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Insp Rept 50-313/97-201 on 970210-0314.Violations Noted. Major Areas Inspected:Design Inspection of EFW & Decay Heat/Low Pressure Injection System
	ML20197B842
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/24/1997
From:	; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20197B838	List: IR 05000313/1997201; ML20197B832;
References
	50-313-97-201, NUDOCS 9803120197
	Download: ML20197B842 (101)
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U.S. NUCLEAR REGULATORY COMMISSION i

OFFICE OF NUCLEAR REACTOR REGULATION

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4-Docket No.: 50 313 License No.: DPR 51 '

! Report No.:

50 313/97 201 j Lical .x Cntergy Operatiotis, In t Facility:

i Arkansas Nuclear One, Unit 1

,. Location: Russellville, Arkansas Dates:

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February 10 March 14,1997 f

Inspectors:

Robert P:,. tis, Jr., Team Leader, NRR Edmund Kleeh, NRR

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Mansoor Sanwarwalla, Sargent & Lundy Corporation i

' Kenneth Steele, Sargent & Lundy Corporation

Leland Rogers, Sargent & Lundy Corporation Rienard Jtison, Sargent & Lundy Corporation

! Augusto Bizarra, Sargent & Lundy Corporation

. Approved by

Donald P. Norkin, Section Chief Specialinspection Branch

Division of inspection and Support Programs Office of Nuclear Rasctor Regt.ation

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9903120197 970724

PDR ADOCK 05000313 G PDR

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TABLE OF CONTENTS E X E C U TIVE S U M M A R Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

El Conduct of Engineering . . . . . . . . . . . . . . .

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E1.1 Inspection Objecilves and Methodology . . . . . . . . . . . . . . . . ......... . . . . . . . . 11 E1.2 Emergency Feedwater System ..............

E 1.2.1 S yst e m ove rvie w . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

................ 1 E1.2.1.1 System Functions

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- E1.2.2 Mechanical Design Review . . . . . . . . . . . . . . . . . . . . . . ..... 2 E1.2.2.1 EFW System Design Requirements for Heat Removal - . . . 3

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E1.2.2.2 Condensate Storage Tank . . . . . . . . . . . . . . . . . . . . . 12

' E1.2.2.3 Pump Redundancy and Flow Capability . . . . . . . . . . . . 14 E1.2.2.4 Net Positive Suction Head , . . . . . . . . . . . 16 E 1.2.2. 5 Piping De sig n . . . . . . . . . . . . . . . . . . . . . .......

. . . . . . . . 17 E1.2.2.6 Environmental Ouslification . . . . . . . . . . . . . . . . . . .1 .

E1.2.2.7 Design Configuration Control . . . . . . . . . . . . . . . . . . . 20 E1.2.2.8 Service Water System E1.2.3 Electrical Design Review .......................23

............................24 E1.2.3.1 Emergency Diesel Generator Loading ... , ....... 25 E1.2.3.2 Protection Coordination ,...................

E 1.2.3.3 Electrical Modifications . . . . . . . . . . . . . , . . . . . . . . . 27

E1.2.3.4 Cable Design and Installation . . . . . . . . . . . . . . . . . . . 29 E1.2.3.5 Wiring and Cable Termination E 1.2.3.8) ..................30 Raceway and Raceway Support instellation . . . . . . . . . 31 E1.2.3.7 Grounding and Cathodic Protection .............,33 E1.2.3.8 Loading of Class 1E Batteries .......,..........34 E 1.2.3.9 DC Distribution System . . . . . . . . . . . . . . . . . . . . . . . 35 E1.2.3.10 AC Distribution System . . , . . . . . . . . . . . . . . . . . 37 E1.2.3.11 EDG Load. Sequencing and Starting Circuits . . . . . . . . 39 E1.2.3.12 Testing of Molded. Case Circuit Breakers . . . . . . . . . . 41 E1.2.3.13 Design Control for Electrical Drawings ........... 43 E1.2.3.14 Discrepancies in Design. Basis Documents . . . . . . . . . 44 E1.2.4 instrumentation and Control Design Review . . . . . . . . . . . . . . . 45 E1.2.4.1 EFIC Actuation Logic

. . . . . . . . . . . . . . . . . . . . . . . 45 E1.2.4.2 ConderMate Storage Tank Level Instrumentation ..... 46 E1.2.4.3 EFW Pump Flow, OTSG Pressure and Level Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . .

E1.2.4.4 EFW Pump instrumentation . . . . . . . . . . . . . . . . . . .. .48 47 E1.2.5 EFW Walkdown Observations and Results . . . . . . . . . . . . . . . . 49 E1.2.5.1 Mechanical Walkdown and in. Plant Observations .. .49 .

E1.2.5.2 Instrumentation and Controls Walkdown and in. Plant Observations . . . . . . . . . . . . . . . . . . . . . . . . 52 E1.3 Decay Heat / Low Pressure injection System . . . . . . . . . . . . . . . . . . . . . . 54 E 1. 3.1 S yst em Ove rvie w . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 54 E1.3.1.1 System Description E1.3.1.2 System Functions .........................54

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i i E 1.3.2 Mechanical Design Revie w . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 E1.3.2.1 BWST Pressure and Vacuum Relief Valve ..........57 E1.3.2.2 BWST Vortexing and Pump NPSH ...............,60

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E1.3.2.3 Post Accident Radioactive Releases from the BWST . . . 61 E1.3.2.4 LPI System Design. Basis Flow and Surveillance

i T e st ing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3 E1.3.3 Electrical Design Review

............................63 E1.3.4 Instrumentation and Control Design Review . . . . . . . . . . . . . . . 63 i E1.3.4.1 EFW Initiation and Control Logic . . . . . . . . . . . . . . . . 64 E 1.3.4.2 BWST Level instrumentation . . . . . . . . . . . . . . . . . . .

! E1.3.4.3 DH/LPl Flow instrumentation . . . . . . . . .66. . . . . . .

- E1.3.5 DH/LPl Walkdown Observations and Results , . . . . . . . . . . . . . . 66 i E1.3.5.1 Mechanical Walkdown and in. Plant Observations . . . . . 66 E1.3.5.2 Instrumentation and Control E 1.4 Exit Meeting . . . . . . . . . . . . . . . . . . . . . . . s Walkdo wn . . . . . . . . . . . 68 i ....................70 i Appendix A

{ List O f Open it e m s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A 1 Appendix B i

Exit Mee ting At t endees . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B1 Appendix C List of Documents Revie wed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C 1 Appendix D Lis t o f A cron ym s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D 1

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EX5CUTIVE SUMMARY From February 10 through March 14,1997, the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation (NRR), performed a design inspection of the emergency feedwater (EFW) and the decay heat / low pressure injectio (DH/LPI) systems at Arkansas Nuclear One, Unit 1 (ANO lh A team leader from the NRR Special Inspection Branch headed the inspection team, which also included an engineer fro NRR and five engineers from the Sargent & Lundy Corporatio The purpose of the ANO 1 design inspection was to evaluate the capability of the selected systems to perform the safety functions required by their design bases, as well as the adherence of the systems to their respective design and licensing bases, and the consistency of the as built configuration with the plant's final safety analysis report (FSARh The team relected the EFW and DH/LPI systems as the focus of this inspection on the basis of probabilistic risk assessment, previous inspection insights, and modification histor The team's evaluation of the EFW systtm revealed that the system provides acceptable operational performance capability and, as installed and operated, the system meets both the original design bases and subsequent licensing commitments. Through the mechanical review, the team considered whether the protected volume of the condensate storage tank was large enough to ensure missile protection; whether the service water system had adequate capability to serve as the assured source of cooling wE :r to the EFW system:

whether the main steam safety relief valves, atmospheric dump valves, and the turbine bypass system valves provided adequate relief capacity; and whether the EFW pumps provided adequate flow capacity. On the basis of these considerations, the team determined that the ANO 1 EFW system can adequately respond to postulated accident conditions to bring the plant to a safe shutdown condition, in addition, the team determined that sufficient rat positive suction head (NPSH)is available for the EFW pumps to derive suction from either the safety related ("O") condensate storage tank (CST) or the service water syste The results of the team's electrical review indicated that sufficient voltage and current are available to power the equipment that comprises the EFW system, in addition, the team confirmed that the electrical equipment has adequate circuit protection and that the EFW pump motor is adequately size The team's review of instrumentation and controls revealed that the EFW initiation control set points were sufficient to ensure automatic actuation of the EFW system when required in addition, the team verified that the CST level indication and alarm in the control room are adequat Walkdown of the EFW system revealed that the overall material condition and general appearance of the system were good, and the majority of the areas containing EFW system components were in good o. der. The licensee has a Site Upgrade Project in progress that will continue throughout 1997 i

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i Notwithstanding the above, the team identified severalissues relative to the EFW system's design of the licensee's implementation of the design. The following examples highlight issues identified by the team outing the inspection or by the licensee during their prepara for the inspection:

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The current bases for the minimum required water levelin the "O" condensate storage tank, as defined in the Technical Specifications, did not agree with the design information reviewed by the team. The minimum water level requirement for the tank was based on it being able to provide, in case of a tornado, sufficient water for 30 minutes for EFW system operation before switch over to the service water system is accomplishe *

The team raised a concern regarding an event that resulted in excessive EFW flow (beyond the design lim',) to the steam generators. Esisting plant procedures do not l specify any maximum allowable flow design limits, which would preclude such excesses, however, an analysis performed by Framatome Technologies, Inc.,

together with the results of the licensee's steam generator tube inspection during the last outage, suggest that there was no immediate operability concern. Nonetheless, the licensee was conducting further analyses to assess the validity of the EFW flow limit this The NRR staff will review the plant specific and potential generic aspects of issu *

Failure of the non *O' steam traps used in the steam supply line to the turbine driven pump (located on the turbine skid in the EFW pump room) could alter the normal environment in the EFW pump room. The licensee has provided adequate

}ustification for continued operation until physical changes can be made to remove the steam traps from the room.

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Althcugh the as installed conditions were satisfactory, the team identified design control weaknesscs for specific processes such as field installed raceway supports, instrument tubing, and sensing lines. The team also identified weaknesses in the manner in which design calculstions are processed and supersede *

Entergy had not established a basis for determining design requirements for instellation of instrument tubing and sensing lines that were found to be inadequate supported. The team noted that the originalinstrumentation for ANO 1 was installed on the basis of the constructor's interpretation of the requirements that were not specifically defined by the architect engineer and the nuclear steam supply system supplier. Entergy attempted to analyze and document existing installations as they arise through the modification process; however, Entergy has not yet implemented a program to analyze and document the remainder of the ANO 1 instrumentatio +

The team noted that for the examples reviewed, the layout of the installed electrical conduits differed from the configuration shown on the conduit and tray layout drawings. The team attributed this deficiency to inadequate control of the field-routed conduit installations, il

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Field routing and diagrammatic drawings for raceway installation give the installing craftsmen considerable latitude. As a result, the team noted numerous instances in which installations deviated from design drawings, engineering standards, or approved procedures and make it difficult to assess the adequacy of the raceway and supports for the EFW and DH/LPl system *

ANO 1 does not electrically test certain safety related molded case circuit breakers (MCCBs). AlthouDh required by Unit 2 technical specifications for MCCBs that supply containment feedthroughs, no similar testing requirement exists for ANO 1, The team considers the testing of MCCBs is needed to verify that a safety related breaker will not prematurely trip, impeding its safety function, and that it will open when required. Entergy is establishing a committee to develop a periodic testing or replacement program as a result of an ANO 1 issue identified prior to the inspectio ANO 1 does perform operational ar.d overcurrent trip tests and mechanical inspections for their 600V load center switchgear and also monitors the functionality for all MCCBsin under recommendations IEEE 308the maintenance rule. FSAR Section 8.317 1971 for Class 1E power systems and breaker testin references t The team's evaluation of the DHILP) system tevealed that the system provides acceptable operational performance capability and, as installed and operated, the system meets both the original design bases and subsequent licensing commitments. Through the mechanical review, the team determined that the borated water storage tank is operable as currently configured with a foreign mater.al exclusion covering with a 3 inch gap installed in place of the original pressure and vacuum relief valve in addition, the team determined that the LPl pumps have adequate NPSH available when drawing from the BWST and the licensee's surveillance testing reflects the actual performance requirements of the LPI pump The team's review of instrumentation and controls identified that both the tank lev instrumentation and the DH/LPI flow instrumentation meet the requirements of the FSAR and of Regulatory Guide 1.97. Walkdowns of the DH/LPI system identified that the overall material condition is generally good. The temporary use of lead shielding is adequately controlled by procedure, which requires evaluation of each installation to ensure that no design conditions are exceede The team identified severalissues concerning the design of the DH/LPI system or the licensee's implementation of that design. The following examples highlight the issues identified the by the team during the inspection or by the licensee during their preparation for inspection:

The licensee's initial operability evaluation of the 1 inch gap in a foreign material exclusion covering on the tank flange was invalid because it assumed that no screening was in place, although screening had been installed which reduced the available flow area through the 1 inch gap. Subsequently, the area was determined to be sufficient as compared to that provided by the original pressure and vacuum relief valv iii

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' Engineering Report 93 R 1002 01, ;

concerning vortexing in the borated water storage >

tank, did not account for instrument error which could under predict the impact of air

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I entrainment on the calculated NPSH available for the ECCS pumps. The licensee '

j addressed this by issuing Condition Report 1 97 0039 whic' Includes a positive i

operability determination for the ECCS pumps, l *

l The licensee's 50.59 safety evaluation, conducted as part of Temporary Alteration ,

971001, did not adequately address radioactive seleases from the bWST which

were discussed in NRC Information Notice 9156. The safety evaluation mentioned

only the offsite doses whereas control room doses may be more significantly affected by these release !

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! Overall based on the above findings, the team found the desy, of the two selected system

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to be good, with adequate design margins. Entergv's understanding of the design basis w t- good as was their inspection preparation and their ability to resolve team identified

concerns. The team also considered the EFW/LPl system self assessments, performed prior j to the inspection, to be a positive initiative which resulted in effective corrective actions, The implementation of the design was found to be adequate with some issues noted. The

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team noted that **1e licensee's staf f were well qualified, professional, and especially helpful l

j in responding to the team's questions during the inspection. This licensee staff's attitude

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and professionalism were reflected in ANO's material condition and appearance.

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111. Engineering El Conduct of Engineering E1.1 Inspection Objectives and Methodology The ob}ectives of this design inspection of the Arkansas Nuclear One, Unit 1 (ANO 1)

emergency feedwater (EFW) system and the decay heat / low pressure injection (DH/LPI)

system were to evaluate the capability of the systems to perform safety functions required by their design basis, adherence to the design and licensing basis, and consistency of t as built configuration with the final safety analysis report (FSAR). These systems were selected for review based upon a review of the plant's individual plant evaluation (IPE)

(probablistic risk assessment), previous NRC inspection insights, and modification his .

The inspection team included a team leader from the NRC's Office of Nuclear Reactor Regulation (NRR), an engineer from NRR, and five engineers from the Sargent & Lun Corporation (S&L). The team included two engineers who evaluated the mechanical a of the selected systems, one electrical engineer, one instrumentation and control engi and oneMarch through field 14,199 engineer. The team was on site for three weeks during the period Februa In conducting the review, the team first assembled the design basis and licensing basis for the selected systems. A review was then conducted of the supporting calculations, analyses, and implementing procedures. Finally, in-plant observations and walkdowns of the plant equipment were performe E1.2 Emergency Feedwater System E1.2.1 System Overview The EFW system at ANO 1 is designed to provide a source of cooling water to the secondary side of the once through steam generator (OTSG) to cool the reactor and its coolant system whenever the noimal flow of cooling feedwater to the OTSG is lost. The original system design was significantly revised / augmented in the period from 1979 th 1984 to meet post TMIlicensing requirements specified in NUREG 0737 and NUREG-057 The EFW system has one steam turbine-driven pump, A, and a motor driven pump, B. The turbine driven and motor driven pumps are adequately sized to remove decay heat from the reactor coolant system. The motor-driven pun o is powered from emergency AC " RED" train and the turbine-driven pump is fed with main steam from either of the two OTSGs and its controls are powered from the " GREEN" AC or DC trai Both the turbine driven and motor-driven pumps supply cooling water to both the A and B OTSGs through the two train headers penetrating the containment and the OTSGs.

Upstream of the headers and the OTSGs isolation valve, each pump has separate lines, each with its own solenoid operated DC control valve. EFW flow is supplied to the steam generators through piping independent of the main feedwater piping and via a separate feed ring located near the top of each steam generato . _ . _ . _ .

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The EFW system is initiated automatically by the emergency feedwater isolatio (EFIC) system from the control room on any of the following signals:

Loss of main feedwater pumps

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Low levelin OTSG

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Loss of all reactor coolant pumps Low pressure in OTSG

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ESAS actuation locations outside the control room.in addition, the system includes provisions For long term cooling of the primary system, the heat from the OTSG is removed v atmospheric dump valves (ADV) to the atmosphere or the turbine bypass valves to the OTSG has one ADV. The controls for both the ADVs an nonsafety-relate The primary water supply for the EFW system is maintained in a

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related ("O") condensate storage tank (CST), T 41B, connected to. the The pumps' suction 3 A and B pumps have a common suction line to the CST. Low water levelin the CST w alarm and annunciate in the main control room. The low level set point provides a time for the operator to initiate make up without compromising plant safety

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E1.2.1.1 System Functions The EFW system is designed to perform the following safety related functions:

} accident (DBA) conditions: Provide feedwater to remove decay heat from the R Loss of coolant flow Loss of normal feedwater Loss of all unit AC power or station blackout (SBO)

Main steam /feedwater line break Small break loss of coolant accident (SBLOCA)

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tsolate the EFW steam and feedwater supply lines from the affected OTSG a steam or feedwater line brea +

Automatically initiate EFW flow upon receipt of an EFIC actuation signal within the time frame specified for the most limiting DBA analyze E1.2.2 Mechanical Design Review The mechanical design review consisted of an assessment of plant design transients establish design requirements and an assessment of thermallhydraulic and fluid mecha calculations to determine if the EFW system is designed to remove the required heat l

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in addition, the inspection team reviewed the plant desiga drawings, modificatio FSAR, technical specifications (TS), operating procedures, NRC bulletins, Information

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Notices (IN), generic letters and engineering evaluations associated with the syste E1.2.2.1 EFW System Design Requirements for Heat Removal E1.2.2.1.1 Scope of Review in this portion of the mechanical design review, the team identified the applicable EFW system design requirements to support removal of decay heat from the primary system through the OTSGs to bring the plant to safe shutdown condition E1.2.2.1.2 Inspection Findings Design _ Basis The EFW system for ANO 1 provides an unlimited safety grade backup source of feedwate to the OTSGs, as required, to assure that core decay heat and primary system residual heat can be removed at a rate such that acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded. The EFW system removes decay and residual heat until the plant has been cooled and depressurized sufficiently to permit the use of the decay heat removal system. The EFW system is assumed to provide feedwater for the plant accident conditions previously mentioned in Section E1.2. The loss of normal feedwater flow is the most limiting event for the EFW system. However, this event was not a part of the original plant licensing bases and is not addressed as part of any design basis accident analysis in the ANO 1 MAR. The main feedwater line break has been enveloped by the MSLB in the ANO 1 accident analysi For all of the above plant accident conditions the EFW system will be required to provide water inventory and heat removal capability for secondary side cooling. The water inventory to the OTSG is provided by the EFW pumps from the "O" CST and the service water (SW) system. The heat is removed from the OTSG to the condenser via the turbine bypass valves in the turbine bypass system (TBS), or to the atmosphere via the main steam safety valves and the atmospheric dump valves (ADVs).

ANO 1 is licensed as a hot shutdown plant, i.e., safe shutdown is achieved when the

reactor is subcritical and the temperature of the primary system is 527*F. For ANO 1, the sate shutdown for all design basis accidents is hot shutdown with the capability to remain at hot shutdown indefinitel The EFW system was originally designed as a nonsafety related system with the capability to remove decay heat from the primary system after a design basis accident, so as to achieve and be in hot shutdown at 5270F indefinitely with potential capability to achieve shutdown or decay heat removal system initiation conditions.

General design criterion (GDC) 34 for decay and residual heat removal requires that methods relied upon for residual core heat removal have suitable redundancy in components and i

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features to 6ssure that the safety functiol. can be accomplished assuming the most limiting single failure. At the time ANO 1 was originally licensed, this was not interpreted to require a safety grade EFW system. In the original design, the EFW system was not actuated by a safety grade system not was it fully independent or redundan NUREG 0732_ Modifications Following the accident at Three Mile Island (TMI), a number of new NRC requirements were generated based on NUREG 0737, NUREG 0578, NUREG 0887, Standard Review Plan 10.4.9 and Branch Technical Position ASB 101. In response to these NRC requirements, the EFW system was redes 3gned to make the following enhancements:

Provide two full capacity indept,ndent loops with diverse sources of motive energy of which at least one EFW pump, flowpath, and associated instrumentation can automatically initiate flow and be capable of being operated independently of any AC power source for at least two hour *

Provide feedwater flow from either pump to both OTSG *

Add system capability to provide EFW flow for a high energy line break assuming any concurrent single active failur +

Provide safety grade indication and control to monitor and control system operatio *

Provide a safety grade source of cooling water of sufficient volume to achieve plant safe shutdow The redesigned system was evaluated for all those plant accident conditions where EFW flow is assumed to provide secondary cooling to determine:

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limiting system flow requirements

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time delays for initietion or isolation of the EFW flow to the OTSGs For the redesigned system, the assessment of condensate required for safe shutdown no longer tecame an EFW design / sizing issue as the assured source of safety grade cooling water (i.e., the SW system)is virtually limitless. The EFW system requirements in response to thr., various plant accidents are shown in Table 1. The loss of main feedwater flow event is determined to be the most limitinD transient with respect to EFW flow requirement . _ _ _

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l Table 1. Accident Analysis

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Analysis Concurrent Safe EFIC EFW Volume of Romerks Reference Limiting Shutdown initiation Event Flo w Condensete Name Single Active Condition p ,;g,,,

/ Required g,,,,,;,, !

