i.; -
# ' ~# .'j g 11 Y f .
DEMA!D kifs . H ca;.on . DE'yf.'
.,y d ELEVATION
- gpm
- a i r; ft ft i
; I i.-j
[ t 2* 7
< .o h j 9 l -
34.25 11G.5
?<,3 ; E . 39.6 pr 4 .. . .,
115.5 i ~i . 3 6 ; i f; ' ~ *.' ; 39.6 .
- .15 .1 l
- .' (,' 39.6 3 3 . .. 3 l
'j pg%
;. 3 V p 0 9.6 l
,9h. j '
9.6 j ;
~ 3 . O '; 114.5
! . $. f D p.,.+. '? 39.6 y .. 24.47 1 ;l ? 17 p 3.3 ISiC @,,f. ;b': 2 g~..!;e. s ., 23.16 105.(
-' . v.
.. 39.6
,,~e '. 100.9 i
39.6 27.08
't FJ' -
1 . :1 5k.i G.44 99.44 f . "}'- N.~ 39.6
.4 s r -
26.37 99.28 .! . t! 39.6 \ }
' . L i.5 39.6 24.46 94.94 i + ?i v.* ' ' S , t 45.' t . p 4.2 9.504 l
- { I- AKE r;.
- O
> 100 19 82.59 f 39.6 f .
6.263 53.77
' 39.6 l i ,' 5, , .g 47.39 3.444 .
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! T ", 'H I 11848 5.753 1.916 4.335 l J h-.
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.? .olumatric flow ratec Fluid properties in the firo require const.cn..nipe.npocif.er . ion were usu.i.
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gpm --
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-lt .; < -
8 , lumetric flow rates require con:: tant fluid propertic:. a r. a J.1 pi pa l .i nec
' , '[ . i.4 pipe speci.'icnie n vere used.
- .J ;3 ' f:? Li V Fluid properties in the first
.1 L < NODE' DEMAND NC"E D - uS:,
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StayNiard'.'ref : MARK'S hat:DDCor. , 1 l CriElcalIPt: 220.55 bars at 270.976 dN -C _ .',,',' l
.h - 'ES' SED Ternperature Vapor Pressure i' - -
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ADOCUD'/M S ". l t
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j l' l.l . i i 5 '*I{ This Addendum to this cal:ulation is prepar' in orcier ro ;)rovidu
- g. . l ',a resulting curve that is similar the result: ty curve in the PSL -'
( j"ICW System Performance" calculation #FC5-2FJM-93-017, wherc {,r .. '? } combined allowable differential pres <ure (.w ) acroo: cCW iiX &
' strainer is plotted vs. ION inict terperr.tu 2 i y .
t- "M4 i
'c j %?Tdf achieve this, the point f rol.1 the combined allowable . 1 '. acrocs gQhCCWlHX".& strainer vs. ICW flowrate curve on ongo 1.: ot : afie th22 -
i .g gpcalculation, where ICW accident flow ic 12,v00 gpm and .-/ l
'tg ;0 5Maiid'(and ICW inlet flow is 95'F), is plotte.. on *. new cr4sh (cact
. I I i C 4Mof;32) with combined allewable w acrons ccW H:, L . trainer'vi. i
{Q 7 4 lAc%[ p. ICW inlet temperature. l i M.iTh'e accident conditions provided to AEU-C vn JP::-??".-s tNP o :-c ol , T ev*
.16.j0.
'k.,Z',9 .> ,D.,'." Input for LOCA Containment Re-Analynic", are.ac tollowc:
.s ?T C it . j.< Design UA = 4. 510X10' btu / hr
- F .
,.[ if '
'- ): .[:.jDesign CCW flow==12000 7305 gpn 3. 53 5X10[ lb/hr. j v
Design ICW flow gpm (9150*F) (995'F) or or 5.973X10" lb/hr. ' hdhiq h fe:h n L - CEj s .,: analysis (#007-AS93-C-004, DEDLS Min SI Case 3) resulted 4 N..fe, lif. , lie 'following outputs:
*i b H.
61 [. $# . Time = 5974 seconds (in recirculation mode). i". CCW HX ICW temp. in = 151.8'F. [* ",- s
- CCW HX ICW temp. out = 116.8'F. l j .h '.;.Ff. s Accident heat load = 35197 btu /sec. ( or '.2 6. X1C' btu / hr) . l
.h I' Y 3 c* q s;To plot another point on the new graph bo '. ng generated in thic 1
l i M calculation at the nore realistic (normal c:ert:ing) condition of 1
; ji[90*F, the above inputs and output were inser:M inte cur CCW systen ;
,Emodel in the therral computer program, *>Ersys.
. .:ith an ICW l
- f-ilfloWrate of 9,000 gpm, the computer run reculted in CCU HX inic
?S and outlet temperatures that ma ch the _:]: v e AEB-CE analvcis
.. ~
1~%resultstowithinl'F(seepages 5-11of32). h.1$
,1 .4FTo . find the allowable de across the CCW HX (at 9,000 gpa and 90'F) i i }':our model of the ICW rystem in the hydrarlic computer program,
}
I c PIPEFLO, was used and the K-value for the CC' Hh was adjurted (see
.ppages12-15of32)untiltheflowthroughitequatedto9,000gpn.
1 , g A resulting ap value of 23 psid was obtainoa wee page 2 of 32). ;
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' te#:that t n the resulting curve (page 1 of 32' , :_.:nc operation 1
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, if [ot'.? recommended combined CCW HX and in CCW the shadowed area under ctrainer differential the curve precoure is where betweenthe
, ": !!n C: 412.5 and 25.5 psid, based on tube degradation concerna.
1
< :. .,< r<t I i ..<The.2.5 psid basis is as stated in tscsumption / ~ ( cac.c 2 of 19) pact :',.wrating
! ! 'l'ifor the CCW strainer and the 10 psid is based cn; r ;
o ' experience that has shown that when the CCW HX .:P is c:r und 10 (1.c. shc11s)
,re foun; :n tne INpali,' excessive marine bicekage I
' ^:tubesheets and in the channel;v: ads, and b) ne: e .:: cad:ng ne tube i I ?l.inaterial velocity limits f or tube crosion.
) i:b . zw . . 3;. . , w C .,
a t es controlled'
- i. j,5 NOTE lt The PEGISYS program is a Westingherre pr:arr is l 7 p under the Westinghouse QA progra ra , and ac such,
['J - acceptable for use in this safety-rclatad application
.Q jf .{; . . (see pages 16-18 of 32). l L a > '
' MOTE 2: The PIPEFLO, Version 4.06 progran .eas ' al dated as shewn
, ( :
- on the included pages 19-32 of 32.
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r r> PSL-1 ICW PERFORMANCE CURVE p j COMBINED ALLOWABLE DIFFERENTIAL PRESSURE (DP) ACROSS CCW HX & STRAINER , ICWINLET TEMPERATURE (DEGREES F) . ~ - . - . - . . - . - . - . . - - - -
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Description:
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I I .: ' FEGI'SYS Analysis - J ' : .n : [ ',- y, PEGISYS TLCMAP Revisicn 3.0 . r.. : (odel: I.o0P2 . .u. ;
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.c Use Count: Il cale 10: Loo?9 [
"al'cj Titto: icw 9,000 gpm 90deg f '~
a . d I.Y uid: : WATER .
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,p i.i G,IN: Analysis package created on hu Aug 9 2 5 : M:,; ; o . .! ,.] "a EGISYS TLOMAP Revision: 2.0 muo 2~
t'Q 8 LOOP 2 Use Count: 11 C61c ID: Loore i kj ;a [ c ., 1hKTiile 'lew 9,000 gpm 90 dog f ~ lu'id s'. ! WATER
,- . ' ' (41t:. 5.pype: Non-Isothermal Globr.1 Ree;, F50 . . 0 .i
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I fr'.c , ., & .. y points for pump from node t
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I $, PEGISYS Analycio packago created on Thu '., 5 15:29: 20 179:. 2 { . ,.f by PEGISYS FLCMAP Povision: 3.0 P. a .- .' l 4 n Model ~ 1,00P2 Use counts 11 Cale ID: 1.cc a
. 1 . % Cale Titles icw 9,000 gpm 90c'eg f ~
f .E Fluid WATER j $ -},h Calc Type Non-Isothorrral Glcbal .;. :at. e c. ;t *
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u
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Input parametern for heat oxchanger frem node ' to nosc 4. j OO '" Flow orientations counter flow
-g ",',, g>N
' O Side 2 Flowrate (1bm/h):
- 4. 4 8 4 e+ 0 E ' f,' '~NJ '*'
t I h ,f.Y
', tj vaJ ,g>. p Side 2 Inlet Presoure(poig): ..
Side 2 Inlet Temperature (F): ( o 0 '-7 to
, .e, ,- .iI,, -,, ,
s gpy.5 , Design UA(BTU /h-F): r 9 L 9.t 51o+06 C 3 7- p *f f p c,. / . m , Leg'C .,,. i -. ,
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" ,O " j :l Mr. Jim Scarola r3..J B-GL- 512 5 E f, i; fhfPSLEngineeringProjectManager July 28, 1993 Or l . p.;
.; ., Florida Power & Light Company
!. : C . f,. St. Lucie Nuclear Plant
'}.g F ~,b(( '
P.O. Box 128
,g.[. Ft. Pierce, FL 34954-0128
.l
, , 4.g .
f4 f , Attention: K. K. Mohindrco
.'{p# ,
[5h5 i FLORIDA POTUE AND LIGHT COMMJ1Y
? St. Lucie Units 1 and 2
; i *.gt,j jq , PEGISYS 3.0 Computer Code Access and Trrining
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Reference:
NATD Letter Number FSE/SS-FPL-7698
"- FPL P.O. D90912-90187
,Ip .
FPL DWA 179107
, . ': Westinghouse GO-JBD10247
.1 L ')
~ 1r't .
Dear Mr. Scarola:
A $ '. rs'u d.^ .. IN hf I NIhDdringJuly13-l'4, 1993 site training wac provided on the
;; ; ~ .
use.of the latest version of the Westinghouse PEGISYS J Computer Code. Attending the training sessions frcu Florida
"'d k; i
Power and Light were Bill Neff and Joc Hege.
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P-*
.: *HCd./ft :Provided : at the training sessions war 2:
dTy I # / @?; l 1) Centrol Copy Number 3'1 of th:. PECISYS User's
.(( Jf: '$ 4 f Manual (Revision 3.0).
h% ~,* L
} j~ Mr ' 2) A copy of the 1cgen procedurs t: access-the
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PEGISYS co=puter code on thc. Wectinghouce NATD ge . SUN Computer Network ucing n perconel computer p ', ;. ,'- 1 (IBM Model 70 or 306/486 windowc coinpatible) and a
.g' - g% modem.
f2 y q'. . . . ,
; 4h I [b.,.D 3) A copy of the emulation software packagas XVision
-'* k .. b2' and XRemote by VisionWaro to accene the PEGISYS I &h }_ Of0+A' w .
.. l.. computer code emulating a UM X X-:crninal using a
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PC and modem. 4 a - Q. e n k, f.' t
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p p.. x '. . { $: n.C 16'
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Ih i ggI%'fsi. - Mr. Jin Scarola S ~ l '7 ?;
, ; 4 ' . ,-: July 28, 1993 v. 3c cL d;}. ' ~ -
3 Page 2 i 'M :$ .h, {t
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I t. V 3 -ild ! . 4) The PC-Connect terml.al emulation software package
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' calculati0n output M.d Plot fil1s frc.T tbc SUN
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F- - The PEGISYS 3.0 computer ccdc has been generattd, verified i i V and validated in accordance with the We stinghouse: :4uclear i and Advanced Technology Division' Quality Assuranca Procram
$ld.M in WCAP-9565. The controlled ccpies of the PrcICY3 Uner's
(,a il' Manual provide the r.ecessary inforretion neerled to w.rforn a
,j PEGISYS calculation and clso have been documented =
t e 5'.i O ,jaccordance with WCAP-9565. I
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fj.AnyindividualassignedacontrolledcopyoftheP20!SYS
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' -j Following any future revisions of this manual of the PEGISYS computer code, all the necessary infor-E(;;m@ationandpagerevisionswillbeissuedtothecontrolled f pjj ., -* tcopy holder.
Also attached is a copy of the torm used to 4
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Fidentify a problem with a Fluid Systems computer code. If N 9; jthere is a possible problem in the PEGISYS computer code,
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W , yr , Systems Computer Code Problem Report Form". h W ' A.'. :. ' b)
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4 7" 5d The completed form should be sent to Rogor Waters, h p [' Manager, Safeguard Systems, Mail Step 4-16 East, I.it i ~P. O. Box 355, Pittsburgh, PA, 15230-055 uno will 2" p R, lk? see that the cognizant engineer receives and acts
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*p~ jththetransmittalofthis inf or :.ation, '?AL is censidered 1 j; [y .
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{ 0 inriginal to you for your records, t < yptp ... ! l .! I? there are any questions, pleara fesi ::. .:e :: r ;:rc n t
.I 'e Marvin R. Wengerd at (412) 374-5C23.
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DESCRIPTION OF VERIFICATION MfiQop (o; .;/;.- u+ 3p,
'In5accordance with JPN-QI 7.1 (for Limited Occ ;.i urchcr.ed cc:tputer 1,
[J ; sof tware) , a cursory review of the venacr' a . ':ng .r m.rn'J ' 1c f tw rc , Inc . )
,M i- focumentation package revealed ev;.dcace of;cnior:aaoce verifi :cu.; gn or p;PEFLO, ersion 4.06, and that it demonctratdd t t, cc::cptable
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tandard engineering practices.and methods. P1c L [j, [ j fs ' . h ,J* I7 DESCRIPTION OF VALIDATION MElHOD (or M/A) EI ..The data for the "hvacfin1" (11/09/92, attacned to thic SWP) computer
< -model by Engineered sof tware, :nc. vill be re-entered ir. .a entirety, l f "jyli Jz the problem re-run and the recults cc= pared to the origina "hvacfini" f
i W.1 computer model verification results. The resultc should compare 2d and any differences should be reconciled as part of tho
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, Inter Office Correspondence l JPN-PSLP-96-0112 To: S. A. Valdes Date: [
St. Lucie Plant From: D. J. Denver Department: JPN/JB Nuclear Engineering.
