IR 05000348/1997201
| ML20141H373 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 05/13/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20141H316 | List: |
| References | |
| 50-348-97-201, 50-364-97-201, NUDOCS 9705230286 | |
| Download: ML20141H373 (100) | |
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U.S. NUCLEAR REGULATORY COMMISSION l
OFFICE OF NUCLEAR REACTOR REGULATION Docket Nos.: 50-348 and 50-364 License Nos.: NPF-2 and NPF-8 Report Nos.: 50-348/97-201 and 50-364/97-201 Licensee: Southern Nuclear Operating Company, In Facility: Farley Nuclear Plant, Units 1 and 2 i Location: 7388 North State Highway 95 l Columbia, AL 36319 Dates: January 27 - March 14, 1997 )
Inspectors: R. K. Mathew, Team Leader, Special Inspection Branch l C. J. Baron, Stone & Webster Engineering Corporation !
P. Bienick, Stone & Webster Engineering Corporation R. B. Bradbury, Stone & Webster Engineering Corporation D. Schuler, Stone & Webster Engineering Corporation i
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Approved by: Robert M. Gallo, Chief Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation
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TABLE OF CONTENTS ;
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EXECUTIVE SUtMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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i l El Conduct of Engineering ....................... I j E Inspection Objectives and Methodology . . . . . . . . . . . . . . . 1 El.2 Auxiliary Feedwater System .................... 1 E1.2.1 System Description and Safety Functions . . . . . . . . . . 1 E1.2.2 Mechanical Design Review ................. 2
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E1.2.2.1 Condensate Storage Tank . ... . . . . . . . . . . . . 2 E1.2.2.2 AFW System Performance ............... 7 E1.2.2.3 AFW System Valve Operation ............. 8 E1.2.2.4 AFW System Testing ................. 9 1 E1.2.2.5 AFW System Modifications .............. 12 E1.2.3 Electrical Design Review ................. 13 E1.2.3.1 Turbine-Driven AFW (TDAFW) Uninterruptable Powe System (UPS) .................... 13 E1.2.3.2 AFW Electrical Loads ................ 15 E1.2.3.3 Electrical AFW Modifications ............ 16 ,
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E1.2.4 Instrumentation & Controls (I&C) Design Review ...... 17
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E1.2.4.1 Condensate Storage Tank . . . . . . . . . . . . . . . 17 E1.2.4.2 Instrument Loop Uncertainty Calculations ...... 18 E1.2.4.3 Instrument Setpoint Uncertainty Program . . . . . . . 19 ,
E1.2.4.4 Modifications and Other Reviews . . . . . . . . . . . 19 E1.2.5 System Interface ..................... 21 E1.2.5.1 Service Water (SW) System . . . . . . . . . . . . . . 21 i Instrument Air (IA) System
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E1.2. ............. 21 E1.2.5.3 Main Steam (MS) System ............... 22 E1. System Walkdown ..................... 23 E1.2.7 FSAR, FSD and Other Reviews . . . . . . . . . . . . . . . . 26 El.3 Component Cooling Water (CCW) System ............... 28 E1.3.1 System Description and Safety Functions . . ........ 28 E1.3.2 Mechanical Design Review ................. 29 l E1.3.2.1 CCW System Performance ............... 29 i E1.3.2.2 CCW Surge Tank ................... 32 1 E1.3.2.3 CCW System Containment Isolation .......... 33 )
E1.3.2.4 CCW System Testing ................. 34 E1.3.2.5 CCW System Modifications .............. 36 E1.3.2.6 Other Related CCW System Review . . . . . . . . . . . 39 E1.3.3 Electrical Design Review ................. 40
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E1.3.3.1 CCW Electrical Loads ................ 40 E1.3.3.2 CCW Swing Pump Operation .............. 41 E1.3.3.3 Electrical Modifications .............. 42 E1.3.4 Instrumentation & Controls (I&C) Design Review ...... 43
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E1.3.4.1 - CCW Surge Tank Level Setpoint . . . . . . . . . . . . 43 El.3.4.2 Instrument Loop Uncertainty Calculations ...... 43 E1.3.4.3 Setpoint Indexes and Other Reviews ......... 44
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I El.3.5 System Interface Design Review .............. 45 E1.3.5.1 Service Water (SW) System . . . . . . ._. . . . . . . 45 E1.3.5.2 Instrument Air (IA) System ............. 45 i El.3.5.3 Auxiliary Building Ventilation (ABV) System . . . . . 46 El.3.6 System Wal kdown . . . . . . . . . . . . . . . . . . . . . . 46
E1.3.7 FSAR and FSD Review . . . . . . . . . . . . . . . . . . . .
El.4 Other Related Electrical Systems Review . ." . . . . . . . . . . . . 48 El.4.1 AC System . . . . . . . . . . . . . . . . . . . . . . . . . 48 i
El.4.2 125 Vdc Battery Systems . . . . . . . . . . . . . . . . . .
El.4.3 Modifications . . . . . . . . . . . . . . . . . . . . . . .
El.4.4 System Walkdown . ... . . . . . . . . . . . . . . . . . . . 53
E1.4.5 FSAR and FSD Review . . . . . . . . . . . . . . . . . . . . l
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El.5 Control of Calcul ations . . . . . . . . . . . . . . . . . . . . . .
Exit Meeting Summary ....................... 57 XI APPENDIX A - OPEN ITEMS. . . . . . . . . . . . . . . . . . . . . . . . . . A-1 -
APPENDIX B - EXIT MEETING ATTENDEES. . . . . . . . . . . . . . . . . . . . B-1 APPENDIX C - LIST OF ACRONYMS USED . . . . . . . . . . . . . . . . . . . . C-1 APPENDIX D - LIST OF DOCUMENTS REVIEWED. . . . . . . . . . . . . . . . . . D-I l
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l' EXECUTIVE SUMMARY ;
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From January 27 through March 14, 1997, the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation (NRR), i j 'Special Inspection Branch, conducted a design inspection at Units'1 and 2 of ,
the Joseph M. Farley Nuclear Plant. Specifically, this inspection included i visits to,the licensee's engineering offices and the plant site on February 3- !
7,' February 17-28, and March 10-14, 1997. The inspection team consisted of a i team leader from the NRR and four engineers from the Stone & Webster
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Engineering Corporatio . .
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As'the focus of this design inspection, the team selected the Unit I auxiliary i
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feedwater (AFW) system, the Unit 2 component cooling water (CCW) system, and
- their support systems, because of the importance of these systems in ,
mitigating accidents at Farley. The purpose of the inspection was to evaluate .
t the capability of the selected systems to perform safety functions required by
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their design bases, as well as the adherence of the systems to their 1 i- . respective design and licensing bases, and the consistency of the as-built !
. configuration and system operations with the final safety analysis report (FSAR). To achieve this purpose, the team followed the engineering design and
- configuration control section of Inspection Procedure 93801. The team also selected and reviewed relevant portions of the FSAR, design-basis documents,
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Technical Specifications (TS), drawings, calculations, modification packages, a procedures, and other plant-related documentatio ;
!- Overall, the inspection team determined that the selected systems are capable l
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margins. The interaction and communication between the plant, home office i
' engineering, and engineering contractors were well managed. Moreover, the
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engineering staff exhibited adequate knowledge of th'e systems evaluated and provided excellent support to the inspection team. The team also found that
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the Farley design and licensing bases have been adequately implemented in all i
but a few instances, and the quality of calculations and design changes was j generally good. Although we consider the safety system self-assessments for
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AFW and CCW systems performed by the licensee to be a positive initiative, the j
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assessments did not identify and correct many of the issues raised by the ,
i team. In addition, the system design documents reviewed by the team J
adequately supported the design, with the exception of the items discussed in 1
the following paragraph l l
The licensee's evaluations of plant modifications, conducted in accordance ,
with 10 CFR 50.59, were generally adequate. However, one 10 CFR 50.59 safety i i
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evaluation performed as a result of a licensee self-initiated safety system assessment' associated with an FSAR change deleting the requirement for tornado )
missile protection for several condensate storage tank (CST) piping '
connections failed to identify the existence of a potential unreviewed safety question. As a result of the team's questions, the licensee reported this issue, in accordance with 10 CFR 50.72, on February 27, 1997, and implemented i interim corrective actions to maintain the operability of the system until the issue can be resolved. In a similar instance, the 10 CFR 50.59 evaluations for a surveillance test procedure and FSAR revision did not identify that a j
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technical specifications change was required as a result of changes to an Auxiliary Building battery profile and associated battery voltage requirements.
The team also observed inadequate tornado missile protection of the turbine-driven AFW (TDAFW) pump vent stack and the CST level transmitters, and their ,
associated electrical conduits and cables. The as-built plant configuration did not conform to the design and licensing bases requirements concerning tornado missile protection. In addition, the exhaust silencers for the diesel generators (including the station blackout diesels) were only protected against horizontal tornado missiles, although the equipment is also susceptible to vertical and other non-horizontal missiles. The NRR staff will review this issue associated with the diesel generators to determine whether the tornado missile protection in the Farley Unit'I and 2 design and licensing bases included missile spectra other than horizontal missiles.
The TDAFW pump discharge check valves were not tested in the reverse direction, and the surveillance test procedure acceptance criterion for forward flow testing of certain AFW check valves was incorrect. In addition, the corrective action for a notice of violation for a similar issue did not thoroughly address the check valve reverse flow testing deficiency. Another testing issue included the lack of a service test for the Class IE TDAFW battery.
The team identified certain design control weaknesses for calculations.
Superseded calculations were not always identified on the calculation index, and design-basis calculations were not always updated when design conditions changed. The licensee implemented controls for new calculations that would prevent such occurrences in the future. Howeter, no action was taken to correct deficiencies with existing calcuhtions.
The team's other concerns included : the design-basis differential pressures for several motor-operated containment isolation valves were incorrect; the fluid temperatures used in the CCW piping stress analyses did not use the maximum operating temperature; the setpoint calculation for the CST low-level alarms did not consider drift and deadband errors; process and documentation discrepancies in the modification to coat portions of the CCW heat exchangers with epoxy and stabilize / plug tubes, as well as the lack of specific documentation demonstrating the adequacy of a fire protection seal; and the installed configurations of the TDAFW battery rack and the CST level transmitter were not in accordance with design requirements.
The team identified discrepancies between the FSAR and other documents, such as functional system descriptions (FSDs), procedures, calculations, and drawing In some cases, the licensee had not updated the FSAR or performed ;
safety evaluations to assess the possibility of unreviewed safety question l The licensee's corrective measures effectively resolved the immediate concerns l identified by the team, and the team did not have any unresolved operability l concern l f
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Reoort Details f
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III. Eneineerine
j El Conduct of Engineering !
, E Insoection Ob.iectives and Methodoloav
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The primary objective of the design inspection at the Farley Nuclear Plant was .
to evaluate the capability of selected systems to perform their safety i
. functions and adherence to their design and licensing bases, and the ,
- consistency of the as-built configuration and system operations with the final !
safety analysis report (FSAR). The team selected the Unit I auxiliary j 1 feedwater (AFW) system, the Unit 2 component cooling water (CCW) system, and 1'
their support systems,.because of the importance of these systems in i mitigating accidents at Farley. The inspection focused on the engineered l safeguards functions of these systems and their interfaces with other system .
For guidance in performing the inspection, the team followed the applicable i engineering design and configuration control portions of Inspection Procedure !
(IP) 93801, " Safety System Functional Inspection." !
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l Appendix A identifies the open items resulting from this inspection, while
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Appendices B, C, and D identify the exit meeting attendees, the list of l acronyms used, and the list of documents reviewed, respectivel '
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El.2 Auxiliary Feedwater System
- E1.2.1 System Descriotion and Safety Functions
! The AFW system consisted of two motor-driven pumps, .one turbine-driven pump, and the associated piping, valves, instruments, and controls. The system was designed to use water from the condensate storage tank (CST) or the service
- water (SW) system backup supply to provide high-pressure feedwater to the
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steam generators through the main feedwater (MFW) system. Each of the motor-
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driven AFW (MDAFW) pumps was sized to supply the steam generators with 100% of the required feedwater flow for a normal safe cooldown of the reactor coolant system (RCS). The turbine-driven AFW (TDAFW) pump was sized to supply the i steam generators with 200 percent of the required feedwater flow for a normal safe cooldown of the RC Under accident conditions', both MDAFW pumps (assuming a failure of TDAFW ) were required to provide the necessary feedwater fl o The- function of the AFW system was to supply high-pressure feedwater to the steam generators during plant startup, cooldown, and emergency conditions when the normal feedwater supply is not available. The system was designed as.an engineered safety feature to provide redundant reans of removing decay and sensible heat from the reactor coolant system via the steam generators during i emergency conditions. In addition, the systeo was designed to meet the single failure criteria so that no single failure will prevent the supply of sufficient feedwater to at least two of the three steam generator The AFW l
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system design was contingent on providing sufficient flow to prevent loss of pressurizer vapor space during a feedwater line break with a loss of offsite power.
El.2.2 Mechanical Desian Review El.2.2.1 Condensate Storace Tank (CST) Inspection Scope The mechanical design review of the CST included a detailed evaluation of the related equipment specification, drawings, manufacturer's information, and applicable calculations, in order to assess consistency with the design and licensing bases. Observations and Findings Structure of the CST The CST was designed to serve as the normal source of water for the AFW system. The total CST volume was approximately 500,000 gallons. The lower 150,000 gallons was reserved for the decay heat removal and cooldown requirements of the AFW system, and this portion of the CST was designed to withstand the forces resulting from the impact of a design-basis tornado-generated missile without loss of pressure boundary integrity.
The team reviewed the CST manufacturer's drawings U1616930, Revision A0, and U161703B, Revision 2, and verified that the volumes of both the entire CST and the lower portion of the CST were correct. The team also verified the tank nozzle locations and the tornado missile criteria sh.own on drawing U161693D, which indicated that the lower portion of the CST was fabricated from 1-inch thick steel plate. Specifically, the team reviewed the CST equipment specification, S5-1111-4, Revision 2, and manufacturer-provided CST design calculation, Charge No. 72-4859-Memphis Engineering. As a result of this review, the team questioned the basis of the equation used in the manufacturer's CST design calculation. Specifically, the manufacturer used a ballistics equation intended to determine the required thickness of a steel plate to prevent perforation. The manufacturer did not provide a basis for using this equation to demonstrate the capability of the tank to withstand the impact of a design-basis tornado-generated missile without a loss of pressure boundary integrity. Moreover, the equation was not based on a specific grade of stee In response to the team's concerns, the licensee performed an independent verification calculation (SM-97-S19108-001) to assess the adequacy of the CST shell thickness.for the most limiting design-basis tornado-generated missil This analysis verified that the maximum membrane stress from the missile would be less than the allowable stress for the tank material. As a result, the licensee's independent calculation confirmed that the CST was capable of withstanding the impact of a design-basis tornado-generated missile without a loss of pressure boundary integrit ,
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- The team also reviewed calculation 1.10, " Condensate Storage vs. Decay Heat,"
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Revision 1, and calculation BM-95-0961-001, " Verification of CST Sizing :
[ Basis," Revision 1. On the basis of this review, the team verified that the CST volume was adequate to perform the required functions during normal and accident: conditions.
- Unorotected CST Connections On October 21,1994, during a self-initiated safety system assessment (SSSA)
of the AFW system, the assessment team discovered that the following.-
- connections on the lower portion of the CSTs in both Unit I and Unit 2 were
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4 .: drain connections on the Unit I and Unit 2.CSTs :
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. sensing lines for the level transmitters on the Unit I and Unit 2 CSTs j i . 6-inch diameter nozzle and isolation valve on the Unit 2 CST l . 6-inch diameter vacuum degasifier suction line on the Unit 1 CST
To resolve this issue, the licensee issued Incident Report 1-94-299, Licensee ;
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Event Report (LER 94-005), and an FSAR change and associated 10 CFR 50.59 l Safety Evaluation. According to LER-94-005-00, dated November 18, 1994, i probabilistic risk assessment (PRA) techniques indicated that the increase in ;
the probability of an accident resulting from the lack of_ missile protection
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on the subject connections was negligible. The LER further stated that the i licensee had revised the FSAR to include the PRA results and to delete the i requirement for missile protection of the subject CST connections. The PRA :
concluded that the impact frequency with which a I
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exposed CST piping was on the order of 1.0perx yea ', tornado An estimat missile could strikeJ the resulting core damage frequency (CDF) was on the order of per 1.0 x 10'g of year. The licensee also documented the PRA in calculation REES-F-94-034, i
"Farley - Estimate of Core Damage Frequency from a Tornado Missile Striking ;
Exposed CST Piping," Revision 0. The team noted that NRC Inspection Report j
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95-20, dated January 23, 1996, documented the related LER revie j
l Based on the 10 CFR 50.59 safety evaluation associated with the FSAR change, i
- the licensee concluded that the lack of missile protection on the CST tank i connections did not place the plant outside its design bases. Further, the l
- licensee concluded that the FSAR change could be implemented without prior ;
approval from the NRC because it did not involve an unreviewed safety question l
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-(USQ) or a change to the Technical Specifications- (TS). The team noted that 1
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NRC Inspection Report 96-07, dated September 27, 1996, documented a review of l this safety evaluatio !
The team also recognized that the licensee's conclusion regarding the 10 CFR 50.59 safety evaluation was founded on Regulatory Guide (RG) 1.70, Section
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2.2.3.1, which states that design-basis events external to the nuclear plag't are defined as accidents that have a probability of occurrence of about 10 ,
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per year or greater with potential consequences serious enough to affect the i safety of the plant. The licensee's safety evaluation compared the calculated 1
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impactfrequencywithwh,ichatornadomissilecouldstrikeexposedCjTpiping
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(approximately 1.0 x 10' per year) to the RG 1.70 value of 1.0 x 10' per year 1 and concluded that this postulated tornado event was not required-to be .
analyzed as an " accident" in the FSAR. .However, the team concluded tpat this
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- comparison was not appropriate because the RG 1.70 value of 1.0 x 10' per year applied to the probability'of occurrence of an external event, such as an :
explosion or a toxic gas release, that could affect the plant as a whole, not ,
the probability of a tornado missile striking a specific target in the plan ;
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FSAR Section 3.5.4 states that Category I equipment and piping outside ,
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containment are either housed in Category I structures or buried undergroun :
FSAR Table 3.2-1 (Footnote 24) supplements that general statement by i
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addressing the tornado missile design of the CST. In addition, FSAR Section 9.2.6.3 addresses the potential of tornado missiles striking the CST. At the !
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time that this condition was discovered on October 21,'1994, FSAR Sections 4 9.2.6.3 and 9.2.6.6 stated that the lower section of the CST was designed to ;
withstand tornado missiles.
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Criterion II of Appendix A to 10 CFR Part 50, Appendix A, requires that i important to safety shall be
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. designed structures, systems, and components to withstand the effects of natura (SSCs) l phenomena such as tornados without loss of capability to perform their safety functions. The Farley licensing bases included this criterion.
- Consequently, the licensee's 10 CFR 50.59 safety evaluation dated November 17, 1994, included a question (question number 6) asking, "May the proposed activity create the possibility of a malfunction of equipment important to
> safety of a different type than any previously evaluated in the FSAR7" The ;
! licensee answered this question "No," in accord with the presumption that a
' CST failure caused by a missile was less likely than other malfunctions considered by the FSAR. However, as discussed above, it was not appropriate to compare the fre
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withthe1.0x10'puencyofatornadomissilestrikeonspecificequipmentper year cri 1 basis event In addition, Question 6 in the safety evaluation addressed the ,
i- possibility of an equipment malfunction; it did not address the probability compared to other malfunctions considered by the FSAR. The team therefore L stated that the use of a PRA to determine the frequency of a tornado missile .
strike on specific equipment was not an appropriate justification to determine l
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that a USQ did not exist when the as-built plant configuration did not conform to the FSAR.
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The 10 CFR 50.59 safety evaluation referred to the NRC Standard Review Plan (SRP), Section' 3.5.1.4, " Missiles Generated by Natural Phenomena," which !
states that the methodology used to identify appropriate design-basis missiles l generated by natural phenomena shall be consistent with the acceptance l criteria defined for the evaluation of potential accidents from external :
sources in SRP Section 2. On the basis of this cross-reference, the l
. licensee's safety evaluation cited SRP Section 2.2.3, " Evaluation of fotential l Accidents," and RG 1.70, Section 2.2.3.1, to establish the "1.0 x 10' per year" criterion. -However, both of these references addressed potential accidents involving hazardous materials or activities related to nearby industrial, military, and transportation facilities in the vicinity of the
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plant. The cross-reference from SRP Section 3.5.1.4 to SRP Section 2. addressed the identification of appropriate design-basis missiles generated by
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l natural phenomena, not the frequency of a tornado missile strike on specific ,
equipment. Consequently, the team obs
2.2.3.1,asabasisforthe"1.0x10'prvedthattheuseofRG1.70,Section per year" criterion was not appropriate because PG 1.70 addresses potential accidents involving hazardous
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materials or activitlas in the vicinity of the plant, not tornado-generated '
missiles.
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4 The licensee stated that both deterministic and probabilistic: analyses were performed in support of the safety evaluation. However, the licensee did not
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document the deterministic analysis in the safety evaluation. The licensee ;
also stated that both the LER and safety evaluation had previously been
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reviewed and found to be acceptable by the NRC. In addition, the licensee i
stated that it was not their normal practice to use PRA techniques as the
- primary justification for safety evaluations.
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9 After reviewing other 10 CFR 50.59 safety evaluations and discussing the t matter with the licensee, the team determined that the inappropriate use of PRA techniques was not a widespread problem Nonetheless, the team identified
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a potential.USQ because the licensee effectively changed the plant from that i described in the FSAR. The installed condition of the safety-related CST tank
- did not conform to the original design bases documented in the FSAR, and the
- facility change was made without prior approval from the NRC. In addition, the team determined that the licensee's corrective action for self-initiated safety system assessment (SSSA) observation was not adequate to resolve this l
< deficiency (unprotected CST piping) and did not meet the requirements >
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- specified in Criterion XVI of Appendix B to 10 CFR Part 50.
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As a result of this condition, a tornado missile could damage some CST ,
connections, thereby resulting in a loss of inventory from the safety-related i tank and affecting the operability of the entire AFW system. Tank damage resulting from a tornado missile would be considered a malfunction of equipment important to safety of a dif,ferent type than any previously >
evaluated in the FSAR. As defined in 10 CFR 50.59, an activity that creates the possibility of a malfunction of equipment important to safety of a i different type than any previously evaluated in the FSAR is considered a US ,
Further, 10 CFR 50.59(c) requires that the licensee submit an application for amendment of its license pursuant to 10 CFR 50.90 when changes are made which L involve a USQ. In addition to Criterion XVI of Appendix B to 10 CFR Part 50, the requirements of 10 CFR 50.59 were apparently not met. This issue was j identified as Unresolved item 50-348; 50-364/97-201-01.
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In response to this item, the licensee initiated Occurrence Report Number 1-97-047 and issued a non-emergency notification LER in accordance with 10 CFR
50.72(6)(1)(ii)(b) during the inspection. The Occurrence Report documented interim actions required to support the operability of the tank. These
, actions included verification that the CST connection valves were closed and
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tagged and revision of the abnormal operating procedure to carefully monitor the CST level during adverse weather conditions and to take action if the CST develops leakage. The licensee committed to implement the action required to bring the plant condition into conformance with the licensing base .