Time

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l No Each EFW pump and i Loss of FSAR N/A Hot EFW EFW flow assessment associated train is coolant Section shutdown flow not not of amount

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flow at 527*F independently capable !

modeled modeled in of of supplying adequate (occurs 14.1. for in safety safety with loss- ULD -1- condensate flow to maintain plant indefinite analysis analysis required of-offsite TOP-11 period at hot shutdown I power has been condition done as the (LOOP) .

assured '

source of condensate from Service

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Water i system is virtua!!y '

limitless i

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EFW Volume of Romerks

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Loss of all FSAR N/A Hold at Hot 40s-60s 250gpm !

No The turbine-drwen Unit AC Section - shutdown (max.) per SG i

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Power or assessment EFW pump has !

for a EFW _earl Station 14.1. of amount adequate capacity minimum time ' Lower EFIC of Blackout 4 ULD -1- period of 2 required to ma.ntain _ j

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delay controIIed condensate plant at hot shutdown TOP-11 hours with flow for required condition an long term has been expected decay heat j done as the period of 4 removal assured

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[ hours, if 4 source of required condensate from Service

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Ard, sis Cc,rgurrent Safe EFIC EFW Reference Umiting Volume of Romerks Shutdown initiation Flow Single Active Condensate Condition /

Failure Required ;

isolation Time FSAR Single active Hot 40s-60s 500 gpm No Section failure of the shutdown This EFW flow 14.3 , (max.) from assessment motor 4 riven at 527'F EFW establishes the design ULD -1- turbine- of amount requirement flow for

EFW pump for time TOP-11 driven EFW of the EFW system indefinite delay pump perio condensate Loss of required normal has been feedwater done as the flow assured source of condensate from Service Water I

system is virtually limitless

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Analysis Concurrent Safe EFIC EFW Volume of Romerks Event Reference Limiting Shutdown initiation Flo w Condensate Single Actrve Condition / Required Failure Isolation Time Main steam line FSAR Main hot EFW break EFW flow No For the eluration of the Section feedwater Shutdown flow not not assessment safety analysis. the (Also 14.2. isolation at 527'F modeled modeled in of amount envelops ULD-1- valve (MFIV) for OTSG tevel remains in safety safety of Feedwater TOP-02 on affected indefinite above EFIC actuation analysis analysis condensate line break) OTSG fails period level. EFW required required for long term decay to close has been heat remova done as the assured source of condensate from Service Water system is virtually limitless

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Analysis

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Crssosient Safe EFIC EFW t

Reference Limiting Volume of Romerks

, Event Shutdown initiction Flo w Single Active Condensete Nerne Condition / Required i

Failure isolation Time i FSAR Failure of 1 Hot Small 40s-60s EFW flow No Section emergency Shutdown EFW required for long i Break (max.) versus assessment i 15. diesel at 527'F EFW term cooling.

! LOCA ULD- OTSG back of amount for time pressure TOP-01 of indefinite delay assumed in perio condensate analysis required l described has been (

in UL done as the Minimum assured '

flow source of !

required is condensate !64-125 from '

gpm Service l Water '

system is t virtually limitless I t

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Completa LosamLCoolantRow The licensee has performed a transient analysis for a complete loss of coolant flo analysis. A review of the licensee analysis for * Loss of Co 14,1.2.6 showed that the plant design limits are not exceeded. However, for this event to the primary coolant and to remove decay heat. The team that either of the EFW pumps and its associated train is capable of providing th flow; and a limitless supply of cooling water to remain indefinitely at hot shutdown i available from the SW system. The heat generated in the secondary side is removed through either the safety related main steam safety valves (MSSVs), or the non-related ADVs to the atmosphere; of to the condenser, if available, through the TB Loss of NormaLFeedwater_ Flow As stated earlier, the loss of normal feedwater flow event was not part of the orig addressed in FSAR Section 14.3. As part of the post the system requirements for providing minimum flow to the OTSGs e were re exa licensee. The loss of normal feedwater flow is Caracterized as a plant heat up ev to other plant heat up events, i.e., loss-of offsite power and station

. Forblackout reevaluation, a value of 500 gpm at a OTSG pressure of 1050 psig was established as condition of 527'F. conservative minimum flow for the EFW system to bring the plant to a h This value considered a single failure and allowed for pump recirculation flow

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and pump wear. The 500 gpm flow at a OTSG pressure of 1050 psig and a 40 secon delay in initiation of the EFW system was then used to analyze the effect of a loss

{ feedwater from reactor power at 102% of full power. In the analysis, heat from t coolant pump was included and the decay heat was assumed to be 1.2 times A heat . No credit was taken for any anticipatory reactor trip and the reactor was as to trip on high reactor coolant pressure as a consequence o,f primary -

. Heat system heat from the secondary side was removed either to the atmosphere via the MSSVs or to the condenser via the Turbine Bypass valves. The team's review of B&W docum 51 1206919 09 and USAR Section 14.3 showed that the plant design limits were n dose release exceed the 10 CFR 100 limits. exceeded, the reactor coolant The impact of a 500 gpm EFW flow at a OTSG pressure of 1050 psig on oth events was reviewed by the team. The team's review determined that the loss of to the flow requirements for the EFW system, and pro the EFW syste T

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For a complete loss of all AC power or SBO, the EFW flow in conjunction with the steam discharge to the atmosphere from the MSSVs or the ADVs provides the cooling on the secondary side for the primary coolant to maintain the reactor at hot shutdown condition The EFW initiates on low OTSG level and provides coolant flow from the condensate a storage tank to the OTSGs with the EFIC system raising the water levelin the OTSGs at a j controlled rate until the 26 foot natural circulation set point is reached. This enables decay

heat removal by the natural circulation characteristics of the system after coast dGwn of th RCPs. The long term flow to the OTSG is controlled by EFIC. EFIC ensures that the flow is sufficient to remove decay heat and maintain the reactor at the hot shutdown condition.

i The licensee's analysis of this event,in USAR Section 14.1.2.8.4, showed that for this event there is no fuel damage resulting from reduced RCS flow as this minimizes heat transfer from the primary side. The station batteries at ANO 1 provide sufficient power for indication period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and control with to maintain an expected the reactor period of 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months in a hot shutdown condition f Main. Steam /Feedwater Line Break The main feedwater and main steam system piping failures are plant cooldown events. For ANO 1, the safety analysis for the main steam piping failure envelops the feedwater pip failure, and hence, the safety analysis only addresses main steamlino failures. Upon

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detection of a main steamline break, EFIC will be actuated on low steam generator pressure

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initiating closure of the main steam isolation valves and the feedwater isolation valves. The OTSG in the steam loop associated with the failed line blows dry after feedwater isolatio Decay heat is removed by the EFW flow supplied to the unaffected OTSG steaming thro the main steam safety valves and the atmospheric dump valves. EFW flow to the ,

unaffected OTSG is not initiated until the law level set point of 30 inches is reached. EFW flow is not modeled in the safety analysis as the OTSG level teniains above the EFIC actuation level. For conservatism in the cooldown analysis, offsite power is assumed available so that a greater primary to secondary heat removal capability is provided by the ,

reactor coolant pumps and feedwater is modeled to continue to flow untilisolated by EFIC

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on low steam generator pressure. Additionally, a single failure of the main feedwater isolation valve for the effected OTSG is assumed. The team concluded that the safety

- analysis for the main steam line break accident was conservative, and plant safety was not affected even by not considering the EFW flow untillate into the acciden Smalbataak i nce of. Coolant _AccidenLISELOCA)

!

For a SBLOCA, the EFW is required to perform three main fuw. ions:

(1)

Establish natural circulation flow in the primary system following an RCP tri (2) Increase OTSG liquid levels to provide a condensation surface late in the event for boiler /condensei heat transfe (3) . Provide long term decay heat remova . . - . _ _ _ . - -- - . _- - .- . . --

_g,_ypsseMme-44+mAm' *''~

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l The EFW system injects water near the top of the steam generator tubes to increase the length of the cold leg of the natural circulation loop. With this feedwater inject. ion locati the available EFW flow rate is sufficient to promote long term core heat removal by natur circulatio The EFW response time used in modeling the overall secondare system response is 40 seconds, although B&W has provided justification for EFW response times u to 60 seconds (an EFW flow varying with 01SG pressure was assumed in the analysis).

The limiting flow rate of 500 gpm at a OTSG pressure of 1050 psig envelopes the ,

requirement for S8LOC i

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E1.2.2.1.3 Conclusion !

The analysis done by ANO 1 for the plant accident conditions where EFW flow is required for secondary cooling, has demonstrated the capability of the EFW system to achieve pl safe shutdown conditions. The evaluation of these events has demonstrated of feedwater flow event is the most limiting with respect to the flow requirements of the EFW system. The limiting EFW design requirement is to provide a flow of 500 gpm at a steam generator back pressure of 1050 psig within a time delay of 40 seconds, though analysis done by B&W has shown that a time delay of 60 seconds is acceptable. No .

assessment has been done for the volume of condensate required as the assured source since the alternate supply of cooling water to the SW system is from Lake Dardanell E1.2.2.2 Condensate Storage Tank E 1.2.2. Scope of Review Evaluate sizing snd the impact of missile protection of the safety-related ("O") CST T 41B since the minimum water level requirement for the tank is based on it being able to provic. .

in the case of a tornado, sufficient inventory for 30 minutes of EFW system operation before switch over to the SW system is accomplishe E1.2.2.2.2 Inspection Findings Cooling water for the OTSG secondary is provided from the following sources:

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'Q" condensate storage tank, T-418. This is the safety related preferred source of domineralized water to the OTS *

Safety related service water. This water is supplied from the lake and is not the preferred source. However, this is the safety related back up source to che CS The TS limit for the level of water in the "O" condensate storage tank is 11.1 feet or 107,000 gallons (the CST is 42 feet in diameter). Thic quantity is adequate for 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months of operation based on an estimated average flowrate of 390 gpm However, there is an administrative limit to maintain 321,000 gallons or 31 feet of wcter in the tan The existing TS bases for the minimum water levelin the 'O' CST is based on beint, able to provide a sufficient amount of water to bring thu unit to decay heat removal system

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by the licensee to support the above TS bases, in the o suction from the "non-Q" T 41 CST. To meet the requirements of NUREG 0737 a , -

related source of cooling water had to be provided for the EFW system in the red system, a stfety-e6ted "O" T 418 condensate storage tank was constructed, and for the is a safety grarfe sMtem meeting all the requirements system water d

Tha "O" T418 con @nsate storage tank was sdsmically qualified but was only pa protnted fig.'n v toinado. Since unlimited safety related service water is used as the ultimate source of cooling water, a partial tornado pronction for T-418 provides a 30 minute supply of water, which is considered to provide sufficient time for the operator to switch 82 suction D 2986 01, of the EFW pumps from the CST tank to the SW system. ANO 1 calcu operation considering a horizontal missile, and calculation 82 D 208 Revision 00, Hank that creates a 7.2"x12" hole 6* from the bottom of the tan As per ANO 1 ca!rVation 82 D 2986 01, to remove decay heat for the first 30 minutes is about 14,000 gallons and water required to recover OTSG levelis about 7,300 gallons. The total volumt of water

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As per ANO 1 calculationrequired 82 0 2086-01, is about 21,300 gallons, This volume equates to a ta

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n the tank is, therefnre, adequate to provide a 30 minute supply of water as re .

The volume of cooling water required to remove decay heat for the first 30 minutes for

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The combined level of water required for both units for 30 -

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feit. The existing 5 foot high missile shield wall around the tank will supply water for

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units in case of a tornad As per ANO 1 calculation 82 D 2086149, in the condensate tank to remove decay heat for the first 30 minutes plus re level and account for the leakage through the 7.2).12 inch hole created by the vertical missile is abot.t 10.5 feet. The existing TS level of 11.1 feet is, therefore, adequate, to provide 30 minutes of cooling water before the operator can switch suction to the SW system. Also, as per the above calculation the CST level required to remove decay heat for the first 30 minutes for both units and account for the leakage through the hole is abo 22.6 feet. It is however not necessary to increase the TS levelin the 'O' CST as Unit 2 EFW pumps do not take any suction from it. Flow to Unit 2 pumps is isolated by locked closed manual valves. The design bases for the condensate storage tank and the TS bascd on ine volume of water required for decay heat removal for 30 minutes which a sufficient time for the switch over of suction to the SW system (reference NRC letter to ANO 1 dated May 17,1984, ANO File # 1CNA058406). For all safety analyses the SW ,

system is relied upon to provide safety grade water. Adequate procedures are in place and adequate time is allowed to assure that on low levelin the CST, switch over is done to t GW system. The switch over can be accomplished by control switch in the contrni roo

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less than 5 minutes. There is, however, a 3/4 inch tube SW leakoff detection monitor in the SW suction piping that is twquired to be manually isolated once the switch over is complete, Redundant CST levelindication. Iow CST level and low EFW pump suction pressure alarms are provided in the control room. The CST low level alarm set points allow at least 20 minutes for operator action. Since the licensee did not confirm by analysis test that they could meet the TS bases for the minimum volurra in the CST, the licensee proposes to revise the existing TS bases for the minimum volume of water required for the CST and to state that this minimum volume corresponds to the volume that is required 30 minutes of EFW operation before switch over to the SW system is accomplished, in c of a vertical or horizontal tornado missile. This item is identified as inspector Followup 50 313/97 201 0 E ... 2.2.3 Conclusion

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The existing TS bases for the minimum water levelin the "O" CST is based on being abl

provide sufficient amount of water to bring the unit to decay heat removal system initiation

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conditions in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or less No evaluation or performance data could be provided by the licensee to support the above TS base The team's review of ANO Calculation 82 D 2086149 and 82 D 2086-01 determined that the existing TS level of 11.1 feet provides sufficient volume of water to enable switch over to SW system within 30 minutes in case of a vertical tornado missile. For a horizontal tornado missile, the 5.1 foot level of water in the CST protected by a missile shield wall provides adeouate water inventory to enable switch over to SW system within 30 minute E1.2.2.3 Pump Redundancy and Flow Capability E1.2.2.3.1 Scope of Review Determine if EFW system is designed with sufficient drive diversity and flow capability t respond timely and effectively to the various plant accident conditions, to remove decay and residual heat and bring the plant to a safe shutdown conditio E1.2.2.3.2 Inspection Findings To meet the requirements of NRC branch Technical Position AS8101 for pump drive diversity to provide system flow, the EFW system has been designed with one motor driven and one steam or turbine driven pump. The motor driven pump and the associated controls and instrumentation for this train are powered from the emergency " RED" AC and DC sources. The turbine driven pump uses steam from either of the two OTSGs to drive it and the associated control and instrumentation for this train are tied to the emergency " GREEN" AC and DC source The design case for the steam or turbine EFW pump is the S80 plant condition or loss of normal feedwater. The motor driven pump is designed to meet requirements for steam line breaks and has a design flow capacity of 780 gpm at a discharge pressure head of 2600 feet (Reference calculation 88E-0086-01, motor driven pump performance curve). The turbine driven pump has a design flow capacity of 850 gpm at a discharge pressure head of 2825 feet (Reference calculation 88E 0086-01, turbine driven pump performance curve).

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The original design basis for the EFW pumps was to be able to deliver 700 gpm of EFW !)ow to the OTSG at 1050 psig within 65 seconds of system initiation signal. Calculation 80 D-10838102 verified that a flow of 650 gpm could be delivered to the OTSG at a OTSG pressure of 1100 psig which is the highest pressure setting for the MSSVs. Table 1 identifies the current flow requirements for the redesigned EFW system for the various plant i accident conditions. The limiting flow requirement for the EFW system design has been !

determined to be 500 gpm to the OTSGs at a OTSG back pressure of 1050 psig. The original calculation,80 D 10838102 and the current calculation 92 E 0077 04 were reviewed against the original pump test performance curves and the latest quarterly pump surveillance test data to determine the delivered flow to the OTSG. The review determined that the EFW system was capable of providing the limiting design flow of 500 gpm to the steam generators or 250 gpm per OTSG at the 1050 psig OTSG pressur Based on the flow capability of the pumps, as determined from the pump curves, the pressure drops in the pump discharge line to the OTSG and the OTSG back pressure for certain operating conditions, it was determined that the EFW system could deliver flow in excess of 1500 gpm. This recommended maximum limit, which was based on OTSG cross flow velocity, was identified to ANO 1 in a 1991 B&W engineering report (92 R 1019 01). If this crossflow velocity limit is not adhered to, it may cause damage to the steam generator tubes. However, existing plant procedures were never revised by ANO 1 to incorporate such limits. The team concluded that the failure to incorporate maximum EFW flow limits into plant procedures to monitor and preclude exceeding such recommended limits represents a weakness with respect to 10 CFR Part 50, Appendix 0, Criterion ill, ' Design Control,' and is identified as Unresolved item 50 313/97 201 0 A review of the operation log showed that for the ANO 1 transient on May 19,1996 the flow limit of 1500 gpm was exceeded for OTSG B for over a minute and for OTSG A the flow exceeded 1400 gpm for over 0.5 minutes, in response to the team's question regarding excessive flows to the OTSG, the licensee's discussion with Framatome Technologies, Inc. (FTI), formerly B&W, determined that, based on a similar incident which occurred at Crystal River, Unit 3, FTl recommended administrative cumulative time limits over the life of the OTSG for operation beyond 1400 gpm. The following cumulative limits were recommended by FTl:

Below 1400 gpm No time limit

1400 gpm to 1650 gpm 350 hours0.00405 days 0.0972 hours 5.787037e-4 weeks 1.33175e-4 months

. 1650 gpm 36 minutes or 133 minutes if there are no flawed tubes in front of any of the EFW nozzles

Over 1850 gpm 10 seconds (Inspection of the tubes is recommended at the first opportunity if EFW flow exceeds 1850 gpm limit for 10 seconds)

Entergy stated that exceeding these limits does not mean that the tubes have failed, but rather that further evaluation of the specific event is neccesary and based on an evaluation performed by Framatome for ANO, there were no immediate operability concerns. As per

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FTI, the effect of high EFW flow rates can be categorized into wear at the upper tube sh or tube sheet plate, fatigue wear failure due to tube to tube impacting, propagation of existing tube flaws, or exceeding the allowable ASME Code usage factors. All the degradation modes except for fatigue can be detected by an inspection of the OTSG tube During the ANO 1 refueling outage in November I 1996,100 percent of the OTSG tubes were '

inspected. The inspection indiceud no unusual degradation of the tubes in the area of the EFW norrles, it was, therefore concluded that neither the May 1996 event, nor the i

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I cumulative effset of EFW initiations over the course of plant life has caused any measurable detrimental effects on the tubes. Again as per FTl, the tube stresses are below the fa endurance limit for EFW flows less than 1750 gpm, even when existing flaws are considered. Since the maximum flow during the event was less than 1750 gpm and at no time in the operating history of ANO 1 has EFW flow exceeded 1750 gpm, the fatigue life of the tubes in the periphery of the EFW norries has not been shortene Entergy issued CR 197 0081, dated February 28,1997, to review implementing recommendations to document a formal FTl analysis discussing OTSG maximum EFW flow limits specific to ANO 1, and to implement a formal tracking mechanism to evaluate and record excessive EFW flow occurrence E1.2.2.3.3 Conclusion The review of the hydraulic calculations for the flow to the OTSGs determined that the EFW system was capable of providing the limiting design flow of 500 gpm to the steam l generators or 250 gpm per steam generator. However, it was determined that the EFW system could deliver flow in excess of the steam generator tube cross flow design limit of

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1500 gpm. Operation's logs showed that for the May 1996 ANO 1 transient, the flow limit of 1500 gpm was exceeded for OTSG B for over a .ninute and for OTSG A the flow exceeded 1400 ppm for over 0.5 minutes. This flow limitation had been identified in a 1991 Et&W engineering report, but existing plant procedures ( e.g., Procedure 1010.010,

' Unit 1 Transient Cycle Logging and Reporting') were never revised to reflect such recommended limit. Based on an analysis performed by FTl for ANO 1, there were no immediate operability concerns. The licensee has issurd CR 197 0081 to further review the issue and take followup corrective action E1.2.2.4 Net Positive Suction Head E1.2.2.4.1 Scope of Review Determine if sufficient net positive suction head (NPSH)is available for the motor driven and turbine driven EFW pumps with suction from the "O" T 41B CST or the SW syste E1.2.2.4.2 inspection Findings NPSH calculation 82 D 2086 02, applicable to both rnotor driven and turbine-driven EFW pumps, was reviewed by the team. A comparison of the pump performance curves for both pumps indicaten that the NPSH requirement for the pumps (700 gpm at 19 feet) are nearly identical. The NPSH required at 130 percent capacity, i.e., a flow of about 900 gpm, is 30-feet (Reference calculation 88E 0086 01, motor driven pump performance curve and