Subject:
ST. LUCIE PLANT UNIT 1 , REA/ PROJECT # SLN-85-081-12 TITLE: EDG VOLTAGE, FREQUENCY SPEED PERMISSIVE SETPOINTS PC/M/ FILE # 120-194 Attached for your review, approval, and use is the MEP for the subject Plant Change Modification. This package provides the details necessary to document settings for the Unit 1 Emergency Diesel Generator VMR, FMR and SPEED SWITCH devices. This completes our efforts on REA SLN-85-081-12. l If you have any questions, contact Poul Christiansen at 691-2115 or Rick i Raldiris at 691-2104. DJD/PE 4 Attachment Cao.esto: T.M. Gerstner - OM/PSL ___ . . N .,/0C0. w r3t. (w/Reprot- , )
- 5. Kozlin - JDC/JB (w/ orig dwgs) +' 4 C.M. Spalter - JPN/JB .
R.L. Kulavich - SCE/PSL ' j K.K. Mohindroo - JPN/PSL .
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Form 1008 (Stockedi Rev tal
i ! 013-PR/PSL-1
,~ Revision 39 April,1996 Page 22 of 26 l
ATTACHMENT 2 PC/M REVIEW FORM (Page 1 of 2) PC/M Number 9 f / 20 Supplement Number O Expiration Date 82 /3'/ 9c PC/M
Title:
EOG Vo Itas e Fng s>&n c <f Speed Pe an,'s s aw Seyo;s n PC/M Classification: Major Modification or nor Modification Safety Classification:
/ Safety Related Quality Related Not Nuclear Safety Related Administrative PC/M Category:
dormal Either/Or As-Requested Package As Fail Generic Does the PC/M contain a Safety Evaluation? O Yes or E No is this a proposed change or modification to a L : stem or equipment that affects nuclear safety? 8 Yes or O No if either of the above questions is YES, FRG approval is required. FRG Review Required? @ Yes or O No~
/I4 h Date: 5/3 - /9G Conf @uration Management WY/5 /s TfA&dd Date: f/3 / 9W Plant General Manager FRG Number: % -/ l'T FRG Secretary: Nw wt Date: f / 3 / 96
'I A~ Jo G ,
FRG Comments: ( CixI
l 013-PR/PSLo1 i Revision 39 j April,1996 i Page 23 of 26 i ATTACHMENT 2
- PC/M REVIEW FORM (Page 2 of 2) '
\
implementing Documentation: Department ER/NfWO # W/O e
.C m CoS/7'30 0 C,31!5IL!
ErY I G6 / EM 3 os O// 5// j I I PC/M Review and Approval: '3 Yb Date 6 / 4 /'75 Configuration nagsqwnt ,
/ DiullLb A
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AU Date C / I I [.0 I Quali$ Control ! Comments: U i i Areas Affected: Yes or No Descnotion Operator Training Yes - As per Training Department Requirements Operating Procedures Ve s I _ C e u m.+ f . a s. Surveillance Procedures Ee's Maintenance Procedures [e s E W /06 9. C/ 4 4 M / E 9, o E ! Spare Parts Mo Drawings / Technical Manuals Yes Paqe d SRDs?h/ N FUSAR Chance No TEDB Change Ye s #7+aekmea+ b Human Factors (CREDIT) No Environmental Concems - No in-Service inspection NO Mainte' nance Rule NO Plant Restrictions No Others No Mh Date ' 5/_3 J ff,p Configffraton Management l 1
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~
PC/M 120 - 194 Revision 0 of Page 1 7 MINOR ENGINEERING PACKAGE (MEP) PLANT ST. LUCIE UNIT 1 PC/M HUMBER 120-194 SUPPL 0 ORIGINATING DOCUMENT REA-SLN-85-081-12 EXPIRATION DATE 12-31-96 PC/M CLASSIFICATION X SR QR NNS ADMIN TITLE EDG VOLTAGE. FREOUENCY AND SPEED PERMISSIVE SETPOINTS i ! ADDITIONAL REQUIREMENTS / INSTRUCTIONS YES NO AS-BUILDING TO COMMENCE UPON ISSUANCE OF PACKAGE? X THIS PACKAGE HAS THE POTENTIAL TO SIGNIFICANTLY X IMPACT PERSONNEL RADIATION EXPOSURE (See QI 3.13). IF YES, JPN FORM 72 IS REQUIRED. 10CFR50.59 SCREENING YES N0 1)DOES THE CHANGE REPRESENT A CHANGE TO THE I FACILITY AS DESCRIBED IN THE SAR? X 2)DOES THE CHANGE REPRESENT A CHANGE TO PROCEDURES AS DESCRIBED IN THE SAR7 X 3)IS THE CHANGE ASSOCIATED WITH A TEST OR EXPERIMENT NOT DESCRIBED IN THE SAR? X 4)COULD THE CHANGE AFFECT NUCLEAR SAFETY IN A WAY NOT PREVIOUSLY EVALUATED IN THE SAR7 X 5)DOES THE CHANGE REQUIRE A CHANGE TO THE TECHNICAL SPECIFICATIONS? X NOTE: IF THE ANSVER TO ANT OF THE ABOVE 10CFR50.59 5CREENING OUESTIONS l$ YES, THE MEP CANA. BE USED I REVIEW / APPROVAL: INTERFACE TYPE GROUP i l PREPARED VERIFIED APPROVED FPL APPROVED
- INPUT l REVIEW l N/A j MECH x ELECT x kkk% g7.12. & W
!&C x CIVIL x [
NuC - x L-47heuG/ l Esi x
/ I NUC FUEL x
* ** Review Interface As A Minlaan On All Non-ArWn MEPs F r Contractor Evals As Determined By Projects FPL PROJECTS APPROVAL: A DATE: MW/2 mm -
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PC/M 120 - 194 Revision 0 Page 2 of 7 ENGINEERING JUSTIFICATION DESCRIPTION This MEP documents setpoints and tolerances for EDG permissive control functions. permissives: Setpoints and tolerances will be provided for the following i Voltage Monitoring Relay (VMR): Tag Numbers K48/954 and K48/964 I Frequency Monitoring Relay (FMR): Tag Numbers K49/954 and K49/964 Speed Switch: Tag Numbers SPEED SW/959 and SPEED SW/969 The PSL-1 EDGs are required to reach full speed and voltage within 10 seconds after receiving a start signal. The VMR and FMR relays are used as i permissives in the automatic starting circuits of the EDGs for closure of the EDG output breakers (FSAR Section 8.3.1.1.7.d). The VMR and FMR relays ! i monitor the EDG voltage and frequency, respectively, and allow EDG output breaker closure when these values are greater than or equal to 90% of rated { voltage and frequency (FSAR Figure 8.3-5). The speed switches provide i permissives for several functions at three different EDG engine speeds, 50, 200 and 870 rpm (8770-10295 and 8770-11438). l j In addition to documenting setpoints as described above", this MEP also correctsdocumentation. existing miscellaneous documentation errors and makes clarifications to RATIONALE FOR SAFFTY CLASSIFICATION FSAR Section 8.3.1.1.7 states that the EDGs are needed to achieve safe shutdown of the plant or to mitigated the consequences of a loss of coolant accident in the event of a coincident loss of normal AC power supply. Therefore, the subject devices and this MEP s classified as Safety Related. 10CFR50.59 SCREENING JUSTIFICATION i This PC/M will not change safety related relay and switch settings as i described in the FSAR and other controlled permanent plant drawings. This ! PC/M documents the settings on the relay setting drawings (series 8770-A-452) and provides tolerances for the settings. This PC/M is not a change to the facility or procedures as described in the SAR, is not a test or experiment not described in the SAR, nor does it require a Technical Specification change. This PC/M does not involve an unreviewed safety question. Therefore, this PC/M does not fall under the requirements of 10 CFR 50.59, and it can be processed via the MEP format. DESIGN BASIS AND ANALYSIS ' I The settings and tolerances for the FMR r'elays are documented in PC/M 116-192M (Reference 5), which replaced the originally installed Westinghouse relays with Wilmar Electronics model 20-050X. The setting is 55 1 0.2 Hz. ' PC/M 116-192M did not incorporate vendor information into the EDG vendor manual (Reference 4). The vendor information is included in Attachment 2 for completeness, and will be incorporated into the manual upon asbuilding. The VMR relays are GE model 12PJVllA21, contact code 20 (two normally open contacts) according to Reference 7 and TEDB. FSAR Figure 8.3-5 states that the voltage permissive for EDG output breaker closure is greater than or equal to 90% of rated voltage (4160 Volts), which is 3744 Volts. The 4200:120
~
s PC/M 120 - 194
" Revision 0 Page 3 of 7 potential transformer (PT) reduces this value to 107 Volts. Based upon vendor data (Reference 4, pages 843-852) the repeatability of the relay on pick-up is 5% of the setpoint. To ensure that the relay pick-up is at least 107 Volts, 4
the setting must be at least 107/0.95 = 112.6 Volts, which incorpe'ater the 55 repeatability setpoint error. shall be 113.0Allowing a calibration tolerance of 0.4 Volts, the relay 0.4 Volts. The speed switch is a Synchro-Start model ESSB-3AT according to Reference 8 ' and TEDB. The speed switch settings are factory set and are listed on the side of the speed switch cover. They are 221, 883 and 3843 Hz (increasing) and correspond to speeds of 50, 200 and 870 rpm, respectively. The three speed switch settings in units of rpm are shown on CWDs 8770-B-327, sheets 959 l and 969, and schematics 8770-B-326, sheets 959 and 969 (see included 1 drawings). t j To verify the correctness of the speed switch settings, the methodology for i determining speed in Hz, using a magnetic pickup for the Hz signal, was applied as follows (see Attachment 3): i Speed (Hz) = (No. of Gear Teeth) x (Encine RPM Setooint_). ! 60 The 265. number of flywheel ring gear teeth, according to Reference 4, page 1043 is i Based upon the methodology, the settings are correct. - Based upon vendor documentation (Attachment 3) the error associated with the speed per 0.01% switch
'F, settings is 0.8% over the voltage range (100 to 140 V de) and typical.
Conservatively assuming a 100 *F temperature range { i results in 1.0% error. Using the square root of the sum of the squares
- method, this results in a 1.28% error. Adding a calibration tolerance of j l
0.72% results in a total tolerance of 2% of the setpoint. The resulting speed switch settings with tolerances, and the c- v ording EDG speed ranges expressed in rpm are as follows: ,
- i SWITCH SETTING EDG SPEED SW #1 221 4.4 Hz 49 - 51 rpm ,
SW #2 883 17.7 Hz 196 - 204 rpm i SW #3 3843 1 76.9 Hz 853 - 888 rpm 8 Attachment 3 also states that the automatic reset of the switches is between 80% and 90% of the setpoint. The reset feature is not adjustable. , 4 The EDG vendor manual (Reference 4) contains incorrect information regarding the speed switch. Reference 4, pages 747-758 pertain to Dynalco speed switches which were replaced with Synchro-start switches by PC/M 357-178 (Reference 6). The incorrect information will be replaced with pertinent information from Synchro-start (Attachment 3). DESIGN INTEGRATION REVIEW The Passport Safety Evaluation Database, was reviewed on April 5, 1996, to ensure that this modification does not invalidate the conclusions of any active safety evaluations. The "PC/M Log" and the "PSL Affected Drawings" CMS Data Bases were reviewed on April 5,1996, to determine if other outstanding PC/Ms or designs in progress will be affected by this PC/M. PC/M 030-196 (Reference 12), which replaces the relays in the EDG Idle Start /Stop Panel, and PCM 177-195 (Reference 13), which changes the control logic associated with the 200 rpm speed switch contacts may affect this PCM. A statement regarding the PCM's is included in the Special Instructions Section. 4
. - - . _ - -. - - -- -- - ~. - . _ _ -- -= --- --
PC/M 120 - 194 Revision _ 0 Page 4 _ of 7 OTHER AFFECTED DOCUMENTS _ Document Proposed Chances 1-EMP-59.01 and 1-EMP-59.01 Incorporate settings into procedures. j SPECIAL INSTRUCTIONS \TMPLEMENTATION REQUIREMENTS 1. Inspect wiring on VMRs, FMRs and Speed Switches to ensure that the wiring is in accordance with controlled drawings. Re-label wire between t speed IA and switch IB Idle (SS) and relay Start-Stop RF from wire number "1060" to "1008A" in the panels.
- (See PC/M drawings JPN-120-194-010
. and -012.) 2. Calibrate manual the VMRs (Reference (K48/954 4, pages 843 - 852)and K48/964) in accordance with ven and JPN-120-194-003. and PC/M drawings JPN-120-194-001
- 3. ,
Calibrate and PC/M drawings the JPN-120-194-001 FMRs (K49/954 and K49/964) and JPN-120 194-003. in accordance with i 4. Calibrate the speed switches (SPEED SW/959 and SPEED SW/969) in accordance 120-194-004. with Attachment 3 and PC/M drawings JPN-120-194-002and JPN-
- 5. t CalibrateJPN-120-194-005 drawings Tachometers on the EDG Idle Start /Stop Panels (see PC/M Attachment 3. and JPN-120-194-006) in accordance with f
6. Coordinate which replaces implementation the relays in the of EDCthis PC' 'ith DCM 030-196 (Reference 12), Start /Stop Panel, and PCM 177-195 (Reference 13), which changes the control logic associad with the 200 rpm speed switch contacts may affect this PCM. The dra, ng corrections and wire number changes shown on PCM drawings JPN-120-194-005 through -012 affect some of the same devices that are affected by PCM 030-196 and 177-195. POST MODIFICATION TESTING 1. Perform functional testing of affected circuits by starting the EDG limit unnecessary starting (preferably during surveillance testing o and verifying that the permissives allowing EDG output breaker closure)(K48 and K50) actuate in less than of receiving a start signal (see 8770-B-327, sheets 953 and 963)The 10 second start criteria is a requirement of Technical Specifications 4.8.1.1.2.e.3.b and 4.8.1.1.2.e.5.b. AS-BUILDING NOTES 1. The EDG vendor manual (Reference 4) contains incorrect informatio regarding the Synchro-start speed switches. Reference 4, pages 747-758 pertain to Dynalco speed switches which were replaced with Synchro-start switches by PCM 357-178 (Reference 6). Replace the incorrect information with information from Synchro-start (Attachment 3). 2. Incorporate Attachment 2 (Wilmar frequency relays) into Reference 4.