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j CST Level Transmitter Enclosure l During the walkdown of the Unit 1 CST, the team observed that the safety-related CST level transmitters and enclosures (fabricated from 1/4-inch l steel plate), as well as the associated cables and conduits, were outside, and 4 routed above ground around the tank perimeter. This equipment appeared to be l susceptible to tornado-generated missile l FSAR Sections 3.2.1.5, and 9.2.6.1 and Table 3.2-1 state that the AFW system instrument and control (I&C) system equipment and CST equipment are Category ,
I, respectively. FSAR Section 3.5.4 states that Category I equipment and l piping outside containment are either housed in Category I structures or l
buried underground. FSAR Table 3.2-1 (Footnote 24) supplements this general statement by addressing the tornado missile design of the CST. In addition, FSAR Section 9.2.6.3 addresses the potential of tornado missiles striking the CST. Criterion II of Appendix A to 10 CFR Part 50 requires that SSCs I important to safety be designed to withstand the effec'. of natural phenomena such as tornados without loss of capability to perform their safety functions.
The Farley licensing bases included this criterio l
The licensee responded that neither the level transmitter enclosures nor the ridged conduit were designed to provide tornado missile protection. Further, the licensee stated that the operators would be informed of the loss of these )
instruments by level indicators LI-4132A or LI-41328 failing high or low.
Consequently, the licensee concluded that the functional impact or consequences of a tornado-generated missile destroying the level transmitter tubing connections would be similar to a missile destroying the conduits that contain the transmitter control circuits. These transmitter circuits were intended for indication and did not perform an automatic control function.
The licensee also stated that the AFW pump suction pressure transmitter (PT-3211A) provided a diverse means for determining CST level. AFW flow transmitter (FT-3402) and steam generator level transmitter (LT-477) also provided indications to ensure that CST inventory was available. On that basis, the licensee concluded that the level transmitters were not essential to safely shut down the plant during events involving missiles.
Nonetheless, the installed condition of the safety-related CST tank level transmitters, and the associated cables and conduits, did not conform to the Farley design and licensing bases. As a result, these instruments could be damaged or destroyed by a tornado missile. In addition to Criterion II of Appendix A to 10 CFR Part 50, the requirements of 10 CFR Part 50, as specified in Appendix B, Criteria III and V which state that applicable regulatory requirements and design basis are correctly translated into drawings, procedures and instructions and that activities affecting quality be prescribed by and performed in accordance with instructions, procedures or drawings were apparently not met. This issue was identified as Unresolved Item 50-348; 50-364/97-201-0 To address this item, the licensee issued ABN 97-0-1043. Other issues related to tornado missile protection of safety-related equipment are addressed in Section El.2.6 of this repor l
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.. Other CST Issues The tuu identified a conflict involving the CST-related AFW Operator License Training Objectives, OPS-52102H, dated June 21, 1992. Specifically, Question / Answer 5, stated that the CST low-low level alarms were set to inform !
the operator when the AFW had 20 minutes of normal supply left, with three AFW i pumps operating at full capacity. However, the team observed that this stated i
. basis was not consistent with the discussion of the CST low-low level alarm j setpoint in the AFW functional system description (FSD), A-181010, Revision 1, -
Section 3.21.6.4, and calculation SM-87-1-4380-001, Revision 0. Specifically,
'
both the FSD and the calculation indicated that the setpoint was dependent on l the TDAFW pump operating at 700 gpm for 20 minute ;
The licensee confirmed this dependency on the basis of calculation l SM-87-1-4380-001, "Setpoint Calculation for Level Alarm Replacement," Revision !
- The licensee stated that the level alarm setpoint, as documented in this calculation was appropriate to inform the operator when the AFW had 20 minutes ,
of normal supply left. Further, the licensee cited plant procedures indicating that the AFW system supply would be aligned to the SW system when the setpoint was reached. However, the team observed that OPS-52102H was not 1 correct and did not provide the information required for the operators to understand.the basis of the setpoint. The licensee issued REA 97-1410 to j revise OPS-52102 j
' Conclusions
'
Overall, the CST was capable of performing its required functions during both normal and accident conditions. With the exception that some safety-related equipment was susceptible to tornado generated missile damage, the mechanical 4 aspects of the CST were consistent with the design and licensing base However, the team noted one instance in which the licensee's safety evaluation for an FSAR change regarding unprotected CST piping failed to identify a potential USQ and corrective action for the SSSA observation regarding this issue was not adequat El.2.2.2 AFW System Performance Inspection Scope The mechanical design review of AFW system performance included a detailed evaluation of the related system drawings, specifications, manufacturer's data, and applicable calculations, in order to assess consistency with the design and licensing bases.
- Observations and Findings The team reviewed calculation 29.01, " Auxiliary Feedwater Pumps Minimum Flow Evaluation," Revision 1, and verified that the minimum flow lines for both the MDAFW and TDAFW pumps had adequate capacity. The team also reviewed calculation 11.13, "As-Built NPSH for Auxiliary Feedwater Pumps," Revision 0, j
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-and verified that, even with the' CST empty, net positive suction head (NPSH)
' exceeded that required for the' pump runout flow experienced during the initial !
!
,
period of a steam or feedwater pipe break even .
Next, the team reviewed calculation 40.02, " Verification of Auxiliary l
Feedwater Flow Basis," Revision 3, to verify the system flow capacity for !
'
various normal and accident conditions supported by'the AFW system. This calculation showed that all of the AFW system minimum flow requirements were '.
met by the minimum available AFW pump performance, considering up to 5%
degraded performance. The calculation also showed that the system did not ;
- exceed the maximum allowable AFW flow limit for main steam line break (MSLB)
and accidental depressurization events, with 5 % enhanced pump performanc The team found these results acceptabl l The team also observed that calculation 25.3, "Overpressurization.of Auxiliary 'I Feedwater. Piping During Overspeed Testing," Revision 0, addressed the maximum
'.. AFW system pressure resulting.from the TDAFW pump operating at 125% of its .
rated speed. By contrast, the licensee indicated that the current overspeed
!
trip setpoint for the TDAFW pump was 115 % of the rated speed, and verified that the system pressure corresponding to 115 % was acceptable for the pump, piping, valves, and flanges in the system. Section E1.5 of this report,
'
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" Control of Calculations," addresses the issue that the licensee had not '
revised calculation 25.3 to reflect the current setpoint.
" Conclusions ,
,
The team found that the mechanical aspects of the AFW system were consistent with the design-basis requirements and the system was capable of performing
,
the required functions during both normal and accident conditions. In addition, the team found that the current mechanical calculations related to system performance were of high quality, with inputs, assumptions, methods, and results clearly stated and well documente ,
,
E1.2.2.3 AFW System Valve Operation Inspection Scope j The mechanical design review of AFW system valve operation included a detailed evaluation of drawings, specifications, manufacturer's data, operating
.
procedures, and applicable calculations associated with various AFW system motor-operated, air-operated, manual, and check valves. Section E1.2.2.4 of
- this report addresses the licensee's testing of the AFW system valve l
, Observations and Findings i
"
The team reviewed calculation 23.5, "FNP Auxiliary Feedwater MOVs
. (IE Bulletin 85-03)," Revision 1, to verify that the licensee used the ,
'
appropriate maximum differential- pressure.value for each motor-operated valve (MOV) in the AFW syste In addition, the team verified that the differential :
. pressure values for each valve represented the most limiting system operating i conditions, and that the calculated values were consistent with the Unit 1 M0V Design-Basis Document,-U418109, Revision .
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The team then reviewed the AFW System Process & Instrumentation Diagram
- (P&ID),D-175007, Revision 22, and the Main Steam (MS) and Auxiliary Steam System P&lD, D-175033, Revision 18, to verify the normal position of each valve in the AFW system and the steam supply to the AFW pump turbine driv The team also found that Surveillance Test Procedure FNP-1-STP-22.5,
> " Auxiliary Feedwater System Flow Path Verification," Revision 17, was j- consistent with the P& ids. These normal valve positions were appropriate for
-
the function of the. valve Procedure FNP-1-AOP-6.0, " Loss of Instrument Air," Revision 20, included a
. list of air-operated valves. In reviewing this procedure, the team noted that
- Table 1 (page 16 of 29) included valves 1-AFW-FCV-3212A,1-AFW-FCV-32128, and -
1-AFW-FCV-3218, which were removed by Production Change Notice (PCN) 1-5582 and are no longer installed in .the plan In response to the team's question, i
the licensee stated that a temporary change notice (TCN) was issued to remove i these valves from the procedure, and FNP-1-AOP-6.0, Revision 21, was issued on February 19, 1997, to incorporate the required change.
-
During the walkdown of Unit 1, the team observed that electrical components: l
'
associated with the safety-related AFW discharge valves (HV-3227A, B, C; and HV-3228A, B, C) were located in the main steam valve room approximately 4 feet above the floor. This elevation was above the stated flood elevation for the room, but was not consistent with the FSAR. The licensee contended that the location of the AFW' valves was acceptable, and issued ABN 97-0-1043 to correct
'.
the FSAR. Section El.2.7 of this report discusses this FSAR discrepancy in
- greater detai .
, Conclusions
.
The team found that AFW system valves were capable of performing the required functions during both normal and accident conditions. With the exception of
.
deficiencies in an abnormal operating procedure and in the FSAR, the
!
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mechanical aspects of the AFW system valve operation were consistent with the design-basis requirements.
~
E1.2.2.4 AFW System Testina Inspection Scope The team evaluated the licensee's surveillance test procedures (STPs) to verify that the licensee had appropriately tested the AFW pumps and all of the
required system valves.
, Observations and Findings The team reviewed the results of the following STPs completed during 1995 and 1996:
. FNP-1-STP-22.1, "lA Auxiliary Feedwater Pump Quarterly Inservice Test,"
,
Revision 23
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-FNP-1-STP-22.8, " Auxiliary Feedwater Inservice. Valve Exercise Test,"
4 Revision 13 l
.- FNP-1-STP-22.16, " Turbine-Driven Auxiliary Feedwater Pump Quarterly '
Inservice Test (TAVE>/=547 F)," Revision 25 ;
4 . FNP-1-STP-22.21, " Turbine-Driven Auxiliary Feedwater Pump Automatic
'
Valve Test," Revision 5 . ;
.
These: test results were all satisfactory, with data within the acceptance !
'
criteria for the procedur ;
! The team also reviewed one example of a surveillance test failure and the '
,
associated corrective action. On-February 29, 1996, the "B" motor-driven AFW pump discharge check valve failed to, pass a reverse flow closure operability '
"
test-(STP-22.24). This valve had also failed the reverse flow closure operability test on December 1, 1992. The licensee's subsequent evaluation revealed that the root cause for both failures was an accumulation of sediment on the check valve seating surfaces. As a result, the licensee issued Work Order (WO) 96001338 to inspect and clean the valve during the Unit 1 RFIS refueling outage, which began in March 1997. The licensee stated that-if the
,
'
sediment accumulation was deemed to be significant by mechanical maintenance '
engineering, a preventative maintenance task would.be established to inspect '
and clean the valve during every other refueling outage. The team determined that the licensee's root cause evaluation and corrective action were appropriate and sufficient. No other similar conditions were identified
.
during the inspection.
t In a similar fashion, the team's review revealed that the licensee had not i tested TDAFW pump discharge check valves V003 and V002D, F, and H in the l reverse directio (Valves V002A and 002B, located in the discharge lines :
from the MDAFW pumps, were tested in the reverse direction.) Check valve V003 i was located in the discharge line from the TDAFW pump, while valves V002D, F, _
j and H were located in the supply lines from the TDAFW pump discharge header to j
.
each feedwater line. The AFW FSD, A-181010, Revision 1, Section 5.27.1, i stated that the basic function of valve V003 was to prevent reverse flow into
,
!
the TDAFW pump when the pump was idle, j
Either check valve V003 or check valves V002D, F, and H were required to perform a safety function in the closed position. Failure of these valves to ,
close could potentially result in failure of the MDAFW pumps to perform their
.
!
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safety function. In the event that one or both of the MDAFW pumps were
, operating and the TDAFW pump was not operating, failure of the TDAFW pump i
discharge check valves ~ to close could result in some flow from the MDAFW pumps .
failing to reach the steam generators. That flow from the MDAFW pumps could j return to the CST though the TDAFW pump minimum flow line, or could be j discharged through the TDAFW pump suction relief valve, QV068C. In addition, i
.
overpressurization of the TDAFW pump suction piping couls' c ur if the reverse flow though the check valves exceeded the capacity of the TDAFW pump. suction a relief valve, QV0680. The team noted that 10 CFR 50.55a(f) and TS Section 4.0.5 require inservice testing (IST) of valves in accordance with Section XI and applicable Addenda of the Boiler and Pressure Vessel (B&PV) Code .
i 10 l
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promulgated by the American Society of Mechanical Engineers (ASME). In particular, the purpose of this testing was to verify the operational I l
readiness of valves that must function to ensure plant safet Reverse flow testing of check valves was addressed in NRC Information Notice .
'
(IN) 88-70, dated August 29, 1988. In a memorandum dated September 12, 1989,
'.
responding to this IN, the licensee listed the AFW system valves that were
- being subjected to . reverse flow testing at the time. The memorandum did not ;
address the lack of reverse flow testing .of the TDAFW pump discharge check -
valves. In addition, during an inspection conducted on October 2-6, and i 16-20,.1989, the NRC identified a violation involving failure to test the MDAFW pump discharge check valves, V002A,and V002B, in the reverse direction.
i Subsequent to the 1989 inspection, the licensee expanded the IST program to.
4 include reverse flow testing of the MDAFW pump discharge check valves, V002A -
.
and V002B, but did not address the TDAFW pump discharge check valves. The licensee subsequently issued REA 97-1410 to correct this discrepanc The team noted that TS Section 4.05 requires inservice testing of ASME Code
Classes 1, 2, and 3 pumps and valves in accordance with Section XI of the ASME '
'
- Boiler and Pressure Vessel Code and applicable Addenda. The licensee is committed to inservice testing in accordance with the 1983 Edition of the Code 1 and Summer 1983 Addenda.Section XI, Subsection IWV-3522 requires valves whose function is to prevent reverse flow to be tested in a manner that proves that the disk travels to the seat promptly on cessation or reversal of flow.
!' The lack of adequate inservice testing was identified as Unresolved Item 50-348; 50-364/97-201-03.
l
'
In addressing this item, the licensee stated that either check valve V003 or check valves V002D, F, and H were required to perform a safety function in the
- closed position. Only one isolation barrier was required in the TDAFW pump ~
l
injection line to prevent diversion of flow from an operating MDAFW pump. The
'
licensee also stated that the results of performance testing during each j unit's last refueling outage indicated that check valves V002D, F, and H were
,
sufficiently closed to allow the MDAFW pumps to perform their safety functio I In addition, to verify operability, the licensee disassembled and inspected
.
the V003 check valves for both Unit 1 (Work Order No. 97001743) and Unit 2 ;
! (Work Order No. 97001744) on March 7, 1997. These activities provided an acceptable confidence level.that the valves were operable. Further, the ,
licensee stated that check valves V0020, F, and H would be identified in the ;
'
IST program and plan documents as having a safety function to close, and Surveillance Test Procedure FNP-1/2-STP-22.29 would be issued to test check valves V002D, F, and H in the reverse direction. In addition, the licensee
issued FNP Occurrence Report 1-97-048 on March 3, 1997, to document this
' condition and to ensure that all corrective measures would be implemented
,
correctly and in a timely manne The team also reviewed Surveillance Test-Procedure FNP-1-STP-22.13,." Turbine-Driven Auxiliary Feedwater Pump Check Valves Flow Verification," Revision 14,
'
dated May 7, 1996. The purpose of this procedure was to verify the capability
of check valves in the TDAFW pump system to pass the, full design flow.
'
(Section 2.2 of the STP defined an acceptance criterion of 625 gpm flow for check valves V006 and V003.) -The-surveillance test involved aligning the
.
d
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TDAFW pump to'the steam generators and verifying that the pump flow was at ,
-
least 625 gpe, measured by either FE-3218 or FI-3403A. These flow instruments were located in the pump suction piping and measure the total pump flo :
!
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. Check valve V006 was located in the pump suction line and would be required to i pass the full pump flow during the test. However,. valve V003 was located in 4 the pump discharge line downstream from the open pump minimum flow -
- recirculation line. Therefore, valve V003 would pass less than the full pump
- flo l The difference between the measured pump flow and the valve flow was the-
'
recirculation line flow. According to Calculation 29.01, " Auxiliary Feedwater Pumps Minimum Flow Evaluation," Revision 0, the recirculation line flow was
- approximately 95 gpm with the pump running at 3960 rpm. Therefore, the flow
- through check valve V003 would be on the order of 530 gpa when the pump was
- operated at 625 gpm. The current acceptance criterion of 625 gpa, presented in Section 2.2 of FNP-1(2)-STP-22.13, was not consistent with the actual test
+
}
fl ow.- Criterion V of Appendix B to 10 CFR Part 50 requires that activities
'
affecting quality be prescribed by and performed in accordance with '
,
instructions, procedures or drawings which include appropriate acceptance
"
criteria for determining the activity is satisfactorily accomplished. The team identified the inadequate acceptance criterion for check valve as
- Unresolved Item 50-348; 50-364/97-201-0 Amendment 112 to TS 3/4.7.1.2 states that the TDAFW pump is capable of
- delivering a total flow of 450 gpm. The check valve test flow of
- approximately 530 gpm was significantly greater than the TS requirement of 450
- gpm. Therefore, the team did not have any operability concerns. In i responding to this item, the licensee stated that U1-TCN-14A and U2-TCN-12A would be issued to correct Section 2.2 of FNP-1/2-STP-22.13 for the forward
,
flow tes Conclusions
The team concluded that the licensee's AFW system testing was generally
]
sufficient to verify that the pumps and valves were capable of performing
, their required functions. However, the team noted weaknesses in the check
- '
valve testing. In one instance, the licensee's corrective actions in response to a previously identified violation did not thoroughly address the check
,
valve testing deficiency.
J
^
El.2.2.5 AFW System Modifications i
- Inspection Scope In this portion of the mechanical design review, the team selected 7 AFW system modifications for detailed e!aluation. In particular, the team
'
assessed each selected modification in detail to verify that the problem i
identification and justification for the modification were clearly stated; to determine if the subject was generic and potentially applicable to other
' systems;:to determine the impact on the design and licensing bases; to verify that the safety evaluation, if required, was correct and consistent with the
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applicable procedures; to verify that the post-modification testing was complete and appropriate; and to verify that the required plant documentation -!
was updated to reflect.the modificatio ,
. .0bservations and Findings- ;
The team had no concerns regarding the modifications reviewed. The majorit l
- of the modifications reviewed were minor in nature, addressing such issues as '
!- replacement of control valve actuators, addition of as-built information to
'
P& ids, acceptance of check valve replacement parts of a different material, ;
~
. acceptance of replacement mechanical seals for the AFW pumps, and replacement
,
of valve trim. The system modification packages were of high quality. The j problem identification and justification for each modification were clearly ;
stated, the safety evaluations were correct and consistent with the applicable '
procedures, post-modification testing was complete, and appropriate, and the ' !
required plant documentation was updated to reflect the modifications. None of the modifications were generic in nature or applicable to other systems, 2 and none of the modifications significantly impacted the design or licensing -
bases of the syste !
'
The most significant modifications reviewed addressed reducing the overspeed trip setpoint for the TDAFW pump and replacing a section of carbon steel i service water supply. piping with stainless steel. The team reviewed modifications in detail and found them to be acceptable.
h Conclusions ,
The team concluded that the licensee's AFW system modifications were consistent with the design-basis requirements and did not adversely affect the i ability of the system to perform its required functions during both ncrmal and
'
accident conditions.
'
El.2.3 Electrical Desian Review P
'El.2.3.1 Turbine-Driven AFW (TDAFW) Uninterruotable Power System (UPS) :
,
- Inspection Scope
) In this portion of the electrical design review, the team evaluated the TDAFW 4 batteries (Units 1 and 2) to assess their ability to supply adequate l electrical capacity, as well as their conformance to the FSAR, installation
,
documents, and maintenance and testing program Observations and Findings j Calculation 07597-E-106, " Battery Capacity Calculation for TDAFP-UPS," ;
~
Revision 1, provided the duty cycle analysis for those emergency loads required for AFW operation when battery charger support was not availabl ,
Calculation"E-106 determined that adequate battery capacity from the TDAFW l Class IE battery was available to supply the AFW electrical load ,
-
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Review of the maintenance and testing activities for the TDAFW batteries l indicated ~ that the licensee had performed weekly and quarterly inspections, ,
<
yearly battery equalization,18-month general battery cleaning, and 54-month UPS battery performance testing. However, the licensee did not' perform battery service testing to verify the ability of the battery to meet the design duty cycle specified in calculation 07597-E-106, as identified in !
Sections 5.3 and 6.6 of the "IEEE Recommended Practice for Maintenance, '
Testing, and Replacement of Large Lead Storage Batteries for Generating !
.
' Stations and Substations," of Standard'450-1980 promulgated by the Institute
>
of Electrical and Electronics Engineers (IEEE). The team noted, however, that other Class IE batteries in the plant were tested in accordance with IEEE 450 recommendations. FSAR Section 8.3.3.2 states that all TDAFW UPS components
, are designed to conform with Class IE electrical system design criteri !
FSAR Sections 8.1.4 and 8.3.2.1.5 identify the Class IE battery testing
requirement ,
The team did not have any immediate safety concern, since the licensee's
'
maintenance and testing (mentioned above) provided reasonable assurance-that i the battery could support the assigned loads. A test program with an !
' appropriate written procedure was not identified and performed in accordance with written procedures incorporating the requirements s 3cified in Criterion
XI of Appendix B to 10 CFR Part 50, and TS Section 6.8.1.a. The team therefore iduw.ified this issue as Unresolved Item 50-348; 50-364/97-201-0 The licensee issued REA 97-1410 to evaluate this item, and stated that battery
- service testing will be performed in refueling outage RF14 for Unit I and RF12 1 for Unit The team compared the installed configuration of the TDAFW UPS system to FSAR '
l Section 8.3.3, Design Change Packages (DCPs) 96-1-9008-0-001 and 96-2-9009-0- ;
001, Production Change Notices (PCNs)81-917 and 81-2125, and applicable design drawings. The installed configuration agreed with the design 1 l
documentation of the previously mentioned DCPs, PCNs, and engineering drawings. The six output fuses for the UPS rectifier had been changed from 3 l amperes to 5 amperes. Section El.2.7 of this report discusses an FSAR l discrepancy associated with this change. Adequate engineering was performed to ensure that the equipment was protected and coordinate Review of the TOAFW battery installation for Unit 2 indicated that various structural and electrical components were not installed in accordance with the manufacturer's drawings and instructions (7597-20-E10.9-29, Revision 1, U265645A, Revision 2, U263212, Revision D). The following deficiencies were noted:
I
. Five polystyrene spacers were installed between the battery cell and the end rail where none were require l
-
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Structural steel bracing in the rear of the rack did not agree with the drawin '
. Bolts were missing from the upper-and lower-tier tie rod bracket ,
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- .- Silicon bronze bolting hardware was utilized at the cable terminations in lieu of stainless steel hardware as shown on vendor drawing U265645A, Revision ,
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The intercell battery connections were torqued to 75 in-lbs instead of l the required 125 in-lbs specified in the battery manufacturer's l
instruction manual U263212, Revision D.