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calculation 88E 0086 01, turbine driven pump performance curve). The available NPSH for suction from "O" T 418 CST for 700 gpm flow is 39 feet and for 130 percent or 900 gpm flow is 33 feet. This calculation is conservative for it ignores any static head whereas a larger static head is available. Also this calculation assumes failure of a check valve in the suction line thereby increasing the pretsure drop in the line across the parallel check valve At the limiting flow of 500 gpm, because of the reduced losses in the piping, a much NPSH would be available. The SW pumps are located at an elevation of 356.5 feet, which is about 18.5 feet higher than the location of the EFW pumps at an elevation of about 338 -

feet. The maximum normal service water discharge pressure is 90 psig (Reference calculation 88 E 010016). Therefore, the maximum suction pressure available for the EFW pumps is the sum of the static pressure plus the SW pumps discharge pressure which is equal to about 99 psig. Adequate suction pressure is available for the EFW pumps for operation with the SW system. The higher suction pressure with SW operation also increases the operating margin for the EFW pump discharge pressure hea The suction piping configuration used to determine the pressure drops in the pump suction flow to EFW pumps in calculation 82 2086-02 does not match with the current configuration shown in the isometncs (Drawings 3 EFW 1,2,108111,113116,118119 and 12 CON 141 144 The pressure drops in the piping are negligible and are expected to have little impact on the pump NPSH. The need for ANO 1 to review piping configuration dif ferences is identified as Unresolved item 50 313/97 201 03. The licensee is aware of the differences in the configuration changes that have evolved over time, and in their response to an NRC 50.54(fl design basis assessment request, have initiated licensing information request (LIRI L97 0035 to resolve deficiencies related to maintenance of calculations. Also as partconfiguration resolve of their response CR C 97-0058 and 59 were also initiated to implement a plan to discrepancie E1.2.2.4.3 Conclusion Adequate NPSH is available for EFW pump operation for suction from either "O" T-418 CST or from the SW system, in a corrective action plan to revise calculations to resolve configuration discrepancies, ANO 1 has initiated LIR L37 0035 and CR C 97-0058 and 5 E1.2.2.5 Piping Design E1.2.2.5.1 Scope of Review in this portion of the inspection, the team reviewed the system piping design, class boundsries and safety class break E1.2.2.5.2 Inspection Findings The EFW system was originally designed as a nonsafety related system, and hence in the original design, except for the interface with containment and OTSG penetration, all the piping was designed to conform to the requirements of ANSI B31.1. The piping penetrat the containment and the OTSGs was designed to ANSI B31.7 Class 11. Also a portion of the later designed piping from the "O" CST to the pump suction was designed as ANSI C31.7, Class 11 The safety-related portion of the EFW pipmg is all termed as " critical" and has

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been seismically qualified to withstand both operating and nvironmental stresses. A review of ANO 1 calculation 88 E 010016 for piping temperature and pressure showed inconsistencies between the pressure and temperature values identified in th' body of the evaluation and the final values summarized in the tables at the end of the evaiuation, e.g.,

the maximum pressure at EFW pump "A" discharge at maximum punip speed for suction from the SW system is about 1932 psig. which is the value shown in the calculation y summary table, whereas the value in the body of the evaluation shows 1850 psig, which corresponds to suction from the "O" CST, but is not the maximum value for pressure determination. The team noted that this and other differences were not addressed in the evaluation. The licen..ee's failute to develop adequate piping pressure and temperature specifications is identified as Unresolved item 50 313/97 01 0 The licensee initiated ER 973848 to review the above calculation to determine the optimum method of confirming discharge l'iping design values, and also to review the consideration of speed control calibration tolerances with respect to determination of the discharge pir ng design pressures. The discharge pressure during overspeed test of the "A" turbine will i

exceed the maximum design pressure for the discharge piping. Therefore, during overspeed testing of the turbine, ihe turbine / pump coupling in disconnected.

l ANO 1 calculation 88 E 010016 and the piping list in the Safety issues Management System (SIMS) were reviewed to verify the GFW system operating temperature and pressure. The following differences were noted between the calculation and the piping list in SIMS:

Dif ferences exist between identification and description of lines in SIMS and P T calculations, e.g., line TCD-001, " pump bearing drain line," as identified in calculation 88 E 0100-16, Revision 1 dr.es not appear in the p; ping list /SIM *

Line GCD 004 is identified in piping list /SIMS as a " morpholine add line," whereas in calculation 88 E 010016 it is identified as a "P7A Lube Oil Cooler Line."

SIMS does not agree with the P T calculation regarding design pressure and temperature for nearly all of the EFW line The first two discrepancies identifi6d above were self identified by ANO 1 and as a result, IRF 8167, dated October 9,1996, .vas initiated to correct the discrepancies. The licensee has directed that until completion of this project, the P.T calculation should be used for design data instead of SIM E1.2.2.5.3 Conclusion The team determined that the safety related or " critical" portion of the EFW system piping is designed to meet the seguirements of ANSI B31.1 o, ANSI 831.7, Class 11 and Ill. However, the consideration of speed control calibration tolerances with respect to determining the discharge piping design pressures needs to be further reviewed by the licensee. The licensee has initiated ER 973848 to determine the optimum method of calculating discharge piping design values and to resolve differencas between the P T calculation and the piping list in SIMS,

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E1.2.2.6 Environmental Qualification E1.2.2.6.1 Scope of Review Determine the effect of the failure of steam traps located on the turbine driven pump skid on equipment performance in the EFW pump roo E1.2.2.6.2 inspection Findings The steam supply piping for the EFW turbine driven pump has four steam traps, two high pressure and two low pressure, located on the pump skid in the EFW pump room. The steam treps ensure that water does not build up in the steam supply piping. As described by P&lD M 204 and verified through walkdown the steam traps vent directly to atmosphere

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into a common floor drain located directly bent ^ the missile shisid wall at the base of the motor driven pump in the EFW pump room. The steam traps are classified as nonsafety-related and were not analytically or otherwise seismically qualified. Failure of the non *Q'

steam traps or pipirig downstream of the traps could result in the continuous blowing of steam in the EFW pump room when the EFW system is operating. For ANO 1, both the motor and turbine-driven pumps are located in the same room. The increased temperature and humidity resulting from the blowing steam could adversely affect operation of both EFW trains. This problem was self id ntified by ANO 1 prior to the inspection and CR C 97 0048 was initiated to take corrective action. The operability assessment determined that the EFW pump and pump room equipment remal...J operable based o'1 the following conditions:

One of the high pressure steam traps currently isolated and not contributing to the steam blowdown, remain isolated until further additior:al evaluatio *

The licensee removed Door 187, which isolated the EFW pump room, as part of this corrective action, as documented in a plant change reques *

A preliminary calculation, using the conditions identified , indicated that the temperatures produced from the blowdown would not exceed the maximum room

- design temperature of 148' On the basis of the review of calculation 86 D 1026 07, the team noted that the steady-state temperature in the EFW pump room after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with both pumps running was 159.5'F, while it was 151.4'F with only the motor driven pump running and 126.4'F with only the turbine-driven pump running, in this calculation, the door to the EFW pump room was considered closed, and this calculation did not assume any failure of the steam trap The limiting continuous operating temperature for the neotor in the motor driven pump is 148'F and thic constitutes the limiting deeign temperature for tbc room. This limiting long-term temperature of 148'F is associated with the pump motor sleeve beerings. Until the door was removed, the plant had been operating beyond the design bases for the EFW pump rooms. The team also reviewed calculation 87 E 0026-09, which the licensee recently completed as part of the followup actions to Cft C 97-OO48. This calculation considered the existing plant configuration (i.e., one high pressure steam trap isolated and

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Room 38, Door 187, removedh and assumed that both pumps are in operation for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the failed trap gets isolated by operator action at the end of the first hou The maximum peak temperature of 149.5'F occurs for a short duration of 8 minutes when one of the high pressure steam traps is assumed to failin the full open position. The steady state temperature calculated for the roorn is 134'F. There is ne, meacuraible increase in room pressure due to the steam trap failure. The steam trap failure wil! cause relative humidity in the room to approach 100 percent, thereby resulting in considerable ,

condensation on cool surfaces in the roo Electrical components in the room considered vital to the operation of the EFW system, Westinghouse pump motor, Rosemount flow transmitters, Limua:que MOVs, Target Rock flow controllers and valves, the turbine control panel, the governor valve and the magnetic speed sensor, have al, previously been shown qualified to HELB conditions that were originally postulated in Room 38 as a result of a main feedwater line break in Room 77. As per Attachment 2 to CR 191304, a maximum bearing temperature of 220'F has been determined acceptable for short duration for the Westinghouse pump motors. A walkdown of the EFW room was conducted by the team on March 11,1997, of sealed enclosures to ingress of steam / moisture. From the enclosures inspected it wast concluded that these were adequately sealed to prevent any moisture ingression. Hence the effect of increased humidity in Room 38 from failure of the steam traps would have no significant impact on equipment performanc Since the environment in the room changes significantly from the normal environment, the Class 1E electrical equipment located within should be considered under the plant's EQ program in accordance with the requirements of 10 CFR 50.49. The licensee at one time considered the e.quipment in the room under its EO program but subsequently removed such equipment from consideration. The licensee stated that the above justification for continued operation is an interim measure enly until physical changes to the plant configuration can be made that will eliminate the concern of potential steam trap failures in Room 38 altogethe Currently one of the high pressure traps ST 75,is isolated and tagged out. The need for the licensee to perform an evaluation under CR C 97 0048 to justify the current design basi for the ANO 1 EFW pump room is identified as Unresolved item 50 313/97 01 0 E1.2.2.6.3 Conclusion Failure of the non *Q' steam traps used in the steam supply line for the turbine-driven pump as located on the turbine skid in the EFW pump room, will cause the environment in the room to change significantly from the normal environment. Justification for continued operation has been provided by ANO 1 as an interim measure until physical changes are made to the plant to remove the steam traps from the room. The licensee stated that an '

assessment will be performed to justify past operation for the EFW pump roo E1.2.2.7 Design Configuration Control E1.2.2.7.1 Scope of Review in this portion of the EFW mechanical design review, the team evaluated the effectiveness of the licensee's design configuration control practices and procedure _ _ _ _ _ _

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E1.2.2.7.2 inspection Findings The team's findings for this portion of the mechanical design review focused on inconsistencies in the valve position indicated on the P8;lDs as well as inadequate configuration controls in the licensee's handling of design calculation E 1.2.2. 7.2.1 Valve Position Indication on P&lDs The team's review of P&lD M.204, Revision 27, showed an inconsistency in valve position indication for the motor operated OTSG isolation valve. The position of the valve was indicated as "CLOSE" instead of "OPEN" as per ANO 1 convention that valve positions depicted are typica! power operation (Mode 1). The team noted that the error was self-identified by ANO 1 and corrected in Revision 2 The guidelines formerly used by ANO 1 to determine valve positions was the valve line during plant start up. During start up when the primary temperature is less than 280'F these valves are closed and are required to be openad when the primary temperature

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exceeds 280'F. AND 1 stated that the error was made as part of corrective actions in response to findings identit'ied in a 1986 NRC inspection. The licensee expressed that there may be a small population of valves whose position may need to be changed from their start-up position and these may currently be shown in a position that may not be consistent with their position during Mode 1 power operation. A review of the start up procedures identified a small population of valves that were required to change position from > 2%

I be made:to 100% power. As a result, the licensee stated that the following changes would power

Revise the convent!on to state inat " valve positions depicted are typical full power operation (Mode 1) "

Review the heat up and power escalation procedures on both units to identify those valves tnat are manipulated af ter initial syctem lineup and confirm that the positions indicated for these valves conforms to typical full power operation position *

Revise P&lDs to change valve positions for valves identified above whose positions are not consistent with the revised conventio E 1.2.2. 7. Design Calculations The review of the design calculations were classified into the following categories: Des calculations that do not reflect existing plant configuration, and design calculations warranting enhancements. For the EFW system, the following design calculations were identified that do not reflect existing plant configuratinn:

Calculation 82 D 2086 01, " Volume of T 418 Requiring Tornado Missile Protection,"

indicates the tank design as being 2 feet underground whereas the as built tank is designed all above groun . __J

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L Calculation 82 D 2086-02 " Determination of Pipe Size and NPSH for EFW Pump 7A and 78 Suction from CST T-418," evaluates the pump NPSH based on piping configuration that is different toda The following calculations were identified that warrant enhancements:

Calculation 88-E 0086 01, "lE Bulletin 88-04 Review for '*/A and P78 minimum flow evaluation," contains a discussion about the effects on minimum recirculation of operating both EFW pumps in parallel. The calculation accurately recognizes the fact that the motor driven pump recirculation flow will be less than the turbine driven pump when both pumps are operated in parallel, since the turbine driven pump has a higher performance curve (due to its higher rotational speed); however, the evaluation is not adequately documented. The team's calculations confirm that the overall conclusio existing calculation is correct, i.e., the motor-driven pump minimum recirculation flow will not be below the minimum design value when both pumps are operating in parallel. ANO 1 initiated ER 973874 to revise the calculatio .

Calculation 88 E 010016. "P T Calculation for Unit 1 Emergency Feedwater System," (discussed previously in section E1.2.2.5).

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Calculation 92 E 0077 04 for EFW pump performance requirements credits a level oi 11.1-feet in CST T-418 (which corresponds to the minimum TS level) to detcimine the pump developed head. However, in the case of EFW actuatian due to LOOP following a tornado, the level of protected volume in tank potentially available corresponds to a tank level of 5.1-feet. The most challenging pump parformance requirements would occur at the tank level of 5.1-feet, however the pump performance requirement was not determined at this level in the tank. Although the evaluation performed by the licensee is not conservative, sufficient pump capacity is available such that the pressure drop due to this immediate act change io in height (< 3 psig) is not significant ewuch to warrant any

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Calculation 92-E-0074-04 supersedes calculations 80-D 1083B-102 and 80-D-1083B-102A, but the latter calculations were not voide In response to the 50.54(f) design basis assessment request, the licensee initiated LIR L97-0035 to resolve deficiencies related to the maintenance of calculations. Also as part of their response, the licensee initiated CR C-97-0058 and 59 to implement a plan to resolve configuration discrepancies. The team did rsot identify any operability concerns associated with changes to these calculations. The need to revise certain drawings and calculations to accurately reflect existing plant configuration represents a weakness with respect to 10CFR50, Appenda 8, Criterion 111, * Design Control,' and is identified as Unresolved item 50-313/97-201 0 l l

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E1.2.2.7.3 Conclusion The inspection team reviewed many calculations and other related design basis documents to determine the design bases for the system. Some of the calculations and design bases documents were redundant or superseded by later documents. The licensee initiated several corrective action documents to resolve the issues identified by the tea E1.2.2.8 Service Water System E1.2.2.8.1 Scope of Review in this portion of the mechanical design review, the team sought to verify that the SW system is capable of supplying the assured source of cooling water to the EFW syste E1.2.2.8.2 !.1spection Findings The SW system consists of two independent flowpaths, Loop ; and Loop 11, providing ~

cooling water and an alternative EFW supply to two 100 percent capacity trains of safety-related equipment. Three air cooled SW pumps (P4A. F4B, P4C), each rated at a nominal flow rate of 6315 gpm at 170 feet, with submerged pump suction in individual bays, are physically located in the ANO 1 intake structur The ANO 1 containment EQ is based on a maximum SW temperature of 95'F with the lake available. However, for accident conditions (i.e., LOCA), the maximum temperature in the

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emergency cooling pond (ECP) may be as high as 120'F. Consideration of the SW-l temperature of 95'F is under review by the NRC.

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'During the last several years, the ANO 1 SW system has been subjected to extensive assessment by the licensee's staff NRC inspection teams, and ANO 1 contractors. As a result, the licensee has implemented several modifications and significant improvements in the system's operating performanc As part of this inspection, the team reviewed the most recent assessment performed by ANO 1 (March 1994) as well as previously performed NRC inspections. In reviewing these documents, the team sought to identify items developed in those inspections and assessments that may affect the design and operability of the EFW system. A concentrated effort has been completed by the licensee regarding the ECP and its capability to provide a 30under is day review post transient by the NR water supply, assuming a loss of lake event. The capacity of the ECP The licensee has en ongoing service water integrity program (SWIP) which continually evaluates the system's performance and implements corrective and predictive upgrades to the system and components. Examples of the SWIP are coating the intake bay's structure with epcxy to preclude Zebra mussels attaching to the bays, (currently being implemented in tha circulating water "B" bay), remove silt and debris from the Lake Dardanelle intake bays and intake canal, (currently in the preparation phase coordinating with the U.S. Army Corps of Engineers regarding the discharge of the dredging operations), develop more reliabic screen wash and debris removal system, improve the chemicalinjection systems for

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service water application, evaluate service water pump strainer design, and service water system crossover and boundary valve replacement, which is an ongoing progra Severalidentified system conditions causing water hammers, one which occurred during bi-weekly required testing of the reactor building (RBi coolers, has been resolved by implementation of design changes. The problem developed due to the rapid operating times of the cooler isolation valves, ::ausing significant pressure cycles (water hammer), A similar condition could occur following a LOOP event that allowed the coolers to drain because the inlet isolation valves were open and failed open on the event. When system operating conditions were restored, the partially drained cc,olers were subjected to a rapid refilling with a subsequent pressure surge in the system. The design replaced the existing fast operating inlet and outlet isolation valves with valves having sfc'wer operating times. To protect the cooler from thermal binding, the design also employed a bypass line with the manual and solenoid valves and restricting orifice across the inlet isolation valve. Another water hammer source was in the SW and ASW discharge lines, which has been resolved by creating air gaps in the lines to preclude pressure surgea following a LOOP event restoratio Two vacuum breakers were added to the ASW system to provide air cushions to minimize pressure surges following a LOOP The licensee has indicated that these modifications have eliminated the occurrences for water hammer during testing and concerns for water hammer surges following a LOOP even With respect to periodic flow testing from the service water system pumps to the EFW pump suction, the licensee stated that full flow testing through the EFW pumps P7A/B is not performed. An operation supplying service water to the EFW pumps would comaminate the EFW lines with take water and require extensive cleanup following suen a test seqeenc However, performance of OP 1106.006, Supplement 6, requires the qualification of service water system flow rate of 520 GPM from the pumps discharge through the interconnecting lines through a normally capped tap off, and directs the flow to the Unit 1 discharge fium This performance verifies deliverability of SW to the EFW system through check valves SW-11&13; up to the motor-operated isolation valves CV 2805 & 2806 (isolation between the service water and EFW pumps common suction line). This test verifies the design requirement for adequate service water system redundant supply to the EFW following depletion of the "O" CST T418, and the non 'O' CST 4 The team verified that the service water system isolation valves and check valves interfacing with the EFW system are properly tested to ASME Section XI requirements E1.2.2.8.3 Conclusion The SW system is designed, tested and performs in a manner consistent with the requirements of the EFW system alternate unlimited supply of cooling water to the OTSGs from Lake Dardanelle in the event of an accident or off-normal operational conditio E1.2.3 Electrical Design Review The team assessed various documents including the FSAR, upper-level documents, procedures and drawings. The implementation assessment consisted of field inspections of the power supplies, raceways, and cables supporting both the EFW and LPI system i s

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E1.2.3.1 Emergency Diesel Generator Loading E1.2.3.1.1 Scope of Review in this portion of the electrical design review, the team sought to determine whether EDGs K4A and K4B have sufficient capacity and capability to supply the power required to maintain the plant in a safe shutdown condition during postulated accident E1.2.3.1.2 Inspection Findings The EDG will normally be in a standby condition ready for a start signal. The lube oil and jacket water in the engine will be preheated to standby temperature with the turbocha at room temperature. During the first 3 minutes from initial start to rated speed, the engine capability is rated at 2650 kw due to EDG turbocharger loading (The turbocharger takes heat from the exhaust and begins to heat up).

For the first 3 minutes of operation from a

" cold" start, the turbocharger is driven by the engine gear train and requires om the power fr engine output. Af ter three minutes of operation, the turbocharger is hot enough and rota free from the gear train driven by the exhaust gas, in addition, the following ratings apply to the diesel generators at ANO-1:

Actual Continuous Rating 2600 kw Service Rating 2750 kw

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2000 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months (83-day) Rating 2850 kw 168 hour0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months (7-day) Rating 2950 kw 4-hour Rating 3000 kw 30-minute Rating 3050 kw The use of 2750 kw for a load limit foi the accident events is conservative s duration of the accidents is assumed to be 30 days (720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months ).