PC/M 120 - 194 Revision 0 Page 5 of 7 REFERENCES
- 1. St. Lucie Unit 1 FSAR, Amendment 14.
- 2. 1-EMP-59.01, Revision 9, "1A Emergency Diesel Electrical Periodic Maintenance and inspection."
- 3. 1-EMP-59.02, Revision 9, "1B Emergency Diesel Electrical Periodic Maintenance and insPaction."
- 4. 1/M 8770-6703, Revision 10. " Emergency Diesel Generator System."
- 5. PC/M 116-192M, Revision 0, "1A and IB EDG Automatic Starting circuit Frequency Relay Replacement."
- 6. PC/M 357-170, Supplement 1, " Diesel Generator Idle Start Modifications."
- 7. 8770-2421, Revision 19, " Diesel Generator' Material List."
- 8. 8770-10292, Revision 2, " Diesel Generator Idle Start /Stop Panel Physical Arrangement."
- 9. 8770-B-327, Sheet 953, Revision 18, "CWD Diesel Generator 1A Breaker."
- 10. 8770-B-327, Sheet 963, Revision 20, "CWD Diesel Generator 1B Breaker." ,
- 11. PSL Unit 1 Technical Specifications, Amendment 141.
- 12. PCM 030-196, Revision 0, "EDG Relay and Mounting Socket Replacement."
- 13. PCM 177-195, Revision 0, "DG 1A & IB Protective Trips Design Change."
ATTACHMENTS Attach Number Number Description / Title of Paaes 1 ALARA Checklist 1 ; 2 Wilmar Frequency Sensitive Relay 2 3 Synchro-Start Speed Switch 6 4 TEDB Change Package 6 l l l l I
-=
- \
i PEG ko : *20194 l REV 00 PAGE I0F) AHTCTED DRAWINGS PE j SMEET REV DESCRIPTION / TITLE DS AFFECTED DWG SMEET REV PR RV l JPN DWG # 0 TOTAL EDUIPMENT DATABASE TEDS 1 00 l EP ATTACMMENT 4 l 0 DNR PC/M 120 194 RELAT SETTING E 8770 A 452 1028 NEW 2 00 JPN*120 194 001 0 DNR PC/M 120 194 RELAT SETTING E 8770 A 452 103 NEW 2 00 dPN 120 194 002 0 DNR PC/M 120-194 RELAT SETTING E 8770 A 452 1068 NEW 2 00 ePN-120 194 003 JPN-120 194 004 0 DNR PC/M 120 194 RELAT SETTING E 8770 A 452 107 HEW 2 00 l JPN 120-194 005 0 CONTROL WIRING DIAGAAM DIESEL I 8770 8-327 959 14 1 00 l JPN*120 194 006 0 CONTROL WIRING DI AGRAM DIESEL. I 8770 8 327 969 13 1 00 l JkN-120 194 007 0 SCMEMAf tC DI AGRAM DIESEL-GENER I 8770 B 326 959 12 1 00 JPN-120-194-008 0 SCHEMATIC DI AGRAM DIESEL-GENER I 8770 B 326 969 12 1 00 JPN 120-194 0D9 0 D.G.1A IDLE START /STOP PNL SCM 8770 10295 2 2 00 JPN 120-194 010 '0 D.G.1A IDLE START-STOP PNL LEF 8770 10298 1 2 00 JPN 120-194 011 0' DIESEL GEN.18 IDLE START-STOP' 8770 11438 0 2 00 , 0 D.G.1s IDLE START STDP PNL.LEF 8770 11440 0 2 00 j JPN 120 194 012 1 1 707AL AFFECTED DRAWINGS REPORTED . - . - 13 l l l l l i l h
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sss!CTED vf400e uswuaL5 t I i - 9tG , REV VEN00R/ EQUIP DISC REMARKS REV ! 4: - PLANT DOC No $NEET t 8770 6703 EMERGENCY D!ESEL GENERATOR $YSTEM GENERAL DES- E SEE A$8UILDING NOTES 00 > [10 t TOTAL AFFECTED VENDOR MANUALS REPORTED * -- 1 1 1
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1 $ PC/M 120-194 REv. O ; ATTACEMENT 1 1 ALARA SCREENING F o; 1 OF 1 1
- 1. Is this PC/M Administrative only, i.e., no plant hardware is changed. ,
_X__ yes, Further ALARA screening is not required. no, Continue screening. j
- 2. Does this Pc/M involve a location in the Radiation Controlled Area (RCA;7 X yes, Location: Diesel cenerator sida Continue ~ screening.
no, Further ALARA screening is not required. ; i
- 3. Does the implementation of this PC/M involve any of the following? l I
-(Cross out all "no" responses)
- a. Movement of radioactive material.
- b. Potential for personnel exposure to a radiation field of > 1r/ hour as a result of l implementation, operation or maintenance (assuming current ares dose rates).
- c. A total lifetime estimated dose due to the modification (installation, operation, maintenance, removal) greater than or equal to one (1) man rem.
- d. Diving' operations associated with systems containing radioactive material.
- e. Entrance into containment during power operation for maintenance, surveillance, etc.
- f. , Movement of, or modifications to, existing permanent radiation shielding.
- g. Modification of systems containing radioactive fluids or resins such that system integrity is affected.
For items (a) thru (g): yes, This PC/M has the potential to significantly impact personnel radiation exposure. Complete Form 72 to ensure total radiation dose is minimized by design. , __X_ no, This PC/M has little or no impact on personnel radiation exposure. Form 72 not required, however, normal ALARA precepts should be followed to minimize radiation exposure. Prepared by Reviewed by Note: If the preparer and reviewer are the same as for the PC/M, duplicate signatures are not required. JPN Form #71, Rev. 7/90
J j . . BULLETIN 20 000
.D .All
. . .kiPV
. R .NQ
. \t . D . .
P7R 11.yU .11U I u .nu , i . ii . a u 9C M t w-i4 4-l ATTACNEMT Z REv. O
- PMG i of 2 l g \ ADJUSTABLE gg MODEL FREQUENCY TRIP RANGE Q tsN l
20-040 40-50 HZ. i 20-050 50-60 HZ. t 20-060 60-70 HZ. b 20-350 350-400 HZ. J l , 20-400 400-450 HZ. q . l SPECIFICATIONS LINE VOLTAGE:. ....... 120VAC 120%. VOLTAGE DRIFT: . ......11% maximuss frequency j error for input voltage of I'
- NOMINAL FIEQUENCT:.. 50, 80 and 400HE. 120VAC 110ll PICK.UP ADJUST:.......... The frecuency trip point OUTPUT CON 1 ACTS:......One set N.O., one set
) i is acrowdriver adjustabla N.C. I The adjustment range for I _ . . . . . ths var - - - ~*-i- = CQNTACT RATINGS:.. .. sam.msinthe_st.2RyDC l shown in the table am. or 120VAC. ! TEhPERATUflE RANGE:.... -55*C to 485'C. CONSTRUCTION:........... Solid state sensor with l l relay output in a sealed ! oirrnENTiAt:............. The frecuency pick u io steei can. l drop out differential is 1% maxinnsa. OPERAT10N:......... ...... The normally open con. l tects close, and the , j '
- TDrERATUIE DRIFT:....11% maxinun frequency normally closed contacts l
error over temperature open. at frequencies i - range. . ae me sei ,oini. 4 I i WILMAR :
"" "e<213332s.sssa WE ELECTRONICS, INC.
1 Revised December 1975 2430 AMSLER STREET TORRANCE. CALIFORNIA 90505 ]
i i . _ 4
- 9CM 120-le l CONNECTIONS DIMENSIONS "jd"U ,
- ., a = ~ = %e zoF1 e
? ? t e.
x d i- = h . . l: . o "o o[o _
! y OO GO
. m - ..,
j aa m n aa== "arancT" A eSA i o t r=s i .._e a j . W
-=
l NOTES: ..
- 1. Contacts are shown in the de-energiaod M l position.
l (Solow the trip set point)
- 2. Remove screw for access to pick g adjust y p
[ . = 11 i i MOUNTING CONPGURATIONS i m00n5
% @W /%
l 20 N0 2M4M 20 050 IN 20 0501 l-20 060 20 0601 4 b- ~ j 20-350 h. ' '=6 20 3501 = .1
- 20 400 20-4001 i
j OTHER WILMAR PRODUCTS l e VOLTAGE SENSITIVE RELAYS e PARALLELING IIELAYS 4 o CURRENT SENSITIVE RELAYS e IEVERSE POWER ltELAYS e PHASE SEQUENCE RELAYS e VOLTAGE IEGULATOltS I e PHASE FAILUllE RELAYS e HELAY TEST SETS
4
= .
j t l \ ESSB i i . 5. .g ..
. mL& FEATURES / OPTIONS RD 'F'9
O
! . -. vi f,,;r . . .~ e Adjustable Swrtch Posnts
! O' I s Sealed NEMA 4 Enclosure h.h O
. Trenseent and Reverse Potarity Protection FA6E l oPh 3
g.;.M. M '.i .t . e Precase Repetrtive Speed Switch Points l { , @@ s Autometsc, Manual or Remote Manual Reset j q e Power Loss Protectson
- D.C. Output for Model 30 Tech (1 ma)
! 3
, ?h y a 8-40 VDC: 64 VDC;100-140 VDC: or 100-130 VAC Power e Broken Signal Lead Protection j isi3 e Standard or Reverse Relay Logec
! . El a Lockwiro Seal e Vibration ! 'ation Krt Y ' s Conduit Cover 7F. R M o rigeW ;?.y.N EB?& SPECIFICATIONS ! .,% NW mi $3.~.6 .,,,a ELENCAL
= Powerinput j "'*gq:1.af$cn +.shrtit aM S @ ifr 7 . '?,W
. .;m + 8-40 VDC 64 VDC 100-140 VDC
. 100-130VAC j .- .-
- Reverse Polarity Protection j g . ,,,- 1000 VDC t ..-- e Maximum Current 12 Volt 24 Volt
. .. ~ .. .u _ (2 ch.8-40 VDC)
Operating 200 ma 220 ma ~ 4 Standby 50 ma 60 ma ! : Femmammeemed 4 l ,,,,m,, g=* i""""" Vy ~ p ;: % ' h i e Signallnput Minimum- 0.7 VRMS l
' j% ,4; ? L, 3 9 :
impedance-2K OHMS
'J d l ; 73#9sssssssss4
. .... ..WL.3 ....
- 7. ,g i
r --
= Relay Contact Ratings Resistance Load-0.1 to 10 amps 28 VDC Inductive Load- 0.1 to 8 amps 28 VDC l .# ... . ~ , ...
" Overspeed Response Time 4 T.I' M ' . . . . . . . . .
100 Milliseconds Typical e Setpoent Roset l .,.w,
-. Automatic- 80% to 90% of Setpoint i
! J... w-.,. ..i==u Adjustable-(Variabm Automatic) 50% to 98% of Setpoint l j e Setpoint Stability 0.8% over voltage range and 0.01% per 'F Typical l The mooel ESSB Electronic Speed Switch has many years
- of proven service.it is engineered to provide a wide vanety.
- of features and options including up to four switch points.
l adjustable reset,a selection of vottage inputs and DC cur- AECHANICAL/ENVD2ONAENIAL rent for tachometers. o Operating Temoerature l
- The ESSB Speed Switch is molded in a rugged NEMA -40*C to +85'C (-40'F to +185'F) i 4 enclosure which provioes protection against the rough e Vibration
! environmental conditions associated with diesei engines. Ag's 9 to 200 Hz 4
= Shock 4 Foot Drop Test i e mm I NEMA 4 High impact Molded Enclosure i e Weight
! 4.5 lb. (2 kg.) i l l
- - ~- ., -
G i l' UnderstaneHng the Electronic Speed Switch What is a Speed Switch? How does it work? . o Electronic device e Obtains signal from sensor oevice imag pickuo. Mint-Gen 3 o Senses rotatenai motion and soeed Signal Generator alternator or ignrton coin in tne torm of a Can be set to swaten a toad or control oevice a treauency. i e Frecuency is monitored electronically within the switch. 1
' s When the speed or frecuency reacnes the oesired setting or setpoint.the output relay tied into the control circuit is tnggered or switched.
l
- fts i?_ow
' &w 9 j SelectingctSpeed Switch R%n < in selecting the best speed switch for controlling a particular applicaton,the following factors should be considered: I o Number of Switch Poets dependent upon the selecton of the signal source,whch Most appicatons reauire only 1 or 2 switch points in turn depends upon the operating speed and physical 1 l' (channets). However, several Synchro-Start Switches of fer . mounting timrtatens. uo to 3 or 4 switch points. - 4 ' s Frequency Range of Apphcstion,Setpomts The vanety of signal sources wnich may be used with most' I and Signal Source SSPI Switches include: The treauency range of the applicaton and setpoints are i j e Magnetic Pickup-Commonly used for higher speed - applicatons.The output is a function of the " gap" between i the pickup and the gear tooth and the penpheral velocity of , the gear. For a Magnetic Pickup Signal Source:
- Setpoint Frecuency in Hertz = No.of Gear Teeth x Engine RPM Setooint 60 1
l e Min 6 Gon Signal Generator-Designed to provide outstand- ' ing signal output at speeds as low as 20 RPM. For a Mini-Gen Signal Generator Signal Source: - Setpoint Frequency in Hertz = Mini-Gen RPM at Engine RPM Setooint ' 2 e Alternstor Output-Used in applicatens unable to accom- ) modate a magnetic pickup or Mini-Gen Signal Generator. i For Alternator Signal Source: - Setpoint Frequency in Hertz = Pulley Ratio x No.of Afternator Poles x Engine RPM Setpoint ' 120 i e Ignition Output-Commonly used as a source for gasoline f and natural gas type engines. i ForIgnition Signal Source:
! Setpoint Frecuency in Hertz = No.of Cylinders x Engine RPM Setoomt 120
)
)
I 'l 4
PCM 610-lcl4 Wn WT ~:5 Selecting ct Speed Switch < cont.) Each signal source produces a different Hertz RekryOperation
- setpoint frecuency for identical engine RPM setpoints. Standard Reney Logec-With power apphed and a signal Example Reowred crank disconnect 300 RPM and over- below the setpoint.the relay will remain de-energtzed until speeo of 2500 RPM.This 8 cytinder gas engine has an 80 the setpoint is reacneo. At setpoint tne relay will be ener-tooth flywneel and 8 po6e attemator with a 2
- 1 pulley rato. gazed and remain energtzed until reset by one of the aDove methods.