. .The' battery rack steel rails and tie rods exhibited corrosio !
' . The licensee reviewed the installed structural configuration of the' TDAFW a
battery rack and determined by. calculation that the corrosion, additional- )
~
polystyrene cell spacers, and structural member. configuration had not
< compromised the. seismic design of the battery installation. .On the basis of
- discussions with the battery manufacturer concerning the utilization of )
'
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silicon bronze fastener material in lieu of stainless steel and the lower r torque value of 75 in-lbs, the licensee concluded that the capability of the
,
battery to perform as designed had not been compromised.
.
The licensee issued Deficiency Report #537766 for the battery rack corrosion; >
REA 97-1408 to reconcile the differences between the installation and the !
design drawings; and REA 97-1410 to revise TDAFW battery maintenance procedures and appropriate documentation to clarify acceptable fastener ;
- material . The team determined that the TDAFW battery structural and i
electrical component installations were not accomplished in accordance with ;
drawings, procedures or instructions as required by Criterion V of Appendix B to 10 CFR Part 50. This issue was'therefore identified as Unresolved Item 50-
' 364/97-201-06,
- Conclusions ,
- Even though the team. identified discrepancies with the structural and
'
electrical aspects of the TDAFW battery installation, the team concluded that {
providing electrical power during the various plant operating modes for which
! it was designe ,
! El.2.3.2 AFW Electrical Loads ,
,
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} Inspection Scope i In this portion of the electrical design review, the team evaluated the
electrical loads required for the AFW system to perform its design functions :
under both normal 'and accident conditions. This evaluation addressed cable sizing, protective coordination, electrical bus loading, and de battery l loading _ calculation Observations and Findings e
?
To ensure that the. AFW loads had been identified, the team reviewed calculations SE-94-0470-001, " Unit 1 As-Built Load Study Update," Revision 1; SE-94-0470-004, "As-Built Load Study Summary Calculation," Revision 1; E-42,
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" Steady State Diesel Loadings Calculation for LOSP, SI and SBO," Revision 8; E-35.1.A, " Setting of Protective Relays.for FNP Unit 14.16 kV Auxiliary Power
- System," Revision 2; and E-95, " Battery Capacity Calculation for LOSP and
- LOSP+LOCA Situations and Limiting Battery Load Profile," Revision 7. All i
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major electrical loads had been accounted for within these calculations for !
- normal and accident conditions. The frequency and voltage ratings of the AFW electrical loads were compatible with the electrical design analyses. The team t
' determined that the methodology and assumptions used were appropriat ,
Calcu1'ation SE-94-0-0378-001, " Instantaneous Trip Settings for MCCBs in the MDV Setpoint Document," Revision 0, was reviewed. The team observed that the '
source of the overload relay time vs. current curve used for MOV-3406 was not documented. ' Additionally, the curve was used by extrapolation beyond its
- published data. The licensee contacted the overload relay manufacturer during the inspection and received information which supported the extrapolation used ;
within the calculation. The licensee stated that a copy of the information 4 from the manufacturer was filed for future reference. No additional concerns '
i were. identified with this calculation.
.
Cable sizing for the 4 kV Motor-Driven AFW (MDAFW) pump motors was reviewe The review compared the actual cable data for the MDAFW pump motors against
' the " Electric Power Cable Sizing Guide," Revision 0. This review concluded
- that cable sizing for the MDAFW pump motors was adequate and was in i
conformance with the Farley cable sizing criteri ;
o
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l Conclusions
.
The electrical design for components that perform the normal and emergency '
l'
functions of the AFW system supported the design-basis functions of the system. The electrical distribution system provided independent, redundant,
,
safety-related (Class IE) power to the AFW electrical load .
E1.2.3.3 Electrical AFW Modifications i Inspection Scope-In this portion of the electrical design review, the team evaluated electrical modifications to the AFW system and verified that the changes were in
'
conformance with the AFW and electrical design bases.
j Observations and Findings The following modifications associated with the AFW system were reviewed by
the team: ABN 96-0-0986, " Add Circuits to TDAFW Pump Control Panel"; PCN 89-1-6106,'"AFW HFA Relay Drawing Change per 88-0-4980"; and PCN 89-1-6354,
'
" Lighting Around TDAFW UPS Batteries." The document review for the referenced modifications indicated that appropriate procedures were followed in their 1 preparation and that proper design review and verification was employe !
,
- 10 CFR 50.59 evaluations associated with the modifications were technically complete and prepared in conformance with administrative procedures. The team did not identify any concerns with these modification .
16 m
q,- w% ww 7- pi rm r
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. Conclusions
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The modifications reviewed by the team were designed in accordance with the
<
system design base ;
1 El.2.4 Instrumentation & Controls (I&C) Desian Review .
El.2.4.1. Condensate Storaae Tank-
] Inspection Scope l In this portion of the' I&C design review, the team evaluated the licensee's )
determination of the CST low-level alarm ] Observations and Findings
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The purpose of calculation SM-87-4380-001, Revision 0, " Condensate Storage 3 Tank Low-Level. Alarm Switches," was to prove that the condensate storage tank )
low-level alarms, Q1P11LA4133A and Q1P11LA4133B, would allow at least 20 minutes for operator action. It also verified that a " Moore Industries" switch could be used to replace an existing "Rosemount" switch to provide the alarms.
.The calculation was reviewed for adequacy and setpoint uncertainties. The l
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team observed that drift error was not addressed in the calculation. The I&C design criteria described this variable as being highly important in selecting
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setpoint tolerances. Rosemount defined the drift error for the Model 1152DP transmitter used in this loop as i 0.2% upper range limit (URL). Although the
- use of this drift error did not add significantly to the loop error, it should have.been addressed in the calculation. The team found that the available level was conservative and a drift of 0.2 % URL (less than one inch in 30
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- - months) was very minima The team also observed that the calculation provided a deadband specification
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for the replacement switch which was analyzed in this calculation for acceptability. The total instrument tolerance calculated by the square root of the sum of the squares method did not include the deadband of 1% of span.
4 In "I&C Design Criteria for -Joseph M. Farley Nuclear Plant-Units 1 and 2,"
page 15, Steps 12.6.1 and 12.6.2 Southern Company Services, Inc. noted that deadband should be assessed. The team found that it was not clear as to the circumstances to use it ( i.e., when to use deadband in uncertainty
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calculations for a particular type of switch).
The licensee issued REA 97-1407 to review this uncertainty calculation and clarify the application of deadband as a factor in' scaling calculations. The
- review of uncertainty- calculation to address drift and deadband errors and e clarify the application of deadband as a factor in scaling calculations was
. identified as Inspection Follow-up Item 50-348/97-201-07.
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, Conclusions 4 The'absenc of drift and deadband in the setpoint uncertainty calculation did
. not significantly affect the trip setpoint. The design basis was maintained.
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The set point had adequate margin to alert the operators for switchover to S El.2.4.2 Instrument Loon Uncertainty calculations .
I a.- Inspection Scope In this portion of the IAC design review, the team evaluated the consistency
> and adequacy of the licensee's setpoint bases and uncertainty calculations for .
various instrument loops in the AFW syste .
j Observations and Findings The licensee was asked to provide various design-basis documents definin ,
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analytical design limits and setpoint uncertainty for selected instrument i loops in the AFW system. Only one calculation was provided, SM-87-1-4380-001,
"Setpoint Calculation for Level Alarm Replacement," Revision 0. This was a ;
calculation developed before the current setpoint program (GO-M-1, Revision 3)
was initiated. The team found that it did not contain a complete instrument i
loop uncertainty calculation nor did it consider drift and deadband error The licensee could not produce additional instrument calculations to review.
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The lack of calculations on some instrument loops made it difficult to judge l whether the methodology that went into the development of the original setpoints conformed to the current setpoint program. The licensee developed a preliminary uncertainty calculation for the AFW system, Loop F3402, " Motor- i
- Driven AFW Pump Suction Flow" (high-flow setpoint), .to confirm the validity of
the original margin between the setpoint and the analytical limit. This <
calculation showed that there was adequate margin in the alarm setpoint, ;
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considering loop uncertainties for a high-flow condition in the suction of the
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motor-driven AFW pump. The preliminary calculation for Loop F3402 was
- adequate and consistent with Farley's current setpoint program, GO-M-1,
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Revision 3. The methodology in the uncertainty calculation was also , ,
consistent with the I&C design criteria. The terminology in the calculation '
was adequately defined and included appropriate supporting reference Conclusions
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- The calculation provided confidence that there was built-in conservatism and margin with the data found in surveillance test procedures. The setpoint c program was adequate and consistent with industry standards and consistent with Westinghouse methodolog ?
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El.2.4.3 Instrument Setooint Uncertainty Proaram ;
- Inspection Scope l
. In this portion of the I&C design review, the team evaluat'ed GO-M-1," Designer
- Interface Document," Revision 3, which provided design guidance for the Farley l instrument setpoint uncertainty progra i
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- Observations and Findings The licensee's setpoint p ogram was generally consistent with recommendations ,
from the Instrument Society of America (ISA), and the setpoint calculation
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methodologies followed the ISA standards ISA-567.04 Part I and II. .The licensee had incorporated the above methodologies in I&C design criteria document. The random uncertainties were combined using the square root-sum-of-squares, and non-random uncertainties were combined algebraicall !
, The team's review of design interface document GO-M-1 showed that when design
. changes impacted setpoint calculations that had been performed by Westinghouse ;
i in emergency operating / response procedures (EOPs/ERPs), the changes had to be evaluated by Westinghouse. Examples included uncertainties for considerations :
of sensor and instrument rack drift, accuracy and environmental effects, l
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indicator accuracy, and calibration accuracy. The team found that the flow
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chart in Attachment 1, page 69, of GO-M-1, which depicted the instrument i setpoint control, lacked the flow path for sending the change to Westinghouse !
for evaluation. The licensee stated that GO-M-1 would be reviewed to determine ',
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if changes were appropriate and issued REA 97-1410 to revise the I&C design criteri !
l Conclusions .
4 -
The team found that design document adequately described the method to update i Westinghouse design changes, but the flow chart did not adequately reflect tie 3 text of the document. The setpoint uncertainty program was adequat El.2.4.4 Modificatigns and Other Reviews l Inspection Scope l In this portion of the I&C design review, the team evaluated 12 AFW system l
design modifications and associated 10 CFR 50.59 safety evaluations to assess
their adequacy and consistency with the relevant design bases. In addition, the team reviewed ATWS (anticipated transients without scram) mitigating system actuation circuitry (AMSAC) logic design, AFW actuation circuits I
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design, and post accident monitoring instrumentation in accordance with l Regulatory Guide (RG) 1.97, " Instrumentation for light-water-cooled nuclear l power plants to assess plant and environs conditions during and following an j accident," Revision 2 desig I
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- .0bservations and Findings i The team found eleven modifications adequate. The quality of the modification l
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documentation was generally good and consistent with the design bases. The team found the associated 10 CFR 50.59 evaluations acceptable. However, one :
i discrepancy with PCN 84-1-2518, " Auxiliary Feedwater Check Valve Temperature !
i Monitoring System,"_ Revision 3, was found. This modification installed l
temperature elements on either_ side of various MDAFW and TDAFW pump discharge i
check valve The orientation'of temperature elements TE-2293K and TE-2293L on P&ID D- !
2 :
. 175007, Revision 22, were reversed from the orientation in the PCN. The'
licensee issued ABN-97-0-1053 to revise the P&I i
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The team verified that the AMSAC logic design for the MDAFW pumps was l
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consistent with design requirements. The team also verified that appropriate l
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isolation devices were installed between Non-lE AMSAC circuits and IE turbine- I driven auxiliary feedwater trip circuits, and between non-lE and the IE:
~ auxiliary feedwater start circuits. The steam generator level output signal ,
to AMSAC (loops L485 and L496) was isolated properly with design requirement !
' AFW automatic initiation /it,olation logic design was also consistent with the ,
!
! guidance provided in NUPF6 0578, 'TMI-l Lesson Learned Tasks Force Status !
Report and Short Term Recommendations." The team determined that both the
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MDAFW and the TDAFW pump would automatically start on a low-low steam l 4 generator level as sensed by the AMSA The team reviewed Farley's computerized Scaling Manual Unit 1, Section 10, ,
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" Steam Generator Level Control & Protection. The team found that this document which calculated the low-low level Reactor Trip was consistent with '
- design requirements. However, the team did not review the Westinghouse l
uncertainty calculation to evaluate loop error as part of the square root-of-sum-of-squares (SRSS) methodology. The automatic actuation of the AFW system l
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met the requirement of NUREG 0578.
d
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In addition, the team reviewed the Regulatory Guide 1.97 Category 1 compliance checklists for the condensate storage tank level transmitters LT 515A/5168, c
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! Auxiliary Feedwater Flow transmitters, FT 3229A, B, and C. Regulatory Guide 1 1.97 stated that the indication for the AFW flow was considered a Category 2,
1 type D variable with a required scale range of 0 to 110% of the design flow.
- The team verified that the scale for the AFW indicators met this guidanc !
!
The team found the checklists and instrument design consistent with RG 1.97 e d
' guidanc :
! Conclusions ;
- - The team concluded that the licensee's design modifications maintained the
. plant's design bases and no concerns were found with the safety evaluations.
The licensee's design for AMSAC, post-accident monitoring instrumentation and !
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.AFW system initiation met the design-bases requirement !
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El.2.5 System Interface Desian Review El.2.5.1 Service Water (SW) System Inspection Scope In this portion of the. system interface design review / the team evaluated the SW system P&!D to verify that the interface between the SW system and the AFW system was coriastent with the AFW system design bases. .The team also reviewed Production Change Request (PCR) 92-1-8234, which replaced the carbon steel SW piping to the AFW system with stainless stee Observations and Findings
A backup source of water for the AFW pumps was provided from the safety-related portion of the SW system. The SW supply was isolated from the normal suction piping by two closed MOVs. These valves could be operated remot manually from the control room or by using the manual handwheel at the valv The SW system also supplied cooling water for the motor-driven AFW pump room ,
cooler The team had no concerns related to the SW system interface. A review of the SW system P& ids, D-175003, Sheet 1, Revision 32, and D-175003, Sheet 2, l
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Revision 27, indicated that 8-inch, safety-related SW system lines were '
provided from redundant SW system headers to the AFW system and to the AFW pump room coolers. This system configuration was consistent with the AFW design base The team questioned the potential for galvanic corrosion at the dissimilar weld interface between the stainless steel piping added by PCR 92-1-8234, Revision 1 and the existing carbon steel valve body. In response, the licensee had evaluated this condition and determined that galvanic corrosion was not likely to be significant. The licensee also stated that the last J examination of the equivalent weld on Unit 2, performed in April 1996, did not show any significant degradation. The team found this condition acceptable, Conclusions
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The interface between the SW system and the AFW system was consistent with the AFW system design base El.2.5.2 Instrument Air (IA) System ) Inspection Scope In this portion of.the system interface review, the team evaluated the AFW l system P&ID and the main steam (MS) and auxiliary steam system P& ids to verify l the normal and failure positions of each air-operated valve in the AFW system l and the steam supply to the AFW pump turbine drive. The team also reviewed ;
the service air system P&ID, D-175035, Sheet 2, Revision )
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- ' Observations and Findings >
The team had ru: concerns related to the IA system interface. The IA system :
- provided air for the air-operated valves in the AFW system and the steam :
supply to the AFW pump turbine drive. A review of the AFW system P&ID, !
D-175007, Revision 22, and the MS and auxiliary steam system P&ID, D-175033, >
- Revision 18. . indicated that~ each of the safety-related air-operated valves was
' designed to fail in a safe position upon loss of air or was provided with an
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air reservoir.to allow continued operation of the valve. This design was .
consistent with the AFW design bases.
- Conclusions
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The interface between the'IA system and the AFW system was consistent with the ;
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i- AFW system design base E1.2.5.3 Main Steam (MS) System '
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! Inspection Scope 1 ,
2 In this portion of the system interface review, the team evaluated the AFW system P&ID and the MS and auxiliary steam system P&ID to verify that the t interface between the MS system and the AFW system was consistent with the AFW
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system design bases.
i Observations and findings *
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The MS system provided a steam supply to the TDAFW pump driver. The interface '
, between the systems was upstream from the safety-related MS isolation valves F in the MS valve room. A review of the MS and auxiliary steam system P&ID, D-175033, Revision 18, indicated that two redundant steam supply lines were .
provided from redundant steam generators to the AFW pump driver. This system l configuration was consistent with the AFW design base l
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i The team questioned if NRC Information Notice 91-75, " Static Head Corrections Mistakenly Not Included In Pressure Transmitter Calibration Procedures," was .
taken into account in Farley's steam generator level control scaling i
)
- documents. The response to this information notice discussed Vogtle Plant's
> discovery that a static head correction of approximately 25 psig had not been l applied during the calibration of the pressurizer pressure transmitters. The l
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same concern could apply to steam generator _ level measurement loops. The team reviewed Farley's Computerized Scaling Manual for Unit 1, Section 10, " Steam Generator Level Control and Protection." The document was reviewed for
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. adequacy and consistency.between the setpoint program and industry standard The team found that the Westingh:ose Computerized Scaling Manual was consistent with the methodology of the I&C design criteria document and F industry standard ISA-567.04, Part II. The scaling manual was adequate and
- setpoint studies were consistent with the requirements of the I&C design
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criteria.
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. Conclusions i The interface between the MS system and the AFW system was consistent with the AFW system design base c El. System Walkdown
' Inspection Scope During the course of the design inspection, the team walked down all of the 1 accessible portions of the AFW system, portions of interfacing systems, and
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the control room. Consistency of calibration intervals with licensing
. documents and calibration data were also checked during the walkdown.
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Verification of the as-built designs was conducted and compliance to RG 1.97 was observe Observations and Findings
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- TDAFW Pumo Vent Sigi During the team's walkdown of Unit 1, it was observed that the safety-related i- TDAFW pump vent stack was installed outside, on the roof of the Auxiliary Building. .This vent was not protected from tornado-generated missiles. FSAR Section 3.5.4 states that Category I equipment.and piping outside containment
are either housed in Category I structures or buried underground. FSAR Sections 6.5.1, 3.2.1.3, 3.2.1.5 and Table 3.2-1 state that AFW system
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equipment and piping are Category 1. FSAR Table 3.2-1 addresses the tornado missile protection of the AFW pumps. .This stack could be damaged by a tornado missile, which could restrict the steam flow path from the turbine drive and
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adversely affect the AFW pump operability. Criterion II of Appendix A to 10 CFR Part 50 requires that SSCs important to safety shall be designed to withstand the effects of natural phenomena such as tornados without loss of capability to perform their safety functions. The Farley licensing bases included this criterion. The installed condition of the safety-related TUAFW
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pump vent stack did not conform to Farley design and licensing basis >
requirements. The team noted that the requirements of Criteria III and V of Appendix B to 10 CFR Part 50 which state that applicable regulatory requirements and design basis are correctly translated into drawings, procedures and instructions and that activities affecting quality be prescribed by and performed in accordance with instructions, procedures or drawings were apparently not met. This issue was identified as Unresolved
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Item 50-348; 50-364/97-201-0 The licensee stated that the AFW system could withstand the loss of the TDAFW pump in conjunction with a loss of offsite power (LOSP) and a single failure of one MDAFW pump and that the remaining MDAFW pump would have sufficient capacity to safely cool down the plant. The licensee identified the following fpreliminary action plan to address this issue:
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The licensee prepared a draft safety evaluation, which reflected the use of
one MDAFW pump to shutdown the plant in the event that the vent stack was '
damaged. The licensee stated that this approach could result in a potential )
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USQ, requiring prior NRC approva An analysis will be prepared to evaluate the effects of the applicable tornado i
missiles striking the exposed vent stack. This analysis would determine if
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i the TDAFW pump would remain operable after the event. If this analysis shows j
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j that the TDAFW pump would remain operable, a safety evaluation will be prepared to revise the FSA { !
If neither of the above items are favorable, then appropriate modifications to :
i the vent stack would be implemented to: prevent missile damag l f
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! The licensee issued ABN-97-0-1043 to address this ite l
!
- Diesel Generator Exhaust S.ilencers !
i The team also observed that the safety-related emergency diesel generators and j the station blackout diesel generator exhaust silencers for both units were i
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installed outside, on the roof of the diesel generator building. This j j
equipment appearad to be protected from horizontal tornado-generated missiles ;
by the buildir; walls. However, the equipment was susceptible to vertical ;
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missiles and other non-horizontal missile FSAR Section 3.5.4 states that Category I equipment and piping outside i containment are either housed in Category I structures or buried undergroun FSAR Table 3.2-1 addresses the tornado missile protection of the emergency j
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diesel generators. FSAR Table 3.2-1 did not address the exhaust silencers 3
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associated with the generators. FSAR Section 3.5.241 discusses the types of i j
4 missiles considered for missile protection. However, the missile spectra were
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not clearly defined. The team asked if the plant's design bases distinguished l
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between horizontal and vertical /non-horizontal tornado missiles, and if the i c current design was consistent with the design base j The licensee stated that, although the FSAR did not oistinguish between i i horitontal and vertical /non-horizontal missiles, a horizontal missile was
- evident as the design basis for the plant. However, the licensee could not
>rovide any documentation clearly supporting their statement. This issue will
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i se further reviewed by the NRC to determine if the tornado missile protection f in the Farley design and licensing bases included missile spectra other than horizontal missiles. This issue was identified as Unresolved Item 50-348; 50-364/97-201-09. The licensee issued REA 97-1409 to address this item.
! AFW Flow Control Valves During a Unit 1 AFW system walkdown on February 6,1997, the team observed that one of the safety-related, air-operated AFW flow control valves (HV-3227B) had been tagged with a Deficiency Report (DR) 547310 indicating that screws were missing. The team observed that the missing screws were required'to hold a solenoid valve to a bracket on the valve actuator. The g
solenoid valve was temporarily attached to the actuator with tie wrap. The
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licensee stated that the DR had been initiated on January 10, 1997 )
(approximately 3 weeks prior to start of inspection). Following questioning i by the team, the condition was corrected on February 6, 1997, via Work Order ]
Number 547310. The team questioned the timeliness of the repair for this i safety-related valv !