The licensee performed calculation 90-E-0062-02 to provide the basis for the large electr loads associated with the 4160-Volt buses, as well as selected loads from the 480-Volt .

buses. In general, these loads are associated with ESF pumps and various other equipm credited in the safety analyses. The calculation specifically examined for three time intervals after the EDGs start, including 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (short time),2 to 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months (medium time), and above 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months (long time). Events considered were MSLB with LOOP large ,

break LOCA with LOOP, SBLOCA with LOOP, and LOOP onl In a second calculation (86E 0002 01), the large electrical loads derived in calculation 0062-02 90 E-are combined with other loads to tabulate the " steady state" loading profiles for the l'.4A and K4B diesel generators. The load lists included motors, battery chargers ,

inverters, transformer losses, and cable losses. Motor-operated valve and other loads were not considered due to their short operating duration (less than five minutes). The totalloads for EDG No.1 " RED" are:

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MSLB LOCA' SBLOCA LOOP O to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2669 kw 2665 kw 2665 kw 2282 kw 2 to 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months 2411 kw Beyond 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months 1477 kw 2489 kw 1996 kw 1521 kw 1448 kw 2065 kw 1739 kw The short time MSLB, LOCA load value is shown to exceed the three minute load limit (2 and SBLOCA events. Totalloads for the EDG No. 2 " GREEN" did not indicate any overloads, in a the third calculation (92 E-003-01), the licensee performed the EDG transient loadin analysis. The purpose was to calculate the voitage and frequency dips for the EDGs when ESF loads are sequenced onto the diesel and to determine if the load sequence profile is acceptable and complies with RG 1.9 (1978). The stated purpose was limited to the portion of RG 1.9 related to frequency and voltage excursions and recovery. The !oad sequence profiles used in this calculation were based upon the MSLB 0 to 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months loads from calculation 86 E-0002 0 included in this calculation. All MOV loads that receive any automatic art signal were The transient loading analysis indicates that a number of loads will not be connected to the EDG until after the initial 3 minutes of operation. It indicates that there are only two load applications during which the starting load exceeds the " cold" diesel rating of 2650 k These take place at 35 seconds when the EFW pump P7B starts and at 65 seconds when the RB cooler fans start. These applications result in brief dips in diesel speed and generat frequency. For the 35 second load, the drop is calculated to be 2.84 Hertz (Hz) which is within the limit of 3 Hz. The 65 second load drops the frequency 2.67 Hz. After the motors

"pullin", the rated frequency will be restored. Since the 2650 kw rating is the load that the diesel can reliably accept in one step, the brief excursions are not of concern. The inrush caused by motors starting at 3 mmutes after EDG initiation causes the load to peak at 2827.5 kw after which it settles to a 2682 kw running load. Additional motor starts at 10 minutes peaks the load at 2984.5 kw and then levels to run at 2800 kw. These peaks cause small very brief dips in the frequency of 0.2 Hz. and 0.3 Hz. The load profiles show that, for allloads, the voltage transients are well within the limits specified by RG 1 E1.2.3.1.3 Conclusion The review of the calculations provided indicates that the EDGs have adequate capability supply 0002 01 power for the design basis accident loads. The next revision of calculation 86-E-discussedis planned abov to clarify the first three minute load profile and the transformer loading E1.2.3.2 Protection Coordination E1.2.3.2.1 Scope of Review The inspection team evaluated the design criteria, calculation methods, and settings for protective devices associated with the EFW and DH/LPI power supplie r -

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E1.2.3.2.2 Inspection Findings The general criteria requirements for the engineered safeguards protective device provided in calculation 84-E-0083-00 The applications include medium voltage systems 480 V load centers. 480 V MCC's,120-VAC and 125 VDC panels. The protective cr

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are fairly general. A clearer understanding of the protection coordination is gained reviewing the individual protective relay and breaker calculations. The team reviewed setting load calculations for the EFW pump, the decay heat pumps, and the 4160 to 480 center in the section of 84 E 0083-001 covering the 480-Volt load center transformer protection .

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O This time delay over current (TDCQ device trip n. point is full-load current (FLC) of the steaay state loads plus the locked rotor currents (L inrush must be for motors that are starting coincidentall Therefore, the team concluded that the TOOC pick up setting E1.2.3.2.3 Conclusion The protective devices for the 4160-V breakers and the 480-V load center breakers supplying correc the EFW and DH/LPl systems are demonstrated by criteria and calculation E1.2.3.3 Electricai Modifications E1.2.3.3.1 Scope of Review translated into actualinstallations by performing field e packages and procedure E1.2.3.3.2 Inspection Findings The team selected a representative sample of six cables (three each from EFW to provide an organized approach to the field walkdowns. To further diversify the sa two of these were power cables, two were control cables, and two were instrumentatio

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cables. Three of the cables terminated in control or instrument panels, one termina 4160-Volt switchgear, and two terminated in MCCs. In addition, four were insta of the original plant construction and two were installed more recently. On the basi sampling walkdown, the team observed that fire stops in wall and floor penetrations are identified by unique numbers Genciled near the openings. The cable routings in th Data Management System (PDMS) do not presently include these fire stop identifica database routings. This will be an aid for future cable ins installers that special fire procedures (including fire watches and qualified repairs) w 27 \

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. 4 required. The team noted that the limit switch used on the P7A EFW pump turbine t'p and throttle valve is a Honeywell Model LSP1 A micro switch. The licensee stated that none of the switches has a safety related function and that they would not be replacing them with Namco EO limit switches since they only operate alarms or localindicating light The team requestel the cable fill and acceptability of the conduits and cable t.ays associated with the power and control cables in the sample. There were three cable tray sections and two conduits overfilled and the remainder were within the allowable ranges. In accordance with the ANO 1 FSAR, the licensee provided engineering reviews of each of the five cases. The analyses were reasonable and no corrective action was necessar The team noted that conduits attached to the wall of the EFW pump room, drawing E 661, containing redundant cables were supported on the same Unistrut support. The team questioned the independence of the safety related raceways. The licensee stated that because the supports are designed to withstand seismic design loads and are passive, redundant trains may be supported from a common support from a seismic standpoint provided Appendix R and other criteria are addresse During the walkdowns, the team noted that the layout configuration of the installed conduits was somewhat different than as shown on the Conduit and Cable Tray Layout drawings. The licensee stated this was due to conduits being field routed. There were no drawings or records of the seismic support details that were installed. The installation procedure for raceway systems (6030.112) states that the " layout drawings are diagrammatic in form. It shall be the responsibility of the craft supervisor to supplement the layout drawings with additional details to support the installation of the raceway system "

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The seismic raceway support procedure (6030.113) recommends that a proposed conduit installation should be walked down for the selection of supports and locations. These words are followed by a note stating that "it is suggested that a simple single line sketch be developed during the walkdown with dimensions shown for future use and reference."

Pre approved supports are identified in procedure 6030.113, drawing E-2080 " Class 1 Conduit and Cable Tray Supports, Notes and Details." and the original Unit 1 drawing E-59 During the walkdowns, a number of installations were noted which were not in accordance with the raceway procedures 6030.112 and/or 6030.113. Because of the lack of installed support records and the diagrammatic nature of the layoth drawings, it could not be confirmed that engineering reviews of the discrepancies had been performe The quantity of non-standard support details and deviations from procedure or drawing requirements found during the walkdowns demonstrates that the present field routing practices represent a lack of design configuration control. Modification procedure 6030.005 discusses a "constructability walkdown" involving the Modification Engineer, Modification Supervisor, Craft Supervisor and Quality Engineer. In a discussion relating to this issue, the licensee stated that they would document the decisions arrived at during the

"constructability walkdown" by including these in the form of work steps in the contmiled work package (CWP). This willinclude conduit routing and seismic support detail choice When documented as sign-off steps, this will give clear direction to the craftsmen for correct installation and to the quality controlinspectors to assure compliance to procedure requirements and approved designs. The need to revise the modifir ation procedure to

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address field routed installations is identified as inspector Followup Item 50-313/97 201-07 .

The team questioned the lack of certain identifications on the. These layout drawings level x", * contaminated area *, " hazard area", " limited haza ,

the layout drawings have no labels for these. The complia identification labels on drawings which the designer has in handpro this information is contained in the computer data base and does not agree that. The lice identifications on drawings are necessar E1.2.3.3.3 Conclusion The field walkdowns of the six cables sampled demonstrated that there is insufficien the work procedure to incorporate documenting a on and inspe This revised procedure will be reviewed to ensure that adequate control is achieve E1.2.3,4 Cable Design and installation

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E1.2.3.4.1 Scope of Review Review and DH/1.Plthe adequacy system of the procedures and practices for installing the cables in the E1.2.3.4.2 Inspection Findings During the review of cable installation procedure 6030.109, the team identified an issue concerning maximum cable sidewall pressure and jamming limitations as specified b vendors of the nuclear grade cable. The licensee issued a procedure change request March 7,1997, to include guidance and restrictions to prevent exceeding the sidewall, pressures and to prevent jamming, it appeared that other types of cable vendor data are obtain their data on diameters, weights per hundred feet, m ,

minimum bending radii during pulling (dynamic), minimum bending radii during tr (static), and the maximum sidewall pressures. After compiling the data the lic

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ensee will include the most conservative values in the cable information attached to procedure 6030.109 or place it in the PDMS database with the necessary notes in the procedur .

As part of the discussion about maximum pulling tsnsions, the licensee was asked to bends and changes in elevaticn. The calculations included and pulling tension progression through the various conduits and bends.esProcedure demonstrated their ability to correctly perform the pulling calcu

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Procedure 6030.112 for the installation of raceway includes a table of conduit bend or radii f manufacturer's standard bends but does not provide radii for long radius (large-swe segmented) conduit bends. The licensee was asked which cables require long radius as a result of the minimum bending radius and maximum sidewall pressure restrictions during pulling. The licensee prot <ided a calculation which compared the cable's m bending radii (dynamic) to the bend radii of the conduit that is appropriate for a with a filllimitation of 53% of the interior area. The team questionede the values sho the radii of the inside surface of the conduit bends. The accuracy and maximum cable bend radii were also questione ers of the cable diam the calculation accordingly. The preliminary meter calculati cables will require long radius bends. A preliminary check of the PDMS records indic that no pulling safety-related (dynamic). cables had exceeded the limit of minimum bending radius It was noted during the fictd walkdowns that the original design employed long sweep conduit bends for large cables (e.g., DHR pump motor feed ca The need to obtain additional design information from the cable vendor to perform .

cable pulling calculation is identified as inspector Followup Item 50-313/97-201-0 E1.2.3.4.3 Conclusion in general, the cab'es for the EFW and DH/LPI systems are adequately design d e and installed. However, the licensee agreed to contact cable vendors aato obtain their d and upon receipt of the information, revise the cable pulling calculation which compa cable's minimum bending radii (dynamic) to the bend radii of the condui E1.2.3.5 Wiring and Cable Termination E1.2.3.5.1 Scope of Review in this portion of the electrical design review, the team evaluated the adequacy of the licensee's cables procedures of the EFW and practices for wiring panels and devices and for termin and DH/LPI system E1.2.3.5.2 Inspection Findings There were several discussions regarding procedure 6030.110 for the termination of for each termination category, but are not grouped by the (e.g., medium voltage, low-voltage, coax instrumentation, thermocouple , panel and wiring or high potential tested or when certain complex terminat inspected during the termination process due to critical steps or portions on of the termi an in-process inspection is required and the procedure all However in power circuits, these terminals have a history of loosening over long time

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the issue by providing to the team pcrtions 1412 054 of maintenance proc 1412.057,1412.061. 2307.008, 2307.022, and 2412.074,

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wl.ich require re-tightening of

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review of the procedures indicated that the issue .

.A was adequa During the field walkdowns, several control panels and severalinstrument panels w examined, including panels C09, C18, C27 and C543 for the EFW and DH/LPl syst wiring and configurations appeared to agree with the wiring diagrams. The require the terminating procedure were generally adhered to. Field cables a were supporte strain was not applied to the terminals, except for one or two conductors in panel C18 The strain on the terminal for the unsupported cable was not very excessive as t

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out. The design basis separation criteria were being complied with, including flex:ble conduits installed where sufficient distance separation was not possible. Cable related cables associated by routing with safety division that associated cables are not labeled as such in thecable listing field but are not E1.2.3.5.3 Conclusion installations indicate that the connections are adequately .

E1.2.3.6 Raceway and Raceway Support Installation E1.2.3.6.1 Scope of Review in this portion of the electrical design review, the team performeds and field examination reviewed work packages and procedures to determine if the licensee is effectively the EFW and DH/LPl systems into actual installation ports of translating desi E1.2.3.6.2 Inspection Findings During the field walkdowns, many examples of raceway installations were observed were not in accordance with drawings, engineering standards or approved procedures Examples included conduits supported from conduits, breathing and instrumen

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supported from electrical raceway, HVAC ducts suspended from cable tray hangers, conduits supported from the rungs of ladder-type cable tray, and many more. During inspection, Entergy produced several calculations that demonstrated acceptability configurations. The team requested a seismic analysis of one safety-related installation (calculation no. 97-E-0015-01). This was a trapeze conduit hanger of which one end was suspended by a 1/2-inch threaded rod from the rung of cable tray EB215. The location w at the west end of MCC B61 on Elevation 386'0" (drawing E-686: "C" line between "1" a

"2" lines). The cable tray rung is P3300 Unistrut channel and suppcm eleven conduits including two 4-inch, three 3-inch, five 2-inch and a 3/4-inch diamete, . The

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conduit threaded rod was connected to the cable tray rung with a 1/2-inch as Unistrut a

nut which h maximum allowable pullout of 2000 pounds, which exceeds the 862 pound load from th intent of existing seismic requirements and is acceptable." conduit an

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Another analysis was requested for a safety related conduit with flexible seismic supports on a threaded rod which was installed near rigidly seismic supported conduits. The licensee was asked if they have criteria for evaluating the swing angles or interaction between rigidly supported raceway and flexibly supported raceway. The installation involved the '4 inch safety related conduit for the power feed to the DHR pump P348 on elevation 317'0" at "C" line between lines "3" and "4" (drawing E 666). The analysis concluded that the movement of the 4 inch conduit during a seismic event would not jeopardize the support for the 1-inch and 3/4-inch conduits, which was about one inch away. The licensee stated that current standards and procedures dealing with design and installation of conduits do not mention nor require consideration of swing angles with adjacent conduits of structures. The current procedures do not prohibit the routing of a new rigidly supported conduit inside the swing envelope of existing flexible seismic support installation In the Support Devices paragraph of procedure 6030.109, it recommends that " cables be supported in vertical trays with cable ties ...". The licensee was asked if radiation qualified cable ties were required in areas of elevated radiation. Drawing E 59 requires that " Cable grips shall be used for supporting cables on vertical risers of more that 20 FT." The licensee stated that the use of Tyraps is to keep a cable tray neat. The licensee also stated that the radiation in the vicinity of the Tyraps is not sufficient to degrade the strength or cause embrittlemen They concluded that Tefzel cable ties are not needed at ANO 1. It is noted that radiation resistant insulated lugs are required by the termination procedure. The team noted that the Support Devices paragraph is incorrect and misleading since it describes the use of cable ties and does not describe the required use of cable grips for suppor The need to review the Support Devices paragraph of Procedure 6030.109 and the followup to the resolution of CR-C-97 0105 is identified as inspector Followup item 50-313/97 201-09 .

The licensee presently uses Anaconda Type UA (Sealtite) (or equal) for conduit sizes up to 4 i. ches and Anaconda Type EF (or equal) flexible conduit for sizes larger than 4 inche Since stainless-steel flexible conduits are often used in radiation applications, the licensee was asked if ANO 1 requires flexible conduit to be qualified for a radiation environment.

! Their response stated that the radiation threshold for polyvinyl chloride (PVC) materialis around 1E8 rad Although radiation does not appear to degrade PVC, the licensee initiated CR C-97-0105 to discuss and document the possible impact on the containment sump screen if the PVC material becomes loose and migrates to the sump. The analysis concludes that the transport of the materialis not expected and discussed the possible impact of PVC material debris being ingested into the ECCS suction header. The PVC is compared to debris previously analyzed in licensee report 93-R 2017-01 and concludes that the PVC will not impair the ECCS components from performing their required function. Since PVC is combustible, the analysis discusses the impact on fire loading assessments. ANO-1 stated that the jacketed flex conduit has not been evaluated in the current fire loading calculations inside containment and does not0 jud e it to be a significant contributor to a fire. The evolution acid was not ofanalyze chlorine gas during burning and its combination with water to form hydrochloric

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E1.2.3.6.3 Conclusion Due to the age of ANO 1, the licensing commitments, regulations, and requirements concerning the raceway and supports are few and non specific. The licensee is aware of field as building of supports and modification to standard designs which has occurred the past 25 years. Field routing and diagrammatic drawings give the installing craftsme considerable latitude and the low levels of seismic input forces allow the acceptance of certain as-built supports by analysis. Although the licensee has not changed materials or raceway practices that were employed during original construction, the team found that for thedesigns the raceway weraand supports acceptabl reviewed during the inspection, for the EFW and DH/LPI sys

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E1.2.3,7 Grounding and Cathodic Protection E1.2.3.7.1 Scope of Review in this portion of the electrical design review, the team evaluated the cathodic prote safety related underground piping, and assessed the effectiveness of the licensee's maintenance practices for the station grounding syste E1.2.3.7.2 Inspection Findings The cathodic protection system applies a reverse galvanic DC voltage supplied by a DC rectifier which is polarized to prevent ions from leaving the surface of the ste in this system, the piping is the cathode. A series of sacrificial metal anodes are burie

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the vicinity of the piping and connected to the rectifier, lons are forced to leave the an years ago. them to deteriorate over the years. The system was apparently buried about 30 causing The inspection team reviewed the periodic maintenance requirements foi

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taken at the test stations at 24 week intervals. The task rec

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" equal to or less than +.250 volts when measured from the protected structure to a properly buried and located zinc electrode reference source." It also records current 2 readings in amperes for each buried anode wired through the test station. The g rea rectifiers, replacing or " refreshing" the anodes, etc.sent to system engine Th3 team reviewed the readings for the past three years for Test Station 10 which is located at the CST. This area has buried safety-related piping which runs from the CS the auxiliary building and also piping for the diesel generator fuel oil. Four sets of " semi -

annual" readings were provided, dated August and February 19,199 ,1994: December 7,1995: May 1,1996; The reference volts were all approximately 0.70 volts. Amperage readings for seven anodes wired through the Test Station were listed. At times, the curre flow to two or three anodes was at or very close to zero. The team noted that ANO-1 d not have a procedure describing the review and trending of the readings; what levels woul require corrective actions; anode or rectifier maintenance recommendations from the manufacturers and methods for cleanin ;

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The team reviewed Procedure 6030.111 for the installation of electrical grounding systems including clarifications included in Revision 4 which was issued on March 7,1997. The procedure does not include, nor exclude, the cathodic protection system and does not include any testing requirements or test equipment. As with cathodic protection, the grounding system has been in the ground for about 30 years. The high voltage switchyard ground grid is periodically tested, as are most substations, ANO 1 does not currently have any requirement to test ohmic value of the ground mat, either interior or exterior, and there is no requirement for periodic inspection of the connections between the mat and the building steel. The installation of new ground rods does not include a test of ohmic value Generally, a generating station ground grid is required to be at 0.25 ohms or below but no values are stated in the procedure. As with the maintenance and inspection of the cathodic protection system previously discussed, ANO 1 stated that they would contact other nuclear station organizations and review industry standards for information on maintenance, inspection and periodic testing of grounding (ER 973866) and cathodic protection systems (ER 973866).