Overspeed Reverse Reisy Logic-W'thr power apphed and a segnal below
. Sigiani Source i Crank Disconnect } 1 300 RPM i 2500 RPM l the setpoint.the relay will be energtzed and remain ener- < gtzed until the setpomt is reached. At setpoint or interrupbon Mag Pickup 80 x 300-400Hz 80 x 2500=3333Hz of power.the relay will de-energize and remain de-energtzed 60 60 until reset by one of the above methods or power is reap-a g s mm n us as part of a W
) Mini-Gen 300 ,35ogg M=1250Hz safe system to assure power is apphed to the relay J 1 2 2 dunng operaton. Alternator 2 x 8 x 300=40Hz 2 x 8 x 2500=333Hz l I 120 120 Both relay operatons are shown in the diagram below: 4
, Ignition 8 x 300=20Hz 8 x 2500=166Hz
- 120 120 TTPICALRELAYOPERATION
. 1 Condetson i stonnero noney LO9c i Reveres Metey Lage i
- ResetRequirements PowerOtt NO .NC NO NC There are four reset options available for resetting the l C' l C#
speed switch: Power On NO NC N NC
]
i Automatic Reset-With automatic reset option,the switch will At Setootnt NC NO NC automatically reset if the frequency of the input signalis l' NO'c l C/ lowered to 80% to 90% of the setpoint. At Reset l FJO j NC NC EXAMPLE: The setpoint is set to 1000 Hertz.With a 1000 Hertz signal appled *he relay will energtze.The unit will . automatically reset (relay degenenze) if the input frecuency is gjef l
- eM l W[
4 towered to 800 Hertz. (The unit may reset anywhere between 800 and 900 Hertz for the automatic reset if specified.) PowerRequirements identify the power source vanatons. Determine the maxi-Electncal Latch-With the electncal latch option.the relay mum and minimum voltage as well as transients of the will energize (after the setpoint has been reached) and power source, remain energized even if the input signal frequency has Determine the voltage and current switching requirements been lowered to 0 Hertz.The only way to reset the unit is of the extemal control circuit. Consider the inductrwe or resis-to remove power. tive nature of the load.lf load currents in excess of 10 amps Menual Reset-With the manual reset option, the switch is are expected. interface relays should be used. supphed with a reset button. By oepressing this button.tne umt wiH be reset. RDVIIOnmental COndiffons Identify the environmental conditions. Determine the maxi-Adjustable Roset-With the adjustable (automatic) reset opton, mum and minimum ambient temperatures. Also consider
- the switch will automatically reset at the frequency deter. vibration and shock.
mined by the setting of the supplied reset pot. By adjusting the pot.the reset can be selected anywhere between 25% Standard Features & OPffons Verify or Test Circuit-Standard on many models which and 95% of the setpoint on the majonty of models. enables the testing of the complete system while the engine is operating at a safe speed. When used.the switch will trip
! at approximately 67% of the setpoint.
Signal Loss-if protecton from possible loss of signalis
! desired.the ESSB offers optional broken signallead protec-tion bGontinuously monttonng for an open or high resis-tance in the signal source circuit.
Power Loss-if protection from " loss of power"to the elec-
! tronic unit is desired, specsfy the reverse relay logic opton.
Terminal Protection-Conduit covers and terminal protec-tors are available to ehminate exposed connectons and danger from high vottage on select mooels. 5
i l N M t],C - 8 4_ IN ST A*,LAT T ON & C USTwENT TNSTRUCT*TNS
*T :
' MODE *. ESSB ELECTRON!O SPEE0 SWITCH
$5% O I NGE.4 cFG !
;. woUNT!NG *NSTRUCT!ONS i The unit may ce mounted l Mounting dimensions are shown on drawing #4562.Install 4 eacn is4- colt 3 in either a horizontal or verttcal position.
in the .31" diameter mounting holes, i 4 i 2. ELTOTRICAL CONNEC'"!ONS TNSTRUCTIONS:
- a. General: STUDY the terminal locations and terminal layout l marxings on the unit. f18AWG, er larger diameter, stranced cepper wire must be used for ALL interconnecting wiring.
l
- b. POWER TNPUT terminais 19 and 20 are to connected to the power supply. The required power supply voltage and polarity (if D.C.),
The power supply may be l is marked on the side of the unit. grounded if desired.
- c. HZ SIGNAL !NPUT terminals 11 and 12 are to be pair twisted cennected te the cable j Hertz signai source. Ungrounded 13 j If it is impractical to route the intercennecting l
- recommended. '
wiring away f:cm high current sources and carr:ers, snielced twisted pair cable.should be used. Connect the snield to terminal 14 only. Cable must have an insulating jacket over the shield to prevent grounding of the shield at any other point. I
- d. SWITCM OUTPU* terminals are to be connected to the electrical cevices Deing controlled using wire sized to carry the required l current. Switch output contacts are isolated and rated 1C amperes i l
.) resistive load (approximately 5 amperes inductive) at either 115V A.C., or 2BV D.C. Consult factory for contact ratings at cther )
voltages.
- e. PIMOTE PISET (Local) (Optional) terminals 17 and 10 are to be ;
connected to the customer furnished momentary contact push button l only when this option is supplied. The push button contacts must be normally closed or normally open as indicated by the markings on the side of the unit. These markings also indicate tne j applicacle switch output number number,
- f. MANUAL RESET (Local) (Optional) push button is proviced an the i unzt cover only wnen this option is supplied. The unit marxings STD relay logic indicate the applicable switch output number. l' requires a jumper across terminals 17 and 18. REV relay logic in coes not require the jumper. If REMOTE RESET is desired addition to cover mounted MANUAL RESET, wire as follows: Remove j i
Jumper f rom terminals 17 and 18 on STD logic units and wire to the remote normally closed momentary push button; on REV logic units l wire terminals 17 anc 18 to the remote norma 11v eoen momentary push button. Switch relay (STD or REV) is marked on side of i: nit.
- g. FEVERSE P.E L AY LOGIO WITM wANUA* FESET (*ccal) (Optional) after power up, tne manuaA :eset cutten must =e pusned to energ:,:e relay i
fer correct operation. On un :s with reverse relay logic, the relays are energized below set point while with stancard relay logic, tne relays are energizec acove set point.
/ continued a
l SYNCHRO-START PRODUCTS, INC.
, . - gg*53.i, 6250 WEST HOWARD STREET, NILES,IL 60714-3433
' ::;;;;=;l~.S. -
Telephone: 708/967-7730 Fax: 708/967-7832 ,.,,,,,,,, l
. . . . .. A c 05AL408 Rev A Site
~
[7 ! .C - !%-
/MT ?
h.2 \; C St%&E S OF S h, TACH, OUTPU- (Optional) terminals are to be connected to an Nominal electronic tachometer only vnen this option is supplied.100 :.::rr-a-rerer and meter rating at full scale must ce betwect 1 mil 11 ampere. For a meter have a nominal rating of 500 micro- For a exceed "150 ohms. amperes, the meternominal resistance must of rating . not 1 milliampere, the meter meter having a Connect terminal 13 to the resistance must not exceed 400 ohns. meter terminal, and terminal (-) 14 to the negat positive (+) meter terminal.
- 3. ADJUSTMENT INSTRUC* IONS Tield adjustments are not
- a. General Adiustment !nst ruct ions ? Should reconumenced unless ACCURATE authorized adjustments be required, instrumentation is used.
remove the metal plateany by unscrewing the 2 #4-40 screws. Before attempting to make locations on the adjustment, carefully study the potentiometer unit to locate the CORRECT potentiometer. Remove the #B-32 screw over the CORRECT potectiometer which are 20 turn units. If a
' click' is heard and.no adjustment affect is observed, Potentiometer adjustmentsrotate the should potentiometer the opposite way.
be done slowly. Af ter completion of adjustments, replace seal screws and metal plates,
- b. SET H2 ADJUSTMENT,1 The side of the unit is marked to indicate the applicaele potentiometer identification letter and f actory SET HZ for each switch number. Clockwise rotation of the potentiometer raises the set point Hertz: the reset Hertz will also be raised.
- c. RESET HZ ADJUSTMENT * (Optional) If this option is supplied, the I
is marked to indicate the applicable side of the unat potentiometer identification letter and Clockwise the factory rotation RESET HZ for of the potentiometer each switch number. lowers the reset point Hertz (increases differential): the set point Hertz will also be lowered to a small degree. Cycle through both the set point and reset point several times to verify that settings are correct. l ELECTRONIC TACHOMETER CALIBRATION! (Optional) The *ACH OUTPUT potentaometer is inclucen wnen the TACH Output circuit is ! d. l supplied. To obtain access to this potentiometer, remove the #8-32 seal screw over hole 'J' . The potentiometer should be adjusted i l so that tachometer indicates the correct speed when the proper L' Hertz signal is applied to terminals 11 and 12. Adjustment is usually made at the speed requiring the most accurate meter reading. Clockwise rotation of the potentiometer increases the speed indicated by the meter.
- 4. MERTZ SIGNAL SOURCE The Hertz signal source must have a minimum output of 1.4 volts Peak-to-Peak for suitable operation above 100 Hertz. The signal must be a minimum of 3 volts Peak-to-Peak if either the set point or reset point is in the 60 to 100 Hertz range. The approximate input impedance is 4.5K ohms at al Synchro-i KH: and 3K ohms at 3 KHZ. Suggested signal source are:
Start Signal Generator or, b) Magnetic Pickup of suitable output.
~
[C M (Z C - A L As:Us Astr uNGEs reR c= M00EL E!SB E*.!TT* cNTO S PEEO 51'!?OH
\ Ph&E G oFG F>SGE NA ! 70 H: to 240 H y
- *mr *:0 M 030 H: to 525 H:
P.ANGr No. . 625 H: to 21:5 HZ RANGE No. 4 1300 HZ :o 4650 H: RANGE No. 5 2350 H: to 8400 H: p
/
/
'IGNAL GENERATOR USED fer HERT:.S:GNAL
" MINI-GEN" SI AL GENERATOR OUTPUT HERTE'- 1/2 " MIN -GEN" RPM
/ l The f =11 ewing chart indicat the applicable'Hert: settings to provice fifteen percent overspeed protection. ,- l
/
l MTIO Of TACH. I RATED SMD OF ENGINE I I ?AKECTF- ENGTNEI 900 RPM '.000 Rihd 6 ;;00 RP M i *500 epa e
'.900 RPM i 5- ; a 258 HD 2 8T H hs e 34$ 'I ' 431 H: .
517 H I i - 1 517 wa- 575 H: T( 490 H: i 962 H: '.035 wz l i i :- 1 i1035 HL ;250 H: i NQ380 H: i 'i25 H: 2070 H: t I The following chart indicates the applicable rt: settings f or cranking motor (starter) disconnect: , RATIO of TACH. I STARTER DISCONNECT \ SPEED l TAKEOFT? ENGINE I 500 RPM i 600 RPMi NO RPM i
.5: 1 1 125 H: 1 150 HZ l 17K HZ l
- 1: 1 4 250 Hz 1 300 HZ a 350 NZ I 2? . t 500 wz e 600 H: ' 700 H 2A i r N I d' i MAGNET!C PTCKUP used for HERT2 STGNAL If a agnetic pickup is used to monitor the flywneel ring gea: teeth, use the f ellow ng f c mula to cete.mine the applicaole Hert: setting:
HERT: - we. ef Geer ?eeth x Eneine r? w se::ei"- 60 BATT! h MARGING ALTEPRATOR used fer VE??" OURCE Usually two wires can be con. ted to a battery
/
.arging alternator system to obta:n single phase A.C. or hal. - ve rectifi single phase.