The licensee stated that DR 547310 had been electronically entered into the DR l'
system when the condition was discovered. This DR had not yet been scheduled
- for repair as of February 6, 1997. The licensee reviewed the condition of the valve and concluded that the valve had been operable before the' repair. Valve HV-3227B was in the supply line from the MDAFW pumps to the "B" steam i generator. The AFW flow control valves were normally maintained in a fully open position. These valves were required to be open during an accident to
- perform their safety function. A failure of the-solenoid 'would result in the valve failing to the open position. In addition, redundant MOVs were located- )
in series with HV-3227 ]
i The licensee stated that the remaining items in the DR system were reviewed l and it was verified that no other potentially significant items related to safety-related equipment existed. After DRs were entered into the DR system l they were reviewed and prioritized by a dispatcher. The dispatcher was
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- responsible for identifying any potentially significant items. The team observed that the potential significance of DR 547310 was not identified
, during the revie . CST Level Transmitters l
. The team inspected the installation of the CST level transmitter Q1P11LT516 :
! and the associated freeze protection. The inspection revealed that the inside
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walls of the freeze protection enclosure were partially insulated and in need
- - of re' pair. Drawing A-170256 Sheet 61, Revision I showed the sides of the
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freeze protection enclosure insulated. Eight inches of process tubing at the '
transmitter housing and transmitter housing were missing the required heat
tracin l
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The team also observed that the he'at tracing power cable appeared not to be '
securely. fastened along the run to the thermostat mounted on the freeze protection enclosure. The cable was observed to be attached to the CST ,
foundation by aluminum bonding tape and not run in conduit. This method was j approved by plant drawing B-172374. The review of licensee's actions to j address freeze protection installation issue was identified as Inspection
'
Follow-up Item 50-348/97-201-10. The licensee issued Work Request (WR)
97001089 to correct this proble ;
i In addition, the team verified that QlPllLT515, LT516 instrumentation installation met the guidance provided in RG 1.97, " Instrumentation for Light- ;
Water-Cooled Nuclear Power Plants to Assess Plant and Environs- Conditions i During and Following an Accident," Revision j
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I' Other Issues Reviewed i
The team observed that cable channel IVBDDA0A was not connected to its lower
> support above MDAFW pump motor 18. Cable IVBDG10P was routed through IVBDDA0A l '
to the IB MDAFW pump motor. The licensee reviewed the installation and concluded that the power cable would be able to perform its intended function .
t
- in the configuration observed since the seismic accelerations in this area
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were small. The licensee initiated deficiency report 537759 to correct the
, channel attachmen During the inspection of the Unit 1 TDAFW pump, a flexible conduit on the turbine skid was noted as being disconnected from its connector at its :
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junction box. The cable within the flexible conduit fed the manual trip
- solenoid on the turbine. The licensee evaluated the situation and concluded ;
that the operation of the TDAFW pump was not compromised since the circuitry was used in the manual tripping of the pump. DR 537762 was initiated to ;
repair the disconnected flexible condui ,
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" Conclusions f In general, the AFW system design observed during the walkdown was consistent with the design-basis requirements. However, the team identified that in some
. cases the design and licensing bases requirements were not fully implemented ,
as required by regulations. The team also identified that further NRC staff review was required to determine if the tornado missile protection in the '
, Farley design and licensing bases included missile spectra other than
. horizontal missiles E1.2.7 FSAR. FSD and Other Reviews
.
, Inspection Scope ;
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During the Farley design inspection, the team reviewed the FSAR sections and FSDs related to the AFW system and the associated electrical and I&C system In addition, the team reviewed licensee's self-initiated safety system assessment (SSSA) for AFW system, Observations and Findings The team identified the following discrepancies in the FSAR:
~ FSAR Section 6.5.2.2.4 stated, "All valves in the AFW flow path from the ;
condensate storage tank to the steam generators are normally open, with the exception of the fail open AFW flow control valves." This statement implied that the AFW flow control valves were not normally open. In accordance with the plant's operating procedures, these valves were normally maintained in the open position. The FSAR statement was not ,
correc :
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, The team observed that electrical components. associated with the
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safety-related AFW discharge valves HV-3227A, B, C; and HV-3228A, B, C j were located in the main steam valve room approximately 4 feet above the i
, floor (elevation 131 feet - 0 inch). The control valves were not above ,
133 feet, 3 inches as stated in the FSAR. This elevation was above the !
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-stated flood elevation for the room; but was not consistent with the i
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[ FSA In addition, the team observed that the limit switches associated with the main steam isolation valves (MSIVs) were located less than one !
foot above the floor, which was below the stated flood leve '
i !
FSAR Section 3K.4.1.2.7(F) stated that the lowest safety-related !
equipment in the main steam room was the atmospheric relief valves ;
, located at elevation 133 feet, 3 inches. The floor elevation was 127 i
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supply valves were also located in the main steam room. FSAR Section :
3K.4.1.2.7(F) also stated that the flood elevation in the room would be 3 feet, 5 inches above the 127 feet, 0 inch elevation before the AFW I
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pumps were isolated, j i
The licensee stated that the MSIVs were environmentally qualified for i
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submergence and that the location of the auxiliary feedwater valves was acceptabl ;
, Section 8.3.3.2 did not reflect the installed fusing for the TDAFW UPS i system. The FSAR stated that the rectifier output fusing for the TDAFW
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changed the fuse size to 5 amperes.
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The above discrepancies had not been corrected and the FSAR updated to ensure l that the information included in the FSAR contained the latest material as :
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required by 10 CFR 50.71(e). This issue was identified as Unresolved Item
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50-348; 50-364/97-201-11. The licensee issued ABN 97-0-1043 to correct and ]
clarify the FSA The team also identified the following discrepancies in AFW FSD A-181010, .
Revision 1: l
' Section 3.10.2.3 stated that the overspeed trip for the TDAFW pump would prevent pressurization of the system in excess of the design pressure rating. During a TDAFW pump overspeed condition, the system pressure
, would exceed the design pressure rating, but would be within allowable l
- limits for infrequent operation, as specified in Section III of the ASME ,
B&PV Cod l l Table T-2 identified the power supply for SV3234B as panel B instead of l panel D and did not show the power supply for FISL-3212A as Panel ;
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The licensee issued ABN 97-0-1043 to revise the FS l i
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..The licensee has' performed 13 SSSAs to'date to determine the system functional i
r performance. .A review of the AFW SSSA identified that the licensee's
. assessment identified many good issues. The open issues resulted from this il
assessment was. satisfactorily resolved except in one instance. Section i E1.2.2.1 addresses this issue in greater detail. Although, the assessment
- identified many issues, licensee's assessment did not identify all of the !'
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issues' discussed in this repor !
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l Conclusions j j The team concluded that, in several instances, licensee did ~not revise the f FSAR in accordance with the requirements of 10 CFR 50.71(e). The team also ;
identified several discrepancies during the FSD review. These weaknesses .
either have been, or are currently being, evaluated by the licensee. Although . i the safety system self-assessments performed by the licensee failed to !
- identify many of the issues raised by the team, these assessments led to the !
identification and correction of many other issue !
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El.3 Comoonent coolina Water (CCW) System j
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El.3.1 System Descriotion and Safety Functions l The CCW system was a closed cooling water system which transferred heat to the !'
i (SW) system from components which process radioactive fluid. The system was designed to function during normal operation, plant cooldown, refueling, and l accident and post-accident conditions. . Those portions of the system which i
! cool safety-related components were redundant and safety-relate l The system consisted of three CCW pumps, three CCW heat exchangers, a two- 1 section surge tank, interconnecting piping, and instrumentation. Two CCW I
- pumps, heat exchangers, and both sections of the surge tank were dedicated to i two separate trains. The third pump and heat exchanger were capable of being
aligned to either train. Manual isolation valves were used to separate the l two redundant trains. One pump and one heat exchanger were normally operated
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to provide cooling for various components in the Auxiliary Building and containment, with another pump on standby. A standby heat exchanger was available should it become necessary to isolate the operating heat exchange '
The third pump and a heat exchanger were isolated from the operating components and on standby to cool engineered safeguards system components if l needed. Two CCW pumps and heat exchangers were normally used during cooldow i
'
~ During emergency operation, the non-essential loads were shed and the i dedicated train was automatically started. Although both trains were used for {
accident conditions, only one train was require l t
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. El.3.2 Mechanical Desian Review
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'El ...3 2 1 CCW System Performance !
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' Inspection Scope
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. In' this portion of the mechanical design review of the' CCW system, the team i evaluated the ability of the system to cool safety-related components during
'
both normal operation and accident condition . Observations and Findings : CCW Heat Removal Capability Calculation 37.7, " Component Cooling Water (CCW) Flow Balance," Revision 0, -
,
documented a hydraulic computer model of.the system and determined the flow through components cooled )y CCW for various modes of operation. The model was calibrated in order to reproduce the test data obtained during normal condition and cooldown condition validation testing. -The modes of operation :
a considered were normal, cooldown, loss-of-coolant accident (LOCA) injection, l
' and LOCA recirculation. Single. failures and a pipe break in the nonsafety-related portion of the system were considered for each mode as appropriate to i cover limiting conditions. This calculation used the pump vendor head-flow
- curves as input. The team reviewed calculation 37.7 and determined that the methodology and assumptions used were appropriate. Validation for the i
" BALANCE" computer program was included in the calculatio Calculation 37.4, "CCW System Heat Exchanger Models and Heat Removal Capacity !
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Calculation," Revision 0, developed models for the heat exchangers cooled by
'
the CCW system. These models were integrated with the flow balance models developed in calculation 37.7 to evaluate the heat r'emoval capacity and'the temperatures in the CCW system for various modes of operation. The modes ;
included were normal cooldown, abnormal cooldown (cooldown in one unit, LOSP, '
and loss of an electrical bus in the cooldown unit), LOCA injection, and LOCA
,
recirculatio .
!
Heat loads for each of these modes were used to calculate CCW temperatures at the outlet of the CCW heat exchangers, which were then used to calculate i i process temperatures of the components cooled by CCW. The maximum calculated process inlet temperature for the reactor coolant pump (RCP) bearing oil
- coolers was above the design allowable during a cooldown mode with the loss of j an electrical bus. The nuclear steam supply system (NSSS) vendor stated that RCP operation during the cooldown was not a system requirement and that i 4 shutting down the RCP before the design limits were reached would be acceptable.
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l l RCP seal water return temperature was also calculated to be higher than the i published design allowable during the abnormal cooldown mode. An evaluation !
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.by the NSSS vendor determined that the seal water temperatures were acceptable j
'
considering the lower reactor coolant temperature when elevated CCW temperatures would be experienced and_the component design limits. The ability to cool down the RCS to 200*F in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> was verified. Calculation ;
, 1
.29 !
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g- y-- , + --- ,.
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37.4 also verified that the design basis SW flow rate of 10,000 gpm to the CCW heat exchanger was adequate. The team reviewed calculation 37.4 and found it adequate. Validation of. the " BALANCE - Heat Exchanger Performance Utility" computer program used was included in the calculatio 'The above two calculations were dependent on the CCW pump performance curve supplied by the pump vendor. Calculation BM-95-0776-001, "CCW System Evaluation Using Degrade CCW Pump Curves," Revision 0, evaluated the performance of the CCW system with the CCW pumps degraded approximately 10%.
This pump performance was referred to as the minimum analyzed pump curve and was used with the in-service testing program. This calculation used the flow model developed for Unit 1 in calculation 37.8 as it resulted in slightly lower flow rates than the Unit 2 calculation did and thus would produce conservative results for_the Unit 2 CCW heat exchangers. The thermal model from calculation 37.4 was used to predict process and CCW fluid temperatures using CCW flows calculated from the minimum analyzed pump curves during operating modes of two train cooldown, abnormal' cooldown, LOCA injection, and LOCA recirculation. The calculated process and CCW temperatures were evaluated as acceptable. The team reviewed this calculation and found it adequat . Net Positive Suction Head (NPSH)
The team reviewed calculation 34.5, -" Component Cooling Water System NPSH (ES No. 90-1820)," Revision This calculation concluded that substantial margin existed between the NPSH required and the NPSH available for the CCW pump The calculation stated that the flow through the pipe from the CCW pump suction header to the CCW surge tank would be negligible under normal operation and that there'was no large leak from the system requiring significant flow from the tank. Therefore, no head . loss was assumed in this i pipe. However, calculation 39.3, "CCW Surge Tank Analytical Limit for Level fetpoint," Revision 0, calculated a flow of 801 gpm through this pipe for the period from a pipe break occurring in the nonsafety-related portion of the CCW system and automatic isolation of the break (10 seconds from the low-low level setpoint in the CCW surge tank). This flow resulted in a very small reduction of the NPSH available of less than one %. The team concluded that there was adequate NPSH available to the CCW pumps in all modes of system operatio . Sinale Failure Desian The team reviewed the capability of the system to accommodate a single failure in conjunction with postulated plant accidents. The mechanical design included adequate valving to enable the two trains of the system to be
,
separated for combinations of operating and standby CCW pumps and heat exchangers. The team reviewed the applicable portions of the following operating procedures for the CCW system and SW system, and determined that the system was properly operated to ensure the ability to accommodate a single failur . FNP-2-SOP-23.0A, " System Operating Procedure Checklist, Component Cooling Water System," Revision 5
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! . . FNP-2-SOP-23.0, " System Operating Procedure, Component Cooling Water
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System," Revision 39
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, FNP-2-SOP-24.0, " System Operating Procedure, Service Water System,"
! Revision 31 j
l Floodina of CCW Comnonents - j
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l The team reviewed the potential for flooding of the room which contained the .
CCW pumps and heat exchangers. Bechtel letter AP-17096, dated Dec. 22, 1989, ;
"CCW Heat Exchanger Room Flooding," reported the results of.a flooding :
evaluation and concluded that there was no concern for flooding in this roan since there were no credible line breaks or other sources of flooding present. The team confirmed that the design-basis criteria used to postulate pipe ,
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breaks was appropriate and conducted a walkdown of the room and adjacent areas which further confirmed the results of the evaluation.
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- Pine Stress and Suonort Analyses
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The' team reviewed the CCW fluid temperatures used in the pipe stress and i
support analyses. The licensee had discovered a discrepancy in a Unit 1 ,
stress calculation during analyses performed in support of a snubber reduction !
program. The normal operating temperature was used instead of the higher temperatures that could occur during a design-basis accident. The licensee reviewed the stress calculations for the CCW systems for both units.and i determined that the maximum operating temperature was generally higher than
. the temperature used for the stress- analyses. The licensee stated that, as l
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substantiated by an initial assessment, the CCW system for both units met the j ASME Code allowable limit i The team reviewed the maximum operating temperatures used in the assessment of
'
- the Unit 2 CCW system and noted some inconsistencies in the data. The i licensee reviewed the maximum operating temperature data for both units and
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i l corrected the data as required. The corrected data did not change the
- licensee's conclusion that the systems met the ASME code allowabl ,
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The licensee established a screening criteria to determine actions required ;
'
j for the existing pipe stress calculations. The licensee determined that negligible impact on the results of the existing analysis would exist if the .
difference between the current analysis temperature and the maximum operating !
$ temperature was'less than 20*F. Calculation Change Notices were being ,
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prepared to document the revised temperature for each affected calculation so that the corrected temperature would be used if the calculation was revised in the future. The licensee stated that pipe stress calculations for which the
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difference between the current analysis and maximum operating temperatures was greater that 20*F would be revised and the support loads recalculated to ensure that design allowable values were me :
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The licensee issued Deficiency Notice 97-001, " Incorrect Dimensions and Temperatures Used in CCW Pipe Stress Analysis," to document the conclusions reached and recommended that a Root Cause Evaluation Team be formed to determine the root cause and the broadness of this event. The licensee also issued REA 97-1407 to revise the Unit I and 2 stress calculations. The adequacy of licensee's root cause evaluation and the corrective actions to .
'
r revise the affected calculations were identified as Unresolved Item 50-348; i 50-364/97-201-1 h
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< Conclusions
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The team concluded that the CCW system was capable of performing the design-basis functions of cooling safety-related equipment during the various plant
.
operating modes for which it was designed. The design margin has been reduced because of the discrepancy in the temperatures used in the pipe stress and j support analyse .
El.3.2.2 CCW Surae Tank a.- Inspection Scope
' In this portion of the mechanical design review of the CCW system, the team evaluated the surge tank to determine if it provided an adequate water supply
- until a pipe break in the nonsafety-related portion of the system can be '
isolated. In addition, the team evaluated whether the surge tank was provided with adequate pressure relief and vacuum breaker capability to maintain the required tank operating pressure within design limit , Observations and Findings
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The team reviewed calculation 39.3, "CCW Surge Tank Analytical Limit for Level Setpoint," Revision 0. This calculation determined the tank level for the low-low level setpoint which would result in the tank becoming empty.in the 1
'
event of a pipe break in the nonsafety-related portion of the system. The CCW i
' pumps had adequate NPSH available when the tank was just empty. The minimum
- level setpoint determined by this calculation was conservative when compared l
'
to the actual setting. The calculation used the results of calculation 38.6, ;
" Determine Flow Rate Through Pipe Break in the CCW System," Revision 0,* as i
input for the maximum flow rate out of the surge tank. The team reviewed this calculation also and concluded both calculations were adequat j The surge tank was protected from overpressure by relief valve PSV-3029, which was designed to limit the internal tank pressure to 14 psig when subjected to the flow which could result from a rupture of a reactor coolant pump thermal barrier cooling coil. The team verified that this flow was less than the 300 gpm for which the relief valve was designed. The relief valve discharge was piped to the floor drain tank in the Waste Processing System. The team reviewed the portion of calculation 12.19, " Nuclear Relief Valves Sizing,"
Revision 1, that dealt with PSV 3029 and walked down the routing of the
'
discharge piping. The team verified that the back pressure that could exist on the relief valve at its maximum capacity was less than that assumed in the valve sizing calculation and that the design input to the calculation was
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t correct. The sources of makeup to the surge tank were the Demineralized Water and Reactor Makeup Water systems. The team verified that neither system was capable of providing a flow to the surge tank in excess of the relieving capability of PSV 302 *
The tank was protected against external pressure conditions, which could exist if the tank vent was isolated during an outflow, by redundant vacuum breaker valves which would open at less than 1 psi Conclusions The surge tank provided adequate water volume for the system and was adequately protected against pressures outside of its design base E1.3.2.3 CCW System Containment Isolation Inspection Scope In this portion of the mechanical design review of the CCW system, the team evaluated the licensee's containment isolation provisions for the CCW pipes that penetrate containment. In particular, the team focused on correct implementation of the design bases for such penetration Observations and Findings The CCW system penetrated containment for the supply and return of cooling for the reactor coolant pumps (RCPs) and the excess letdown and reactor coolant drain tank heat exchangers. Motor-operated, air-operated, and check valves were used as containment isolation valves. The team verified that the containment penetration arrangements were designed to accommodate a single active mechanical or electrical failure and that containment isolation capability was' not lost as a result of a failure of the nonsafety-related IA syste The team verified that the containment penetrations and the piping inside containment ~were adequately protected against the effects of overpressure that could occur during LOCA containment temperature The licensee's response to NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," dated January 27, 1997, was reviewed. This response stated that the only CCW penetration requiring review for potential overpressure was the return line from the RCP thermal barrier (penetration 43), and that valve HV-3184 will unseat and relieve pressure in the penetration to a section of piping protected by a relief valve before the penetration piping allowable pressure was exceeded. The team reviewed the valve specification sheet, A-J-300-13, "CCW from RCP Thermal Barrier" and memoranda REA 96-1286 and ENG 51-1 dated January 14, 1997, which documented that the valve lift pressure was less than the piping design pressure. The team found that this documentation supported the conclusion that penetration 43 was adequately protected against overpressur '
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The team reviewed the design-basis differential pressures in the MOV Design-Basis Document, U 418109, Revision A, for the motor-operated containment isolation valves, MOV 3046, 3052, and 3182. The design-basis differential '
pressure was defined in U 418109 as the maximum differential pressure that the valve could have to open or close against. These valves were located in the piping at penetration numbers 42 and 44, which were the CCW supply to the RCPs and the CCW return from the RCP oil coolers. FSAR Table 6.2-31 identified these penetrations as Type II. FSAR Section 6.2.4.1 defined Type II penetrations as serving those lines that connect directly to the containment
. atmospher ,
The CCW lines which run through penetrations 42 and 44 could become directly connected to the containment atmosphere as a result of the dynamic effects of a LOCA inside containment. Therefore, the design-basis differential pressure for the containment isolation valves should consider the maximum post-LOCA containment pressure that could exist when the valves operate. The design-basis differential pressures in U 418109 were 24 psid for MOVs 3046 and 3182, and 9 psid for MOV 3052. The licensee stated that the maximum pressures could
'
be 52 psid for MOV 3052 and 27 psid for MOVs 3046 and 3182 when the effect of containment pressure was considere The licensee stated that the thrust setpoints for these valves were such that they could close at differential pressures of a minimum of 2.75 times greater
, than the higher design-basis differential pressures. (This resulted from the confidence band methodology used by the licensee to determine the closing thrust requirements.) The team noted that the requirements defined in Appendix B to 10 CFR Part 50, Criterion III, regarding verifying and checking the adequacy of the design were not followed for determination of the design-bases differential pressure for the CCW motor-operated containment isolation valves. This issue was identified as Unresolved Item 50-348; 50-364/97-201-13. The licensee issued ABN 97-0-1044 to revise the MOV Design-Basis Document and related documentatio I Conclusions Containment isolation of the CCW lines penetrating the containment generally met the design-basis requirements and would adequately perform its safety i function. However, the team identified that in one instance the licensee did not properly verify or check the adequacy of the design-basis differential
!
pressures used for MOV El.3.2.4 CCW Systeg Testino Inspection Scope In this portion of the mechanical design review, the team evaluated the licensee's testing of CCW system component In particular, the team's objective in conducting this evaluation was to verify that IST program included components intended to perform a safety function (as required by Section XI of the ASME B&PV Code), that the testing adequately verified the design function of the components, and that TS requirements for testing were appropriately implemente l
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. Observations and Findings .
TS 4.7.3 requires periodic verification of accessible valves in the flow paths servicing safety-related equipment that are not locked, sealed, or otherwise secured at least every 31~ days. The team reviewed surveillance procedure FNP-2-STP-23.7, " Component Cooling Water Flow Path Verification Test," Revision 12, and confirmed that it implemented the TS surveillance requirement for flow path testing. TS 4.7.3 also requires periodic verification during shutdown that each automatic valve servicing safety-related equipment actuates to its correct position on a safety injection test signal. The team reviewed surveillance procedure FNP-2-STP-40.0, " Safety Injection with Loss of Offsite Power Test," Revision 25, and confirmed that it implemented this TS surveillance requirement for CCW valve TS 4.0.5 requires testing of ASME Code Class 1, 2, and 3 pumps and valves in accordance with Section XI of the ASME B&PV Code. This code required testing of valves and pumps that perform a safety function. The team verified that i the required CCW pumps and valves were tested by a review of the following surveillance procedures:
. FNP-2-STP-23.8, " Component Cooling Water Valve Inservice Test," Revision
. FNP-2-STP-23.12 "CCW to RCP Thermal Barrier Check Valves Reverse Flow
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Test," Revision 2 l
. FNP-2-STP-23.1, "2A Component Cooling Pump Quarterly Inservice Test,"
Revision 14 (This procedure included testing of the pump 2A discharge ;
check valves and was typical for the three CCW pumps.) ,
l The licensee also performed performance testing for the CCW heat exchanger l Of these tests, the team reviewed FNP-0-ETP-4379, " Performance Test for Unit I and 2 Component Cooling Water Heat Exchangers," Revision 5, and determined I that it yielded appropriate results founded on acceptance criteria consistent with the design-basis calculation Inservice testing of the CCW pumps was performed in accordance with
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FNP-2-STP-23.1, "2A Component Cooling Water Pump Quarterly Inservice Test," 1 Revision 14, and similar procedures for the other two pumps. The procedure was performed with the system in a normal alignment which required the pump minimum flow recirculation lines be in service. Data for pump differential pressure and flow measured downstream from the recirculation line were recorded and compared to a table of acceptable differential pressures and corresponding' flows in the procedur Calculation BM-95-0776-001, "CCW System Evaluation Using Degraded CCW Pump Curve," Revision 0, documented the acceptability of the CCW system performance with the pumps degraded approximately 10% from the vendor test curves and stated that this degraded pump curve will be used for comparing test data to verify the performance capability of the CCW pumps. However, the table of acceptable performance values in procedure FNP-2-STP-23.1 was different than the degraded pump curve used in the calculatio i
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' The' team noted that the acceptance values in the procedure considered the pump ,
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recirculation lines in service and the degraded curve in' the calculation was i
for a condition with the recirculation lines closed and questioned how the ;
acceptable test values'in the' procedure accounted for the design-basis pump
degradation. The licensee stated that-a special test was performed,
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FNP-2-E1P-4416, which obtained flow and pressure data with the recirculation line closed at a single operating point. These data _were compared to the design-basis degraded pump curve. If the margin in pressure was greater than
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10%, the 90% acceptance data in the IST procedure was deemed appropriate; if less than'10%, adjustments were made to the IST acceptance data so that i would be within the design-basis degraded pump curve. (The 90% value was an ASME XI standard reference point.)- These data required adjustment for only ;
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Pumps IA and IB. The team found that the above process resulted in acceptable
' test acceptance criteria for the CCW pumps. This process was not completely '
documented.