The licensee's lack of testing and maintaining the cathodic protection system for underground piping is identified as Unresolved item 50 313/97 201-1 E1.2.3.7.3 Conclusion Readings of the cathodic protection provided during the inspection raise questions about the conditions of the anodes in the vicinity of the CST tanks. The lack of procedures for cathodic protection maintenance and ground system testing are to be reviewed by ANO 1 as part of the closecut to the ERs noted abov E1.2.3.8 Loading of Class 1E Batteries Determine if the DC system and batteries are designed to have sufficient capacity and are capabk to provide EFW flow for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months independent of ar.y AC sourc E1.2.3.8.2 Inspection Findings The team reviewed calculation,92 E 0021-01," Emergency Duty Cycle and Battery Sizing Calculation," Revision 4, dated September 17,1991, for battery loading and sizing. It was performed in accordance with IEEE Standard 4851983. The calculation specified a minimum cell temperature of CO degrees Fahrenheit, which the team verified, and a aging factor of 1.25, but no design margin was used. All nonsafety DC loads were transferred to the black battery increasing each Class 1E battery's spare capacity Unit 1 is not a SBO coping plant, but each battery is sized for a 2-hour discharge during an SBO with the assured AC source unavailable. The calculation was not re eised to show EFW valves CV-2627 and CV 2620 new continuous and locked rotor currents or the additionalloading of

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the reactor building spray pumps' close and trip circuits. DCP 93-008, field complete in 1995, upgraded those motors and DCP 92-1003, field complete in 1992, provided both undervoltage tripping of each pump and verification of bus voltage before sequencing it onto an CDG-backed bu The licensee performed a preliminary calculation that showed these changes to have minimal impact on either Ur.it 1 battery's capacity. kemorandum ANO-97-00225 was issued requiring more diligence by the staff in determining the affect of plant modifications on

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battery and EDG loading calculations. The licensee's failure to update the battery calculation represents a weakness with respect to 10CFR50, Appendix B, Criterion 111

" Design Control," and is identified as Unresolved item 50-313/97 201 0 The present surveillance tests verify a fully charged battery based on the manufacturer'

recommended full-charge specific gravity (SG) at 1.210. Daily pilot cell tests require a SG greater than or equal to 1.195 and quarterly tests require an average SG of 1.200 + / .010 The battery is operable as long as the SG is above the lower limit for daily or quarterl

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between the present requirements and those of the proposed T Surveillance procedures 1307.006, " Unit 1006 Battery Quarterly Surveillance," and 1307.016 " Unit 1 D06, D07, and D10 Pilot Cell Test," refer to the acceptance criteria for voltage and SG in the TS however, the TS do not have acceptance criteria for either of these parameters. The surveillance procedures need to be revised to refer to the actual acceptance criteria for voltage and SG referenced in EE 8418 E1.2.3.8.3 Conclusion The calculation reviewed supports each C' ass 1E battery's present loading and cap each normal during batteryorhas considerable transient operation. margin, due to the recent removal of non 1E loads, either E1.2.3.9 DC Distribution System I

E 1.2.3.9.1 Scope of Review in this portion of the electrical design review, the team sought to verify the Class 1E batteries * short circuit outputs and the fault capability of the DC buses. The team also evaluated the adequacy of the minimum voltages at the DC buses and individual components like the EFW DC MOVs. In addition, the team sought to determine the level of protection afforded to DC buses and components by protective devices, and assessed whether proper coordination exists between specific devices. Finally, the team reviewed each battery charger's functionality and recharging of Class 1E batterie E1.2.3.9.2 Inspection Findings DC calculation January 21,1994, 92 E-0021-02, "D01 DC System Short Circuit Study," Revision 3 dated ,

was reviewed to determine whether DC buses'short circuit ratings can be exceeded by maxirnum fault currents. Maximum charger output and maximum b room temperature were assumed not to affect short circuit currents in agreement with IEEE 946. Battery discharge terminated at a terminal voltage of .5 volts above ground potenti A prototype test on a similar battery-cell type and the vendor's published opinion suppor

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these assumptions. DC calculation 84-E-0083-50, "DC Load Center D01 " Revision 2 date ,

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February 2,1994, was reviewed to determine if red-train protective devices coordinated with each other and protected equipment and cables. Protective devices had adequate fault- capability ratings to clear expected fault currents without causing damage to surrounding equipment and components. At the team's request, the licensee evaluated the

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temperature where require impact on fuse time current plots and determined that coordination e still exist Selectivity did exist where required. Individualload protective devices

" green" train calculation 84 E 0083 51 opened before the protective devices of t produced similar result DC calculation 09 E 0102 01, *

Valves," Revision 8 dated June Terminal Voltage Calculation for DC Motor Operated 28,1996, during Class 1E battery's first minute discharge. Cable conductor temperatures assumed to be 90 degrees centigrade at battery final voltage of 105 VDC and MOV motors had adequate voltage to start and run. After a 1989 NRC inspection determined that improve their terminal voltages above the criterion of 90 V operators were replaced to increase setup margins in agreement with NRC GL 89-10. The for CV 2870 which is 73% of nominal (91 VDC), but the mo scheduled to be replace During this review, it was discovered that a current calculation 80-D 1083A-02 "EFIC D

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, Valve Torque Calculation Under Reduced Voltage Condition," Revision 1, dated Apri 1986, that determined MOV torque requirements under degraded-voltage conditions h

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been superseded. The need to revise calculations to accurately reflect existing plant configuration represents a weakness with respect to 10CFR50, Appendix B, Criterion 111

" Design Control " and is identified as Unresolved item 50-313/97-201 06. DC Calculation 92 E 0021-08, "ANO Unit 1. Class 1E 125, Train 1, DC Voltage Drop Study," was rev to determine if " RED" train components (solenoids, relays, etc.) had sufficient terminal voltage during five accident scenarios (SBO, LOCA with LOOP, LOOP, loss of char capability, and normal plant operation). A DC circuit's worst case current was calculated then summed with those of others at nodal points. These currents facilitated calcul voltage drops back to each DC component. A component's available voltage was com to its minimum pickup value. All calculations were performed in a computer load flow program. The team verified voltages and currents for the devices at buses D11, A3 and ,

B52 to were verify correctly operability of those entered devices. into the program and spot checked the comparisons DC Calculation 92 E 0021-04, " Battery D06 and D07 Recharge Time," Revision 0 dated

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IEEE 946 was followed and adhered to with the results being Charging of the

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Unit 1200A battery chargers were replaced by new 400A c chargers were changed out in 1995 and the " RED" train ones in 1996. Amending calculation 93-D-1010-04," Recharge Time for DOB and D07 with Changes per DCP 93-1010," Revision 0, dated September 7,1994, recalculated the battery charging times a corrected some discrepancies in assumptions discovered t'y the team during inspection. Th need to verify the incorporation of changes into the parent calculation, after the design changes were completed in the field, is identified as inspector Followup 201-1 /97-Item

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E1.2.3.9.3 Conclusion

" RED" train DC buses had short circuit ratings above the available short circuit current " GREEN" train calculation 92 E 0021-03, '00?

DC System Short Circuit Study,' produced s;milar results. On the basis of smart sample; selected for review, the team verified tha components with highest operating and pickup voltages had adequate terminal voltage to function properly. Selective coordination existed between " red' train protective devices a substantiated by their time current plots. No instance was identified where protective

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devices were improperly sized either for voltage, current, or fault capability. Even with minimum terminal voltage, the chargers were able to recharge batteries in about 11 hour1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months A concern regarding non conservative assumptions in DC calculations arose from the licensee's self. assessment of the LPI and EFW systems. Specifically, this concern invo inverter load instrumentation accuracy, load diversity factors for the DC panels worst ca battery discharge voltage, temperature correction factors for the DC cables, and other similar variables. The licensee verified the validity of selected assumptions and concluded that there was no appreciable effect on their respective calculation E1.2.3.10 AC Distribution System E 1.2.3.10.1 Scope of Review in this portion of the electrical design review, the team evaluated the adequacy of the AC distribution system voltages, plant cabling ampacities, and degraded grid voltage relay setting E 1.2.3.10.2 inspection Findings AC calculation 92 E-0009-01, 'AC Motor Operated Valve Terminal Voltage," Revision 7 dated September 5,1996, was reviewed to evaluate exclusively AC MOV's operation unde

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under different scenarios including starting and running at 86 Some MOV motors did not have the requisite 80% or 90% motor terminal voltage du starting and running conditions, respectively. They were reevaluated individually on a case by-case basis with only results given. The team selected CV-1401 & CV 2619 from that group for which the licensee produced the detailed mechanical calculations verifying required torques. Both MOVs developed 60% greater torque than required to start them in either directio The team dated Marchreviewed 12,1997, AC calculation 95 E-000105,'ANO Unit 1 Millstone Study," Revision 0

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of an engineered safeguards (ES) signal when the source is Startup time of transfer to offsite power, etc.) were conservative. This terminal voltages of the MOVs under three different scentaios. This study assumed worst-case " load flow" upon receipt of an ES signal, and then verified sufficient bus voltages for starting and running the required loads. Worst case bus voltages are experienced because of assumed loading patterns and buses being supplied by an offsite source. The team

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requested supporting documentation for AC and DC cable ampacities and determined that there is no overall " master" ampacity calculation. The Unit 1 power cables were sized in accordance with standards issued by the insulated Power Cable Engineers Association (IPCEA), but the AE did not turn over any original calculations to ANO. The team reviewed the ampacities of a sample of 10 cables connected to various type plant loads and found them to be adequate considering applied derating factors for temperature, raceway-loading conditions, and raceway fire wrapping. The licensee developed ampacity calculations for specific loads cabling. The team reviewed and verified four such ampacity calculations (82D 1054A 01, 87 D-1088-09, 88-D 1012 03, and 94-D-6012-03)and requested information about derated cable-ampacities for LOCA conditions at 138 degrees Fahrenhei It was noted that only the reactor building cooling fans continuously operated during a LOCA and their cable ampacities were 100% greater than the required load amperage. The licensee uses Unit 2 ampacity criteria for sizing any new cables or upgrades to existing circuit P54 440 for The cablescriteria consist routed in open top cable of tray IPGA P-46 426 for cables routed in conduits an The team verified the setting of the 480 ESF busses degraded grid voltage relays between the FSAR. TS, and the appropriate calibration procedure (1416.031, Revision 1). The relays are currently side of the set to dropout at about 92.66 % (460 x .9266 = 426.24 VAC) on the primai /

Tech Specs 3.5.1.8.b allows the relays to be set from 423 VAC (9 (93.7%).

The calibration procedure allows the relay to be set to pickup from 426.12 (92.63%) to 427.08 (92.84%),

with the desired setting being 426 6 (92 74 %). The . .

primary concern is that this relay will pickup at or prior to 92% setting and that bus voltages LOO will recover prior to i' timing out while starting ESF loads during a LOCA without a The licensee's calculation assumed a tolerance for the relay setpoint that was 133% times the actual tolerance of 0.6% or 0.798%. That tolerance value applied to the lower allowed setting stated in the calibration procedure allows the relay to pickup below the 92% required setting. The licensee adjusted the lower allowed setting to agree with the present desired set point. The team agreed with the licensee's actio The team reviewed the operation of the swing buses for primary make-up pump P36B and service water pump P48. The transfer switches or manually thrown disconnect switches are not load-break so are positioned before the circuits are energized. This aoministrative control prevents both sources to either pump from being connected together even for a single failure of one of the upstream switchgear breakers .

E 1. 2.3.10.3 Conclusion The calculations reviewed indicated that all Unit 1 AC MOVs had sufficient terminal voltage to start and to run under degraded bus conditions. Some MOVs had to be reevaluated but ,

even they had sufficient voltage to develop the appropriate motor torque. For the worst-case load flow of Startup Transformer 1 supplying auxiliary loads following receipt of ES signal, all buses had adequate voltage to ensure the operation of allloads. Since the licensee did not have historical ampacity data, the team reviewed a sample of Unit 1 cables and found them to be adequately sized for ampacity. Based on the results of those reviews, the ampacity voltage relays were criteria and correctly sampled cables were considered adequate and the degraded-grid calibrate i

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E1.2.3.11 EDG Load Sequencing and Starting Circuits E1.2 3.11.1 Scope of Review

In this portion of the electrical design review, the team evaluated the design adequacy of the load sequencing and starting circuits for the Class 1E EDGs, along with other issues directly related to the EDG E 1.2.3.1 1.2 Inspection Findings The team reviewed the EDG protective functions which upon actuation initiate an EDG output breaker trip. ANO 1 is committed to Safety Guide 9," Selection of Diesel Generator Set Capacity for Standby Power Supplies," dated March 10,1971. which does not specify those protective functions to be bypassed under accident conditions, while RG1.9 specifies all protective functions to be bypassed except for generator differential and engine overspeed. A previous NRC electrical distribution system functionalinspection report concluded this to be a weakness in the current ANO 1 desig The team agrees with that determination because EDG availability is potentially decreased under accident conditions. Each EDG had protective functions consisting of overcurrent, differential current, anti motoring and loss of excitation, including those associated with the engine-protection and system (e.g., overspeed, low lube-oit pressure, crankcase over pressure engine over crank).

All trip functions were single channel with no redundancy or coincident logic provisions. The team's concern is the loss of an EDG during accident conditions due to a minor protective function trip. The need for the NRC staff to review ANO 1's current design for bypassing EDG protective trips is identified as inspector Followup Item 50-313/97-201-1 The team reviewed the loading of the ESF buses to assess the current licensing basis for a concurrent LOOP and LOCA, as well as an event in which one preceded the other. The only difference in the loading patterns is the sequence in which the injection pumps and shutdown cooling loads are started. This reordering of the starting loads involves no change in kilo volt ampere loading. The licensee's single failure criterion, ULD-O-TOP-18, states that special timing of a random single failure does not have to be considered. However, if a LOCA precedes a LOOP, then the load sequencing occurs twice. An ES signal and undervoltage (UV) signal produce similar results in regard to the loading of accident mitigation loads on safety buses except that an ES signal by itself would place load on the offsite power source while a LOOP causes safety bus loads to be picked up by the respective EDG, The undervoltage at ESF buses due to the delayed LOOP will trip running loads off the ESF buses. Subsequently those loads will be restarted and sequenced onto those same buses extending the accident response times of ESF bus load The team determined that the licensee has never evaluated the impact of a delayed LOOP on its LOCA accident analysis nor has the licensee initiated circuit modifications to mitigate its impact. The team's review of the licensee's evaluation of NRC IN 93-17, Revision 1, which specifically dealt with the delayed LOOP issue, stated that it was outside its design and licensing basis. As a result, the licensee decided to take no further action without additional justification. After discussion with the NRR technical staff, the primary concern is the

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resetting of the load sequencer in response to a delayed LOOP. This issue is not a concern to ANO 1 since the licensee uses individual time delay relays that reset upon being deenergize The team evaluated the licensee's review of NRC IN 94-019, " Emergency Diesel Generator Vulnerability to Failure From Cold Fuel." This document concerned the cloud and pour points of EDG fuel oil under cold weather conditions. The cloud and pour points are the temperatures at which crystals begin to form in fuel oil and when fuel oil can no longer be pumped, respectively. The cloud point defined by the American Society for Testing and Materials (ASTM) Standard D79751981 is 6 degrees centigrade ('C) above the tenth-percentile minimum temperature ( 11 *C at ANO 1) or 5'C, and the pour point is slightly lower. The actual ANO 1 test results indicate the allowable range of values for cloud and pour points to be respectively 8 to -12'C and -15 to 23.9*C, The allowable test values for cloud and pour points are below those required by ASTM; thus, the ASTM values are acceptable. The EDG fuel oil storage tanks and transfer system are located below ground, thus the fuel oil should not experience its cloud and pour point temperature The team reviewed the latest surveilla. ice tests of the EDG load sequencing relays. The results showed that all of the relays picked up about mid range in their acceptable time bands. These relays were originally electropneumatic timing relays with a qualified life of 10 years from the date of manufacture. They have been replaced by solid state relays with

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a qualified life of 15 years. The accuracy and repeatability of the new relays is considerably better than the old relays. The time delays on these relays agrees with those in the FSAR for sequencing of the respective loads on to the ESF buse The team also reviewed the topic of suitable missile protection for the EDG exhaust stack This ::oncern arose during the licensee's self assessment of the EFW and LPI systems which was performed prior to the inspection. Section 5.1.5.8 of the FSAR states that all Class 1 structures have been analyzed for tornado loading including tornado-driven missiles. The concern is that a tornado-driven missile could directly strike the EDG exhaust stack causing it to deform inwardly and appreciably reducing the EDG exhaust during operation. Since the original Bechtel calculations could not be produced during the inspection, the licensee agreed to reconstitute the snalysis. The need for the licensee to reconstitute the tornado analysis for missile protection of the EDG exhaust stackr. is idantified as inspector Follow-up Item 50-313/97 201-1 E 1.2.3.1 1.3 Conclusion The team will refer ANO-1's present design for bypassing protective EDG trips to the NRR electrical staff for further review. ANO-1 stated their position that a LOCA followed by a LOOP is outside of its design basis. EDG operation will not be impeded during cold weather conditions since fuel oil temperature will not drop to cloud and pour point temperatures, respecavely. The EDG load sequencine relays operate as intended with maintenance ensuring the load-initiation times stated in the FSAR for LOCA and a coincident LOOP. The new solid state relays have high repeatability and accuracy to ensure their timely operatio ANO 1 is performing an analysis to justify the present configuration for the EDG exhaust stacks with no modifications required to physically protect them from tornado missile ___ __

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E1.2.3.12 Testing of Molded Case Circuit Breakr E 1.2.3.12.1 Scopo of Review in this port!on of the electrical design review, the team sought to determine it the original design bases require testing of the molded case circuit breakers (MCCBs), and whether significant safety impact could result from a failure to test snme critical breaker E1.2.3.12.2 Inspection Findings The team evaluated the licensee's basis for not electrically testing MCCBs at ANO 1. The team also reviewed regulatory requirements and licensee commitments that address periodic testing of those electrical components. The licensee's self asser.sment indicated that the lack of MCCB testing is a concern. For Unit 2, TS Section 4.8.2.5.a.3 requires testing for the MCCBs that supply containment feedthroughs. Each breaker is tested every N x 18 months, where N is the number of devices of each breaker type. A percentage of each breaker type is tested every refueling outage. The only Unit 1 testing requirement for MCCBs is described in Section 3.3-17 of the FSAR which states that the Class 1E power system is designed to meet the requirements of IEEE 308 1971. The licensee takes no exception to IEEE 308, which in Section G.3 requires periodic testing of any electrical componer't which is not exercised during normal operation. IEEE 308 provides illustrative examples of similar components that do require testing and MCCBs are not exclude The licensee interprets Section 6.3 of IEEE 308 as requiring testing only for components in the illustrative examples provided. Other ANO 1 licensing documents including the GDCs and TS do not mandate MCCB testing. Several NRC information notices have discussed this

' issue including IN 92-051, which discussed premature tripping of safety related breakers.

The team reviewed the licensee's evaluation of the IN for applicability to ANO 1, which

indicated that the breakers were properly sized and that any premature operation would be a l conservative position. Periodic testing of these MCCBs was not developed because of their

! satisfactory performance history. Receipt acceptance of these MCCBs was modified to

! include electrical testing to verify their published time current curves. NRC IN 93-064 discusses periodic testing of MCCBs to minimize the age degradation effect. The licensee determined that most vendor technical manuals do not specify electrical testing and that mechanically exercising the breakers will remedy grease solidification since that is the most probable basis for improper operation of the dormant breakers. The licensee's preventive maintenance program for Unit 1 MCCBs every four refueling outages.' quires mechanically exercising a certain number of IN 93-064 referred io four industry reference sources that recommend testing and maintenance of MCCBs. One of those is Electrical Power Research Institute (EPRI) Report NP-7410, " Breaker Maintenance," Volume 3, " Molded Case Circuit Breakers," September 1991. Page 21 of that document r.ates that two-thirds of breaker failures are attributed to the mechanical operating mechanism or the overcurrent trip device. Mechanically exercisin a breaker would be sufficient preventive maintenance to handle the first type of problem for dried lubricant but not overcurrent trip device problems. Calibration drift, improper set point, and degradation of material properties for overcurrent trip devices can only be adequately dealt with by electrical testing. The same document in Table 3-2 suggests a

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penodicity for MCCB testing of every 4 to 6 years for both the overload trip and instantaneous overcurrent trip units. At ANO-1, MCCBs perform other critical functions that warrant electrical L* sting for the same reasons as the ones supplying containment feedthroughs. One breaker supplies a nonsafety related load from a safety-related bu This breaker should trip prior to the bus' incoming breaker to prevent loss of the entire bus for a singular load faul Coordination or selectivity between these breakers is not verified and maintained by electrical testing of instantaneous trip dt: vices, in Unit 2, two redundant MCCBs in series willinterrupt a fault current before their penetra'. ion feedthrough experiences damage. ANO 1 relies on the timely actuation of e single breaker to maintain containment integrity, in Unit 2, conformance to RG 1.63 is ccmmitted to in FSAR Amendment 34. In order to meet NRC RG 1,63, Unit 2 has two redundant breakers in series to supply containment feedthroughs. For Unit 1, the licensee has not committed to RG 1.63 and could not satisfy it without a hardware chang The licensee stated the following reasons for not testing Unit 1 MCCBs supplying containment feedthroughs:

First, the field cable to the penetration feedthrough has a smaller wire gage than the feedthrough, therefore it will fail before the feedthrough, thus preserving containment integrity. The inspection team questioned this design philosophy of permitting wiring to act like a " fuse." Second, other protective devices like the thermal overload relay will actuate, clearing the fault. The team questioned this reasoning because most faults response are time is in toothe time current plot's instantaneous region, for which that relay's slo Regarding upstream feeder breakers, no backup protection is afforded by them. They are load center breakers supplying MCCs and their instantaneous trip settings are higher than the MCCB setting and will not detect the fault current magnitudes of these circuits. Third, mechanically exercising MCCBs three times every four refueling outages, will prevent breaker failures. The team considers this valid maintenance for mechanical failures involving hardened lubricants in the MCCB operating mechanisms, but not for other failure-modes requiring electrical testing. Fourth, potential feedthrough damage will only increase the containment leakage minimally. In addition, the conductor is protected by a stainless steel tube which will not incur damage along its entire 36-inch length. However, the penetrations have never been destructively tested. Finally, any containment leakage will be immediately collected and processed by ther existing penetration room ventilation syste As a result of the self-assessment, the licensee established a committee to review the subject of MCCB preventative maintenance testing and to determine if a testing program, replacement program, or a combination of both should be adopted. ANO-1 is reviewing a wide range of programmatic issues to upgrade their present MCCB program but has not indicated what type of program they plan on adopting. The team concluded that the licensee's failure to perform electrical testing of Unit 1 MCCBs represents a weakness with respect to 10 CFR Part 50, Appendix B, Criterion Ill, " Design Control," and is identified as Unresolved item 50-313/97-201-1 E 1.2.3.12.3 Conclusion The team considers that the testing of MCCBs is called for by lEEE 308 in order to verify that a safety-related breaker either will not trip early impeding its safety-related function and that it will open when require __