Use :ne f=11owing f or::ula to determ- the applicable Hert: se: ing. "P" desagnates the number of alternator les _ r .ple : 14) "P .R. " ces:gnates the Pulley Ra :.o (Example: 2.4). vEnr: - put.Ey :.Ar- x :o. er s :ErsAre= ?c .Es y _ s :E ==- rE Pc:N
/ ;20 K
h* SYNCHRO-START PRODUCTS, INC. 6250 West Howard Street, Niles,IL 60714-3433 iw . - . - Telephone: 708/967-7730 Fax: 708/967-7832 l-SAL-017 Rev A 6/94 Form 1014
o . ~ o Facility PSL Unit 1 PC/M or DCR No. 120-104 Attachment No. 4 l Revmon No. O i Page 1 of 7 i TOTAL EOUIPMENT DATA BASE CHANGE PACKAGE COVER SHEET l i l 1 Description of Change: UPDATED REFERENCES FOR AFFECTED COMPONENTS. . l Basis for Change:
- ADDED RELAY SETTING DRAWINGS. I
References:
i l Enrineerine ., Prepared by 6 7" N Date // d Verified by /7. P. O Date ll!7!9h Approv Date 4' (Disciplin'e Supv/ Lead Engineer) Configuration Manneement Reviewed by: Date: Data Entry by: Date: Data Entry Verified by: Date: For instructions, see QI 2.8-1 JPN Form 53, Rev 10/92
3 Facility: PSL unit: 01 te: Commenent: K69/954
, $d IM-IN Associates Atts h ,
Reva g , Date Printed: 11/07/94 Page of ] TOTAL EQUIPMENT DA',A 6'.tE SHEET EO Tag: N/A EQ Rev N/A EO Doc Pac: N/A System: 59 DIE $EL GENERATOR SYSTEM Seismics I Safety Class: _ Eng Ref O Grows 1E E0 surv Note: N/A E0 $ peer N/A RG197: ,
- EO Related: N EC Scews N/A RG197 Cat: ,
O Sasis EG Remarks: N/A RG197 Typer , Coup Type: RL Sub typet _ safety Channet: _ Pes: 116 292 Names FREQUENCY RELAY FOR DIESEL GENERATOR 1A RELAYING & METERING P Locn Code DG3/0G 1A CNTL PNL startup system: 053 1 Locn Desc: Insti MFG s: WILM WILMAR ELECTRONICS INC. Engineering Verified: Y instl Model 20-050X Rev: 000 orig Po: NY 422253 Coup Grows ~ NPRDS: N Acct No: 530 EO Tabs _ Insutstion Amytt . Train: , Scaffold Reg . Critical Cosps , Control Room Cosps , j Work Group: IST Reqd: N RWP Reqd: N j Maint Pges , ,,,,,,,, Drawings sheet: 8770 2421 8T70 B 327 954 8770 G 3sa . 8770 2417 ano.wa
/TrM o - A - 44 '2_ l o 2.T5 teen nanosts: ~
8770 6703 Procedures:
- 00062 n - sq. 01]
1- l om- , PCM/DCR WISTORY: DCR 5LE 92 253 RM l'ZO /T& M&S 8652 05540-9 Approved Alternate Mfg: Descriction Models tov. fnett. Ene Ver. ( WL5 WE511NGMousE tLau . n/r 000 ~ t y 117C717G03 000 Y (WES WESTINGHOUSE ELECT. Parameter Name Value: UOM: ; 4
.. . _ _ . . - - - _ . . ~ _ . - . .- -. . . . - .- . .
9 Facility: PSL Units 01 t.p , Component K49/ m RH12&44 Associates Att 4 Rev O I Date Printed: 11/07/94 Pege } of ] TOTAL EQUIPMENT DATA BA$E $NEET , l E0 Tag: N/A EQ Rev N/A E0 Doc Pact N/A , .i system: 59 ole 5EL GENERATOR SYSTEM 5eismic: I Safety Class: _ Eng Reft l 1 0 Group: 1E E0 sury Note: N/A EO Speer: N/A RG197: , j l
)
-10 Related: N E0 Scews N/A RG197 Cat: ,
1 1 4 0 Basist _ EO Romerks: N/A RG19T Type ,
, Coup Types RL $wb Types _ safety Channel _
Pca: 116 292
- Names FREQUENCY RELAY FOR DIESEL GENERATOR 18 RELAYING & METERING Locn Code Of.a/DG 1R CNTL Put Start w System
- 053 Locn Dese:
i insti MFG #: WILM WILMAR ELECTRONICS INC. Engineering Werified Y . l Instt Model 20 050x Rev 000 Oris Po: NY 422253 Coup ' Group: NPR0s: N Acct No: 530 EO Teb _ insulation Revis , Train: , j Scf.ffold Req: . Critical Canet ,, Centrol Room Comps , !
- i d
Work Grow: IST Reqd: N RWP Reqd: N ! i 1
- =iate-:_ _ _ _ _ _ _ _ _
! Drawing: Sheet: 8770-2421 . 8770-8 327 m 8770 C 388 8770 2417 f 8770-2419 i n 1 .. _ 8T70 6703 l Pro,cedure.s e .... nme ,e.tae- c .o D PCM/DCR NISTORY: DCR $LE 92 253 M&S 8652 05540, 9 (g j pgg
, k.
Approwed Alternate Efg: Dear r h* i M
- h det? Rev. Insti . End Ver. /
JiEE WESTINGNOUSE ELECT. N/F M , t 7
\WE5 WESTINGNOUSE ELECT. 117C717G03 000 . Y j t
P;rameter
;, Name: value: UOM:
d
i l
- l Facility: PSL unit: 01 LMD: ,
M 12(>-/ M Components K48/9g Associates Att Rev C Date Printed: 11/07/94 Page g of ] fotAL EQUIPMENT DATA SA$E 5NEET E0 Tag N/A EO Sev: N/A EO Doc Pact N/A system: 59 O!EsEL GENERATOR SYSTEM Seismics I Safety Class: _ Eng Ref t l 0 Grow: 1E E0 sury Note: N/A E0 speer: N/A RG197: , EO Related: N EO Scow: N/A RG197 Cat , G Sasist _ EO Remarks: N/A RG197 Type , Cony Type AN $@ f ype s _ Safety Channet: Pca: Name: VOLTAGE RELAY FOR DIESEL GENERATOW 18 RELATING 1 METERING Locn Code: DG8/DG 18 CNTL PNL Startup System: 053 Locn Desc Insti MFG #3 GEN GENERAL ELECTRIC Co. Engineering verified: Y I instl Model: 12PJV11A21 Rev 000 crig Pos NY A22253 Coup Group: NPtos: N Acct No: 530 EO fabs _ Insulation Amvt , Train , Scaffold Reg , critical Comp: . Control Room Comp: , Work Group: !$f Reqd: N RWP Regd: N
** int Pen =3 . .__ _ _ . _
Drawing: Sheets 8770 2421 8770 8 327 9M 8770 G 388 8770 mn.2,417 no ( v 7 70-A - 462 i eco -.. _ - - _ l0G S _ 8770 6703 pe a ..... O-ese-er.o2-) ~ PCMbCRNI$f0RY: CCR SLE 92 253 Approved Alternate Mft:
Description:
Modet: Rev. Instl. Eng Ver. GEN GEhERAL ELECTRIC Co. N/F 000 Y GEN GENERAL' ELECTRIC Co. 12PJV11A1 000 E Y Parameter . Name: VaLue: UOM:
. ~
g l Facility: PSL units 01 Lg , k ~M Campanents c48/954 Assoc ste Atts 4 Rev O " Date Printed: 11/07/94 Page g of j
./ '
TOTAL EOUIPMENT DATA BASE $NEET EQ Tas: N/A E0 Rev N/A E0 Doc Pac: N/A
! system: 59 DIESEL GENERATOR SYSTEM seismics I safety Class: _
Eng Ref 0 Group: 1E E0 sury Note: N/A E0 Speer: N/A RG197: ,,
. E0 Related: N E0 Scews N/A RG197 Cat: ,
e Basis: E0 Remarks: N/A RG197 Type: ,, coup Type: AN 54 Types _ safety Chamet Pem: Names YOLTAGE RELAY FCR DIESEL GENERATOR 1A RELAYING & METERING Locn Code: DG8/DC 1A CNTL PNL startup system: 053 Locn Desc: Ir.stl MFG #3 GEN GENERAL ELECTRIC CD. Engineering verified: Y Insti Model 12PJV11A21 Rev 000 Orig Po: NY 422253 Camp Group: NPROS: N Acct No: 530 4 EQ Tabs _ Insulation Rawls Train: , , Scaffold Req: , Critical Comp: , Control Room Comps . Work Grow: IST Reed: N RWP Reed: N Maint Pens: , , , ,, , , , , ,, 4 l
. Drawing: sheet 8T70 2421 8770-B 327 954
. 8770-G 388 8770 2417
'? "
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~
p 232. .A -442 Tech Manu
/o 2.5 '
8770-6703
.- ~.
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p p.,a . - . (/- EnP- H. o (j l' CPC.m i Lo - s9 + 3 Approvec a nernate - Cfs: Gescription: Modet Rev, instl. Eng ver. GEN GENERAL ELECTRIC Co. N/F 000 , Y ] GEN GENERAL ELECTRIC Co. 12PJV11A1 000 E Y eerameter , Names veLue: UOM: a 4 4 i
I
. ~ \
3 l w Facility: PSL Units 01 Lies ,
- Conoonent: $ PEED SW/969 Associate: Atts i Rev Date Printed: 11/07/94 Petea [ of 7 d
TOTAL EQUIPMENT DATA BASE $NEET EG Tes: N/A EO Reva NA E0 Doc Pact N/A system: 59 DIESEL GENERATOR SYSTEM Seismics I $afety Class _ Eng Ref t 0 Croup: 1E to sury Note: N/A E0 Speer: N/A RG197: _ EG Related: N to scows N/A RG197 Cat: , O Basis _ EQ Remarks: N/A RG197 Type: , Camp Type: CK 52 Typer _ safety Chamet: _ Pce: Name: SPEED SWITCM FOR DIESEL CENERATOR 1B START Locn Coder DG8/DC 15 IDLE PNL Startie System: 053 Locn Desc: Instl MFG #: $YN STWCNRO $ FART PRtm , INC Engineering Verified: Y Inst! Model: Essa.3AT Rev: 000 orig Po: 59169-597468 Camp Croup: NPRD$ Y Acct No: 530 l EG Tab _ insulation Anwls , Train: _ scaf f sid Req: _ Critical Comes . Control Room Cano: , ! I Work Crow: IIT Reqd N RWP Reqd: N I a.intP _ _ _ _ _ ____ l Drawing: Sheets 8770 B-327 969
- 8770 10292 8770 11A38 5770 11040 8770 11441 1776-1M7o '
u Iecn ti770 - 4_ 46 'l
--. ; fo3 8770-6703 Procedur c O-MF-H,02.J cos... l PCM/DCR HISTORT: st r et Mt l
Approved At ter .yC t-1 110- I A 4-
, i l
Cfs:
Description:
Monet: Rev. Insti. Eng Ver. l P;rameter Names Values UOM: i VENDOR COMPARTMENT, N/A l BREAKER RATING AMPS SiEAKER FRAME N/A BREAKER Coll N/A . BREAKER $ENSOR TAP W/A BREAKER TRIP AMPS BREAKER TYPE [,._ N/A l.T. PNASE ,J ,, AMPS 5.T.T. PNA5E ,_ AMPS L.T.T. PNASE - AMPS L.T.A. PNA$E ~ AMPS
$fARTER TYPE N/A STARTER SIZE N/A STARTER WIRING FIC. N/A REMOTE STARTER TTPE N/A REMOTE STARTER SIZE N/A CABLE $12E N/A
_. . .~_. ._ i l l l e ! facility: PSL Unit 01 LMO: , lN" i Consonent SPEED SW/959 I Associates Atts l Rev: Date Printed: 11/07/94 Page]of] j
)
i TOTAL EQUIPMENT DATA SASE $NEET EQ Tag: N/A EO Rev N/A EQ Doc Pac: N/A System: 59 DIESEL GENERATOR SYSTEM Seismics I Safety Class _ Eng Ref O Grows if EQ Surv hote: N/A EQ Speer N/A RG197 , l EQ Related: N E0 scows N/A RG197 Cat , RG197 Typer , j 0 Basist _ EQ Remarks: N/A Camp Type: CK Sut:i Type: safety Channe!: Pce: Name SPEED SWITCH FOR DIESEL GENERATOR 1A START Locn Code: DG8/0G 1A IDLE PNL Startup System: 053 Locn Descs Instl FFG #: SYN SYNCHRO-START PROD, INC Engineering verified: Y Insti Model ES$9 3AT Rev: 000 Oris Po: 59169 597468 Camp Group: NPRDS: Y Acet No: 530 EQ Tab: _ Insulation Amytt . Train: ,, Scaffold Req: , critical Comp: , Control Room Comp: , Work Grows !$f Reqd N RWP Reqd N Maint Pges ,, , , ,,, , , ,, Dtaving: Sheet: 8770 8 327 959 8770 10292 8770 10295 4 8770 10297 ) 8770 10298 8 .inpoo
~
j L ! /O ) f 8770 6703 Procedur m - o 0l PCM/DCR HISTORT: Drs " F N Approved Alternet Rev. Instl. Eng Ver. Mfs:
Description:
Modet: Parameter Name: value LAM: VENDOR COMPARTMENT N/A BREAKER RATING AMPS BREAKER FRAME _ N/A BREAKER Coll N/A SREAKER SLNSOR TAP N/A BREAKER TRIP AMPS BREAKER TYPE N/A I.T. PMASE AMPS 5.T.T. PMASE AMPS L.T.T. PMASE AMPS L.T.A. PMASE AMPS STARTER TYPE N/A
- TARTER $12E N/A STARTER WIRING FIG. N/A REMOTE STARTER TYPE. .
N/,A
. l , . 1 ,, .- r--
_. lg- r-0mee Correspondence nI ' s s.\ t* I pp( U ' td' -l2 7PN-PSLP-96-0110 To: S. A. Valdes Date: APR 2 41996 ! St. Lucie Plant ' From: D. J. Denver o periment: JPN/JB Nuclear Enginee ing subject: ST. LUCIE PLANT UNIT 1 REA/ PROJECT i STAR 2-951391 TITLE: DELETION OF EDG AUTCMATIC START ON CIAS AND CSAS PC/M/ FILE # 96064 i Attached for your review, approval, and use is Engineering Package 96064 for the subject Plant Change Modification. This package provides the details necessary for the disconnection of the
- CIAS and CSAS contacts which provide automatic start signals for the EDG.
Performing this modification eliminates the possibility of running the EDGs in parallel with offsite power with the nonessential EDG trips blocked. j A SIAS trips the EDG breaker, and a bus undervoltage signal trips the 4160V bus supply breaker; in both cases the EDG is separated from offsite power. This modification also makes the EDG starting logic consistent with that l of Unit 2. EDG automatic start on SIAS and undervoltage are not affected by this modification. i By copy of this letter, Document Control is requested to distribute this package to the normal distribution. f 4 If you have any questions, please contact A. R. Dodd at 691-2171 or Rick Raldiris at 691-2104. Gdb DJD/A //(.- j Attachment C*'" *
~~_.~-_
T.M. Gerstner - OM/PSL N H.L. Fagley/DCC - CS/PSL (w/ Repro)
- l ._ O
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i , S. Kozlin - JDC/JB (orig dwgs) .' j K.K. Mohindroo - JPN/PSL ,.,." O r. 1 C.L. Schaeffer - JMC/JB i . l C.M. Spalter - JPN/JB " re- . . . . I Others: / ~~- - evC ,I i F:WISWCWSPSLWOW.009
- _f /
/
s.- inos isaa.s> n., vis ,
i QI 3-PR/PSL-1
. Revision 39 April,1996 Page 22 of 26 ATTACHMENT 2 PC/M REVIEW FORM (Page 1 of 2)
PC/M Number %OG4 Supplement Number O Expiration Date 12/3l /_36 PC/M
Title:
DELETroM OF EDs ALSTornATic. STAW oN CIAS AND c3As PC/M Classification: Major Modification or Minor Modification Safety Classification:
/ Safety Related Quality Related Not Nuclear Safety Related Administrative PC/M Category:
/ Normal Either/Or As-Requested Package As-Fail Generic Does the PC/M contain a Safety Evaluation? Y Yes or O No is this a proposed change or modification to a uns'ysystem or equipment that affects nuclear safety?