) Calculation BM-95-0776-001 was performed under REA 95-0776. The response to j
' this REA was issued on May 22, 1996, by Bechtel letter AP-21413. A safety
' evaluation and a proposed ABN for revising the FSAR, affected FSDs,-and
affected drawings were attached to this letter. The licensee stated that the IST procedure will be revised to include the degraded curve after the safety .
l evaluation was approved by the Plant Operations Review Committee (PORC). The i licensee stated that an ABN will be issued to revise the appropriate documents l
once PORC approval has been obtained. The review of IST procedure and applicable safety evaluation was identified as Inspection Follow-up Item 50-
,
348; 50-364/97-201-14.
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' Conclusion ,
The team found that surveillance requirements in the TS for safety-related pumps and valves had been appropriately implemented by procedures and that the l t CCW heat exchangers were tested to verify the performance required by the design bases.
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E1.3.2.5 CCW System Modifications
) Inspection Scope j
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In this portion of the mechanical design review, the team evaluated 10 CCW
- - design modifications to assess their consistency with the design and licensing <
bases. In particular, this review considered the licensee's 10 CFR 50.59 l 4 evaluations, post-modification testing, and sample of the affected documents i that. required updates to reflect the given design change.
. .
. Observations and Findings .
The team determined that the 10 CFR 50.59 evaluations for the modifications reviewed were appropriate and changes to required design and licensing basis documentation were identified and implemented. The design bases for the ;
component (s) affected by the change were correctly identified. Except.for the i instance discussed below, appropriate post-modification testing was performe ; !
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Design Change Package (DCP) 96-0-9012-2-006, " Process Coating for CCW Heat !
'
, Exchangers," provided direction for modification of.the.CCW heat exchangers by i
- application of an epoxy coating (Plastocor) to the tubesheets, channel head,
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channel cover, channel head shell relief line, approximately 12 inches into .
- the service water inlet and outlet lines, and 12 inches of the inlet end of l
the tubes. 'The DCP, along with REA 96-1211, also provided direction for ;
j plugging and stabilizing tubes. Procedure FNP-0-ETP-4418, "CCW Heat-Exchanger
.
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Epoxy Coating Application," Revision 1, implemented the. epoxy coating and Work l
! Orders 96001476, 96001477, and 96001478. installed the' stabilizing-rods in the ;
tubes. .The team reviewed calculations SM-96-9012-002, " Effects of Plastocor >
Coating on CCW Hx's Thermal Performance," Revision 0, which determined an .
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equivalent length for a tube partially coated with epoxy, and SM-96-9012-004, ,
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" Effects of Plastocor Coating on CCW Hx's . Thermal Performance," Revision 0,
- and identified no concerns with these calculation The team identified the following discrepancies.with this modification: ; Commercial grade dedication of the Plastocor epoxy material included a r pull-off test of some samples at the vendor facility. The application ;
i procedure used for these coating samples was not documente Such ,
documentation would provide la method to ensure that the commercial grade
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a dedication tests were representative of the modification performed at i Farley. The licensee obtained documentation from the vendor during the i inspection that the application method used for the samples was the same ,
4 as that used for the actual work at Farley. Additionally, the licensee i stated that Plastocor personnel performed the application in both i location ,
Y J
- No post-modification testing to ensure design flow capability had been
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maintained was required by the procedure used to apply the epoxy coating
, nor in DCP 96-9012-2-006. The team noted that an earlier version of the DCP stated that post-modification flow testing would be performed and this statement was deleted in a subsequent revision to the DC Although the application method appeared to limit the epoxy to the ,
intended surfaces, unexpected / unknown occurrences could cause the epoxy
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to hinder flow through the tubes. The epoxy was applied to the first l 12 inches of each tube and this area inspected in-process for correct >
application. The work procedure did not require any inspection of the remaining portion of the tubes (tube length is approximately 29 feet),
nor any flow testin The licensee stated that the heat exchangers were returned to service with no problems indicated, including no abnormal flow indication. The team noted that reduction in flow greater than that which would be equivalent to the tube plugging limit might not be detected on installed flow instruments as the reduction could be within the instrument error band. The only post-modification tests recorded on Work Authorizations for the heat exchangers, numbers 123723, 121390, and 123724, were leak 4 tests. The team was concerned that the design-basis flow capability of the heat exchangers was not verified after the modification was completed and before the equipment was returned to servic ;
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Farley. Support Procedure GO-M-1," Designer Interface Document,"
Revision 3,. states that design input requirements are in accordance with Section 3.2 of American National Standards Institute (ANSI) Standard N45.2.11, which states that design input shall include test requirements including in-plant tests and the conditions under which they will b performed. Farley Nuclear Plant Administrative Procedure FNP-0-AP-8,
" Design Modification Control," Revision 22, Section 9.1.5, requires '
identification of tests which will be required to verify acceptable .
performance of plant components and systems added, modified or affected by incorporation of the design change into the plan ,
i In addition, the Farley Nuclear Plant Administrative Procedure
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FNP-0-AP-70, " Conduct of Operations Plant Modifications and Maintenance -
Support," Revision 4, Section 4.19, requires that the functional test procedure shall include as a minimum, sufficient information and steps to adequately test and document the proper function of the eq.dpn.ent added or modified by the design chang The design control measures for the design change did not ensure that design requirements were correctly translated into instructions procedures, or drawings and accomplished in accordance with drawings, procedures or instructions as required by Criteria III and V of Appendix B to 10 CFR Part 50. The lack of adequate post-modification testing was identified as Unresolved item 50-348;50-364/97-201-1 . The team noted several discrepancies in calculation SC-96-1211-002, "CCW Heat Exchanger Maintenance Repairs," Revision 1. This calculation documented the seismic and mechanical acceptability of the modificatio i
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Section 2.1 determined a net increase in weight of the heat exchanger of 13.85% and compared this increase to a 15% increase stated to be ,
acceptable by the vendor. However, the 13.85% incorrectly subtracted the weight of the water displaced by the modification, as the 15%
allowable limit was defined on the basis of a dry heat exchanger. The licensee stated that the correct weight increase to be used for comparison was 14.86%.
Section 2.3 evaluated the heat eichanger shell and supports and used 14.09% as the increase in weight between the tubesheets following the modification. This value should have been 14.86% as discussed abov i Section 2.4 concluded that the small increase in weight associated with the modification would not have a significant impact on the seismic response of the system, and therefore the foundation anchorage of the heat exchangers would remain ade-" ate. This conclusion was an engineering judgement and the n 0+ . ton in margin for the foundation anchorage was not documented as i,t the foundation base calculation, U-405165. The licensee reviewed the anchorage design at the request of l
the team and stated that margin existed in the anchorage with the added weight of the modification considere .
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.Section 2.5 referenced a data sheet from the plug vendor, EST,
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containing design information for the Perma Pop-a-Plugs used to plug the
- tubes. This data sheet was neither attached to the calculation nor ,
specifically identified therei None of these discrepancies changed the results of the safety analysis or conclusions of the calculation. The licensee issued REA 97-1407 to
-
revise the calculation. The review of revised calculation addressing -
.
'the above discrepancies was identified as Inspection Follow-up Item
- 50-348; 50-364/97-201-1 . Minor administrative errors were made in preparation of the DCP. The '
- Design Input Record indicated that the heat exchangers were not
"
referenced in or identified by any design-basis documents; contrary to
. the existence of several design-basis calculations concerning the heat ,
exchangers. The DCP calculation record omits calculation SM-96-9012-004
' and has the wrong title for calculation 5C-96-1211-00 ;
' Conclusions
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The modifications reviewed maintained the design and licensing bases of the 1
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system and contained appropriate safety evaluations. The modifications were /
properly implemented and documented except for the documentation, calculation, and testing discrepancies identified concerning the CCW heat exchanger
modificatio El.3.2.6 Other Related CCW System Review ' Inspection Scope i
In this portion of the mechanical design review, the team evaluated LERs, NRC Notices of Violation, CCW System SSSA observations, and licensing commitments
, for appropriate follow-up action. Licensee practices for updating drawings '
l were reviewed, as well as valve lineup procedure Observations and Findings
! The team identified no concerns with licensee actions with regard to the LERs, Notices of Violation, and licensing commitments reviewed. The licensee's safety system self-assessment identified many good issues. Follow-up actions noted in these documents were verified. The issues were resolved
satisfactorily except in one instance. This is discussed in the following
- paragraph. Although the safety system self-assessments performed by the itcensee failed to identify and correct many of the issues raised by the team, these assessments led to the.-identification and correction of many other issue Procedures FNP-2-50P-23.0A, " Component Cooling Water System," Revision 5; FNP-2-50P-2.lA.. * Chemical and Volume Control System," Revision 8; and FNP-2-SOP-
'
1.lA, " Reactor Coolant System," Revision 6, were checklists for the normal positions of valves and circuit breakers. The team identified numerous t ' differences between the P& ids h r '.he system (D-205002 Sheet 1, Revision 21;
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Sheet 2, Revision 10;-and Sheet 3, Revision 2) and procedures FNP-2-SOP-23.0A and FNP-2-50P-2.lA concerning the existence of caps on vent and drain line j
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The team noted that item 5 of-SSSA observation CCW-CM-01 was related to this '
item in that it discussed a discrepancy between FSAR Section 9.2.2.3 and the
. checklists in FNP-2-SOP-23.0 that was apparently not resolved. The SSSA was
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issued on April 19, 1990.- The licensee issued ABN 97-0-1044 to update the !
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P& ids, and stated that the SOP valve checklists which included CCW valves
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would be updated to match the drawings and that field verification would be 1
4 performed when the SOP checklists were implemented. The requirements '
specified in Criterion XVI of Appendix B to 10 CFR Part 50 which state that
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conditions adverse to quality such as malfunctions and deviations be promptly J identified and corrected were apparently not met. This issue was identified as Unresolved Item 50-348; 50-364/97-201-17.
- The team questioned if a controlled list of drawings that were maintained as- '
built existed or how such drawings were otherwise identified. The licensee
. stated that drawings which were not maintained as-built were not necessarily ;
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marked accordingly and no procedure existed to identify which drawings were as-built. The team was concerned that design changes and/or decisions could ;
use drawings which did not reflect the as-built plant. The team did not !
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j identify any issues because of the lack of this information. The licensee ,
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stated that only three categories of drawings were not as-built (condui I i
layout for Class IE structures, telephone backboard, and bills-of-material).
The licensee had identified this problem before the inspection, and had started developing a list of all currently maintained drawings as well as
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i those no longer updated for distribution to project personnel. The licensee issued REA 97-1410 to finish developing this list.
[ Conclusions Licensee follow-up actions for the LERs, Notices of Violation, and licensing j commitments reviewed were satisfactory. Weaknesses were identified in the
. control of as-built drawings and corrective action for SSSA observation -
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concerning the existence of m ps on vent and drain line j
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j El.3.3 Electrical Desian Review
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! El.3.3.1 ' CCW Electrical Loads Inspection Scope In this portion of the electrical design review, the team evaluated the
'- electrical loads required for the CCW system to perform its intended design
- functions under both normal and accident conditions. In particular, this evaluation addressed cable sizing, protective coordination, electrical bus loading, and de battery loading calculation Observations and Findings To ensure that the CCW loads had been properly identified, the team reviewed ;
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t calculations SE-94-0470-007, "Farley Unit 2 As-Built Load Study Update,"
Revision 0; SE-94-0470-005, "As-Built Load Study Summary Calculation,"
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Revision 1; E-42, " Steady State Diesel Loadings Calculation for LOSP, SI and *
SBO," Revision 8; E-35.2.A, " Setting of Protective Relays for FNP Unit 2 4.16 kV Auxiliary Power System," Revision 0; and E-95, " Battery Capacity Calculation for LOSP and LOSP+LOCA Situations and Limiting Battery Load Profile," Revision 7. All major electrical loads were accounted for within these calculations for normal and accident conditions. The frequency and
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voltage ratings of the CCW equipment were compatible with the electrical design analyses. The team observed that the methodology and assumptions used were appropriat Cable sizing for the 4 kV CCW pump motors was reviewed. The review compared the actual cable data for the CCW pump motors against the " Electric Power Cable Sizing Guide," Revision 0. The review determined that cable sizing for the CCW pump motors was adequate and in conformance with the plant cable sizing criteria. Protective coordination for the 4 kV CCW motors was reviewed and found to be in accordance with plant drawings and setting sheet Calculation E-95 was reviewed by the team to ensure that the CCW de loading was evaluated. All required loads were identified in the calculatio Conclusions The electrical design for components that perform the normal and emergency functions of the CCW system was adequate. The Electrical Distribution System provided independent, redundant, safety-related (Class IE) power to the CCW electrical load El.3.3.2 CCW Swina pumo Operation
, Inspection Scope
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In this portion of the electrical design review, the team evaluated the ability of the train B CCW pump to be aligned to either electrical redundant power supply (train A or train B) in accordance with the engineering design base ,
, Observations and Findings The team reviewed FSD A181000, " Functional System Description Component Cooling Water System," Revision 6, and logic diagrams B-205810 (sheets '
22,23,100, and 101). CCW pump motors C and A were dedicated to 4 kV Train A and Train B respectively. CCW pump B motor could be aligned either to Train A or Train B class IE 4 kV electrical buses. A review of the logic drawings referenced and a plant walkdown of the installation revealed that a cross-tonnection of the two electrical trains (A and B) could not be achieve Keylock operated disconnect switches were installed and interlocked to prevent j
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cross-connection of the two electrical train :
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. . Conclusions The team concluded that adequate physical controls as well as electrical ;
interlocks existed in the design of the swing CCW pump B.to ensure that the i
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design-basis separation of the redundant class IE Trains A and B would be maintaine ;
E1.3.3.3 Electrical Modifications
, Inspection Scope i
In this portion of the electrical design review, the team evaluated electrical modifications to the CCW system and verified that the changes were in !
conformance with the CCW and electrical design base ; Observations and Findings The following modifications associated with the CCW system were reviewed by .
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the team: ABN 96-0-0930, " Change CCW Pump Room Equipment Maximum Temperature,"
and Procurement Deviation Evaluation (PDE) 93-0-0029, " Replacement Motor for ,
CCW System."
PDE 93-0-0029 evaluated motor replacements for Limitorque MOVs Q2P17MOV3185A-and B. - The evaluation ensured that the original vendor (Limitorque) certified i that the replacement motors met the environmental, electrical, and seismic design as originally.specified. The appropriate documentation was specified :
in the purchase requisition (x253631) and the motors purchased met all design criteria of the original motor ABN 96-0-0930 updated FSDs A-181000 and A-181004 to reflect increased room temperatures when various room coolers were out of service for maintenanc REA 95-0873 documented the results of an engineering evaluation which allowed temperature increases for the motor control center (MCC) room for MCC IB and j 2A, battery charger A, B, and C rooms, and the CCW pump room. The team ;
i reviewed REA 95-0873 and verified that adequate justification existed for the :
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equipment to operate in these elevated temperatures during normal and accident t conditions. Documentation was available to support the engineering evaluation. A 10 CFR 50.59 evaluation was performed which identified changes to the FSAR. The appropriate sections of the FSDs incorporated ABN 96-0-0930.
i ! Conclusions
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The electrical modifications reviewed by the team for CCW were adequate. The !
design bases as delineated within FSDs and the FSAR were maintained, j
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El.3.4 Instrumentation & Controls (I&C) Desian Review E1.3.4.1 CCW Surae Tank level Setooint Inspection Scope In this portion of the I&C design review of the CCW system, the team evaluated calculation CS-L3027C, " Instrument Loop Uncertainty and Setpoint Calculation,"
Revision 0. In particular, the team assessed the adequacy and consistency in relation to the setpoint program, GO-M- Observations and Findings This uncertainty calculation was performed for Unit I to evaluate the existing low-low level setpoint of 20 inches above the tank bottom. The results also applied to Unit 2. The calculation evaluated CCW surge tank low-low level '
actuation requirements and verified appropriate stroke times for CCW isolation valves Q2Pl7HV-3096A and The evaluation was required to ensure adequate NPSH for the on-service train CCW pump without a need for make-up water being supplied to the CCW surge tank. The team.found the calculation adequate and consistent with Farley's setpoint uncertainty progra Conclusions The team determined that low-low setpoint in this calculation was correct and the methodology of the calculation was consistent with Farley's setpoint progra El.3.4.2 Instrument Looo Uncertainty Calculations Inspection Scope In this portion of the I&C design review, the team evaluated the consistency and adequacy of the licensee's setpoint bases and uncertainty calculations for various instrument loops in the CCW system, Observations and Findings The team observed that nearly all of the CCW instrument loops lacked calculations documenting uncertainty. The lack of calculations made it difficult to judge w' nether the methodology that went into the development of the original setpoints conformed to the current setpoint program. The licensee developed three preliminary uncertainty calculations to show that original setpoints and margins demonstrated the conservatism included in the design of the Farley system Loop F3045, "CCW Return Flow from the RCP Thermal Barrier" (high-flow setpoint), calculation documented that there was adequate margin in the setpoint considering the loop uncertainties. The licensee also noted that the i 15 gpm activation range (the setpoint tolerance) could be reduced to conform to a more conservative calibration tolerance consistent with the capabilities of the instrument. This reduction would have a slight effect on the
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i percentage of flow equation on page 3 of the calculation, but would have minimal affect on the total error. The slight effect on the flow equation 1 would yield more margin. The licensee stated that the final version of
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Loop F3045 would correct these deficiencie )
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Loop L3027. "CCW surge tank level setpoints" calculation documented that the r low and low-low. level CCW surge tank setpoints would perform their intende '
function considering the loop uncertainties by alarming and providing actuating interlocks to Q2Pl7HV-3096A/B. Loop P3184, "CCW return pressure l
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from RCP thermal barrier" (high-pressure setpoint), calculation documented '
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that the pressure switch was set to maintain maximum sensitivity to a thermal barrier cooling coil failure considering the loop uncertainties, but high .
enough to prevent spurious actuation as a consequence of normal. proces !
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pressure changes. The team had no concerns with these two calculations.
4 Conclusions t The calculations proved to be adequate and consistent with Farley's current
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.setpoint program.- The methodology was consistent with the I&C design criteria i and industry standards. The calculations had built-in conservatism and margin particularly with the associated surveillance test procedure data sheet .
Future setpoint changes can be adequately calculated and process limits can be :
' reconstituted and used for these changes using the current setpoint progra .
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El.3.4.3 Setooint Indexes and Other Reviews
, Inspection Scope I In this portion of the I&C design review, the team evaluated the consistency l
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l of setpoint index B-205968 and the field instrument.setpoint index from the
. Westinghouse balance of plant (BOP) manual (Farley drawings U-262166 and U-262167). In addition, the team also evaluated licensee's compliance with
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Regulatory Guide (RG) 1.97, " Instrumentation for light-water-cooled nuclear
- power plants to assess plant and environs conditions during and following an l accident," Revision 2.
- Observations and Findings The team observed that there were discrepancies in setpoint data between the
- setpoint index and the Westinghouse BOP manual field instrument index found on
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drawings U-262166 and U-262167 Revision 3. The team found that the column
! titled "setpoint unit" for level instrument LSLL 3027C on drawing U-262166
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noted 27 inches. Similarly for LSLL 3027D the "setpoint unit" was 27 inches on
, drawing U-262167. Both of these setpoints should be 20 inches, as stated in the setpoint index.. The team found that the low-low setpoint of 20 inches had been used correctly in the surge tank level and calibration range calculation
(CS-L3027C). Farley-procedures FNP-2-IMP-210.6, Revision 8, " Component Cooling Water Surge Tank Level Loop calibration" Q2P17LT3027C and FNP-2-IMP-210.7, Revision 8, " Component Cooling Water Surge Tank Level loop Calibration" Q2Pl7LT3027D were also reviewed for consistency. Each associated data sheet !
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in each maintenance procedure correctly noted 20 inches tank level. The licensee issued ABN 97-0-1044.to correct the Westinghouse manual.
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The team reviewed th'e RG 1.97 Category 1 compliance checklists for the CCW heat exchanger inlet flow instruments, FT 3043A, B, and C and the component cooling water heat exchanger discharge temperature instruments, TE 3042A, TE i
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3042B, and TE 3042 RG 1.97 stated that the indication for the CCW flow was considered a Category 2, type D variable with a required scale range of 0 to
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110% of the design flow. The team verified that the scales for the CCW flow :
L indicators met this requirement. The range of the temperature elements (32*F .
to 200*F) was sufficiently sensitive in the normal operating range. The team
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found the checklists and instrument design were consistent with RG 1.97 guidanc I Conclusions The team found that the setpoint indexes were consistent and in accordance with the design basis except for one minor discrepancy which did not adversely
- affect safe plant operation. The-post accident monitoring instrumentation met the RG.I.97 guidanc .
El.3.5 System Interface Desian Review El.3. Service Water (SW) System
) Inspection Scope
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In this portion of the system interface design review, the tear,. cvaluated the ability of the SW system to supply cooling water to the CCW heat exchanger In particular,'the team reviewed FSAR Section 9.2.1, "Statior Cooling Water System," and Section 9.2.5, " Ultimate Heat Sink"; the P& ids; portions of FNP-2-SOP-24.0, " Service Water System," Revision 31; and calculation SM-ES-89-1500-007, " Bounding. Service Water Inlet Temperature Profile," Revision Observations and Findings The team found the documents reviewed satisfactory and did not identify any concerns with the interface between the SW and CCW systems. The SW system
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design for single failure was consistent with that of the CCW system and SW design temperatures and flow rates were appropriately utilized in the CCW design, Conclusions
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The team concluded that the design of the interface between the SW and CCW systems was satisfactory and that the SW system operation adequately supported CCW system operatio El.3.5.2 Instrument t.ir (IA) System Inspection Scope
! In this portion of the design review, the team evaluated the interfaces :
between the IA system and the CCW air-operated valves. In particular, the
, team reviewed FSAR Section 9.3.1, " Compressed Air System," data sheets for the
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CCW air-operated valves from control valve specification SS-1102-36, and FNP-2-A0P-6, " Loss of Instrument Air," Revision 1 l l Observations and Findings
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l The IA system was not safety-related and the team determined that loss of the system would not prevent the CCW system and its components from performing- l their safety functions. The team did not identify any concerns with the l interface between the IA and CCW systems,
! Conclusions 1 The team concluded that the design of the interface between the IA and CCW systems was satisfactory.