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E1.2.3.13 Design Control for Electrical Drawings E 1.2.3.13.1 Scope of Review in this portion of the electrical design review, the team souoht to determine if the existin electrical drawings contain adequate fuse data to support maintenance activities and the fuse control program, in addition, the team considered whether the fuse control program procedures support the replacement of all safety related fuse E 1.2.3.13.2 inspection Findings The team reviewed several electrical schematics related to the DC distribution system and also to components in the two systems being reviewed. Several of the drawings did not have sufficient fuse data to identify the respective fuses and subsequent checking showed that the fuse control procedure, 1025.056, Revision 0, also did not either list the fuses or any pertinent information on them. The affected * RED" train DC and AC panels are the DC distribution panels RA1 and C11,125 VDC MCC D15, and the 120 VAC panels RS1 and RS2. The respective panels for the " GREEN * train would be similarly affected. The following drawings were identified as lacking fuse data:

E 280, Sh.3, SW lsolation Valve SV3812 -Fuses S057 FU 1,2 E 295, Sh.4, EFW Turbine MOV SV2663 Fuses D1512 FU3,4 E 295, Sh.4A, EFIC BYPASS SW.- Fuse M084 F9 E 331. En.31. EFW Turbine Speed Control and Indication Fuses 6428 F24,2,1:FU3,1,2 E 331, Sh.40, Condensate Storage Tank LevelIndication Fuse 1452 F16 The fuses on the above drawings are representative of all affected AC and DC power t

but without further review the exact number was unknown. The de present fuse control program is tied to several NRC commitments. The electrical drawing upgrade project (EDUP) was committed to during a 1989 NRC inspection, it was to as-build and upgrade all ANO 1 Class 1E switchgear, MCC and major control cabinets electrical drawings. A list of electrical cabinets and enclosures were identified for review by walkdown of cabinet wiring against drawings. That list included 4.16 kV switchgear,480-VAC load centers, MCCs, EDG control cabinets, and a number of safety-related control room cabinet As a result of a previous NRC inspection for Unit 2, the licensee committed to develop and-implement a fuse control program for Units 1 and 2. In a later response to the NRC detailing the action plan, the licensee stated that the fuse controllist would encompass the safety related cabinets currently included in the scope of the EDUP. Fuses which are part of internal components such as instruments are not to be included in the Fuse Control List as they are contained in vendor technical manuals and will be subject to like-to-like replacement. However, the fuse list contained about 40% of totalinstalled fuses and 4% of

.

those fuses had been walked down. The majority of safety related fuses were assumed to be on that list. The fuses depicted on the sample of drawings listed above are not considered as instrument internal components but are power fuses located m distribution

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cabinets . In addition, they are in safety related circuits that can not be differentiated from those fuses currently in the fuse control procedure. Procedure 1025.003, " Conduct of Maintenance," discusses the requirements for like to-like replacement of installed fuses and also the requirements for an engineering evaluation if fuse data is not available. The team <

was informed by the licensee that in the last 4 years there have been more than five instances identified at ANO 1 where the wrong fuse was installed. The licensee walked down only 4% of the fuses in the Fuse Control List for Unit 1. The licensee's failure to ,

translate design basis information into drawings and procedures to ensure correct fuse replacement represents a weakness with respect to 10 CFR Part 50, Appendix B, Criterion 111, " Design Control," and is identified as Unresolved item 50-313/97 201 1 E 1.2. 3.13.3 Conclusion The team concluded that since only 4% of the fuses in Unit 1 were walked down to verify they were correct in all aspects (voltage rating, ampere rating, time delay, etc.), the existing method of like for like replacement may not be valid. The fuse control procedure and associated maintenance procedures require an engineering evaluation only for blown fuses with insufficient data or when replacement fuses are different than those installed. Since the fuses supplied in the DC distribution panels, DC MCCs, and vital AC panels have not been verified, the team determined that a like for like replacement for those fuses is not appropriat E1.2.3.14 Discrepancies in Design Basis Documents E 1. 2.3.14.1 Scope of Revie In this portion of the electrical design review, the team evaluated the FSAR and subsequent !

'

plant modifications design base and licensce condition reports to identify any deviations from the plant's E 1. 2.3.14.2 Inspection Findings The team reviewed plant modifications LCP 92 5008, DCP 90-1030~and DCP 82-1050A which encompassed all significant changes to the electrical design bases for the LPI and EFW systems. The modifications were essentially upgrades with equipment being replaced with equivalent or superior equipment. The FSAR was reviewed against the licensing documents, TS, and system operating procedures with one discrepancy found. Section 14.2.2.1. stated that for an EFIC actuation for a steamline break that both steam generators are isolated. This is not ccrrect since in that case only one steam generator would be isolated. The team reviewed CR-197-0043 which dealt with the possibility of damaging safety related cables for a fault on nonsafety-related cable in the same tray and agreed with the licensee's evaluation. The need to correct discrepancies and review the corrective actions related to the accuracy of the FSAR represents a weakness with respect to 10 CFR Part 50.71(e) and as a result, the team identified this issue as Unresolved item 50-313/97-201-1 , ,. ..

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E 1.2.3.14.3 Conclusions The intent of the original design basis remained intact after these plant modifications since they were primarily upgrades and also because the design basis was a higher tier document not influenced by such changes. All the components operated as intended and there were no unreviewed safety questions as a result of plant modifications to the two system E1.2.4 Instrumentation and Control Design Review E1.2.4.1 EFIC Actuation Logic E1.2.4.1.1 Scope of Review in this portion of the instrumentation and control (l&C) design review, the team sought to determine the capability of the EFIC actuation logic to initiate the EFW system on low OTSG level and isolate the EFW flow from the affected OTSG following a main steam or feedwater line brea E1.2.4.1.2 impection Findings in response to NUREG 0578 and NUREG 0737, the EFW system was upgraded from the original design to safety grade, which incorporated an automatic actuation function in the EFIC. The EFIC system monitors selected plant parameters and automatically initiates the EFW or OTSG isolation upon detection of any of the following conditions:

loss of both main feedwater pump *

L low levelin either OTSG.

! +

loss of all four RCPs.

! +

loss of either OTS +

ESAS actuation by high RB pressure or low reactor coolant pressure, initiation of the EFW system includes automatic startup of the pumps, alignment of the valves to direct EFW flow to the appropriate OTSGs, control of EFW flow to regulate OTSG level and control OTSG pressure. The EFW isolation valves to the OTSGs are normally open, but are designed to close on actuation from the EFIC system vector logic when a dngraded SG is detected on the basis of low steam pressure signals from the SG pressure transmitters. The EFW pump start and isolation valve logic is shown on Drawing M-402, Sheets 3 & 5 respectively while the EFIC input logic is shown on Drawing 58526-016 and P&lD Block Diagram M-204, Sheet 4. These documents were reviewed for functionality and were found in agreement with FSAR Section 7.1.4. The EFIC system automatically isolates EFW flow to a OTSG that is degraded (as a result of a main steam or feedwater line break or a break in the EFW line downstream of the isolation valve) and maintains the EFW flow to the intact OTSG. The EFIC vector logic derives the isolation signal from a differential pressure signal between the faulted OTSG and the intact OTS The team reviewed the results of loop uncertainty / set point calculation 80-D-1083C-01 for the OTSG differential pressure and found it consistent with TS 3.5.1, " Operational Safety instrumentation, Bases." The EFW flow rate to the OTSG is controlled by EFIC, which also

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controls level at appropriate set points. The level control set point values determined in calculation 80-D.1083C 01 were reviewed by the team and found to be in agreement with FSAR Section 7. E1.2.4.1.3 Conclusion The EFIC system is designed to meet commitments to NUREG 0578 and NUREG 0737 and the design documents and as-built condition meet the system functional requirements as discussed in the FSA E1.2.4.2 Condensate Storage Tank Level Instrumentation E 1. 2.4.2.1 Scope of Review in this portion of the design review, the team evaluated the instrumentation and controls associated with "O" CST T-41B to assess conformance with the design basis and requirements of RG 1.9 E1.2.4.2.2 Inspection findings The CST levelinstrumentation provides the following functions:

Indicate and record CST levelin the main controi room per RG 1.97 guidanc *

Actuate a low level alarm in the main control room to alert tha operator of a CST low-level condition.

I *

Actuate a 10-10 level alarm in the main control room to alert the operator to manually transfer EFW cooling water supply from the CST to the SW syste Per TS 3.4.1 and 3.4.2, the reactor shall not be heated above 250 F unless a minimum of 11.1 feet (107.000 gallons) of water is available in tank T-418. The TS require that if tank T-41B is inoperable for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor shall be placed in the hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and if it remains inoperable for an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the reactor shall be placed in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Level instrument loops LT-4204 and LT-4205 are used to verify that this requirement is met. The TS do not specify the number of instruments or channels that need to be operable but it is implied that availability of CST levelindication is a prerequisite to reactor startup and operation. This inconsisteacy was discussed with the licensee and the team was subsequently informed

.that the TS improvement project team is currently updating the ANO-1 TS to clarify the operability requirements for all RG 1.97 instruments which will include the C'- 'evel instrument The licensee has indicated that the updated TS will specifically address CST levalinstrument redundancy and channel availabilit Set point basis document 91 R 1018-02, set point report EAR 91-177, and loop accuracy calculation 82 D-2086B-41 for the CST level set points were reviewed against TS 3.4.1 and FSAR Table 7.11 A for conformance. Per instrument data sheets, the low level alarm is set at 333 inches and the 10-1o level alarm at 61 inches. These results are consistent

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code A06, control room levelindicator and recorder are s ,

the data in SIMS for these instruments (LIS 4209 and LRS-4204) show the range as -

,

FT. These instruments were subsequently walked down and were confirmed to be 0 the range in FSAR Table 7.11 A from 0-100% to 0-30 FT ncies and review the corrective actions related to the accuracy of the FSAR represents a issue as Unresolved itemweakness with respect to 10 CFR Part 50.71(e) and as a res 50 313/97-201-1 E1.2.4.2.3 Conclusion The CST levelinstrumentation meets the design intent of FSAR Section 7.14 and the .

guidance of RG1.97 and the licensee has initiated the necessary update to clarify th operability requirement for the CST level instrumentation in TS Section 3. E1.2.4.3 EFW Pump Flow, OTSG Pressure and LevelInstrumentation E1.2.4.3.1 Scope of Review and levelinstrumentation for conformance with ure ,

NUREG-057

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E1.2.4.3.2 inspection Findings TOP-05 provide the design basis for the EFW flow and OT --

instrumentation. In accordance with NUREG 0578 and RG 1.97, OTSG pressure instrumentation is classified as a Type A. Category 1 variable requiring redundancy power source. One pressure indicator and a pressure recorder are provided in main control l

2618B/PR-26188, respectively). Review of the various design -

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installed powered from condition suggested the emergency buse that these instrumentation are redundant,1E qualified a OTSG levelinstrumentation is also required per RG 1.97 to be a Type A, Category requiring redundancy and a 1E power source. For each OTSG, a set of redundant ind and recorders is provided in main control room console C09 for low-range and hig level sensing, a total of four dualindicators and two dual recorders. Both the low -

range level instrumentation are overlapped and adequately spanned to cover OTSG leve ranging from the tube sheets to the separators in accordance with RG 1.97. After that the OTSG levelinstrumentation meets the requiremen The EFW pumps have a design flow of 500 gpm at OTSG .pressure RG 1.97 of 1050 psig

. .

requires a control room flow indication range of 110% of the design flow, or 550 gpm flow indicators on Panel C09 (F1-2645/F1-2647 for Pump P7A and F12646/FI 2

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78), each have a range of 0-900 gpm, which envelopes the guidance in RG 197 . This data is consistent among the various documents that wer .

e reviewed. The flow instrument loops

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I are redundant and safety grade, and are powered from the emergency buses. The licensee has classified these indicators as RG 1.97 Type D, Category 1 instrumentation. The team identified the following discrepancies or incorrect information while reviewing the licensee's documents:

The adapter table in Drawing E 258 Sh. 5, Rev. 2 incorrectly lists PI 2667A as the

" RED" train steam generator E248 pressure indicator. The actualinstrument should be recorder PR 2667A. The team noted that the licensee has initiated DRN 94-11529 to correct this discrepanc * In ULD SYS-08, the ster n generator pressure indicators PI 2618A and PI 26678 on Panel C09 were incorrectly identified as dualindicators. Per SIMS database, these instruments are single indicators. The licensee has initiated ULD Change Request 314 to correct this discrepancy, instrument loop error calculations 85-EO-000318 and 80.D 1083C 01 were prepared for the EFW flow, OTSG pressure, and level instrumentation. The team noted in its review of the calculations that adequate tolerance has been provided to account for indicator and recorder inaccurac E1.2.4.3.3 Conclusion The EFW flow instrumentation meets the design intent of NUREG 0578, FSAR Section

.

I 7.1.4, and RG 1.9 E1.2.4.4 EFW Pump Instrumentation E1.2.4.4.1 Scope of Review in this portion of the design review, ths team sought to verify that the EFW pump suction and discharge pressure instrumentation will provide adequate alarm for loss of pump suction pressure and loss of EFW flo E1.2.4.4.2 Inspection Findings Instruments PDIS 2804 and PIS-2811 monitor the suction and discharge pressures of pump P7A and provide signals for low-pressure alarms in the main control room. PDIS-2806 and PIS-2812 provide the same function for pump P78. A low suction pressure condition warns the operator of an impending pump runout and a high r?scharge pressure condition is an early warning of a blocked EFW flow. Calculation EAR 91-177 was reviewed by the team and determined that sufficient margins have been provided in calculating the set points for the above mentioned instruments. The team also noted that CR 197-0040 has been initiated to address an inadequate low suction pressure set pomt that could occur at certain operating scenarios. The licensee discovered this condition during a self assessment that was performed before this design inspection. This condition does not affect system operability since the RG 1.97 flow instrument loops FT 2645 through FT-2648 will monitor the required pump flo s _ .

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The team also notM on drawing M-203, sheet 3, that press 6 a switch PS 2802, identified i the SIMS database as a spare instrument,is currently connected to the common suction lin of EFW pumps P7A and P78 with a normally open valve (CV 2802C). Its function was to provide low suction pressure alarms for the pumps; however, it has since been superseded by P0lS-2804 and PDIS 2806 which were installed during the 1986 EFW upgrad Although this instrument does not currently perform any function other than to maintain pressure boundary, the fact that it is connected to the process with a normally open valve practice, leaving this spare instrument on line provides no benefi has been discussed with the licensee and ER 97-3862 or isolate PS-2802 and update the drawing accordingly.was subsequently initiated to delete E1.2.4.4.3 Conclusion The set point values calculated under EAR 91177 are acceptable and are consistent with the documents reviewed. The licensee's self assessment, which was performed prior to the inspection, identified low suction pressure set points that could occur during certain operating sc.enarios. The issue was addressed by the beensee in CR 1-97-OO4 E1.2.5 EFW Walkdown Observations and Results E1.2.5.1 Mechanical Walkdown and in-Plant Observations E 1. 2. 5.1.1 Inspection Scope The team performed a walkdown of thc ANO 1 EFW system, including safety grade CST T418 located outside of the auxiliary building (AB), and the associated piping and instrumentation and observed the 5-foot partial missile protection enclosure around the tank. The remainder of the system walkdown included the AB EFW pump room SFW pi ,

penetrations into the reactor building (RB), the ADV area adjacent to the RB main steamline penetrations, MSIVs, the EFW pump P7A steam supply irolation valves and associated piping, and the main control room with associated instrumentation in all the areas observed The initial walkdown was conducted on February 12,1997. Subsequent walkdowns and

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operation observations were conducted throughout the team's site visit E1.2.5.1.2 Observations and Findings Condensate Storage Tanks The ANO 1 safety-grade CST T41B (321.OOO-gallon capacity)is located due west of the Unit 1 RB. The interconnecting pipir.g. valves, and some instrumentation are located in valv pits adjacent to the tank. The cross-connect valves for the ANO-2 AFW pump supplies are closed and locked in the valve pits. The piping and electricallayout is all underground to the AB connections. The team noted that the 5-foot partial missile shield had small drain holes through the wall to preclude any collection of rainwater inside the enclosure. A level transmitter instrument cable conduit also penetrates the shield wall with adequate clearance for draining. The team asked whether these permanent shield wall drains were accounted for when calculating the totalinventory of cvai able CST water following a loss of tank

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integrity resulting from missiles. The licensee responded that the drain losses would n affect tht, calculations for the required 30 minute inventory, since the vv.ame would be grarter inside the shield wallif the tank were to be sheared at the bottom and totally los The draining losses for 30 minutes would be less than the increased inventory retained in the shield wall versus the inventory in the tank sheared at tr,e 5 foot blevation. A nitrog apargespecificatio within and vacuum degasification system is employed to maintain T41B oxygen levels The ANO 1 nonsafety grade CST T 41 (202,000 gallon capacity)is located just southwest of the RB, and serves as the backup source of water for the EFW system. The T 41 tank is isolated by valve lineup from the EFW system during normal plant operations. The tank is interconnected with the Unit 2 CST (2T41) which will serve as the plant supply of demineralized water should T41 become unavailable. Manualisolation valve tie line allow uperating for use of either tank as the demineralized water source. Section 8 of plant procedure 1106.006, " Emergency Feedwater Pump Operation," describes operator actions to use the contents of the CST T41 (nonsafety grade) inventory when the contents of T41B (safety grade) are depleted. The operator will monitor both tank levels and, if necessary to maintain EFW flow, will bring the SW system (unlimited source of EFW to the OTSGs) supply into operatio Main Steam Supply to EFW Pump P7A The valves and associated piping are not easily accessible. During the observation of this area, the team noted significant chattering from the ccrrponents of the steam supply pipinC to EFW pump P7A. The team questioned whether this continual chattering was caused by leakage in the system, and regrested informa. ion un the progressive detrimental effect on these components, possibly leading to an inability to reverse seat if required. The licensee responded that two of the check valves in the steam supply to pump P7A were passing a buildup of steam pressure in the supply lines probably as a result af ieakage from solenoid valves 2613 & 2663, which has occurred in the past. At that timo, the check valves were fitted with stellite seats and disks to preclude excessive wear on the surfaces if the solenoid valves again leaked. Currently, the licensee has Job Order 009u0444 ir place to work these components and perform inservice testing (IST) inspections during 1R14 refueling outage. The team determined that the licensee's ineffective disposition of the solenoid valve leakage issue previously constitutes ineffective 10 CFR Part 50, Appendix B, Criterion XVI, ' Corrective Action." Consequently, :he team identified this issue as Unresolved item 50 313/97 201 1 EFW Pump Room The team noted that the EFW pump room entrance door had been removed. This subject has been identified in the steam trap f ailure review identified earlier in the Mechanical Design Review section of this report. As the team reviewed the EFW System,it was identified that the EFW room cooler was not 1E powered. The licensee stated that since the EFW pump room cooler was not safety grade powered, the EFW pump room would be able io use the adjacent waste handling systems area convection sii volume to maintain the EFW room maximum temperatures within acceptable levels following a loss of power event associated with EFW requirement .

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On March 11,1997, the adequacy of sealing of the electrical enclosures to pr

>

of the electrical equipment with respect to the locatio Sealing adequacy was checked for several electrical enclosures locate room. The enclosures featured a hinged gasketed cover with 4 to 6 retaining c positive seating. The conduit entries to the enclosures were also sealed a the electrical enclosures was adequate to prevent moisture . The ingressionmajority of the drains undar the missile shield wall. The P7A turbine associated with the turbine hstrumentation and controls, except the governor and

, controls, are located close to where the steam would enter the room. The enclosu conduits entering at the bottom of the box. One en above the discharge point, it contains pressure indicator / switch P78. The enclosure was deemed adequate, but if moisture did penetrate into the box. it wa determined that failure of the switch would not affect the performance of the E .

The team noted a section of line identified to be service water directly above th driven EFW pump assembly that had some insulation removed and a work tag a .

There was a 1 inch capped off service water vent line making contact with a 4-inc under Cfv196 0051 water system line. The licensee had identified potentialinterferen and evaluatert the condition in ER 963521R101. The service wate line is seism!c Category I whereas the fire water line is seismic Category 11. The between the two lines is so small that the 4 inch fire water line will move up slide over the top of the 1-inct, vent connection without adversely affecting the 1 in 10 inch service water lines. A seismic Category 11/l review was performed by the lic which concluded that the 4-inch fire water line did not represent a hazard. Sufficient seismic Category I structural barriers exist such that the fire water line would n affect any Seismic Category I systems in the area. The licensee slated that the line be reworked sometime in 1998 to remove the interferenc The team noted that the overall material condition and general appearance of the good and the majority of the areas containing EFW system components are in good o The licensee has a Site Upgrade Project in progress and has funded it to continue throughout 1997. This program considers the adequacy of lighting, paint condition (in relation to requirements), and general overall degraded conditions. The areas prev upgraded are monitored and as any current touch up needs are identified, they would be addressed immediately. The licensee provided touch up kits at strategic locations for the use of the area responsible personnel whenever that area requires attentloa.