8 Yes or O No if either of the above questions is YES, FRG approval is required. FRG geview Required? Yes or O No lb i - Date: 5 / ? / %6
' Configuration Manabont Date: I/ / fI Plant General Manager FRG Number: H -l2.1 W FRG Secretary; Date: T/ 4 / 96 FRG Comments: u Mna4 C y )
l l l i
013 PR/PSL-1
. Revision 39 April,1996 Page 23 of 26 ATTACHMENT 2 PC/M REVIEW FORM (Page 2 of 2) implementing Documentation:
Department ER/NPWO # W/O # IC 63 l T 2A i et L,oi n [M JL 6 sl 54 gi'
%ot t Act (
l i PC/M Review and Approval: b- t.L D ate 5 f 3f G -
. Configuration Manaqtment aL A M% ="*lC M DateifLg Comments:
Areas Affected: Yes or No Description Operator Training Yes As per Training Department Requirements Operating Procedures VES OP l- 04 00050 (see. %2 T oP PC/n) Surveillance Procedures VES (C)Ph) _ Maintenance Procedures Spare Parts VES l[ M lI.tc-) Mb Drawings / Technical Manuals MS SRDs? hl N FUSAR Change TEDB Change YES Human Factors (CREDIT) MO Environmental Concems IdC) in Service inspection MC) Maintenance Rule E Plant Restrictions DM EDE. ter A TME - tu mobEc 5,6. OR CEFUp:f J:D. Others QQQq , ,
. Y0 Date 5 lbb Configuration M gement
. ~
FPL NUCLEAR ENGINEERING ENGINEERING PACKAGE i 96064 REV: 0 SUPPL: 0 PC/M NO: PLANT: St. Lucie UNIT: 1 TITLE: Deletion of EDG Automatic Start on CIAS and CSAS REVISION DESCRIPTION: Issued for Use LEAD DISCIPLINE: Electrical EXPIRATION DATE: 12/31/96 REA NO: STAR 2-951391 DWA NO: PC/M CLASSIFICATION: SR XX QR NNS DESIGN ORGANIZATION: JPN/PSL/ PEG DISC CHIEF REVIEW REQD? YES __ NO X EXTERNAL INTERFACES: None DISC CHIEF SIGNATURE: REVIEW / APPROVAL: INTERFACE TYPE GROUP PREPARED VERIFIED APPROVED FPL APPROVED
- INPUT REVIEV N/A MECH X N/A N/A N/A N/A ELECT X d.M O D b oM N/A
!&C X N/A N/A N/A CIVIL X N/A N/A N _ . N/A NUC'* # N k N[4 N/A
! N/A N/A ESI X N/A N/A hUC FUEL X N/A N/A NA N/A
- For Contractor Prepared EPs As Determined By Projects ** Review Interface As A Ninimum On All EPs !
l l FPL PROJECTS APPROVAL: A'ock [# DATE: #/5Mt
1 PC/M No 96064 i Rev 0 Page 2 of 27 LIST OF EFFECTIVE PAGES 2 REV. PAGE NO. REV. PAGE NO. l 1 0 Attachment 4.4 2 0 1 0 i 3 0 2 0 i 4 0 3 0 l 5 0 4 0 6 0 5 0 7 0 8 0 Attachment 4.5 l 9 0 1 0 10 0 2 0 4 11 0 3 0 12 0 4 0 13 0 5 0 14 0 6 0 15 0 7 0 16 0 8 0 17 0 9 0 18 0 . 19 0 Attachment 4.6 1 l 20 0 1 0 21 0 2 0 I 22 0 3 0 23 0 4 0 24 0 5 0 . 25 0 6 0 ; 26 0 7 0 27 0 8 0 Attachment 4.1 Attachment 4.7 1 0 1 0 2 0 -, 3 0 Attachment 4.8
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l 1 0 Attachment 4.2 2 0
- 1 0 Attachment 4.9 1 0
. Attachment 4.3 1 0
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PC/M No 96064 Rev 0 Page 3 of 27 ABSTRACT While operating Emergency Diesel Generator (EDG) 2A in parallel with the offsite source during the 1995 Unit 2 refueling outage, Containment Isolation Actuation Signal (CIAS)- A was manually actuated as part of the ESFAS (Engineered Safety Features Actuation System) testing per procedure OP 2-0400050 (Ref. 5.13). Upon resetting CIAS-A, the EDG tripped due to actuation of the reverse power relay. Subsequent investigation has revealed that the EDG ran with the generator in a reverse power condition (i.e. generator motoring) for approximately 45 seconds. The reverse power occurred immediately upon initiation of CIAS-A; however, the trip function of the reverse power relay was blocked. When the CIAS-A signal was reset, the relay trip block was removed _ and the EDG tripped. Subsequent to the event, the EDG was successfully tested under load and examined during performance of the 18-month preventive maintenance. No evidence of damage was found. A review of the Unit 2 EDG circuits showed that initiation of CIAS-A caused the EDG governor to change from one frequency preset value to a second preset value. The second preset value for EDG 2A was less than the offsite source frequency; therefore, the fuel racks acted to reduce fuel to the engines and slow down the EDG. When the EDG frequency tried to drop below the offsite source frequency, the generator started to act as a synchronous motor to maintain the speed of the EDG. The direction of power flow changed from out of the generator to into the generator. This actuated the reverse power relay, which was unable to trip the EDG due to the block enabled when CIAS is present. The control circuits of the Unit 2 EDGs were modified by PC/M 156-295 to delete the automatic EDG start and block of EDG non-essential trips on CIAS or CSAS. The EDGs supply reliable power to those electrical loads which are needed to achieve safe shutdown or to mitigate the consequences of a design basis accident in the event that the preferred ac power source is interrupted. Presently, actuation of a Safety Injection Actuation Signal (SIAS), Containment Spray Actuation Signal (CSAS) or CIAS automatically starts the Unit 1 EDGs and blocks all protective trips except for overspeed and generator current differential. Actuation of a SIAS also opens the EDG output circuit breaker if the EDG was running either in parallel with the offsite source during testing prior to SIAS or during a Loss of Offsite Power (LOOP). With either the EDG breaker or the 4160V bus feeder breaker open, the EDG runs completely separated from offsite power and there is no possibility of a reverse power condition for the generator. Automatic actuation of a CSAS requires a SIAS; therefore, the CSAS automatic start of the EDGs is redundant to the SIAS start and the EDG breaker would open and prevent the reverse power condition. However, a CSAS without SIAS, due to a spurious actuation of the CSAS relay, would not open the EDG circuit breaker. This could result in the EDG remaining paralleled to the offsite source with most of the protective trips blocked, increasing the potential for EDG damage. I
' i 1
, s. . ~
i PC/M No 96064 i Rev 0 Page 4 of- 27 i i Similarly, a CIAS without SIAS also does not open the EDG circuit breaker, which could . result in the EDG remaining paralleled to the offsite source with the protective trips ! blocked. CIAS is generated from high containment pre:sure ano high containment
- radiation. The high containment pressure setpoint is the same as that for initiation
! of SIAS. Any of the FSAR Chapter 15 analyzed accidents which would geneme a CIAS in , Modes I through 4 would also generate an approximately concurrent SIAS. CIAS is not i required for Mode 5. During Mode 6, the high radiation setpoint for CIAS is decreased
- in order to monitor for a fuel handling accident. Therefore, during Modes 1 through j 4, a CIAS would not occur without a SIAS, but during Mode 6 a CIAS could occur without
, SIAS. A CIAS without SIAS could also occur due to a spurious actuation of the CIAS
- relay. Under the circumstances of an EDG running in parallel to the offsite source and
- _ a CIAS without SIAS, the EDG would be exposed to an increased potential for damage.
I i The scope of this modification is to delete the CIAS and CSAS EDG start and non-i' essential EDG trip block signals. Following the modification, the EDGs will automatically start and the non-essential EDG trips will be blocked only for SIAS or i for a loss of offsite power. In either case, the EDG would be disconnected from the < offsite source. This would eliminate any possibility of running the EDG in parallel l with the offsite power with the protective trips blocked. Due to the differences between the Unit 2 and Unit 1 EDG control circuit design, a reverse power incident
- similar to what occurred on Unit 2 could not occur on Unit 1. However, the Unit 1 EDG
- logic is being changed in order to provide increased protection against other j potentially damaging situations which may arise when the EDG is paralleled with offsite i power and the EDG trips are blocked. Sufficient post-modification testing is
- incorporated to ensure that CIAS and CSAS no longer start the EDGs and that the automatic EDG start on SIAS remains functional. The design analysis ensures that there
- are no adverse effects to any other component or system required for mitigation of a l CIAS without SIAS event.
! The EDGs provide backup AC power in the event of a loss of offsite power to those l components and systems necessary for a safe shutdown of the plant and accident
- mitigation and are therefore considered nuclear safety related. This modification l affects the starting circuits of the EDGs and is therefore classified as nuclear safety l related.
i
- A review of the changes to be implemented by this PC/M was performed in accordance with ;
- the requirements of 10CFR50.59. As indicated in the Safety Evaluation (Section II), i this PC/M does not involve an unrevic=d safety question, nor does it require t.
j
- - revision to the Plant Technical Specifications. This modification will have no effect
! on plant safety and operation. Prior NRC approval is therefore not required for the ! implementation of this PC/M. i l i 1
_ _ - . _ - _ _ ..~_ _ . _ ._ _ . _ _ . _ _ _ _ . _ . . _ _ . _ _ _ _ _ . _ ). -- PC/M No 96064 ! Rev 0 ! Page 5 f* i of 27 l r ! I. Dfilgg :
- 1. Structure. System or Comoonent (SSC) Involved:
l l i ' This modification affects the St. Lucie Unit 1 EDGs, which are part of the Class i i 1E 4160V ac electrical system, and the ESC-SA and ESC-SB Panels, which are part j of the Engineered Safety Features Actuation System. T
)
- 2. SSC Purpose / Function /Desian Basis l-
_ 2.1 Class 1E 4160V ac electrical system l. ) The Class IE 4160V ac system distributes power to its connected safety
- related loads which are required to achieve and maintain the reactor in a j safe shutdown condition and to mitigate or monitor the consequences of a .
2 design basis accident. ! t l 2.2 Emergency Diesel Generators (EDGs) , i *1 ! 2.2.1 The EDGs supply reliable power to those electrical loads which are needed ; i to achieve safe shutdown or to mitigate the consequences of a design basis l 1 accident in the event that the preferred ac power source is interrupted. l l q 2.2.2 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.
l 2.3 Engineered Safety Features Actuation System (ESFAS) 2.3.1 The ESFAS automatically initiates safety injection, containment isolation and containment heat removal systems in the event of a design basis accident. It is comprised of Safety Injection Actuation Signal (SIAS), Containment Spray Actuation Signal (CSAS), Containment Isolation Actuation Signal (CIAS), Recirculation Actuation Signal (RAS) and Main Steam Isolation Signal (MSIS). The present ESFAS design also starts the EDGs on SIAS, CIAS and CSAS in order to provide rapid availability of an emergency 1 power source for the ESF equipment in the event of a coincidental' or j subsequent loss of offsite power. ; 2.3.2 The design basis for the ESFAS is to actuate those components necessary for mitigation of a Design Basis Accident as analyzed in Chapter 15 of the FSAR. The response times for the various system components is required to i support the maximum system response time, as determined in the Design , Basis Accident analysis. TMs response time must include sufficient time i to reestablish power from the EDGs and sequence on the loads in the event . of a loss of offsite power fo those' components dependent upon ac power. l
- 3. Safety Classification: SR.,Y1 QR NNS l l
PC/M No 96064 Rev 0 Page 6 of 27 , l
- 4. Reason for Desion Chance i
Testing of the EDGs require that the EDG under test be manually synchronized and loaded in parallel with offsite power. During this time, all protective trips, such as reverse power, of the EDG should remain operable. Actuation ; of CIAS or CSAS without SIAS, either manually or by spurious actuation of the l relay, results in blocking most of the protective trips and does not open the , EDG circuit breaker. This would result in the EDG running in parallel with ! offsite power with most of the protective trips blocked. Should an abnormal j condition develop, the EDG would be unable to trip with the resultant i possibility of damage. Deletion of the CIAS and CSAS automatic EDG start and : the protective trip blocks for CIAS and CSAS will eliminate any possibility I of running the EDG in parallel with offsite power with the protective' trips l blocked. The SIAS automatic start and block of most of the protective trips - of the EDGs is not changed by this modification. l l
- 5. Desian Chanae Descriotion The design change consists of removing from the EDG logic circuits the CIAS and CSAS relay contacts which initiate EDG start and block non-essential EDG I trips.
5.1 Panel ESC-SA (Unit I control room) Permanently disconnect two cables, remove three jumpers, and remove and i reconnect the wire of one cable. 5.2 Panel ESC-SB (Unit I control room) )
~
Permanently disconnect two cables, remove three jumpers, and remove and reconnect the wire of one cable. 5.3 DG 1A Control Panel (Unit 1 EDG building) 1
. Permanently disconnect two cables !