E1.3.5.3 Auxiliary Buildina Ventilation (ABV) System Inspection Scope In this portion of the system interface design review, the team evaluated the ability of the ABV system to provide acceptable ambient conditions for )
operation of the CCW system. In particular, the team reviewed FSAR Section 9.4.2, " Auxiliary Building," and portions of FNP-2-SOP-58.0, " Auxiliary Building HVAC System," Revision 2 Observations and Findings The room containing the CCW pumps and heat exchangers was served by the nonsafety-related non-radioactive heating and ventilation system and redundant safety-related coolers cooled by service water. REA-0873, " Auxiliary Building Room Coolers Attendant Equipment Evaluation," documented that the CCW equipment in the CCW room could operate if one cooler was out of service in a design-basis accident situation where a single failure causes loss of the other cooler. The team did not identify any concerns with the interface between the ABV and CCW system Conclusions The team concluded that the design of the interface between the ABV and CCW systems was satisfactor E1.3.6 Systfm Walkdowa
- Inspection Scope The system walkdown included examination of the CCW pumps, heat exchangers, and other CCW equipment in the CCW room; the CCW surge tank; and the CCW valves and piping in the contairment penetration area outside containmen The team also verified the consistency of selected portions of the system with plant drawings, as well as the consistency of the instrument calibration intervals with licensing documents. In addition, the team verified the as-built designs and identified RG 1.97 instrument .-. . - . - . - - . . - -. . . .- - .
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Observations and Findings ;
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The team determined that the overall material condition of the plant areas was good. The equipment sampled matched the design requirements. The team .
' verified that instrumentation installation met the guidance provided in RG 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess :
Plant and Environs Conditions During and Following an Accident," Revision ,
i The team observed that the calibration frequency of the SW pressure i transmitter Q1P16PT30018 had a 3-year calibration interval. This calibration i interval appeared to be excessive. The team was concerned that an instrument calibrated every three years could drift out-of-tolerance. An extended :
calibration interval usually would require a supporting analysis of drift
! error. Since the licensee could not produce an uncertainty calculation for :
the instrument loop PT 3001B, the team questioned the current calibrat. ion :
i interval. The team found that the" M-85" maintenance program was tracking ,
- this device. The program flagged any device if any two of the preceding four calibration intervals exhibited out-of-tolerance conditions. There had been ;
- one calibration performed on Q1P16PT3001B since the "M-85" program was
. implemented and one calibration before that recorded an out-of-tolerance as- !
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found calibration. The licensee issued ABN 97-0-1410 to implement a revised i
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calibration interval.
- Conclusions I The team determined that. generally the CCW system design observed during the walkdown was consistent with the design-basis requirements.
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- !.3.7 FSAR-and FSD Review Inspection Scope
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In this portion of the design review, the team evaluated the FSAR sections and FSDs related to the CCW system and the associated electrical and I&C systems.
, Observations and Findings
The team identified the following discrepancies in the FSAR:
i Table 9.4-6A listed the room temperature for the Component Cooling Pump j Room at the beginning of the post-accident period as 119'F whereas
- Table 3.11-1 indicated a design temperature of 104*F for the same room.
- CCW relief valves Q2P17V153, V154, V155, and V158 were not listed on Table 6.2-39 as part of the containment isolation boundar '. Table 9.3-1 did not include valve HV 2229, which was also a safety-related, air-operated valve that received a safety injection actuation signal (SIAS).
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- Several differences existed between FSAR Tables 9.2-6 and 9.2.7 and Tables T-1 through T-5 in the CCW FSD. For example, FSAR Table 9.2-6 listed the charging pump lube oil cooler flow as 20 gpm and FSD Table T-2 lists this flow as 30 gp The 7.bove discrepancies had not been corrected and the FSAR updated to ensure that the information included in the FSAR contained the latest material as
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required by 10 CFR 50.71(e). This issue was identified as Unresolved Item 50-348; 50-364/97-201-18. The licensee issued ABN 97-0-1044 and REA 97-1410 to revise the FSA The team also identified the following discrepancies in the CCW FSD A-181000: Table T-1 stated the charging pump lube oil cooler total heat load as'
O.036x10E6 instead of 0.096x10E . Sections 2.1.1.1 and 2.1.1.2 did not contain the current system flow rates and heat loads as documented in calculations 37.4, 37.7, and 3 . The results of calculation BM-96-1211-002, which evaluated the system for operation with degraded CCW pumps, were not include The licensee issued ABN 97-0-1044 to correct the FS Conclusions The team concluded that in several instances the FSAR was not revised in accordance with 10 CFR 50.71(e) requirements. Discrepancies were also identified during the FSD review. The weaknesses mentioned above either have been, or Are currently being evaluated by the license El.4 Other Related Electrical Systems Review E1.4.1 AC System Inspection Scope In this portion of the design review, tne team evaluated the emergency diesel generator (EDG) loading calculation, AC system voltage calculations, swing diesel generator operation, and degraded grid relay settings. In particular, the team focused this evaluation on assessing the consistency between the selected systems and the AC system design bases, Observations and Findings The team reviewed calculation E-42, " Steady-State Diesel Generator Loading Calculation for LOSP, SI, and SBO," Revision 8, to verify that the analysis was consistent with the design basis information of the FSAR and FSD A181004,
" Electrical Distribution System," Revision 10, and A-181005, " Diesel Generator System," Revision 10. Calculation E-42 had been maintained current with the f ant installed configuration. Various plant modifications which affected
.lculation E-42 were documented within the calculation and within Table 8.3-1
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of the FSAR. Additionally, calculation E-42 (sheet 4) identified diesel -
generator operating restrictions when lightly loaded (less than 30% load) and that certain motor control center (MCC 1F and 1K) alignments were require The licensee demonstrated that this information was incorporated in plant procedures, FNP-1-EEP-1, FNP-1-ESP-0.1, FNP-1-SOP-36.3, and FNP-2-SOP-3 AC system calculations- SE-94-0470-001, "Farley Unit 1 As-built Load Study Update," Revision 1; SE-94-0470-004 ,"As-Built Load Study Summary calculation," Revision 1; SE-94-0470-007, "Farley Unit 2 As-built Load Study Update," Revision 0; and SE-94-0470-005, "As-built Load Study Summary Calculation," Revision 1, were reviewed by the team. Methodology, assumptions, and inputs were consistent with design criteria. as-built data, and design bases information. Adequate AC voltage was avai',able during normal, maintenance / refueling, and LOCA conditions for those electrical loads reviewe The diesel generator system consisted of five diesel generators (DGs)
supporting both Units 1 and 2. Normal assignment of the diesels was: DG 2A (4075 kW) and DG IB (4075 kW) to Unit 1; DG IC '2850 kW) and DG 28 (4075 kW)to
' Unit 2; and SB0 DG 2C (2850 kW) served both Units. DG 2A and DG IC would be automatically aligned to either Unit 1 or 2 depending on the accident condition. The team reviewed the diesel generator logic and verified that proper diesel generator alignment would be achieved during normal and accident conditions. It was also verified that this operation was consistent with the design bases of the FSAR, the TS, and FSD A-18100 The team reviewed the design bases of the degraded grid protection, as stated in FSAR Section 8.1.1. The team verified that the design bases were conveyed into the engineering design documents. During the plant walkdown, the team verified that the relays were set as stated on the relay setting sheet Conclusions The team concluded that adequate AC supply was available for both normal and accident conditions. The electrical loading of the individual components had been considered in the diesel generatt.e calculations. The AC system design documents were consistent with the derign base l l
El.4.2 125 Vdc Battery Systems a.- Inspection 3 cope In this portion of the design review, the team evaluated the safety-related Class IE 125 Vdc Ausiliary Building battery system. In particular the purpose of this evaluation was to verify that the engineering design and installation were consistent with the relevant design bases and industry standard Observations and Findings The team determined that the 125 Vdc Auxiliary Building battery system described within FSD A-181004, " Electrical Distribution System," Revision 10, and alectrical calculations E-95, " Battery Capacity Calculation for LOSP and
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LOSP+LOCA Situations and Limitina Battery Load Profile," Revision 7; E-144,
" Determination of Battery Capacity Margins for Adequacy of Voltage at Safety-
Related Components'for Various Load Profiles," Revision 4; and E-116 " Minimum d Available DC Voltage and Permissible Control Circuit Lengths for Limiting
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Battery Load Profile," Revision 4;.was consistent with the FSAR. The assumptions and methodology used in calculations E-95,116,. and 144 were in agreement with IEEE 5tandard 485-1983,"1EEE Recommended Practice for Sizing l Large Lead Storage Batteries for Generating Stations'and Substations."
The team noted that FSAR Section 8.3.2 had been extensively revised as a result of the' previous NRC inspection > findings questioning the adequacy of l
f; Auxiliary Building battery duty cycle and. voltage requirements. As a result, the FSAR clarified the design-basis requirement for Auxiliary Building battery ..
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, as a duty cycle of 1 minute with a load profile of 500 r.mperes for LOSP or LOSP+LOCA and a 2-hour duty cycle with a load profile of 350 amperes for LOSP or LOSP+LOCA assuming a battery charger failure. The analytical basis for the :
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change was contained in calculation E- 144, " Determination of Battery Capacity '
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Margins for Adequacy of Voltage at Safety-Related Components for Various Load '
Profiles." The team noted that the 1 minute duty cycle, requirement was not identified in TS Section 4.8.2.3.2. ,
j The team reviewed technical specifications (TS) Section 4.8.2.3.2.c.5 and .
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procedure FNP-1(2)-STP-905.1, " Auxiliary Building Battery Service Test," ,
Revisions 5-10, and noted that the TS criterion of 1.75 Vdc per cell (terminal ;
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- voltage 105 Vdc) for Auxiliary Building batteries IA, IB, 2A, 2B was not in agreement with calculation E-144, Revision 4, which required battery terminal i
. voltage greater than 105 Vdc for the battery load profile specified in !
p calculation E-144 and FSAR Section 8.3.2.1.1.1.2. Specifically, the new i
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voltage requirements in accordance with the above calculation were 108.31, 112.43, 113.34, 112.25 for.first minute and 108.16,110.75,107.80 and 109.35 i
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for the end of 2-hour for batteries lA, IB, 2A and 2B, respectively. The
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individual cell voltage will be calculated by dividing the battery terminal ,
voltage by number of cells ~(60 cells).
Production Change Notice (PCN) B-92-0-8099, Revision 0-2, incorporated the i battery duty cycle and load profile changes in FSAR Section 8.3.2 and !
calculation E-144, Revisions 0, I and 2, determined the required battery l terminal voltages. The required battery terminal voltages at the end of one j
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minute in accordance with the above calculations were: Battery 1A, 112 Vdc; t 18, 114 Vde; and 2A and 28, 112.8 Vdc. The required terminal voltage at the
! end of two hours was 110 Vdc for all four batteries. 10 CFR 50.59 evaluations performed for PCN B-92-0-8099 and the changes to FNP-1(2)-STP-905.1 stated F that TS Section 4.8.2.3.2.c.5 was not affected.
Subsequent revisions of the calculation (Revisions 3 and 4) also did not identify the need to change the TS. The current revision of the surveillance procedure (FNP-1(2)-STP-905.1) showed the acceptance values (design requirement from calculation E144) as " Engineering Acceptance Criteria" and not as a TS requirement. The team determined that the surveillance requirement specified in TS was less conservative (allowed battery cell voltages to decrease to 1.75 Vdc which is equivalent to 105 Vdc battery
- terminal voltige) and would not have met the design requirement for supplying
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adequate voltage for all safety-related components. FSAR Section 8.3.2 states that batteries are designed to provide adequate voltages required for safety-related components during ' normal and accident conditions. FSAR Section 8. also states that s. service test of each battery be performed on the load profiles listed in 8.3.2.1.1.1.2 during each refueling outage or at intervals of 18 month CFR 50.59 rcquires prior Commission approval for any proposed change, test, or experiment that involves a change in the te'chnical specifications incorporated in the license. 10 CFR 50.59 also requires that the licensee shall submit an application for license amendment pursuant to 10 CFR 50.90 for a change in. technical specifications. The team noted that safety evaluations for Auxiliary Building battery service test procedure FNP-1(2)-STP.905.1 and PCN B-92-0-8099 including calculation 07597-E144 for FSAR revision failed to identify the required change for the technical specifications (TS) Section 4.8.2.3.2.c.5. The licensee also did not submit the application for license amendment pursuant to 10 CFR 50.90. This issue was identified as Unresolved Item 50-348; 50-364/97-201-19. The licensee issued ABN 97-0-1043 to revise the TS and the procedures.
. The team reviewed the latest battery capacity testing and service testing results available during the inspection for Auxiliary Building batteries lA, IB. 2A, 28 and determined that the batteries met the requirements of IEEE 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Storage Batteries for Generating Stations and Substations," and calculation E-144 for battery terminal voltage Operator training manual OPS-40204E/52103C, "DC Distribution," dated May 1996, was reviewed by the team to verify conformance to the design bases. Two
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discrepancies were identified. The Auxiliary Building battery room
.high-temperature alarm setpoints on page T-2d for Temperature Switches 3884A/B were higher than the allowable design criteria of Il0*F, and page 10 indicated that the Auxiliary Building batteries were sized with an end voltage of 105 Vdc (1.75 Vdc/ cell), which was not in conformance with engineering calculation E-144. The licensee stated that pen-and-ink changes had been noted in the training manual text used by the instructors as an interim change and issued ABN 97-0-1043 to revise the training manual, Conclusions The team concluded that the design / licensing bases for the Auxiliary Building batteries had been maintained except for voltage requirements for TS battery service test. Calculations were adequate and their assumptions and methodology were consistent with industry standard ... -- - . . - - . . - - . . - - .-. - - - _ - .
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- E1.4.3 . Modifications
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l Inspection Scope '
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In this portion of the design review, the team evaluated two' electrical
' modifications to verify that the design bases and design configuration were !
being maintained and that the associated-10 CFR 50.59 safety evaluations were !
,
- adequate.
2 Observations and Findings The team reviewed PCN-B-92-2-8068, which provided the design for the replacement.of the Unit 2 auxiliary batteries 2A and 28. The design change modification replaced the. existing "GNB" batteries with "C&D" type LCU-27 batteries.. Engineering evaluations were performed to substantiate that no changes to the FSAR were required. The following calculations were reviewed i
"
"c; consistency with the modification: E-95, " Battery Capacity Calculation for l L W tnd LOSP+LOCA Situations and Limiting Battery Load Profile," Revision 7; i E-1M, " Minimum Available DC Voltage and Permissible Control Circuit lengths for Limiting Battery Load Profile," Revision 4; E-144, " Determination of !
Battery Capacity Margins for Adequacy of Voltage at Safety-Related Components
- for Various Load Profiles," Revision 4; and E-42, " Steady State Diesel ,
l
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All calculations except for_E-42 were adequately updated for the battery '
i replacement. The discrepancy in calculation E-42 is discussed in Section E of this report. The team confirmed that combustible loading and seismic 1 analyses were reviewed by the licensee for any impacts. The review indicated
.
i
- that both the overall battery weight and combustible material was less than ,
'
the original installatio !
t l The team also reviewed DCP 95-0-8853-0-001. This modification provided the i design for replacement of the Westinghouse Type SV-1 undervoltage relays in i the emergency DG relay panels. The modification i antified a relay l replacement which had operating characteristics (operating speed, deadband, 1
- contact capacity, burden) similar to the SV-1 relay. The licensee ensured j l that the seismic capability of the new relay was adequate. No concerns were identified.
<
No concerns were identified with the 10 CFR 50.59 evaluations.
' Conclusions n The electrical modifications reviewed by the team were designed in accordance with existing design bases, necessary documents were identified for update,
- and installations completed.
,
. mr- . , - - - - - =u-= m.,~w- w wr w . . - . - .e 3,- -,,.y -1-vy .r * - - - " + 1 er
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El.4.4 System Walkdown Inspection Scope The team performed a walkdown of the Auxiliary Building batteries (IA, 1B 2A, 2B); the Unit 2 CCW pump area; the Unit 1 AFW area; and the Class IE battery charger and UPS area, emergency switchgear rooms, and DG rooms for Units 1 and 2. In particular, the intent of the walkdown was to verify that the electrical installations conformed to the relevant engineering design criteria and the applicable drawing Observations and Findings All auxiliary building batteries were installed as shown on the drawing Auxiliary batteries 2A and 2B reflected the design configuration of PCN-B-92-2-806 Battery room temperatures were within the design range (60*F-110'F), the areas were clean of debris, the battery racks were clean of battery acid, and the individual cells were clean of dust and battery aci The Unit IA battery exhibited slight corrosion on various intercell connectors. The licensee stated that maintenance would correct this
,
conditio The team verified that the electrical installations were in accordance with the design criteria of A-172389, " Cable Working Specifications," Revision 91; A-177541, " Tray and Conduit Details and Notes," Revision 8; and A-177538,
" Electrical M eral Details & Notes," Revision 2 Electrical separation requirements were reviewed to ensure the installations complied with FSAR Section 3A (which discusses conformance with RG 1.75) and various installations involving non-1E loads powered from class IE sources were observed within the DG building. All installations complied with the engineering design criteri The team verified that 4.16 kV breaker cubicles (DF-03, 04, 05 and DG-04,05,10) and MCC cubicles FA-H4 (MCC 1A) and FA-14 (MCC 2A) were installed and configured per the appropriate engir.eering documents. The team '
noted that the calibration due date on the 50G relay on cubicle 1-DF05 was past due. A review of the Preventive Maintenance Task Planning System indicated that the 50G relay actually required calibration by November 25, 1997. The test frequency of the relay had been changed since its last calibration, in accordance with recommendations from the plant's Reliability Centered Maintenance (RCM) Progra The licensee had also taken precautionary measures to remove any 4.16 kV
'
breakers, which were not in the connected position, from the switchgear while evaluating the seismic adequacy of unconnected breakers in the switchgea The team identified housekeeping issues such as debris in the cable trays, unsecured ladders, fans and carts and cables outside the raceways. This was brought to the licensee's attention. The licensee took prompt measures to correct the O
.
Various fire barrier penetration seals were observed. Penetration 45-121-26 was located in the wall between the Unit I hot shutdown panel area (Fire Area 12) and an adjacent cable chase (Fire Area 13) and was sealed with silicone foam. The penetration contained electrical raceways / cables and copper tubin The copper tubing was installed under Change Notice No. BM-3095 (dated August II,1980). The design change provided additional cooling to Room #254 so as to maintain a temperature of less than 80*F. The copper tubing and electrical ,
raceway / cables were installed through the penetration in support of the room '
cooling modificatio A review of the design change did not reveal any engineering evaluation of ,
fire barrier 45-121-26 for the modification. The team questioned the '
qualification of penetration 45-121-26 with copper tubing. In response, the licensee provided Factory Mutual Test Report 27390 as the test results for silicone foam qualification for fire barrier penetration The actual configuration of copper tubing was not tested in the Factory Mutual Repor Change Notice No. BM-3095 did not indicate if penetration 45-121-26 was previously. reviewed for the installation of copper tubing. The licensee's preliminary evaluation showed that there were no concerns with the qualification of this penetration seal. The licensee stated that they used engineering judgement during the initial installation to ensure that the above test enveloped the configuration of the penetration. The licensee also stated that a calculation would be developed to ensure that the copper piping installation would not degrade during a fire and breach the fire barrier.
,
FSAR Section 98.2.2.5.3 stated that for the fire barrier penetrations that were not in the as-designed condition, an evaluation was required to establish its qualification. At end of the, inspection, an evaluation was not available for team's review. The licensee issued REA 97-1407 to resolve this ite Review of testing documentation or analysis to show.that as-built configuration matched the tested configuration for fire penetration seal was identified as Unresolved Item 50-348/201-2 Conclusions In general, the installation of cables, raceways, and equipment was in conformance with design documents. However, in one instance the licensee did not have adequate documentation to show that as-built configuration matched the tested configuration for fire penetration seal. The material condition was generally adequate, but the team identified the need for improvement in housekeeping are E1.4.5 FSAR and FSD Review Inspection Scope In this portion of the design review, the team evaluated the FSAR sections and FSDs related to the electrical system _
,
I t
. Observations and Findings Th'e team identified the following discrepancies in the FSAR: Section 8.3.1.1.3.A.2 stated that the unit auxiliary transformer "B" megavolt-ampere (MVA) rating at 65'F was 47.99 instead of 46.7 as shown on drawing D-20270 . Section 8.3.1.1.9B referred to Section 8.3.1.1.3 for interrupting capacities for distribution panels. However, Section 8.3.1.1.3 did not include interrupting capacity data for distribution panel . Section 8.3.1.2 stated that there were twenty-one 600-V/208-V motor control centers; however, the actual number of motor control centers identified is ninetee The licensee issued ABN 97-0-1043, REA 97-1410, and ABN 97-0-1054 to correct '
these discrepancies. The above discrepancies had not been corrected and the FSAR updated to ensure that the information included in the FSAR contained the latest material as required by 10 CFR 50.71(e). This issue was identified as i Unresolved Item 50-348; 50-364/97-201-2 The team identified the following discrepancies in the DG System FSD, A-181005 Revision 10:
l Section A.4.1.3 stated that diesel generator 1C was only lightly loaded 4 (6% of its continuous rating) in all four scenarios, but it did not agree with calculation E-42, Revision 8. The calculation stated that the load was only 39 kW (less than 1.5%). Section 5.10.8.7 stated that the DG lube oil heaters were powered from 120/208V MCC distribution panels instead of the DG 600-120/208V auxiliary transformer . Open items DG-FSD-006, 018, 019, and 024 resulted from the licensee's review of the FSD were not incorporated in the FSD. These open items dealt with system operating information concerning temperature and pressure for the diesel generator intercooler and lubricating oil system component _
l The licensee issued ABN 97-0-1053 to revise the FS Conclusions The team concluded that in several instances the FSAR was not revised in accordance with 10 CFR 50.71(e) requirements. Discrepancies were also l identified during the FSD review. The weaknesses mentioned above either have l been, or are currently being evaluated by the license )
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O
El.5 Control of Calculations Inspection Scope ,
In this portion of the design review, the team evaluated numerous engineering calculations related to the AFW and CCW systems, as discussed in other sections of this repor In particular, the team reviewed the design control aspects associated with these calculation ' Observations and Findings The team determined that in several cases the calculations that had previously been superseded were not identified as such on the calculation index, and design-basis calculations were not appropriately revised to show the existing design condition. In addition, the licensee did not always revise affected calculations when new calculations were performed. The following examples represent these weaknesses:
.
Calculation 38.06, " Determine Flow rate Through Pipe Break in CCW System," Revision 0, concerning the CCW surge tank low-low level setpoint, superseded calculation 35.5, " Evaluation of CCW Surge Tank Level Setpoints," Revision 0, yet calculation 35.5 was shown as active on the inde .
Calculation 34.5, " Component Cooling Water System NPSH (ES No. 90-1820)," Revision 0, was affected by calculation 39.3, "CCW Surge Tank Analytical Limit for Level Setpoint," Revision 0, as discussed in Section El.3.2.lb of this report, but calculation 34.5 was not revised accordingl .