'

E1.2.5.1.3 Conclusions The CST is designed for partial missile protection and the structure is more than adeq provide the required inventory of water to supply the EFW pumps during the 30 minutes necessary for the operators to line up an attemative suction supply for the EFW pumps. The eldctrical conduit and rainwater drain holes in the shield wall will not preclude the ability supply the minimum 30-minute water inventory. The piping, electrical and l&C support

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components are protected within the shield wall or are in the underground pits and pipe chases to the auxiliary building and the nitrogen supply and degassing system ensures t quality of chemistry required for the CST water inventor The loss of the EFW pump room area cooler will not impair the operation of the EFW room equipment because the adjacent areas have an adequate convection cooling capab The electrical and instrumentation sealing is sufficient to preclude any adverse ef fects of steam and water resulting from a steam trap failure in the EFW pump roo E1.2.5.2 Instrumentation and Controls Walkdown and in Plant Observations E 1.2. 5. 2.1 Inspection Scope in this portion of the inspection, the team conducted a walkdown of the control room, an its instrumentation, and obserywd the methods used to identify indications related to RG 1.97 and to verify the as built conditio E1.2.5.2.2 Inspection Findings During a walkdown of the EFIC input pressure transmitters from the OTSG, the inspectio team noted a missing clip that restrains the isolation valve and sensing line for OTSG pressure transmitter PT 26678. In addition, the team also found that the corresponding clips for the three other redundant instrument channels located in the same vicinity were loose and are not providing the necessary restraint es designed for the associated instrument tubing and isolation valves. The safety significance of the missing clip is a potential failure of the instrument sensing lino during a seismic event that could affect proper functioning of the OTSG pressure controlloop. These instruments are also classified as 8G 1.97 instruments which are required for post accident monitoring. The team considered this condition as a generic problem since all eight OTSG pressure transmitters share a common desig As a result of the team's concern, the licensee initiated CR 197 0058 together with a corresponding analysis to justify that the existing condition without the clips is seismically adequate and the svstem is operable, Also, Action item 217 under this CR was issued to walkdown alllike model pressure transmitters (Rosemount 1153) to verify any generic imp lication. As a result of '.he walkdown,it was concluded that those that were found were of different mountir,g design and that the mounting clip problem on this specific model instrument was not wiuespread. The failure to translate seismic requirements to field installstions of OTSG pressure transmitters represents a weakness with respect to 10 CFR Part 50, Appendix B, Criterion lli. ' Design Control," and is identified as Unresolved item 50-313/97 201 1 When the EFW pump room was walked down, the inspection team found approximately 10 FT of unsupported instrument sensing line for instrument looo PT/PI 2811. The team noted that there was no design basis for supporting instrument sensing lines for Unit 1. From discussions with the licensee, the Unit 2 design allows 21/2 FT maximum unsupported span, which has been exceeded in this case. This instrumentation is a Class O pressure boundary, it moitors the discharge pressure of the turbine-driven EFW pump, provides

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h d

localindication and sends an input signal to the safety parameter display system (SPDS).

The licensee was made cognizant of this issue and CR 1 97 0074 was written to document this condition and justify system operability. With the issuance of two CRs related to the instrumentation supports, the licensee determined that these observations have a gene implication that warranted immediate attention. CR 197 0074 was elevated to a significant condition and subsequently initiated a detailed walkdown of the EFW, OH/LPI and high- <

pressure injection (HPl) systems to evaluate the extent of the condition. As a result of this walkdown, several discrepancies were found, categorized as follows: {

i

+

existing installations that are undocumented or unanalyzed, such as tubing supports exceeding the 30 inch unsupported span and the use of non standard instrument support *

,

lack of design bases or criteria for determining instrument line slope +

drawing discrepancies between drawings and the as built conditio *

missing or incorrect tags on instrument valves.

s CR 197 0074 was scheduled for presentation to the Corrective Action Review Board (CARB) on April 6,1997, to be followed by a root cause analysis of the problem and implementation of corrective action. The inspection team also noted that this effort has been expanded to other ANO 1 systems outside the scope of this inspection. An example

,

of this is CR 197 0087 which was issued concerning EDG instrumentation lines that were

,

inadequately supported. A similar walkdown has also been initiated for ANO 2. The lack of design basis for the OTSG instrument sansing lines is defined as Unresolved item 50-313/97 201 1 A walkdown of the control room was performed to verify the RG 1.97 indicators and '

recorders for the EFW and LPI systems. Types A, B and C, Category 1 & 2, and related i

instruments are identified with green dots on the instrument nameplates, The team noted that Type D and E variables and open/close indicating lights are not provided with any kind special of identification. in response, the licensee indicated that RG 1.97 Types A, B and C, Category 1 & 2 variables, as identified in FSAR Table 7.11 A, "RG 1.97 Instrumentation,"

and Topical Report ULD 0 TOP 05, are the only indicators that have green dots. Per the RG,

no specialidentification is required for Types D and E instruments. The licensee had taken l

exception to the open/close indicators which were previously reviewed and accepted by the NRC (reference ANO 1 correspondence OCNA089320). The following items were also noted during the control room walkdown:

d

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FSAR Table 7.11 A is not consistent with the as installed condition with respect to ranges and units for LIS 4204 and LIS 4205. An LDCR to correct this discrepancy has been initiated as discussed in Section E1.2.4.2.2 in this repor *

Green dots were temporarily installed to identify some RG 1.97 indicators on console C09. The licensee subsequently initiated installation of permanent green dots in accordance with the appropriate ANO nameplate and tagging procedur . _ . . ,_ . - - - - - . . . . . _ _ --.

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During a walkdown of the auxiliary building, the team observed a misaligned limit switch on valve CV 4804. This misalignment caused the actuator cam to drag on the limit switch arm toward the end of the valve stroke instead of riding on the roller bearing. Extended operation with this condition will eventually wear out the limit switch arm which could result in a faulty actuation. The limit switch is Seismic fl/l and its function is to provide a signal for valve position indication for the R8 vent system. The safety significance of this condition is that there could be other safety related limit switches in the plant with similar mounting configuration that could fail and affect operation of associated safety related equipment. Maintenance procedure 1025.033, " Control of Post Maintenance Testing," for each plant safety component esteblishes a unique maintenance plan (including required post maintenance testing). The licensee recognized that in this instance the maintenance plan did not specifically direct the workers to check and verify this limit switch cam assembly. As a remedy, Job Order 00954697 was reopened to implement corrective action and a revision to Work Plan 008626 was initiated to add a step to verify proper operation of the limit switch mechanism. A followup action was performed to walk down alllimit l switches of like mndet numbers which was narrowed down to another limit switch of mounting design. Likewise, a revision was also incorporated in the corresponding work plan for that limit switch. As a proactive measure, the licensee also performed a search for simitar conditions in Unit 2 and there were no limit switches similar to that of Unit team witnessed the monthly surveillance testing of EFIC Channel D per ANO 1 procedure 1304.208.

! This test was performed to satisfy the requirements of TS instrument Surveillance Table 4.1 1. The team's identification of an inadequate post modification test for CV 4804 and the need for the licensee to revise Work008626 Plan to add a step to verify the proper operation of the limit switch assembly is identified as Unresolved item 50-313/97 201 2 E1.2.5.2.3 Conclusion The lack of design basis for instrument tubing installation is being addressed by the license As mdicated by the licensee, a root cause analysis and subsequent corrective action on a caue to casi. basis will be determined and implemented in accordance with the appropriate plant procedures. The identification of RG 1.97 related instruments in the main control room meets the intent of the RG. The limit switch misalignment issue for CV 4804 has been addressed by the licensee by evaluating the impact on all affected components at both ANO 1 and 2 and implementing necessary procedural change E1.3 Decay Heat / Low Pressure injection System E1.3.1 System Overview The ANO 1 DH4PI system is a dual purpose system. It is designed to remove decay heat from the core to provide a > sensible heat from the RCS during the latter stages of cooldown (DH) and as of automatically injecting bc:sted water into the reactor vessel (LPI) for cooling of the core in the event of a LOC E1.3.1.1 System Description The DH/LPI system is described in Chapter 6 and Section 9.5 of the ANO 1 FSA . .

Decay Heat

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The DH system normally takes suction from the reactor coolant outlet line and delivers the water back to the reactor through the core flooding norries after passing through the DH pumps and coolers. The DH system may be opwated when the reactor pressure is within the system design pressure for cooldown of the system to refueling temperatures. During this operation, the pressurizer is cooled by auxiliary spray from the DH system in accordance with operating procedures which prescribe the required cooldown rate limitations. The decay heat is transferred to the SW system by the DH coolers. Two DH pumps are arranged in parallel and are designed for continuous operation during the period required for removal of decay heat for refueling. If two pumps are in service, the design flow can cool the RCS from 280'F to 140'F in 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months . Two DH coolers remove the decay heat from the circulated reactor coolant during a routine shutdown. Both coolers can be operated to cool the circulating reactor csolant from 280'F to 140'F in 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months , if there is only one cooler and pump available, the cooldown period to 140'F is approximately 5 day The borated water storage tank (BWST)is located outside the RB and the AB. It contains a minimum of 2270 ppm boron in solution and is used both for emergency core injection and i filling the fuel transfer canal during refueling. The BWST also supplied borated water for emergency cooling to the RB spray system, the DH system (LPl function), and the makeup and purification system (HPl function). It also supplies makeup water to the spent fuel cooling syste The bulk water temperature in the BWST must be maintained abovo 40"F under all weather conditions. For this purpose four electric immersion heaters with 180 kW total capacity are provided. Each of these four heaters consists of three elements of 15 kW capacity each and each of these elements has its own individual circuit breaker. As a result, the failure of any one heater reduces the total capacity by only 15 kW. Since the insulation on the vessel limits the heat losses to a maximum of 20 kW under the worst conditions, the effect of a BWST heater failure is insignificant. A tank vacuum breaker is provided to ensure proper tank draining. The heating of the tank is considered to be a protection against treezing fo'

the tank as well as the vacuum breaker.- However, to provide additional protection to the vacuum breaker, additional heat tracing is provided on the vacuum breaker itsel The tank is located in a fenced area and, during normal operation, the radioactivity concentration in the tank is maintained so as to limit the dose rate to personnel in the area at or below 1.0 mrom/hr. During refueling operations the dose rate at the surface of the tank is monitored. in the event that this dose rate exceeds 1 mrom/hr, a " controlled area" is established and the radioactivity concentrations reduced via circulation of the tank inventory-through the spent fuel pool domineralizer and spent fuel pool filto Low Pressure injection The LPI system is designed to maintain core cooling for large break sizes and operates independently of and in addition to.the HPl system. Automatic actuation of LPI is initiated by low RCS pressure or high RB pressure. Initiation provides the following actions:

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The valves in the lines connecting the BWST to the LPI pump suction headers ope <

+

The valve in each LPlline open * DHR pumps star *

DHR cooler service water isolation valves open in response to the pump star *

DHR cooler outlet valves open and cooler bypass valves clos LPils accomplished through two separate flow paths, each including one pump and one heat exchanger (cooler) and terminating directly in the reactor vessel through core flooding nozzles located on opposite sides of the vessel. A cros9 connection betseen the two LPI linen provides the capability of injecting an adequate supply of borated water for core cooling even in the event of a core flood line ruptur The initialinjection of water by the LPI System involves pumping water from the BWST into the reactor vessel. With all engineered safeguards pumps operating and assuming the maximum break size, this mode of operation lasts for a minimum of about 25 minute When the BWST reaches an indicated level of 6 feet, the operator opens the suction valves from the RB sump permitting recirculation of the spilled reactor coolant and injection water and closes the BWST outlet valves. An alarm is also annunciated in the control room to indicate a low levelin the BWST. Check valves and the closed BWST outlet valves in the line from the BWST provide redundant isolation to prevent backflow into the BWS E1.3.1.2 System Functions The design basis for the DH/LPI system, as stated in the governing upper level document (ULD 1 SYS 04, Revision 0)is summarized belo E1.3.1.2.1 Safety Related Functions The LPI system has the follow ng safety related functions:

Transfer heat from the reactor core following a LOCA such that fuel and clad damage that could interfere with continued effective cooling is prevented and clad metal-water reaction is limited. Additional criteria for the ECCS are limits on calculated peak cladding temperature, cladding oxidation, hydrogen generation, cootable geometry, and long term coolin *

Provide water to the RB spray pumps' suction piping from the BWST and RB sum +

Supply oper water to the HPl pumps from the BWST and from the RB sump for piggyback

. u *

Provide borated water from the BWST at a sufficient concentration to ensure that there is an adequate shutdown margin under accident conditions and to fill the RB to an adequate level to allow recirculation from the RB sum .

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Provide long term heat removal from the RB post LOCA to reduce RB temperature and pressur E1.3.1.2.2 Regulatory / Safety Significant Functions The DH system at ANO 1 has the following regulatory / safety significant functions:

+

Remove decay heat from the reactor core and sensible heat from the RCS during the latter stages of cooldown at a rate such that specified acceptable fuel design limits and design conditions of the RCS pressure boundary are not exceede *

Circulate reactor coolant to pievent boron stratification and to minimize the effects of a boron dilution acciden +

Provide an alternative supply of borated water from the BWST for volume contraction during cooldown to cold shutdow +

Provide redundant cooling, inventory addition methods, and instrumentation for events involving a loss of DH *

Prov;de a niakeup flow path to the spent fuel poo E1.3.2 Mechanical Design Review E1.3.2.1 BWST Pressure and Vacu"m Relief Valve E1.3.2.1.1 Scope of Review in this portion of the raschanical design review, the team assessed the operability of the BWST in the past with the original pressure and vacuum relief valve installed, during the interim phase with two different foreign material exclusion (FME) coverings on the tank flange, and in its current configuration with the original valve removed and a long term FME covering in plac E1.3.2.1.2 inspection Findings The BWST was originally supplied with a single 8 inch diameter pressure and vacuum relief valve in accordance with the requirements of the original code of construction (AWWA D-100). This valve was to provide protection against over pressurization of the tank or against a vacuum being pulled on the tank (when the levelis lowered) causing it to collaps The tank also has a 4 inch diameter overflow line that is normally open to the atmosphere and can act in conjunction with the pressure and vacuum relief valve to mitigate pressure and vacuum conditions within the BWST (although that is not its design function).

On December 4,1996, with the unit at power, the pressure and vacuum relief valve (PSV-1412) was removed for surveillance testing usin;, Procedure 1306.034 under Job Order 00952943. During testing it was discovered that the valve had a through wall crack in its

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flange and therefore could not be reinstalled. At the time of valve removal, a white plastic material was placed over the tank flange to act as a short term FME covering until the valve could be reinstalled, it is estimated that the white plastic material was on the tank flange for approximately 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months before it was replaced. Since the valve could not be reinstalled after testing, a long term FME covering had to be placed on the tank flange. This initiallo term covering was installed on December 11,1996, under the original job order (00957859 issued on December 5,1996) and consisted of a blind flange separated from the tank flang by a gap of approximately 1 inch with a wire mesh FME screen over the openin On December 12,1996, under Job Order 00957910, the initiallong term covering was modified to enlarge the gap to approximately 3 inch and maintain an FME screen over the opening. On February 18,1997 Temporary Alteration 97 1 001 was af fixed to this covering to allow it to remain in place while a permanent repair or replacement is designed and installed. At the end of the inspection, the covering was stillin plac During this period, three condition reports were written by the licensee to document perceived coverings. CRproblems with the removal of the valve and installation of the required FME 196 0063, written on December 5,1996, documented the through wall crack on the valve flange end performed an operability assessment of the BWST with the valve off for an indefinite period of time. This operability assessment considered the FME covering of a blind flange separated from the tank flange by a 1 inch gap plus the 4 inch ,

diameter overflow line as both being available to provide vacuum relief. However, this operability flow blockage.analysis did not consider the FME screen over the 1 inch gap and the resultant CR 197 0019 was written on January 22,1997 of withthe BWST the use with the white plastic materialinstalled on the tank flange as well as of procedure 1306.034 to remove the valve for testing with the unit at powe The procedure was written for doing testing when the unit is in a refuelir'g outage (BWST empty) and was not intended to be used for testing during power operation (the 10CFR50.59 97 0031 evaluation for this procedure never addressed doing testing at power).

CR 1-was written on February 3,1997, to address the assumption that the 4" diameter flow path provided by the tank overflow line was a backup to the pressure and vacuum relief valve. This assumption affects the operability assessment of the BWST in its origin configuration should this valve have failed. Initial calculations have shown that the 4-inch overflow line will not provide adequate vacuum relief and therefore is not redundant to the pressure and vacuum relief valve. These three condition reports contain many Suggested Corrective Actions this inspection report discusses only those actions assigned to D Engineering involving BWST operabilit Tank operability must be determined for all three configurations involving FME coverin well as for the original configuration with the pressure and vacuum relief valve in place. The most straightforward way to determine operability is to compare the flow area of the original pressure and vacuum relief valve with the flow area created by the FME coveri the FME flow area is equal to or greater than the flow area of the original valve than the

,

tank was operable with that particular FME covering in place. The BWST was considered operable by the licensee with the blind flange end 3 inch gap FME covering in plac Calculation 97 E 0012-01 was performed by the licensee to determine the gap width

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necessary to allow the same flow area as the pressure and vacuum relief valve taking account of the blockage created by the surrounding FME screen. The screen is on the outside of the tank flange and, assuming the worst case mesh size, blocks 48.3% of the available flow area. The remaining open area created by the 3 inch gap provides a flow area greater than that of the pressure and vacuum relief valve. In addition, this current configuratiore could be considered as superior to the original pressure and vacuum relief valve because it is a passive device, whereas the original valve had to change position in order to relieve a vacuu The BWST was operable for the configuration involving the blind flange with a 1 inch ga This FME covering also had a screen over the opening. The initial operability assessment done as part of CR 196 0663 did not consider the flow blockage created by this scree That operability determination included the flow through both the FME covering and the 4 inch overflow line. Summing the two flow areas gave a total flow area slightly greater than that of the original valve. However, if the screen over the 1 inch gap is included, the total flow area is less than that of the original valve. Because of this apparmt flow area discrepancy, the licensee did a more detailed review of the actual co, guration of the original pressure and vacuum relief valve. The licensee determined that the original valve had a screen over the 8 inch vacuum breaker opening that limited the actual finw area to 76% of the total area and that the valve had been flow tested in that configuration by the vendor. Taking into account this reduction in flow area of the original valve, the licensee was able to show that the total flow area of the FME covering with the 1 inch screened Oap plus the 4-inch overflow line was greater than that of the original v'alve and therefore that the BWST was operable. This information will be appended to the original operability evaluation done as part of CR 196 066 The BWST was operable for the 30-hour period of time when the white plastic material was installed over the tank flange (just after the pressure and vacuum relief valve was removed The licensee's basis was that the plastic material was installed with two slits cut in it that would allow the tank to * breathe" through this covering. In the event of a full flow ECCS drawdown of the BWST, the plastic covering may be torn away leaving the full flow area of the tank flange open to relieve potential vacuum conditions. The licensee is evaluating this as part of CR 197 001 Finally, the BWST was operable in its initial configuration with the pressure and vacuum relief valve installed on the tank flange. The BWST was originally supplied with only one pressure and vacuum relief valve and a single 4-inch diameter overflow line. In the event of the failure of this valve during tha D.,:s injection phase of a LOCA, the BWST would be left without a sufficient method of vacuum relief and could potentially fail under the vacuum created by full ECCS drawdown of the tank. Under this scenario, tank failure would not be prevented by vacuum relief through the 4 inch overflow line since this line is not a full fiow backup to the pressure and vacuum relief valve. The licensee response was that the pressure and vacuum relief valve is a self-actuating device like a check valve or safety relief valve and that the licensing basis of Unit 1 does not require postulating the single failure of such a device during the injection phase of a LOCA (see ULD 0-TOP 18, " Single Failure Criteria," and SECY-77 439).

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The team's review determined that the licensee's evaluation for the BWST va removal pursuant to 10 CFR Part 50.69 was inadequate since it did not consider the flow restrictions caused by the actualinstalled configuration of the FME screen. In addition, the neod to review the final resolution of corrective actions associated with 197 0019 andCR CR 197 0031 is identified as Unresolved Item 50 313/97 201 2 E1.3.2.1.3 Conclusion The inspection team agrees with the licensee that the BWST was operable in its initial configuration with the original pressure and vacuum relief valve in place. The DWST was determined by the licensee to be operable with the two temporary FME coverings in plac The licensee further determined that the BWS1 was operable with th) 3 inch gap FME covering installed on the tank flange. This covering is to remain in place until a permanent replacement is selected. The licensee has not committed to a like.for like replacement of the pressure and vacuum relief valve and will, instead, actively consider other modifications

"...to provide additional assurance of adequate design features and design margins." The initial operability evaluation for the 1 inch gap FME covering was incorrect but was being amended by the licensee to incorporate recent information provided by the valve vendo E1.3.2.2 BWST Vortexing and Pump NPSH E1.3.2.2.1 Scope of Review in this portion of the mechanical oesign review, the team evaluated the extent of vortexi in the BWST during the latter stages of ECCS injection flow after a line break, and considered the effects of this vortexing on the NPSH calculations for the HPI pumps, LPl pumps, and RB spray pump E1.3.2.2.2 Inspection Findings The potential for vortexing at the inlet of the suction piping for the ECCS pumps inside the BWST is important because vortexing can cause air entrainment in the suction flow to these pumps. As little as 1% air entrainment by volume in the suction flow to a pump can increase its NPSH requirements by a factor of 1.5 per Appendix A of RG 1.B2. This potential exists only in the Unit 1 BWST: the equivalent tank for Unit 2 (refueling water tank) has a vortex breaker welded into the inlet of the suction piping for the ECCS pumps

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To examine the problem of vortexing in the BWST, the licensee prepared ER 93 R 1002 0

"ANO 1 BWST Outlet Vortex Suppressor," dated February 5,1993. This report uses the information, including test data, presented in NRC publication NUREG 0897 for containment emergency surnps and extends it to large water storage tanks. The report makes a strong case for vortexing causing less than 1% air entrainment in the suction flow to the ECCS pumps. Moreover, air entrainment can occur only at high flow rates in the suction piping (several ECCS pumps operating simultaneously) coincident with low levelin the BWST (jus above the manual switchover level to the RB sump, which is 6 feet). The report concludes

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" .that the NPSH available to the ECCS pumps will not be adversely affected by vortexin for all modes of pump operation and that no changes need be made to the BWST suction piping."