5.4 DG 1B Control Panel (Unit *1 EDG building) Permanently disconnect two cables ' i
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PC/M No 96064 Rev 0 Page 7 of 77
- 6. Desian Chanae Checklist 4
Does the Design Change involve / impact / require justification of: YES !LO REFERENCE Internal / External Flooding X Heavy Load Handling _ X Tornado / Intern:1 Missiles _ X Single Failure Criteria .JL. Human Factors X Paging System Audibility X Masonry Block Wall Interaction X Environmental Criteria X Plant Security Capability X HELBA Criteria / Analyses X Seismic Qualification X Seismic Interaction X Electrical Separation Criteria X Accessibility /Laydown/ Clearance Requirements X Loads Applied to Existing Structures (+ buried) X EDG/ Battery Loading / Load Sequencing X Para. 8.1 Hydrogen Generation In Containment X Heat Sinks in Containment _ X Emergency Plant Operating Procedures (E0Ps/0NOPs) X ASME Code X Emergency Lighting Criteria X Snubber Criteria X Material Compatibility / Hazardous Materials X Electrical Equipment Grounding X Cable Tray Seismic Loading _ X Instrument Setpoints X Hurricane / Tornado Wind Loading X Thermal / Hydraulic Performance X Coatings Inside Containment __ X Emergency Response Data System (ERDS) X Emergency Planning X NHL Property Insurance Requirements X Es91Msiiie~htiKQ6silficifiun X Att. 4.2 Fire;!Protectionicipabilitj X Att. 4.1 Safe /Al ternateishutdswn( Caji~atiil i ti X Att. 4.1
.ALARAiExposureltri_tsria"~^^^~ _
X Att. 4.3 FoiliDEE3FjlMJQ{iH{_ls'3Majliidjg[jMiffjyE365Elli~s]yfjf6K}hife]t(siiis?
l
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- PC/M No 96064 i
Rev 0 [* Page 8 of 27 L
- 7. Desian Evaluation / Justification
. 7.1 EDG Control Circuits j' 7.1.1 Control Circuit Description
- The EDG automatic start circuit is presently actuated by CIAS, CSAS, SIAS, or i a. 27Z relay contact, which is actuated when there is an undervoltage condition on the 4160V bus via a 27X-11 (Train A) or 27X-1 (Train B) relay, or if the 4160V bus feeder breaker opens (CWDs 957 & 967, Ref. 5.3 & 5.4).
! These inputs energize relays ASR, K16 (operate coil) and K16X/0. Contacts 1- from both the ASR and K16X/0 relays actuate relay ESR which, in turn, j actuates the air start solenoid valves to start the EDG. A contact from l l relay K16 actuates the air start solenoid valves directly, in parallel with , the contact from relay ESR. Contacts from relay ASR also release or block ! the idle start relays and block the normal shutdown relays. Separate SIAS, i CSAS and CIAS contacts from panels ESC-SA and ESC-SB are used to block the ! nonessential. EDG trips, such as high Jacket water temperature, low oil ! pressure, reverse power, and overcurrent (CWDs 956 & 966, Ref. 5.5 & 5.6). j A separate SIAS contact trips the EDG circuit breaker upon actuation of SIAS.
- l
- The governor is switched from droop (paralleled with another power source) to i
. isochronous (operating separate from other power sources) mode by contacts l
- l. from the 4160V Bus IA3 incoming feeder breaker from Bus IA2 and the 4160V Bus :
183 incoming feeder breaker from Bus 182 (CWDs 957 & 967 and 8770-2417 & 8770 i -11447, Ref. 5.3, 5.4, 5.7 & 5.8) . These contacts switch out the . load i sensor, insert a droop adjust resistance in the frequency sensor circuit, and l cause the governor to control only on the frequency sensor circuit. The modification deletes the CIAS and CSAS start contacts from the EDG l automatic start and protective relay block circuits. All other control l functions remain the same. j 7.1.2 EDG Operation Presently, the Unit 1 EDGs automatically start on CIAS, CSAS, SIAS or 4160V
- bus undervoltage. The governor is switched from droop to isochronous mode 4
only when the 4160V bus feeder breaker opens. If the EDG circuit breaker was i closed, it would be tripped on SIAS only. This is done to place the EDG in i standby condition if it were running in parallel with offsite power before the SIAS or to load shed and sequentially load the safety loads if there had been a LOOP prior to the SIAS. After modifying the circuit, the Unit 1 EDGs will automatically start on SIAS or 4160V bus undervoltage only. The governor and voltage regulator operation will not be affected. The EDG circuit breaker will trip on SIAS as before. l l I I
, - - - - ,- _s . - . , - nI
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~ PC/M No 96064
~
Rev 0 1 Page 9 of 27 j 7.1.3 Essential Equipment List / Safe Shutdown Analysis The equipment being modified is on the Essential Equipment List (Ref. S.12); however, no equipment is being deleted or modified in such a way that the ? Essential Equipment List is affected. Four cables (CIAS and CSAS start , signals to EDG 1A and 18), which are currently essential cables, are being i made spare. Following a review of. the Appendix "R" Safe Shutdown Analysis, ) it was determined that the only change is to delete the "B" train cables , (10967E-SB & 10967G-SB) from the listed cables in fire zone FZ-57. See Attachment 4.9. 7.2 Other Components I An evaluation was performed to determine the effect of this modification on other components actuated by CIAS or CSAS. The results of this evaluation are given in Attachment 4.4 and are discussed in Section II.2.7. No components were found that affected EDG operation.
- 8. Evaluation of any "YES" responses in Desion Chan' eo Checklist 8.1 EDG/ Battery Loading / Load Sequencing
! The automatic start of the EDGs for CIAS and CSAS are being deleted by this '
- PC/M. This modification does not add any new loads or remove 'any existing loads from the EDGs, nor does it change any of the load block times.
Therefore, EDG loading and load sequencing are not affected by this '
- modification. Battery loading is slightly reduced by the disabling of two control relays in each DG Control Panel; however, the change is not j significant and does not affect the battery loading calculation (Ref. 5.9).
- 9. Desian Verification Statement j The assumptions required to develop the design criteria were adequately described, reasonably and appropriately identified for subsequent reverification if required. The design inputs were correctly selected and incorporated and applied operating and construction experience were considered.
The drawings and documents used in the design were checked to assure the latest revisions were utilized. The codes, standards and regulatory requirements were properly identified and the modifications were shown to ! meet those requirements in addition to the FSAR commitments (FSAR description of automatic EDG start on ESFAS signals required a revision due to the design change). Interface commitments were shown to be satisfied. Periodic and special test requirements were verified to be correct. The output provided by this design was reasonable when compared to the in;:ut. The acceptance criteria for the modification were adequately incorporated j into the design documents to allow verification that the design requirements
- have been satisfactorily accomplished. The design verification has also shown that the EP has been correctly classified as " Safety Related".
4
.. _ _ _ . _ _ _ _ . _ _ _ . . __. _ . . _. _ - _ _ . . ~ _ . . . _ _ _ _ _ _ _ _ _ _
PC/M No 96064 Rev 0 Page 10 of 27 4 II. 18EDI
- 1. Descriotion and Purcose
, This engineering package provides the details necessary to delete the Unit 1 i CIAS and CSAS EDG start signals such that the EDGs start automatically only for SIAS or a LOOP. The purpose of this modification is to eliminate any possibility of damage to the EDGs, due to spurious actuation of CIAS or CSAS relays, which may result from operating in parallel to offsite power with the protective relay functions blocked. ]
~
j 2. Analysis of Effects on Safety l 2.1 Safety Classification
- The modification considered in this EP is classified as Safety Related because it affects the ESC panels and the starting of the EDGs, each of which
! perform a safety related function. Deletion of automatic EDG start on CIAS
- and CSAS represents a change in the EDG control logic and a physical change i in the ESC and DG Control Panels.
1 2.2 Effect on Existing Evaluations and Designs in Progress ! A design integration was performed to assure that the modifications do not 5 affect any other designs in progress or any active Safety Evaluaticns. In performing this review, the following documents were utilized:
- 2.2.1 St. Lucie Affected Drawing and Document Index Computer data base was accessed on 2/17/96.
j Four designs in progress were noted as affecting the drawings included in j this package:
- a) PC/M120-194; documents setpoints and tolerances for EDG permissive
- control functions. No effect on EDG start modification.
! b) PC/M 177-195; modification to make existing DG mechanical trips
- active during the 10 minute idle cooldown period. No effect on EDG
, start modification. c) PC/M 180-195; modification to annunciator system to prevent DG Fire Control Panel alarm or failure from blocking other subsequent DG alarms from being given in Control Room. No effect on EDG start modification. d) PC/M 030-196; replaces sockets & relays in DG Idle Start /Stop Panels. No effect on EDG start modification. 2.2.2 St. Lucie Active Safety Evaluation List data base accessed on 4/17/96. No safety evaluations were listed under the EDG system (system 59) which are affected or which could affect this modification. I i
i . . j PC/M No 96064 Rev 0 i ' Page 11 of 27 2.3 Actuation Logic 1 CSAS automatically actuates the Containr.ent Spray System (CSS). CSAS is initiated by a coincident 2/4 high-high containment pressure signals and a simultaneous SIAS. SIAS automatically actuates the Safety Injection System (SIS). SIAS is initiated by either a coincident 2/4 low pressurizer pressure signals or a ~"" coincident 2/4 high containment pressure signals. l CIAS automatically actuates the Containment Isolation System (CIS). CIAS is initiated by a coincident 2/4 high containment pressure signals, E a ; coincident 2/4 high containment radiation signals , E on SIAS actuation. , l 2.4 Technical Specification Applicability CSAS is required by Technical Specification 3.3.2, Table 3.3-3, to be l OPERABLE in MODES 1-3 for the High-High Containment Pressure Function (s 10 l psig) and MODES 1-4 for the Manual Actuation Logic. In addition, the CSS is required by Technical Specification 3.6.2.1 to be OPERABLE in MODES 1 & 2 and , MODE 3 with pressurizer pressure a 1750 psia. l SIAS is required by Technical Specification 3.3.2, Table 3.3-3, to be ! OPERABLE in MODES 1-3 for the High Containment Pressure Function (s 5 psig) and the Low Pressurizer Pressure Function (a 1600 psia) and MODES 1-4 for the Manual Actuation Logic. In addition, the SIS is required by Technical Specifications 3.5.2 and 3.5.3 to be OPERABLE in MODES 1-4. CIAS is required by Technical Specification 3.3.2, Table 3.3-3, to be OPERABLE in MODES 1-3 for the High Containment Pressure Function (s 5 psig) and the High Containment Radiation Function (s 10 R/hr) and MODES 1-4 for the Manual Actuation Logic and for the SIAS Actuation Logic. Additionally, the High Containment Radiation Function (s 90 mR/hr) of CIAS is required by Technical Specification 3.3.3.1, Table 3.3-6, to be OPERABLE in MODE-6 to automatically close the containment purge valves when high containment radiation is detected while in MODE 6 to limit offsite dose in the event of a fuel handling accident. The CIS is required by Technical Specification 3.6.3, Table 3.6-2, to be OPERABLE in MODES 1-4. The CIS is also required by Technical Specification 3.9.9 to be OPERABLE during MODE 6. Two indeF -dent Shield Building Ventilation Systems are required by Technical Specification 3.6.6.1 to be OPERABLE in MODES 1, 2, 3, and 4. l Two independent control room emergency air cleanup systems are required by !
, Technical Specification 3.7.7.1 to be OPERABLE in MODES 1, 2, 3, and 4.
There is no applicable Technical Specification for the Fuel Handling Building , Ventilation System. , i l i
.. e . j PC/M No 96064 Rev 0 i Page 12 of 27 } 2.5 - Applicable ESFAS Design Functions .j Automatically initiate containment heat removal systems, when conditions ! j indicative of a major high energy line break inside containment are detected, !
- to prevent overpressurization and potential loss of containment pressure
! boundary integrity. ! Automatically initiate the safety injection systems, when conditions l indicative of a LOCA are detected, to limit the extent of fuel cladding l damage. Automatically initiate containment isolation and the shield building i ventilation system, when conditions indicative of a LOCA are detected, to limit the uncontrolled release of radioactivity. l l Automatically start the diesel generators, when conditions indicative of a i LOCA are detected, to provide rapid availability of an emergency power source i for the ESF equipment in the event of a coincidental'or subsequent loss of offsite power. J l Automatically align t'he control room ventilation system for emergency operation, when conditions indicative of a LOCA'are detected, to maintain control room habitability. ) Automatically close the containment purge valves, when high containment radiation is detected while in Mode 6, to limit site dose in the event of a j fuel handling accident. i 2.6 Applicable System Design Functions ! Containment Sorav System (CSS) - Provide, in conjunction with the containment i fan coolers, sufficient heat removal to prevent the containment pressure and ! temperature from exceeding analyzed design values following a LOCA, CEA j ejection event, or Main Steam Line Break (MSLB) inside containment, assuming l a single failure coincident with a LOOP. l j Hiah Pressure Safety In.iection System (HPSI) - Automatically inject i
- sufficient borated water into the Reactor Coolant System (RCS) to recover the 1 core in the event of a LOCA. Inject borated water into the RCS during excess
! heat removal events (e.g. steam line breaks) and CEA ejection. ! Containment Isolation System (CIS) - Isolate fluid systems that pass through 4 containment penetrations such that any radioactivity that may be released into the containment atmosphere following a postulated Design Basis Accident . (DBA) is confined. The containment isolation system is designed such that no 1
- single active failure in conjunction with a LOOP could result in offsite i doses or doses to the operators in the control room in excess of 10CFR100 or j GDC 19, respectively, i
i-4
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- PC/M No 96064 Rev 0 Page 13 of 27 DieldBuildinoVentilationSystem(SBVS) - Limit the pressure rise in the l Lhield Building annulus following a LOCA so as not to exceed the Shield Building internal design pressure, assuming a single active failure, and i controlling and collecting leakage from the containment penetrations. '
Establish and maintain a subatmospheric pressure in the Shield Building
- annulus following a LOCA, assuming a single active failure, and to resist the n.aximum potential for exfiltration under all wind loading conditions .
! characteristic of the site. Provi d scans of reducing offsite doses :
- resulting from post accident leakagt by evacuating air from the annulus and
] routing through the SBVS filtration units, assuming a single active failute. , Provide means of reducing offsite doses by providing fission product removal
- _ capability while evacuating air.from the Fuel Handling Building during a fuel j handling accident (FHA), assuming a single active failure.