Calculation 25.3, "0verpressurization of AFW Piping During Overspeed Testing," Revision 0, determined that an unacceptable piping pressure could develop in the event of turbine overspeed of 125% of rated spee This calculation was not revised to reflect the revised overspeed setpoint of 115% of rated speed nor was a calculation prepared to supersede calculation 25.3. The design bases for acceptability of the overspeed setpoint were documented in modification PCN B-88-1-5003,
" Change Overspeed Trip Setting for TDAFW Pump," Revision .
Calculation E-144, " Determination of Battery Capacity Margins for
- Adequacy of Voltage at Safety-Related Components for Various Load Profiles," Revision 2, reflected a new battery installation but calculation E-42, " Steady-State DG Loading Calculation for LOSP, SI and 580," Revision 8, which used calculation E-144 as an input, was not revised accordingl .
AFW flow calculations 40.02, " Verification of AFW Flow Bases," Revision 3 (Unit 1), and the equivalent calculation for Unit 2 (38.04) provided design-basis information for the AFW system; yet calculation 35.04,
" Auxiliary Feedwater System Head Curves (ES 90-1831)," Revision 0, which provided similar information, was not annotated to indicate it did not contain design-basis informatio ____-_----------____:
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Calculation SC-96-1211-002, "CCW Heat Exchanger Maintenance Repairs," '
- Revision 1, stated that it was judged that the increase in weight- !
resulting from the CCW heat exchanger modifications.would not affect the j acceptability of the foundation anchorages. However, the calculation
,
J for the anchorages was.not revised nor was the increase in heat (
'
L exchanger weight reflected in the appropriate calculatio i The licensee stated that the philosophy used for calculations before the J
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'
implementation of procedure NEP 4-4, " Preparing and Reviewing Calculations,"
'in 1996 was that calculations were performed to evaluate a potential design ' ! parameter on an "as-designed" basis. The licensee also stated that subsequent .
j design decisions that were different but substantiated by a particular calculation may not have been updated into the calculation but should be found l l
.
documented in letters or other documentatio .
2 The team did not identify any systematic method of correlating these letters ;
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or other documentation with the affected calculation. Therefore, the team was
L concerned that existing design-basis calculations could be superseded or not i consistent with the as-built plant and these calculations used inappropriately l'
in subsequent analyses and design decisions. The team did not identify any
- instances where the use of calculations that were not consistent with the as- !
I built plant resulted in an adverse effect on safety. Adherence to current l
,
calculation procedures and practices would have prevented most of the above ;
deficiencies. The team noted that the new calculations were controlled i
- properly, but the licensee did not take any action to correct deficiencies 1
with existing calculation l
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.
The licensee issued REA 97-1407 to appropriately correct the above ,
l
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deficiencies in calculation and design control. The team determined that the licensee's design control measures did not meet the requirements specified in )
- Criterion III of Appendix B to 10 CFR Part 50 and Farley Support Procedure GO-
' M-1, " Designer Interface Document," to ensure that plant design-basis ,
'
documentatior'is maintained current. This issue was identified as Unresolved l l- Item 50-348; 50-364/97-201-22.
L
' j Conclusions
! The team concluded that control of design-basis calculations was weak before J the licensee implemented Procedure NEP 4-4.
- Manacement Meetina l
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l XI Exit Meeting Summary On_ March 14, 1997, the team members conducted a pre-exit meeting with the l
. licensee. On March 26, 1997, the team leader conducted a final public exit i
,
meeting, during which the team's overall conclusions and inspection findings l were presented. The licensee did not identify as proprietary any information !
provided to, or reviewed by, the team. Upon conclusion of the exit meeting, :
the NRC team leader and others answered questions from local media l representative ]
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APPEND 1X A i
i Open Items j This report categorizes the inspection findings as unresolved items and inspection follow-up items in accordance with the NRC Inspection Manual, Manual Chapter 0610. An unresolved item (URI) is a matter about which mere information is required to determine whether the issue in question is an acceptable item, a deviation, a nonconformance, or a violation. The NRC Region II office will issue any enforcement action resulting from their review i of the identified unresolved items. An inspection follow-up item (IFI) is a i matter that requires further inspection because of a potential problem, ,
l because specific licensee or NRC action is pending, or because additional i information is needed that was not available at the time of the inspectio Item Number _ Findina Title LYE 50-348;50-364/97-201-01 URI Unprotected CST Connections l (Section E1.2.2.lb2)
50-348;50-364/97-201-02 URI Tornado Protection of CST Level ;
Instrumentation (Section E1.2.2.lb3) ,
!
50-348;50-364/97-201-03 URI AFW Check Valve Reverse Flow Testing (Section E1.2.2.4b)
50-348;50-364/97-201-04 URI AFW Check Valve forward Flow Testing !
(Section E1.2.2,4b)
50-348;50-364/97-201-05 URI TDAFW Battery Testing (Section E1.2.3.lb)
50-364/97-201-06 URI TDAFW Battery Installation (Section l E1.2.3.lb) l 50-348/97-201-07 IFI CST Level Alarm (Section E1.2.4.lb)
50-348;50-364/97-201-08 URI Tornado Protection of TDAFW Pump Vent Stack (Section E1.2.661)
50-348;50-364/97-201-09 URI Tornado Missile Spectra (Section E1.2.6b2) l 50-348/97-201-10 IFI CST Level Transmitter Freeze Protection (Section E1.2.6b4)
50-348;50-364/97-201-11 URI AFW FSAR Discrepancies (Section El.2.7b) l 50-348;50-364/97-201-12 URI Stress Analysis Temperature (Section ,
E1.3.2.lb5)
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t 50-348;50-364/97-201-13 URI MOV Design-Basis Differential Pressure :
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(Section El.3.2.3b)
50-348;50-364/97-201-14 IFI CCW Pump Testing (Section El.3.2.4b)
50-348;50-364/97-201-15 URI Post Modification Testing (Section ,
El.3.2.5b)
50-348;50-364/97-201-16 IFI Calculation Discrepancies (Section
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El.3.2.5b)
50-348;50-364/97-201-17 URI Drawing and Procedure Discrepancies
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(Section E1.3.2.6b)
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50-348;50-364/97-201-18 URI CCW FSAR Discrepancies (Section El.3.7b)
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50-348;50-364/97-201-19 URI TS change for Auxiliary Building Battery (Section E1.4.2b)
50-348/97-201-20 URI Fire Barrier Penetration Seal
' Documentation (Section E1.4.4b)
50-348;50-364/97-201-21 URI Electrical FSAR Discrepancies (Section El.4.5) i
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50-348;50-364/97-201-22 URI Control of Calculations (Section E1.5b)
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APPENDIX B EXIT MEETING ATTENDEES
- Southern Nuclear Operatina Company. In l
i W.G. Hairston, President & Chief Executive Officer J.D. Woodard, Executive Vice President i D.N, Morey, Vice President J. Carlington, General Manager, Nuclear Support :
R. Hill, General Manager, Farley Nuclear Plant B.D. McKinney, Nuclear Engineering & Licensing Manager F D. H. Jones, Engineering Manager
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J.E. Odom, Operations Superintendent D.E. Grissette, Operations Manager J.J. Thomas', Engineering Support Manager G.S. Waymire, Administration Manager E.F. Bates, Nuclear Engineering & Licensing W.H. Warren, Engineering Support Supervisor !
- G.P. Crone, Training Superviso J.W. McGowan, Safety Audit and Engineering Review Group Supervisor l
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H. Mahan, Senior Engineer
'B. Badham, Safety Audit and Engineering Revie'w Group C.D. Nesbitt, Administration i
U.S. Nuclear Reaulatory Commission ,
R. Mathew, Team Leader, Special Inspection Branch, NRR l
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- D. Collins, Deputy Director, Division of Reactor Safety, RII
. T. M. Ross, Senior Resident Inspector, Farley J. Bartley, Resident Inspector, Farley
- K. Clark, Public Affairs Office, RII ,
,
Public Members
'D. Pearson, WOOF Radio
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J. Davis, WTVY- TV CH 4 D. Corb, WTVY- TV CH 4 I
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APPENDIX C LIST OF ACRONYMS USED ABN As Built Notice ,
ABV Auxiliary Building Ventilation AC Alternating Current AFW Auxiliary Feedwater AMSAC ATWS (Anticipated Transient Without Scram)
Mitigating System Actuation Circuitry ANSI American National Standards Institute ASME American Society of Mechanical Engineers BOP Balance of Plant B&PV Boiler and Pressure Vessel CCW Component Cooling Water CDF Core Damage Frequency CST Condensate Storage Tank DC Direct Current DCP Design Change Package DG Diesel Generator DR Deficiency Report E0P Emergency Operating Proceduro ERP Emergency Response Procedure FNP Farley Nuclear Plant FSAR Final Safety Analysis Report FSD Functional System Description gpm Gallons Per Minute I&C Instrumentation and Control IA Instrument Air IEEE Institute of Electrical and Electronic Engineers IP Inspection Plan ISA Instrument :ciety of America Inservice Testing
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IST kV Kilovolt kW Kilowatt LER Licensee Event Report LOCA Loss of Coolant Accident
, LOSP Loss of Offsite Power MCC Motor Control Center MDAFW Motor-Driven Auxiliary Feedwater MOV Motor-Operated Valve MS Main Steam l MSIV
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Main Steam Isolation Valve MSLB Main Steam Line Break MVA Megavolt-Ampere NPSH Net Positive Suction Head ;
!
NRC Nuclear Regulatory Commission NRR Nuclear Reactor Regulatioq, Office of (NRC)
NSSS Nuclear Steam Supply System NUREG NRC Technical Report Designation P&ID Piping and Instrumentation Diagram :
PCN Production Change Notice ,
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PCR Production Change Request Procurement Deviation Evaluation l
- PDE l PORC Plant Operations Review Committee PRA Probabilistic Risk Assessnent RCM Reliability Centered Maintenance RCP Reactor Coolant Pump RCS Reactor Coolant System RE Request for Engineering Assistance RG Regulatory Guide SB0 Station Blackout SI Safety Injection ,
SIAS Safety Injection Actuation Signal l SRP Standard Review Plan l SRSS Square Root of the Sum of the Squares i SSC Structure, System, and Component ,
SSSA Self-Initiated Safety System Assessment STP Serveillance Test Procedure SW Service Water TCN Temporary Change Notice TDAFW Turbine-Driven Auxiliary Feedwater TS Technical Specifications UPS UninterrJptable Power System URL Upper Range Limit '
USQ Unreviewed Safety Question V Volts Vac Voits Alternating Current l Vdc Volts Direct Current WR Work Request
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APPENDIX D LIST OF DOCUMENTS REVIEWED f.iD1
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Component Cooling Water System Functional System Description, A-181000 Rev. 6 Auxiliary Feedwater System Functional System Description, A-181010, Re Functional System Description Diesel Generator System, A-181005, Rev.10 Functional System Description Electrical Distribution System, A-181004, Rev. 10 SSSAs Component Cooling Water System, Self-Initiated Safety System Self-Assessment,
. November 22, 1989 Auxiliary Feedwater System, Self-Initiated Safety System Self-Assessment, March 3, 1995 FSAR/PSAR (Partial Review)
PSAR Section 5.1.2.5, Tornado Loads, Amendment 6 FSAR Section 3A, Regulatory Guide 1.7b Section, Rev. 9 FSAR Section 6.5, Auxiliary Feedwater System FSAR Section 3.3, Wind and Tornado Loadings FSAR Section 3.5, Missile Protection FSAR Section 3.1, Conformance with NRC General Design Criteria
. FSAR Section 3K.4.1.4.7, Flooding FSAR Section 6.2.4, Containment Isolation System FSAR Chapter 15, Accident Analyses FSAR 7.0 Instrumentation and Control
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FSAR 8.0 Electric Power
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FSAR Section 9.2.2, Cooling System for Reactor Auxiliaries FSAR Section 9.1.3, Spent fuel Pool Cooling and Cleanup System FSAR Section 9.2.5, Ultimate Heat Sink FSAR Section 9.3.1, Compressed Air System FSAR Section 9.2.1, Station Cooling Water System FSAR Section 9.4.2, Auxiliary Building FSAR Section 3.1, Conformance with NRC General Design Criteria FSAR Section 6.2.4, Containment Isolation System FSAR Section 98, Appendix 9B Fire Protection Program, Rev. 12 FSAR Section 3.2, Classification of SSCs FSAR Section 3.K.4.1.2.7, Flooding FSAR Section 9.2.6, Condensate Storage Facilities :
Relav Settinos ,
A-177048 Sht. 254, Relay Settings, Re A-177048 Sht. 38, Relay Settings, Re A-177048 Sht. 909, Relay Settings, Rev. I j D-1 ;
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A-177048 Sht. 44, Relay Settings, Rev. 0 A-177048 Sht. 346, Relay Settings, Rev. 3 A-177048 Sht. 261, Relay Settings, Rev. 2 A-177048 Sht. 275, Relay Settings, Rev. 2 A-207048 Sht. 44, Relay Settings, Rev. O A-207048 Sht. 270, Relay Settings, Rev. O A-207048 Sht. 256, Relay Settings, Rev. 0 ,
A-207048 Sht. 269, Relay Settings, Rev. O A-207048 Sht. 255, Relay Settings, Rev. O A-207048 Sht. 254, Relay Settings, Rev. 2 A-207048 Sht. 38, Relay Settings, Rev. O A-177048 Sht. 505, Relay Settings, Rev. 2 A-177048 Sht. 505A, Relay Settings, Rev. I l A-177048 Sht. 506, Relay Settings, Rev. 2 A-177048 Sht. 506A, Relay Settings, Rev. 1 A-207048 Sht. 505, Relay Settings, Rev. 2 A-207048 Sht. 505A, Relay Settings, Rev. 2 A-207048 Sht. 506, Relay Settings, Rev. 2 A-207048 Sht. 506A, Relay Settings, Rev. 2 J
Calculation 1 8M-95-0776-001, CCW System Evaluation Using Degraded CCW Pump Curve, Rev. 0
'8M-95-0961-001, Verification of CST Sizing Basis, Rev. 1 BM-96-1171-001, Heat-Up of Turbine Drive AFW Pump Room with Loss of >
Ventilation, Rev. O CST Manufacturer's Design Calculation, Charge Number 72-4859 REES-F-91-006, Estimate of Core Damage Frequency from a Tornado Missile Striking Exposed CST Piping, Rev. 0
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REES-F-94-014, Probability of Tornado Striking SW DG Return Lines in Excavation Trench, Rev. O SC-96-1211-002, CCW Heat Exchanger Maintenance Repairs, Rev. 1 SM-87-1-4380-001, Condensate Storage Tank Low Level Alarm Switches, Rev. O SM-96-9012-002, Effects of Plastocor Coating on CCW hx's Thermal Performance, Rev. O SM-96-9012-004, Effects of Plastocor Coating on CCW hx's Thermal Performance, Rev. O SM-ES-89-1500-007, Bounding Service Water Inlet Temperature Profile, Rev. 0 1.10, Condensate Storage vs. Core Decay Heat, Rev. I 2.18, AFW System Pump Sizing Calculation, Rev. 0 5.24, AFW Pump Room Air Handling, Rev. 0 7.14, AFW System - PCR-78-088, Rev. 0 8.13, AFW Restriction Orifices, Rev. 0 8.14, AFW Restriction Orifice Sizing, Rev. I
11.13, Available NPSH for AFW Pumps, Rev. 0 12.19 Nuclear Relief Valves Sizing, Rev.1, (partial. Review)
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17.8, AFW Flow Analysis, Rev. 0 17C-9, Auxiliary Feedwater Pump Room, Rev. 0 ,
170-10, Turbine Driven Auxiliary Feedwater Pump Room, Rev. 0 25.3, Overpressurization of AFW Piping During Overspeed Testing, Rev. 0 23.5, AFW MOVs (IE-Bulletin 85-03), Rev. I 29.01, AFW Pumps Minimum Flow, Rev. 1 34.5, Component Cooling Water System NPSH (ES No. 90-1820),Re .04, AFW System Head Curves (ES 90-1831), Rev. 0 35.5, Evaluation of CCW Surge Tank Level Setpoints, Rev. 0 36.06, Minimum AFW Flow to SGs During an Inadvertent Opening of the TD Pump Full Flow Recirc. Line (ES 91-2039), dev. 1 i D-3 l
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37.4, CCW Heat Exchanger Models and Heat Removal Capacity Calculation, Rev. O 37.7, Component Cooling Water (CCW) Flow Balance, Rev 0 38.6, Determine Flowrate Through Pipe Break in CCW System, Rev. 0
.
39.3, CCW Surge Tank Analytical Limit for Level Setpoint, Rev. 0 40.02, Verification of AFW Flow Basis, Rev. 3
E-35.1.A Setting of Protective Relays for FNP Unit 1 4.16 kv Auxiliary Power System, Rev. 2 E-35.2.A, Setting of Protective Relays for FNP Unit 2 4.16 kv Auxiliary Power System, Rev. 2 E-42, Steady State Diesel Generator Loading Calculation for LOSP, SI, and SB0, Rev. 8, E-95, Battery Capacity Calculation for LOSP and LOSP+ LOCA Situations and Limiting Battery Load Profile, Rev. SE-94-0470-004, As-Built Load Study Summary Calculation, Re SE-94-0470-005, As-Built Load Study Summary Calculation, Rev. 1 SE-94-0470-007, Farley Unit 1 As-Built Load Study Update, Rev. I SE-94-0470-007,Farley Unit 2 As-Built Load Study Update, Rev. O SE-94-0-0378-001, Instantaneous Trip Settings for MCC'B's in the MOV Setpoint Document, Rev. O SE-88-ll26-1, Battery Study-Reevaluate Batteries for 59 Cell Operation, Rev. 2 SE-88-ll26-2, Service Water Battery Load Profile, Rev. 1 E-98, Minimum Available DC Voltage & Permissible Control Circuit lengths For
Existing Battery Load Profile per E-95, Rev. 3 E-106, Battery capacity Calculation for TDAFP-UPS, Re E-IIS, Excessive Voltage Drop in DC Circuits, Rev. 3 E-Il6, Minimum Available DC Voltage and Permissible control Circuit Lengths for Limiting Battery Load Profile, Rev. 4 E-144, Determination of Battery Capacity Margins for Adequacy of Voltage at Safety-Related Components for Various Load Profiles, Rev. 4 D-4
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Procedures FNP-0-AP-8, Design Modification Control, Rev. 22 FNP-0-AP-88, Nuclear Safety Evaluations, Rev. O NEP 4-0, Processing Design Change Requests, Rev. 3 NEP 4-15, Project Deficiency and Corrective Action Reporting GO-M-1, Designer Interface Document, Rev. 3 FNP-0-ETP-4379, Performance Test for Unit 1&2 Component Cooling Water Heat '
Exchangers, Rev. 5 FNP-2-ETP-4416, CCW Pump Flow Variation with Mini-Flow Valve Position, Rev. O FNP-0-MP-7.3 Turbine Driven Aux Feed Pump Overspeed Trip Setpoint Check, Rev. 3
' NP-2-IMP-210.1 May 1995, Component Cooling Water Surge Tank Level Loop F
Calibration Q2Pl7LT3027A, Rev. 6 FNP-2-ARP-1.1, Main Control Board Annunciator Panel A, Rev. 17 PDI 005.4-4, Project Desk Instructions for Calculations, Rev. I GO-NG-42, 50.59 Evaluations, Rev. 3 FNP-0-50P-0, General Instructions to Operation Personnel, Rev. 40 FNP-0-SOP-0.4, Fire Protection Program Administration Procedure, Rev. 31 FNP-0-50P-23.1, Component Cooling Water Pump Lubrication Procedure, Rev. O FNP-1-50P-22.0, Auxiliary Feedwater System, Rev. 37 FNP-1-SOP-22.0A, Auxiliary Feedwater System, Rev. I FNP-2-50P-1.1A, Reactor Coolant System, Rev. 6 (partial review of CCW valves)
FHP-2-SOP-2.lA, Chemical and Volume Control System, Rev. 8 (partial review of CCW valves)
FNP-2-SOP-23.0, Component Cooling Water System, Rev. 39 FNP-2-50P-23.0A, Component Cooling Water System, Rev. 5 FHP-2-SOP-24.0, Service Water System, Rev. 31 (partial review)
FNP-1-A0P-2.0, Steam Generator Tube Leakage, Rev.18 D-5
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FNP-1-A0P-5.2, Abnormal Operating Procedure Degraded Grid, Rev. 8 ,
'FNP-1-A0P-13.0,' Loss of Main Feedwater, Rev. 10 FNP-1-A0P-14.0, Secondary System Leakage, Rev. I ;
FNP-1-A0P-29.1, Plant. Stabilization in Hot Standby and Cooldown without "A" Train AC or DC Power, Rev. 9 ;
FNP-1-A0P-29.2, Plant Stabilization in Hot Standby and Cooldown without "B" Train AC or DC Power, Rev. 8 FNP-2-A0P-5.2, Abnormal Operating Procedure Degraded Grid, Rev. 7 FNP-2-A0P-6.0, Loss of Instrument Air, Rev.15 FNP-2-AOP-9, Loss of CCW, Rev. 9 FNP-1-STP-22.1, IA Auxiliary Feedwater Pump Quarterly Inservice Test, Rev. 23
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FNP-1-STP-22.2, IB Auxiliary Feedwater Pump Quarterly Inservice Test, Rev. 23 FNP-1-STP-22.5, Auxiliary Feedwater System Flow Path Verification, Rev.18 FNP-1-STP-22.6, Auxiliary feedwater Pump Train 8 Functional Test, Rev.15 FNP-1-STP-22.7, Auxiliary Feedwater Pump Train A Functional Test, Rev.18 ;
FNP-1-STP-22.8, Auxiliary Feedwater Inservice Valve Exercise Test, Rev.13
FNP-1-STP-22.9, Auxiliary Feedwater Pumps IA and 18 Auto Start Test, Rev. 8 !
l FNP-1-STP-22.10, Turbine Driven Auxiliary Feedwater Pump Blackout-Start Test, !
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Rev. 18 FNP-1-STP-22.11, Auxiliary Feedwater Pump 1A LOSP Test, Rev. 11
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l FNP-1-STP-22.12, Motor Driven Auxiliary Feedwater Check Valves Flow *
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Verification, Rev. 9 FNP-1-STP-22.13, Turbine Driven Auxiliary Feedwater Check Valves Flow Verification, Rev. 1 FNP-1-STP-22.16. Turbine Driven Auxiliary Feedwater Pump Quarterly Inservice Test (TAVG >/- 547 F), Rev. 26
- . FNP-1-STP-22.18, Auxiliary Feed Automatic Valve Position Verification, Rev. 5
'FNP-1-STP-22.19, Auxiliary Feedwater Flow Path Verification, Rev. 13 FNP-1-STP-22.20, TDAFW Pump Steam Admission Valves Air Accumulator Test, Rev. 7 D-6
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FNP-1-STP-22.21, Turbine Driven Auxiliary Feedwater Pump Automatic Valve Test, Rev. 5 FNP-1-STP-22.22, Motor Driven Auxiliary Feedwater Pump Automatic Valve Test, Rev. 5 FNP-1-STP-22.23, Turbine Driven Auxiliary Feedwater Pump Trip and Throttle Valve Indication Operability Test, Rev. 3 FNP-1-STP-22.24, Aux. Feedwater Check Valve Reverse Flow Closure Operability Test, Rev. 8 FNP-1-STP-22.26, Auxiliary Feedwater Pump 1A Cold Shutdown Inservice Test, Rev. 5 .
FNP-1-STP-22.27, Auxiliary Feedwater Pump IB Cold Shutdown Inservice Test, Rev. 4 FNP-1-STP-22.28, Aux. Feedwater Pump Suction Check Valves Reverse Flow Closure Operability Test, Rev. 5 FNP-1-STP-256.1 Reactor Safeguards Response Time Test, Rev. 13 FNP-1-STP-256.15, Loss of 0FF-Site Power Response Time Test, Rev. 14 FNP-1-STP-256.18 Turbine-Driven Auxiliary feedwater Pump Response Time Test, Rev. 11 FNP-1-STP-45.0, Refueling Valve Inservice Test, Rev. 13
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FNP-2-STP-23.1, 2A Component Cooling Water Pump Quarterly Inservice Test, Rev. 14 FNP-2-STP-23.2, 2B Component Cooling Water Pump Quarterly Inservice Test, Rev. 17 FNP-2-STP-23.3, 20 Component Cooling Water Pump Quarterly Inservice Test, Rev. 12 FNP-2-STP-23.7, Component Cooling Water Flow Path Verification Test, Rev.12 FNP-2-STP-23.8, Component Cooling Water Valve Inservice Test, Rev.18 FNP-2-STP-23.12, CCW to RCP Thermal Barrier Check Valve Reverse Flow Test, Rev. 2 FNP-2-STP-40.0, Safety Injection with Loss of Off-Site Power Test, Rev. 25 FNP-1-STP-914, Auxiliary Building Battery Charger Load Test FNP-0-EMP-1352.06, TDAFW UPS Power Supply Cleaning and Inspection , Rev. 2 D-7
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- FNP-0-EMP-1352.06,TDAFW UPS Power Supply Cleaning and Inspection, Rev. 1
' FNP-0-EMP-1370.02, Installation and Repair of Penetration or Conduit Seals, Rev. 5 ,
i FNP-0-EMP-1513.01, ~ Electrical Maintenance Procedure ITE Magnetic $tarters and l Overload Relays, Rev.10 ,
l FNP-1-EMP-1352.05, Turbine Driven Auxiliary Feedwater (TDAFW) UPS Battery i Performance Test, Rev. 1
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FNP-1-STP-905.1, Auxiliary Building Battery Service Test, Rev. 9 :
FNP-1-STP-905.2, Auxiliary Building Battery Performance Test IA, Rev. 1 - ,
- Performed on 10/9/92 ,
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FNP:-1-STP-905.2, Auxiliary Building Battery Performance Test IB, Rev.1 - .
Performed on 10/20/92 l FNP-1-STP-905.2 Auxiliary Building Battery Performance Test 2B, Rev. 1 - ,
Performed on 3/22/95 ;
<
FNP-1-STP-914, Auxiliary Building Battery Charger Load Test , Rev. 4 l l
FNP-2-STP-905.1, Auxiliary Building Battery Service Test 2B, Rev. 6 -
F Performed on 10/21/93 FNP-2-STP-905.1, Auxiliary Building Battery Service Test, Rev. 10
FNP-2-STP-914, Auxiliary Building Battery Charger Load Test, Rev. 5 FNP-2-STP-914, Auxiliary Building Battery Charger Load Test, Rev. 4, Performed l on 5/8/96 FNP-0-STP-906.1, Service Water Building Battery Performance Test, Rev. 3 .
(Battery # 3 Performed on 2/29/96, Battery # 4 Performed on 3/13/96, Battery # l 1 Performed on 2/22/96, and Battery # 2 Performed on 3/6/96) !
FNP-1-ENP-1341.08, Auxiliary Building Battery Equalization, Rev. 3 i t
Desian Chanaes PDE 93-0-0092, CCW Valve Wedge l DCP 96-0-9012-2-006, Process Coating for CCW Heat Exchangers l
PCR 84-2-2911, Installation of Curb for CCW Pump 2C i PCR 89-2-5829, Replacement of CCW Surge Tank Vacuum Breaker Valves PCR 91-2-7251, Isolation of Swing CCW Heat Exchanger l
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PCR 92-0-8120, Replacement of AFW Control Valve Actuator, Rev. O PCR 90-1-6503, Acceptance of Replacement Mechanical Seals for TDAFW Pump, Rev. O PCR 92-1-8234, Replace 8 inch Service Water Supply to AFW Pump A, Re PCR 85-2-3322, Replacement of CCW Flow Indicators Q2P17FISL 3048 A, B, C, N2PlFI3036, FI 3049, FI 3066, FI 3077-PCR 87-4073, IA MDAFW Pump Discharge Pressure Loop PT 3213A Range change PCR 89-2-5721, Replace Q2P175V3028A/B PCR 90-1-6934, Replace a non-lE flow indicator with a IE indicator PCR 91-2-7432, Evaluates CCF Surge Tank Low-Low Level Actuation Requirements, Verifies Stroke Times for CCW Isolation Valves Q2P17HV-3096A/B, and Deletes the Automatic Pump Trip PCN 89-2-5721, Replacement of CCW Surge Tank Vent Valve Solenoid Valves PCN B-90-0-7100, Replacement of CCW Relief Valve PCN 91-0-7656, CCW Pump Motor Bearing Replacement PCN 92-0-8369, CCW Pump Bearing and RTD Replacement PCN B-88-2-5425, Revised P&ID to Reflect As-Built Plant Condition PCN B-90-0-6871, Equivalency Evaluation for Valve Pa'rts PCN B-88-1-5249, Equivalency Evaluation for Valve Trim Parts PCN B-88-1-5003, Change Overspeed Trip Setting for TDAFW Pump PCN B-90-1-6743, Drawing Change for Agastat Relay Replacement PCN B-84-1-2518, Auxiliary Feedwater Check Valve Temperature Monitoring System PCN-85-2-3422, Removal of auto-manual stations FIC-3009 A, B, C PCN S-87-1-4457, Adds Aux F.W. solenoid Valves Q1N235V3227AC, BC, and CC PCN 88-1-5200, PT 3213B Range Change PCN B-89-2-5721, CCW Surge Tank Relief Valve Solenoid Replacement PCN B-91-0-7190, Replace the TDAFW Pump Speed Indication Converter / Transmitter PCN B-91-1-7431, Instrument Loop Uncertainty and Setpoint (L30270)
D-9
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87-1-4073 1A, MDAFW Pump Discharge Pressure Loop, PT 3213A Range Change DCR 85-2-3322 Replacement of Flow Instruments that were being overranged DCR 92-1-7934, Relocates the CCid flow meters in the control roo N1Pl7F13043AA, BA and C 'DCR 96-1-9067 TDAFWP Steam Admission Valves Modify the Control circuits of ,
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Q1N23HV3235A&8 i KN 89-1-6106, AFW HFA Relay Drawing Change per 88-0-4980 ,
KN 89-1-6354, Lighting around TDAFW UPS Batteries DCP 96-1-9008, TDAFW UPS Fuse & Indicating Bulb Replacement
'
KN B-92-2-8068, Unit 2 Auxiliary Building 8attery Replacement
' KN S-88-1-5130, Unit 1 Auxiliary Building 8attery Rack Configuratio I Correction PDE 93-0-0029, Replacement Motor for CCW System .
DCP 95-0-8853, Replacement of SV-1 Relays for Diesel Generators ABN 96-0-0930, Change CCW Pump Room Equipment Maximum .
ABN 96-0-0986, Add Circuits to TDAFW Pump Control Panel ABN 94-0-0520 Incident Reoorts IR No. 1-89-414, 12/89 IR No. 1-89-275, 8/15/89 IR No. 1-86-304, 7/30/86 IR No. 1-93-231, 10/7/93 .
IR No. 1-87-251, 8/29/87 IR No. 1-88-122, 4/6/88 IR No. 1-86-273, 7/14/86 '
IR No. 1-94-299 Licensina Commitments LC 0401 Clarification of TM1 Action Requirements (NUREG-0737)
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LC 554 Licensing Correspondence dated 1/14/81 LC 555, Rev. 008 Title: II.E.1.2 Auxiliary Feedwater System Automatic
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Initiation and Flow Indication D-10 :
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LC 1154, Anticipated Transient Without Scram (ATWS)
LC 1712, II.K.2.19 Sequential Auxiliary Feedwater Flow Analysis LC 2448. Evaluation of IE Notice 83-55
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- LC 3320 Licensing Correspondence dated 2/10/82 l
LC 3613 Engineered Safety Features Bypass, Override and Reset Circuits LC 3947, Undetected Unavailability of the Turbine Driven Auxiliary Feedwater -
Train j
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LC 3971, Licensing Correspondence' dated 7/29/80 LC 3991, Licensing Correspondence dated 9/30/80 LC 7322, Miniflow Evaluation - NRC Bulletin No. 88-04
'LC 743B, IE Information Notice (IEN) 86-14: PWR Auxiliary Feedwater Pump Turbine Control Problems LC 7514, Resolution of NUREG-0578 Cat A Requirements LC 8217, Miniflow Evaluation - NRC Bulletin No. 88-04 :
LC 8412, Licensing Correspondence dated 12/12/89 f LC 8505, NRC Inspection Report 90-01
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LC 8549, LER 89-004, Supp #1 LC 8815, NRC Generic Letter 90-03 LC 8902, NRC IN 85-33 ,
LC 9532, Rev. 2, LER 91-011-00 LC 10773 Licensing Correspondence dated 6/13/94 LC 11038, Loss of Fill-011 in Transmitters Manufactured by Rosemount Licensee Event Reports81-024 82-006
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'82-039 94-005-00 0-11
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Notices of Violation -
NOV dated 2/11/81 NOV dated 9/13/93 NOV dated 5/2/95
NOV dated 1/23/96 NOV, dated 9/27/96-Drawinas
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D-207618 Sht'.1, Elementary Diagram 575V Motor Operated Valves, Rev. 6 B-204610 Sht. 18, Conn. Dia. Motor Operated Valves, Rev. 7 D-204920 Sht. 1, Power Penetration WA10 Inside CTMT Connection Diagram-
'.Q2T528016-B, Rev. 18'
D-204948 Sht.1, Control Penetration Wall Inside CTMT Connection Diagram-Q2T528038-B, Rev. 17 D-207374 Sht.1, Elementary Diagram Solenoid Valves, Rev. 2 B-204607 Sht.15, Conn. Diag. Solenoid Valves, Rev. 6
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D-204936 Sht. 1, Control Penetration EB01 Inside CTMT Connection Diagram-Q2T528019-A, Rev. 18 D-207855 Sht.1 Solenoid Valves, Rev. 5
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B-204606 Sht. 30, Conn. Diag. Solenoid Valves, Rev. 6
' D-204942 Sht. 1, Control Penetration WB03 Inside CTMT Connection Diagram-i Q2T52B020 -B, Rev. 15 B-205810 Sht. 23, Logic Diagram, Rev. 3 B-205810 Sht. 22, Logic Diagram, Rev. 3 B-20S810 Sht. 100, Logic Diagram, Rev. 1 .
B-205810 Sht. 101, Logic Diagram, Rev. 5 D-207185 Elementary Diagram Component Cooling Water Pump 2B-Train "A", Rev. 9
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D-207187, Elementary Diagram Component Cooling Water Pump 2b-Train "B",
Rev. 13
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D-12 i
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D-203266, Tray & Conduit layout E1. 100'-0" Safe Shutdown Raceway Identification and Location of Kaowool Wrap, Rev. 11
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.D-180533 Sht. 1, Tray & Conduit Layout El. 100'-0" Area 2 Sa ef Shutdown Raceway Ident and Location of Kaowool Wrap, Rev. 9 D-180533 Sht. 2, Tray & Conduit Layout El. 100'-0" Area 2 Safe Shutdown Raceway Ident and Location of Kaowool Wrap, Rev. 5 l
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D-207183, Elementary Diagram Compnent Cooling Water Pump 2C
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D-207184, Elementary Diagram Compnent Cooling Water Pump 2A, Rev. 12
D-177183; Elementary Diagram Compnent Cooling Water Pump IC, Rev.14 ,
D-177185 Sht. 1, Elementary Diagram Compnent Cooling Water Pump 1B Train A,
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Rev. 16
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D-N7185 Sht. 2, Elementary Diagram Component Cooling Water Pump 2B Train A),
Rev. 9 ;
D-107187, Elementary Diagram Component Cooling Water Pump 2B Train B), Rev.13
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D-207119, Interlock Schematic Component Cooling Water Pump 2B D-177186 Sht. 1, Elementary Diagram Auxiliary Feedwater Pump 4160V No. lA, Rev.'20 D-177186 Sht. 2, Elementary Diagram Auxiliary Feedwater Pump 4160V No. 1B, Rev. 10 ,
D-207186, Elementary Diagram Auxiliary Feedwater Pump 4160V No 2A & 2B, Rev. 14
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~D-177229, Elem Diag - AFW Pump Room Cooler Fan Motors, Rev. O D-207229, Elem Diag HHSI & AFW Pump Room Cooler Fan Motors, Rev. 8
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D-177243, Elem Diag - Component Cooling Water Pump Room Cooler Fan Motors, Rev. 11 D-207243, Elem Diag - Component Cooling Water Pump Room Cooler), Fans, Rev. 8 !
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D-173096 Sht. 1, Loads Diagram, Rev. 28 D-203096 Sht. 1,. Loads Diagram, Rev. 19 l
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D-203096 Sht. 2, Loads Diagram, Rev. 7 i
D-173096 Sht. 2, Loads Diagram, Rev. 13 ;
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D-177032, Logic Diagram Diesel IB Auto Start & Loading, Rev.15 D-13
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' D-177033, Logic Diagram Diesel.1-2A Auto Start & Loading, Rev.18 ,
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D-207032, Logic Diagram Diesel'2B Auto Start & Loading, Rev. 13
D-207033, Logic Diagram Diesel 1-2A Auto Start & Loading, Rev. 16, D-207036, Logic Diagram Diesel IC Auto Start & Loading, Rev.10 D-177036, Logic Diagram Diesel IC Auto Start & Loading, Rev. 11
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D-202700, Main Single Line Diagram Generator and 4160V Transformers, Rev. 13 -l t
D-207001, Single Line-Electrica1' Auxiliary System (Emergency 4160V & 600V),
Rev. 14 :
D-177001, Single Line-Electrica1' Auxiliary System (Emergency 4160V & 600V),
Rev. 16 l D-172700, Main Single Line Diagram Generator and 4160V Transformers), Rev. 17 D-1770ll, Single Line Protection & Metering 600V Load Center IE (EMERG), '
Rev. 1 l D-177010, Single Line Protection & Metering 600V Load Center ID (EMERG), .
Rev. 1 l
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D-2070ll, Single Line Protection & Metering 600V Load Center 2E (EMERG), ,
Rev. 04 D-207010, Single Line Protection & Metering 600V Load Center 2D (EMERG),
Rev. 10
- D-177944, Single Line Diagram Turbine Driven Aux. Feedwater Pump UPS), Rev. 3 U-263193A, Schematic Diagram 3C NO-Break System Type CT2B48B100CR3CA3 B-177556 Sht. 18, Mcc Schedules - 600V MCC - IT), Rev. 9 B-177556 Sht. 18A, Mcc Schedules - 600V MCC - IT), Rev. 7 B-177556 Sht. IBB, Mcc Schedules - 600V MCC - IT), Rev. 9 D-172213 Sht.1, Cable Tray Layout & Exp Conduit Diesel Bldg Sht. 3), Rev. 31 D-172540, Connection Diagram-Motor Control Center IT - Sections D, E, &.F),
Rev. 8 D-173092, Single Line Cable & Conn Diag - 600V lZ & 120/208VAC Dist Cab's Circ Wtr Chlorine Struc), Rev. 0-D-14
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.D-172863, Elementary Diagram - Motor Control Center IT (Diesel Building),
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Rev. 12 D-175007, Auxiliary Feedwater System, Rev. 22
- D-175033, Sheet 2, Main Steam and Auxiliary Steam Systems, Rev. 18
D-175003, Sheet 1, Service Water System, Rev. 32 D-175003, Sheet 2, Service Water System, Rev. 27
D-1750ll, Sheet 3, HVAC-Radwaste Area, Rev.' 9-D-175029, Sheet 1, HVAC Process Flow Diagram-Radwaste Area, Rev. 11
.0-175035, Sheet 2, Service Air System, Rev. 6 !
i i
, D-205002, Sheets 1, 2 and 3 1
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CST Manufacturer's Drawings U161693D, Revision A0; and U161703B, Rev. 2
(ther Documents Reviewed-
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Applicable sections of the Technical Specifications including 3/4.7.1.2, ,
- 3/4.7.1.3, 3/4.3 and 3/
- MOV Design Basis Document (partial review), U418109, Rev. A REA-0873, Auxiliary Building Room Coolers Attendant Equipment Evaluation f
- ES 1617; Flooding Analysis of CCW Pump Room I Bechtel letter response AP-21413 dated 5/22/96 to REA 95-0776, Minimum ,
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Analyzed CCW Pump Performance Data and Operation of CCW Pump with Minimum Flow Line Closed, Rev. 1 Original NRC SER Section 9.3.1, Auxiliary Feedwater System j l
Auxiliary Feedwater Operator License Training Objectives, OPS-52102H, 6/4/92 ;
- . .
? Work Order Numbers 97001743, 97001744, 97001743, and 547310 t
i Nuclear Generation Department Memorandum, IE Information Notice (IEN) 84-66, ,
i dated 5/12/87 l Nuclear Generation Department Memorandum, .IE Information Notice (IEN) 89-48, dated 11/14/90 ;
Nuclear Generation Department Memorandum, IE Information Notice (IEN) 88-70,
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dated 9/12/89-
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Intracompany Correspondence, CST Unprotected Connections - Evaluation of Acceptability per GL 91-18,2/28/96 System Specialist 3 Year History Review for AFW System, 2/6/97 Alabama Power Company Letter, Resolution of Safety Issue 93, " Steam Binding of .
Auxiliary Feedwater Pump .* ' Generic Letter 88-03),5/24/88 1
Alabama Power Company Letter, Resolution of Safety Issue 93, " Steam Binding of Auxiliary Feedwater Pumps" (Generic Letter 88-03),9/9/88 l NRC Letter, Completion of MPA 8-98, " Steam Binding of Auxiliary Feedwater !
Pumps," Generic Letter 88-03, for FNP, 10/28/88 10CFR50.59 Evaluation, Condensate Storage Tank Missile Protection, Rev. O, 11/16/94 Equipment Specification, SS-1111-4, Field Erected Steel Tanks, Rev. 2 U2130930 Differential Pressure Transmitter 59DP1 Veritrak U-262166, 262167, 262168 Field Equipment List Cabinet J and Cabinet K.
U279269, U263864 Plant Specific Setpoints for Emergency Operation Procedures l Rev. 12 , Dec. 1990 t Section 10, Steam Generator Level Control & Protection (computerized scaling manual)
FNP-1-EEP-1 Nov,1995, Loss of Reactor or Secondary Coolant, Rev.15 WCAP-13794 Bases Document for Westinghouse Setpoint Methodology for Protection Systems, July 1993 1 I
WCAP 13751 Westinghouse Setpoint Methodology for Protection Systems, June 1993 WCAP 13945 Bases Document for ERP Instrumentation Uncertainties, Dec 1993 WCAP 13992 Steam Generator Lower Level Tap Relocation Assessment, March 1994, Rev. 13 OPS 40204E/52103C, Systems-Training Student Text DC Distribution OPS 40204E/52103C, Systems-Training Student Text DC Distribution A-350972, Criteria for the Selection / Evaluation of Thermal Overload Heater And Recommendations for Selection of Magnetic Breaker Setpoints for Telemacanique (ITE-Gould) Motor Control Centers Starters Controlling the Motor Operated Valve Actuators, Rev. 1 WO 503405, Work Order for Battery Cell Replacement TDAFW Battery l D-16 i
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7597-20-E10.9-29-1, Battery Arrangement 2 Step Seismic (6) 4 LCY-420 Batteries CA D M-9309 Rev. 0, Unit 2 U-279462 7597-03-E10.9-34-2, Battery Arrangement 2 Step Seismic (6) 4 LCY-420 Batteries C& D M-9309 Rev. O, Unit 1 U-265645A OPS-40204F/52103D, Farley Systems -Training Course for TDAFW Battery /UPS
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REA 96-1093, Load Change for Diesel Generator IC/2C ,
REA 95-1022, Evaluation in support of FSAR Section 8.2.2.4 Change SPI-07-2. Specific Electrical System Parameters to be Reviewed Before Issue of :
Electrical Design Package AP-18775, Diesel testing Material Referenced in Calc. E-42
,
B-87-2-4592 Evaluation in support of FSAR Table 8.3-3 Change Design Guide- Electrical Power Cable Sizing Guide for J.M. Farley Nuclear i Plant-Units 1 and 2, Rev. 0 5S-1102-132, Inquiry for Piping, Instrumentation Tubing, Ductwork, Electrical Raceways and Firewall Penetration Seals, Rev. 9 S/N'27290.1, ASTM E84-75 Fire Tests Surface Burning Characteristics of Dow Corning Q3-6548 Silicone Foam and Dow Corning Sylgard 170 Elastomer Applied I Over Aluminum Designation FH 530 - Factory Mutual Testing j S/N 27290, Fire Test on Silicone Rubber Penetration Seals in Masonry wall - l Factory Mutual Testing i
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BM-3095, Change Notice for Room Cooling for Room 254, Rev. I
A-180580 Unit 1 Safe Shutdown Equipment Report, Rev. 14
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A-180581 Unit 1 Safe Shutdown Circuit Report, Rev. 1
- A-180582 Unit 1 Safe Shutdown Raceway Report, Rev. 15 A-203580 Unit 2 Safe Shutdown Equipment Report, Rev. 16
[ A-203581 Unit 2 Safe Shutdown Circuit Report, Rev. 18
< A-203582 Unit 2 Safe Shutdown Raceway Report, Rev. 16 A-207545, Terminal & Pu11 box Details & Notes, Rev. 35-A-177541, Tray &. Conduit Details & Notes, Rev. 91 A-172389, Cable Working Specifications, Rev. 5 D-17
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A-177538, Elec General Details & Notes, Rev. 24 >
Safety Evaluation for B-92-0-8099, Rev. 2 Safety Evaluation for FNP-1-STP-905.1, Rev. 6 Safety Evaluation for FNP-1-STP-905.1, Rev. 8 Safety Evaluation for FNP-1-STP-905.1, Rev. 7 Safety Evaluation for FNP-1-STP-705.1, Rev. 4A Safety Evaluation for FNP-1-STP-905.1, Rev. 5 Safety Evaluation for FNP-2-STP-935.1, Rev. 6 ,
Safety Evaluation for FHP-2-STP-905.1, Rev. 4A i
Safety Evaluation for FNP-2-STP-905.1, Rev. 6A Safety Evaluation for FNP-2-STP-905.1, Rev. 7 Cable Schedule for IDYFT-F2P, Report 10007 l A-181987 Fuse Replacement Manual Unit 1, Unit 2, & Unit 2 Shared Safety Related Equipment, Rev. 17
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