Control Room Emeroency Cleanuo System - Assure that no single active failure
- coincident with a LOOP can result in loss of functional perfomance.
{ Maintain the control room envelope at an average positive pressure above that ! of the surroundings following a LOCA. Provide means to limit .the 3 introduction of airborne radioact.ivity, smoke, toxic gases, or steam to the j control room envelope. Provide air cleaning for the control room envelope
- atmosphere so that airborne radiological doses, experienced by control room
! personnel following ' a DBA, do not exceed the limits imposed by GDC 19. Assure that makeup air bro'ught in during an event that has resulted in i control room isolation does not bypass the air' cleaning process before mixing with the control room envelope air. l 2.7 Evaluation i
- The Containment Spray System (CSS) actuates on a Containment Spray Actuation l Signal (CSAS). The CSAS will not actuate the CSS without a concurrent Safety l Injection Actuation Signal (SIAS). The Emergency Diesel Generators (EDGs)
! automatically start on receipt of a SIAS. As such, EDG auto-start on a CSAS 3 is unnecessarily. redundant. Any initiation of the CSS will be preceded by or j concurrent with a SIAS initiation and auto-start of the EDGs. Thus, the '
- auto-start of the EDGs will occur without any time delay associated with
- removal of the CSAS EDG auto-start function. Therefore, removal of the EDG i auto-start function on CSAS is acceptable, i
i Similarly, the Containment Isolation System (CIS) actuates on a Containment i Isolation Actuation Signal (CIAS). The CIAS will actuate the CIS on high i containment pressure, the same signal and setpoint which will actuate the
- SIAS. The CIAS will also actuate the CIS on receipt of a SIAS signal.
Since the EDGs automatically start on receipt of a SIAS, two of the three ~
- CIAS actuation conditions result in a concurrent SIAS actuation and EDG start. Thus, the auto-start of the EDGs will occur without any time delay
- associated with the CIAS EDG auto-start function for two of the three CIAS i
- actuation logics.
The only CIAS actuation condition which would not result in a concurrent SIAS 1 (and EDG auto-start) is the "high containment radiation" signal. 'Upon , completion of this modification, such a CIAS would result in initiation of 5 CIS without a concurrent EDG auto-start. Such a situation will result in a
, delay, of up to 10 seconds, in emergency AC power being available should a LOOP occur subsequent to the CIAS.
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j PC/M No 96064 I Rev 0 , Page 14 of 27 i For MODES 1-4, the only possible accident scenario for which the plant could l receive a high radiation signal prior to a SIAS would be a small break LOCA l (SBLOCA) which could possibly initiate the high radiation signal without a ; 1 concurrent or subsequent high containment pressure signal. and thus a l i concurrent SIAS. In this scenario, the break in question may or may not : j cause a loss of coolant in excess of makeup pump capacity. l If the loss of coolant is within the makeup capacity of the charging pumps, i
- the operator would begin an orderly shutdown of the plant without the need 1
. for EDGs. Should a LOOP occur during this shutdown (due to the manual :
l~ reactor trip), the additional 10 seconds required for the EDGs to come up to : i speed would be of no consequence since the break is relatively small when i compared to the makeup capacity. Thus, there is no need for a CIAS EDG auto- ' start for this scenario and the SBLOCA spectrum analysis bounds this event. i I If the loss of inventory was not within the makeup capacity of the charging l 4 pumps, a reactor trip would occur on Low Pressurizer Pressure. Further, i reduction in pressure would generate a SIAS and subsequent EDG auto-start. 4 Although the EDGs would possibly be operating sooner if actuated by a CIAS : auto-start (i.e., high radiation signal initiates CIAS before low pressurizer l
- pressure trips reactor and initiates SIAS), the St. Lucie Small Break LOCA 1 l Analysis (FSAR Section 15.6.6.2.2) conservatively assumes that offsite power i is lost upon reactor trip and that the safety injection pumps must await a 30 1 second delay for diesel startup and load sequencing following the SIAS. In )
essence, no credit is taken for an early start of the EDGs due to receipt of i l the CIAS prior to the SIAS. Therefore, no benefit is derived from a CIAS-
- initiated EDG auto-start in this scenario.
i For MODE 5, CIAS is not required to be OPERABLE. Therefore, this i modification will have no effect on plant operation in MODE 5. i For MODE 6, CIAS is required to function on high containment radiation only, i since there is no capability to generate a high containment pressure signal ! in this mode. In MODE 6, the CIAS safety function is to provide containment [ isolation (close the containment purge valves on a high radiation signal) in ! the event of a fuel handling accident inside containment. Per Attachment l 4.4, the containment purge valves do not require AC power to perform their safety function. In addition, the FHA as described in the FSAR is bounding i for this event. It assumes that no isolation functions occur which require ; i EDG auto start on CIAS. Therefore, removal of the EDG auto-start function on j 1 CIAS for MODE 6 will have no adverse effect on plant safety. ! 1 Finally, to further validate that there is no effect on CIAS/CIS actuated I components (and thus the applicable accident analyses) as a result of
- removing the EDG auto-start capability on receipt of a CIAS, a review was 4 performed of the components that are actuated on CIAS. Per Attachment 4.4, only the following CIAS actuated components require EDG AC power
- a Shield Bldg Vent Exh Fans HVE-6A & 6B j (Design Req'at. LOCA; Assumes LOOP) a Control Room Emerg Filtration HVE-13A & 13B -
i Fans f (Design Req'at. LOCA; Assumes LOOP) i l 1
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, - - . - -. - . - -- - . - - _ . - - .. - ~ . - . - . - - _ _ - - . . . - - . - l4 PC/M No 96064 l Rev 0 , j Page 15 of 27 l l .
- a Control Room Isol Valves FCV-25-14,15,16,17,18,19 j 24, & 25 '
j (Design Req'mt. LOCA; Assumes LOOP) i j u Control Room Emerg Filtration D-18 & 19 ! Dampers : (Design Req'at LOCA; Assumes LOOP) )i i ! e Primary Water Isol Valve MV-15-1 i (Design Req'mt. LOCA; Assumes LOOP) ; i : i ~ e Instrument Air Isol Valve MV-18-1 ! I (Design Req'at. LOCA; Assumes LOOP) { i As stated in Section 2.6, each of these systems / components' design basis i ! assumes these components are capable of performing their safety function in 8 the event of a LOCA and/or a LOOP. The EDGs will continue to auto-start on i ' either a LOCA or LOOP after this modification. Therefore, no loss of safety 1 function will occur as a result of this modification for any of these i i components actuated by CIAS. Since the remaining components actuated by CIAS ) l do not require EDG AC power, this modification will have no adverse impact on l . their safety function. i
- i j No components are being added or removed from the ESC or DG Control Panels.
Therefore, the seismic characteristics and qualifications of the ESC-SA, ESC-SB and EDG 1A and IB Control Panels are not affected by this modification. i 3. Failure Modes and Effects Analysis (FMEA) i
- The contacts of the CIAS and CSAS logic output relays used to start the EDGs
! and block the EDG lockout relays are being disconnected in the ESC-SA and i ESC-SB Panels and the DG-1A and -18 Control Panels by this modification. The associated auxiliary relays in the EDG-1A and -1B Control Panels and two
- diodes in each of the EDG Idle Start /Stop Panels and associated wires are being abandoned in place. No new components or systems are being added. An evaluation of all components actuated by CIAS and CSAS has been performed, as i discussed above and documented in Attachment 4.4, with the result that no i other systems or components are affected by this modification. Therefore, i
the failure modes and effects of the EDG, Containment Isolation System, and
- Containment Spray System are not affected by this PC/M. -There are no new 1
- failure modes created by this modification.
l l ! l
- 4. Effect on Technical Specifications 3
The Technical Specifications were reviewed for potential effects associated , with this modification. The applicable Technical Specificatic,ns (for ESFAS,
- Emergency Diesel Generators (AC Power), and those components actuated by CIAS i and CSAS) were identified and a determination was made that this modification j will have no effect on the PSL-1 Technical Specifications. A review was
- performed on MODE applicability, as documented in Section 2.4, to ensure that
{ plant operation, in the MODES for which CIAS and CSAS are required, also j remains unaffected. i
4 . . . - PC/M No 96064 Rev 0 Ptne 16 of 27 l
- 5. Unreviewed Safety Ouestion (US0) Determination The Code of Federal Regulations, specifically IOCFR50.59, states that the holder of a license authorizing operation of a production or utilization facility may (i) make changes in the facility as described in the Safety Analysis Report (SAR), (ii) make changes in the procedures as described in the SAR, and (3) conduct tests or experiments not described in the SAR, without prior NRC approval, provided the proposed change, test, or experiment does not involve a change in the Technical Specifications or pose an Unreviewed Safety Question (USQ). A proposed change, test, or experiment shall be deemed to involve a USQ (1) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety a previously evaluated in the SAR may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any previously evaluated in the SAR may be created; or (iii) if the margin of safety as defined in the Basis for any Technical Specification is reduced.
NSAC-125, " Guidelines for 10CFR50.59 Safety Evaluations," prepared by the Nuclear Management and Resources Council and the Nuclear Safety Analysis
! Center of the Electric Power Research Institute (EPRI), is used as a guidance document to determine whether an Unreviewed Safety Question is involved in the activities associated with this modification.
l 5.1 Does the proposed activity increase the probability of occurrence of l
- an accident previously evaluated in the SAR? '
I The probability of occurrence of an accident previously evaluated in the SAR has not been increased. The applicable accident analyses, LOCA and FHA, were reviewed and it was determined that these analyses are bounding for the new plant configuration. The CSAS and
- CIAS are not accident initiators. Therefore, they can have no effect on the probability of occurrence of an accident.
i 5.2 Does the proposed activity increase the consequences of an accident previously evaluated in the SAR? The consequences of an accident previously evaluated in the SAR have not been increased. The CSAS and CIAS are used for accident mitigation for LOCA, HELB inside containment, and FHA. As stated previously, elimination of the EDG auto-start function on actuation of CSAS will have no.effect on the capability of CSAS to perform its mitigative functions. Actuation of CSAS cannot occur without prior or concurrent SIAS. Since EDG auto-start occurs on SIAS, no reduction occurs in the mitigative ability of CSAS by elimination of the EDG auto-start function on CSAS actuation. e
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PC/M No 96064 Rev 0
- Page 17 _ of 27 i
i Similarly, elimination of the EDG auto-start function on actuation 1 of CIAS will have no effect on the capability of CIAS to perform its mitigative functions. As stated previously, in MODES I-4, the only j condition which could result in a CIAS without a concurrent SIAS is ' a SBLOCA where a high radiation signal could be generated prior.to a high containment pressure signal. In this event, CIAS would : actuate prior to SIAS and no EDG auto-start would occur until the i subsequent SIAS . However, this event is bounded by the SBLOCA i spectrum analyses, where a LOOP is assumed on reactor trip and the safety injection pumps are assumed to await a 30 second delay for i diesel startup and load sequencing following a safety injection signal. For MODE 6, the removal of the EDG auto-start capability from CIAS is bounded by the FHA previously analyzed. No credit is i taken for the CIAS EDG auto-start function. l 5.3 Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated ; in the SAR? l P The probability of occurrence of a malfunction of equipment l; important to safety has' not been increased. The purpose of this modification is to prevent the possible occurrence of a malfunction ; of equipment important to safety, namely the EDGs. This ; modification will delete EDG start and lockout relay block on CIAS : and CSAS such that the EDGs start automatically and . the lockout l relay will be blocked only for SIAS and LOOP. This will eliminate , any possibility of running the EDGs in parallel with offsite power ;
.with the protective trips blocked. This modification will create no ;
new system interactions or failure modes. The EDGs will continue to ; perform as assumed in the FSAR accident analyses. The FSAR accident , analyses remain bounding for the applicable equipment configuration. i The evaluation of components actuated by CIAS and CSAS, as 2 documented in Attachment 4.4 and discussed in Section 2.7, has l determined that no components other than those described in this l PC/M are affected by this modification. ; i 5.4 Does the proposed activity increase the consequences of a ! malfunction of equipment important to safety previously evaluated in I the SAR7 The consequences of a malfunction of equipment important to safety ! previously evaluated in the SAR have not been increased. As stated , above, no equipment important to' safety will be adversely impacted ! by this modification. This modification removes a potentially ; adverse equipment configuration which, if uncorrected, could cause i damage to the emergency diesel generators. No new failure modes or j system interactions are being introduced. The EDSs will continue to i function as assumed in the FSAR accident analyses. The FSAR ! accident analyses remain bounding for the applicable configuration. i Since there will be no increase in the probability of a malfunction l of the EDGs or CIAS and CSAS, and no components other than those described in this PC/M are affected, there can be no increase in the i consequences associated with such a malfunction. ' .- , , ~
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l PC/M No 96064 j .. Rev 0 l* Page 18 of 27 2 5.5 Does the proposed activity create the possibility of an accident of l- a different type than any previously evaluated in the SAR? t The possibility of an accident of a different type than any I previously evaluated in the SAR has not been created. This modification does not degrade the reliability of the EDGs or the . ESFAS components. All equipment will continue to function as ! assumed in the FSAR accident analyses. This modification removes an j unnecessarily redundant ESFAS function. While the accident design
- function of the CIAS and CSAS logic will be modified slightly, the
! plant will continue to operate as before. The FSAR analysis for 1 l 58LOCA and FHA will continue to remain bounding for the new i j' configuration, since no credit is taken in the analysis for EDG ! start on CIAS. ! f ! 5.6 Does the proposed activity create the possibility of a malfunction ] of equipment important to safety of a different type than any l previously evaluated in the SAR7 l 4 i
- The possibility of a different type of malfunction of equipment '
- important to safety than any previously evaluated in the SAR has not ,
- been created. No new equipment interfaces or interactions have been l i established. This modification will have no adverse impact on l 5
equipment important to safety. All affected equipment will continue ! to perform its safety function as assumed in the NRC-approved FSAR i i-accident analyses. While the specific design of the ESFAS logic has i
- been modified slightly, the overall design function and operational
- characteristics assumed in the accident analyses remains unchanged. ]
1 ! 5.7 Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification? 1 The}}