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While preparing for the inspection, the licensee issued CR 197 0039 against the above engineering report. The condition report states that instrument error was not included in the report which could result m a lower actual BWST level than that indicated. Therefore the report under predicts the impact of air entrainment on the NPSH calculations for the ECCS pumps when drawing suction from the BW4T at low level, The condition report includes an operability determination which concludes that the ECCS pumps are still operable based on the following:

The only pumps that could have NPSH concerns are the HPl pumps because of their high NPSH requirements (approximately 38 FT, compared to approximately 10 to 15 FT for the LPI and RB spray pumps). The HPl pumps may have an NPSH problem nnl when the LPI pumps and RB spray pumps are also operating (high flow in the suction pip and the BWST levelis just above the sw!tchover point to the RB sump. In this situation the HPl pump can be shut down because the LPI pumps alone can orovide adequate core cooling. Part of the requirements to disposition CR 197 0039 is the issuance of a comprehensive corrective action plan. This plan is expected to require completion of Revision 1 of the ER and consideration of any impact of the results of that report on other calculations such as the latest revision of the NPSH calculations for the LPI and RB spray pumps when taking suction from the BWST (refer to calculation 89 E 0010 26, Revision 5).

The need to review Revision 1 to the engineering report and the revised NPSH calculetions for the ECCS pumps is identified as Unresolved item 50 313/97 201 2 Design Engineering has committed to contact the vendor of the HPl pumps to determine if any testing was done for the pumps running under conditions of inadequate NPSH. If such testing exists and if the pumps ran for a reasonable length of time with no damage, then this would give additional confidence that the HPl pumps could run for a specific length of time with air entrainment in the suction flow before they are shut down by Operation E1.3.2.2.3 Conclusion Upon completion of the calculations, Revision 1 of the ER warrants licensee review of NPSH calculations for ell ECCS pumps considering all suction sources and modes of operatio The team noted that the licensee review of all appropriate calculation input parameters such as the correct recirculation flow and post LOCA sump water temperature was warranted so as to properly document this portion of the design basis of the ECCS pumps.

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E1.3.2.3 Post Accident Radioactive Releases from the BWST E1.3.2.3.1 Scope of Review During this portion of the mechanical design review, the team assessed the possibility of radioactive leakage into the BWST in a post accident situation. This is important because the BWST is vented to the atmosphere, and a contaminated BWST would become a source of unmonitored radioactive release to the environment after an acciden E1.3.2.3.2 inspection Findings Post accident release from the BWST was discussed in NRC IN 9156, " Potential Radioactive Leakage to Tank Vented to Atmosphere." The licensee has evaluated the IN for applicability and is currently in discussions with the NRC concerning the applicability of the issue with

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respect to ANO 1. The team noted that CR C 96 0135 has been opened to track this issue howewr, the team did not review it during the inspection. High levels of radioactivity can be introduced into the LPI system when it is transferred into its recirculation mode af ter a design basis LOCA. In this mode the DH pumps draw suction from the RB sump and circulate this water through the OH coolers and back to the RPV. The water in the RB sump can be highly contaminated since most of it was spilled from the break in the RC Radioactivity from the LPI system can then be introduced into the BWST through leakage through the valves in the test lines that connect the DH cooler discharge piping to the BWS In the post accidant situation, there will be two closed gate valves in series in each of the test lines back to the BWST. Having this double barrier against leakage in each test line is important in reducing the possibility of getting significant quantities of radioactivity into the BWST. One of these valves, DH 9, is open approximately 35% of the time that the unit is in normal operation to allow for tank purification and mixing through the spent fuel cooling system. Operator action is required to close this valve in the accident situttion. When the ECCS pumps begin to draw suction from the BWST, a BWST low level signal will be

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received in the control room. This sigralis generated when there is approximately 39 feet of water Stillleft in the BWST. Upon receipt of this low levelindication, the operator will check the valve abgnment with interconnecting systems to prevent inadvertent draining of the tank (see Procedure 1203.012H). If valve DH 9 is open, the operator will close it as part of shutting down tank purification (see Procedure 1104.006). This valve will be closed before the LPI system switches to the recirculation mode and will thereby establish the double barrier agair.r leakage into the BWST.

l

! The team reviewed the 50.09 safety evaluation conducted as part of Temporary Alteration 071001 (discussed in Section E1.3.2.1.2 of this report) and concluded that it did not adequately address radioactive releases from the BWST that could potentially occur through the 3 inch gap FME covering on the flange of the tank vent. The evaluation addressed only offsite exposure, despite the fact that control room exposure may be more significantly affected by releases from the BWST. In addition, the evaluation made no mention of the response to IN 9156 and, therefore, did not consider the licensee work related to this issu The failure of the safety evaluation to adequately consider the effects of control room dose is identified as a weakness with respect to 10 CFR Part 50.59 and is identified as Unresolved item 50 313/97 201 2 E1.3.2.3.3 Conclusion The team noted that the issue of potential unmonitored radioactive releases from the Unit 1 BWST has been discussed previously between the licensee and the NRC staff and the team is also satisfied that the proper operating procedures are in place to close valve DH 9 in the post accident situation before the LPI system switches to the recirculation mode. The valve is a manually operated valv x

. 4 E1.3.2.4 LPI System Design Basis Flow and Surveillance Testing E1.3.2.4.1 Scope of Review in this portion of the mechanical design review, the team identified the design basis flow the LPl system and considered how these requirements are reflected in the licensee's surveillance testing of the LPI pump '

E1.3.2.4.2 Inspection Findings The design basis for LPI flow is referenced in Table 1 of ER 96 R 1003-0 This table lists the required flows into the RCS as determined at the injection nozzles on the RPV and th Calculation 92 E 0077 03, corresponding RCS pressures which were used as input t using these design basis flows into the RCS and the RCS pressuras as input, calculated the effects of system resistance, recirculation flow, and suctiors head to determine the flow at the LPI pumps and the differential pressure across t pumps nozzles onwhich would be required in order to provide the necessary flow at the injection the RP Section XI of the ASME B&PV Code defines surveillance testing acceptance criteria and allows up to 10% degradation in pump performance before declaring the pump inoperable When the pump performance requirements from Calculation 92 E 0077 03 are compared to the vendor certified pump curve, the minimum requirement would not be satisfied at some l

flows with the code allowable 10% degradation. The acceptance criteria established in Section XI of the Code with regard to inservice testing states that the greater of 10%

degradation or the requirements from Calculation 92 E 0077 03 be med. Since the smalles margin between the vendor pump curve and the Section XI accoume criteria is at 995 gpm and the surveillance test is actually performed at 3000 gpni, the acceptance criteria at 3000 gpm was chosen to limit pump degradation to that which would be allowed at 995 ( gpm. This acceptable pump degradation curve is greater than the 10% pump degradation

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curve basis design allowed flow.by the Code and ensures that the pump will always provide the required E1.3.2.4.3 Conclusion The team finds that this method of determining the acceptance criteria for inservice

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reflects the actual performance requirements of the LPt pumps and is in compliance with Section XI of the ASME B&PV Cod E1.3.3 Electrical Design Review Section E1.2.3 includes the electrical review for the DH/LPI system E1.3.4 instrumentation and Control Design Review in the I&C design review of the DHILPI system, the team assessed various design and

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licensing basis documents that included the FSAR, TS calculations, and other specific attribute _ _ _ _ -

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t E1.3.4.1 EFW Initiation and Control Logic E1.3.4.1.1 bcope of Review i

in this portion of the l&C review, the team sought to verify the capability of the DP/LPI actuation logic to automatically start the DH pumps and initiate isolation vah e line u E1.3.4.1.2 Inspection Findings The LPl system is automatically initiated by the ESAS activation by low RCS pressurt, o'

high RB pressure. Initiation of the LPI consists of automatic startup of the DHR pumps and alignment of valves to direct LPI flow from the BWST to the RCS in response to pump start, the corresponding DHR cooler service water isolation valves open and the room cooler fan starts. The isolation valve logic is shown on Drawing M 418, Sheets 1,2, & 3, and the P&lD is shown on Drawing M 230, Sheet 1. These documents were reviewed for functionality and are in conformance with Sections 6.1.2.1 and 7.1.3 of the FSA Emergency operation of the DH/LPIis initiated by a low RCS pressure of 1585 psig or hig RB pressure of 4 psig, as indicated by input signals to the ESAS from pressure transmitters PT 1020, PT 1022, PT 1040 (RCS) and PT 2405, PT 2406, PT 2407 (RB). The team's review 0003 19)of the results of uncertainty and set point calculations (83 D 1049 05 and 85 EO-for the RCS and RB pressure instrument loops was consistent with the SIMS database and other documents reviewed.

E1.3.4.1.3 Conclusion The DH/LPl system initiation and controllogic is designed in accordance with the FSA The design documents and as built condition meet the system functional requirements as discussed in the FSA E1.3.4.2 BWST Levelinstrumentation E1.3.4.2.1 Scupe of Review in this portion of the l&C design review, the team sought to verify the conformance of the level of instrumentation with the guidance of RG 1.97, and assessed the capability of the tank to provide high , low , and lo lo level alarm E1.3.4.2.2 Inspection Findings The BWST instrumentation provides the following functions:

Indicate and record BWST levelindication and recording in the main control room to

, satisfy RG 1.97 guidanc *

Alarm high and low BWST levels within the TS limit . - .- - - - _ . - - . - . _ _ - - - -- _ _ - _ - -

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Alarm low GWST level to alert the operator to manually transfer water supply from

the BWST to the RB sum >

As defined in TS 3.3.1, the design basis for the BWST levelinstrumentation requires two channels to be operable whenever containment integrity is established, if found inoperable, it has to be restored to operable status within 7 days with the redundant instrument channel in operable conditione otherwise, reacto operation is permissible only during the succeeding seven days. If this condition cannot be met, the reactor shall be in hot shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months or in cold shutdown within an additional 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if not corrected. The BWST level instrumentation consists of transmitter loops LT 1411 and LT 1421. These redundant transmitters provide separate signals to level recorder LR 1411 and indicator LIS 1421 on

main control room panels C14 and C18. The instrument loops consist of level transmitters,

< signal converters, isolators and control room instruments which are all classified as safety-related and powered f.om a Class 1E sourc Per TS 3.3.1, the BWST is required to contain a level of 40.2 + /1.8 FT. FSAR 6.1.2. states that the operator manually transfers the cooling water supply from the BWST to the RB sump when the BWST reaches a level of 6 FT. Logic drawing M 418, sheet 2. and calibration procedures 1304 012 and 1304.197 show the following alarm set points:

High level Set point 41.1 FT Low-level Set point 39.3 FT Lo lo Level Set point 7.0 FT The team reviewed set point basis ducument 91 R 1018-02, set point repr '1 R 1017-16, and loop accuracy calculation 83 D 1 * 53 01 for the BWST levelinstrumer.. uon. The team found that the calculated values are consistent with the drawings and calibration procedures. These set points envelope the TS and FSAR requirement The team identified a discrepancy between the FSAR Table 711 A and data sheet M 51 sheet 287, Revision 0, with regard to instrument ranges for instrument loops LT 1411 and LT 1421. FSAR Table 711 A, Type A05, shows these loops with a range of * bottom to top." As indicated on drawing M 516, sheet 287, and instrument data sheets, the associated instruments (LIS 1411. LIS 1421 and LR 1411) all have a range of "0-45 FT

, H3 0." As later verified during a walkdown of the control room, the actualinstalled

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instruments have a range of 0-45 FT H3 0. The need to correct discrepancies and review the corrective actions rclated to the a,ccuracy of the FSAR represents a weakness with respect to 10 CFR Part 50.71(e) and as e result, the team identified this issue as Unresolved item 50 313/97 201 1 E1.3.4.2.3 Conclusion The BWST levelinstrumentation meets the design intent of FSAR Section 7.11 A, the requirements of RG 1.97 and the level alarm instrumentation is in agreement with FSAR Sections 6.1.2.1.2 and TS 3. _ _ __ _

_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

_ - - - -

. .

E1.3.4.3 DH/LPI Flow instrumentation E1.3.4.3.1 Scope of Review During this portion of the l&C design review, the team sought to verify the conformance of the DH/LPl flow instrumentation with the requirements of NUREG 0578 and RG 1.9 >

E1.3.4.3.2 Inspection Findings FSAR Table 7.11 A, drawing M 232, sheet 1, and Topical ULD 0 TOP 05 provide the design bases for the DH/LPI flow instrumentation. The DH pumps have a design flow of 3000 gpm at a pressure of 450 psig. RG 1.97 requires a control room flow indication range of 110%

of the design flow, or 3300 gpm. The flow indicators on Panel C09 (F11401 for Pump P34A and FI 1402 for Pump P348), each have a range of 0 4500 gpm, which envelopes the RG 1.97 requirement. This data is consistent among the various documents that were reviewed. The flow instrument loops are redundant, safety grade, and powered from o Class 1E source. The licensee has classified these indicators as RG 1.97 Type A, Category 1 instrumentation, in reviewing the various documents provided by the licensee, the team noted a discrepancy in FSAR Table 711 A where Type Code A03 control room recorders are

,

identified as *2 dual pen" recorders. Actualinstalled instruments are 2 single pen recorder Subsequent to this observation, the licensee initiated action to correct this discrepanc Instrument loop error calculations (85 EO 000318 and 80 D 1083C 01) were prepared for the DH/LPI flow instrumentation. These calculations were reviewed by the team and the team found that adequate tolerance has been provided to account for instrument loop I

inaccuracy. The need to correct discrepancies and review the corrective actions related to the accuracy of the FSAR represents a weakness with respect to 10 CFR Part 50.71(e) and

{ as a result, the team identified this issue as Unresolved item 50 313/97 201 1 E1.3.4.3.3 Conclusion The DH/LPI flow instrumentation meets the design intent of NUREG 0578, FSAR Table 7.11 A and RG 1.9 E1.3.5 DH/LPI Walkdown Observations and Results E1.3.5.1 Mechanical Walkdown and in Plant Observations E1.3.5.1.1 Scope of Review The team performed a walkdown on the AND-1 DH/LPI system, including the BWST outside the AB, and associated piping and instrumentation in that area. The remainder of the walkdown included the AB DH/LPl vaults, the piping penetrations into the RB, the main control room, and all of the instrumentation associated with the system in all areas observed. The initial walkdown was conducted on February 12,1997. Subsequent walkdowns and operation observations were conducted throughout the team's site visit The sodium hydroxide and sodium thiosulfate drawdown addition tanks were also observed adjacent to the BWS j

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.

E1.3.5.1.2 Inspection Findings The ANO 1 BWST (T 3)is located just outside the RB and the AB with access through the post accident sampling station (PASS) area. The tank is a seismic Class 1 structure erected with a plastic liner and storage capacity of 380,000 gallons. The tank is bsulated and has internalimmersion heaters to maintain the tank between 45 and 50' F. The licensee described an existing problem with the pressure / vacuum relief assembly designed to be installed on the tank. The relief assembly was removed because the licensee identified a crack on the component's flange during a recent maintenance calibration check of the pressure / vacuum relief assembly. Several CRs have been issued against the multiple problems identified in the replacement proces During the walkdown, it was noted that the sodium thiosulfato tank adjacent to the BWST had the isolation valves connecting to the system closed and locked. The original design required this tank to supply chemicals during drawdown of the building spray system operation. The licensee stated that the system was deleted from need several years ago, and is no longer in use. The system and tank are still connected to the BWST/ sodium hydroxide drawdown piping, and isolated by procedure with locked valves. Drawing M-236 Revision 82, currently shows the sodium thiosulfate tank with piping connected and valves closed, to the RB spray injection system. The licensee has drained the tank contents and revised the P&lDs to indicate these valves connecting this system to be closed, and operating procedure 1015.035 requires the valves to be closed and locked. Administrativo controls are relied upon to keep the system isolated. When questioned why the piping connections are not cut and capped to preclude any possible misalignment, the licensee responded that there was a potential future use of the tank that never has been followed up and processed into a modification. The licensee also stated that the tank and associated piping will remain on the drawing as long as it is physically in the plan The team also noted what appeared to be significant dried boron residue on the DH/LPIlines in the "A" DH/BS vault. The licensee's initial response to the team's concern of the condition reported that the residue is a result of accurnu!sted minor spillage that occurs whenever the lines are broken and flushed in an area ahon this vault, following an outag The licensee initiated a job request to clean up the material. Since the area is a significant

" radiation area," the work will be carefully planned to follow "ALARA" requirements during any cleanup process. ANO 1 initiated a chemical analysis of the material deposited on the piping which revealed a compound of calcium / manganese carbonate and not boron it appears that the matenal may have come from water used during repair of the seal area and not leakage of the vault penetration, it was noted during the walkdown that to provide auxiliary spray to the pressurizer when on losv pressure injection following an accident, operators are required to manually position seven valves to initiate spray flow. Of these valves, three are in the "B" DH vault and two are in the "A" DH vault. Procedures direct the operators to align these valves before switching the pumps' suction from the BW3T source to the RB sump source. The team questioned the time involved and the potential for delaying the shif t of pump suction in an emergency scenario as conflicting actions, possibly complicated by radiation concerns. ANO 1 responded that the auxiliary spray was one of three means for controlling boron concentration (precipitation mitigation) in the RCS, and was therefore not required to be initiate _- . _ - _ . -_ _ _ _ _ _ _ _ -

_

_ _ _ - _ - _ - _ _ _ - _ _ . ..

. .

With respect to cooldown requirements, the licensee stated that Abnormal Operating Procedure 1203.013, * Natural Circulation Cooldown," describes pressurizer pressure control operations using the heaters' on/off controls, which will cause surging in and out, interchanging water with the RCS during cooldown. During events in which HPl pumps continue to operate, the auxiliary spray system can be supplied from HPl discharge pressur With no sprays or level cycling by pressure swings, the pressurizer will have no direct

, cooldown method, the heaters will be de energized, and the temperature will drift down toward ambient. During the walkrfown, the team noted several areas of the DHrLPl system piping and components where lead shield blankets were attached to reduce local radiation exposures. The team questioned whether the ficansee had analyzed the structures for the additionalloading on supports and hangers of the associated shielded components. The licensee responded that procedure 1601.003, " Control of Temporary Shielding," Revision establishes a method for administrative control of temporary shielding used to reduce personnel radiation doses. The procedure accounts for reviewing all effects of attaching shielding to piping and components and the team noted that the engineering evaluations were performed as require E1.3.5.1.3 Conclusions The BWST pressure / vacuum relief valve assembly poblems have been discussed in the mechanical section discussion of this DH/LPIinspection report. No other team concerns were noted on the BWST portion of this walkdow The team's concern regarding the no longer used sodium thiosulfate tank was adequately addressed by the licensee's administrative controls requiring the isolation valves to be locked closed as part of the systems valve lineup checklis The concern over the residue on the DH/LPI lines in the "A" vault, below the penetration through the overhead, was resolved by performing chemical analysis to determine that the material was not boron and that the vault isolation requirements were not compromise The concern regarding operator ability t's perform valve lineup changes in providing DH/LPI pump initiated spray flow to the pressurizer following an accident or off normal operating condition within a reasonable time ar causing increased exposure was been adequately addressed by the licensee. The DH/LPIinitiated spray flow is not the only method available for controlling the mixing and precipitation mitigation of boron in the reactor vesse Cooldown of the pressurizer is not critical and may be accomplished using several options if a cooldown of the component is required. The use of temporary lead radiation shielding in the plant was adequately controlled by procedure which evaluates each mstallation of the shielding materials to ensure that no design conditions are exceeded on the systems piping and or components by attaching the lead blankets in areas which require radiation reductio E1.3.5.2 Instrumentation and Controls Walkdown E1.3.5.2.1 Inspaction Scope The team conducted a walkdown of the control room to verify actualinstalled instruments and to observe methods to identify RG 1.97 related indicators and recorders for the DHILPl

-. . . _ _

.-

,

syste and Subsequent CH pump rooms. walkdowns and observations were also conducted on the BWST tank E1.3.5.2.2 Inspection Findings A walkdown of the control room was performed to verify the RG 1.97 indicators and recorders for the DH/LPI system. Type A, B, and C (Category 1 & 2) related instruments are identified with green dots on the instrument nameplates. The team noted that DH cooler outlet temperature indicators TI 1432 and TI 1403, classified as display Type D variables were not provided with any kind of specialidentification, in response, the licensee indicated that only thost RG 1.97 Category 1 & 2 variables as identified in FSAR Table 7.11 A, "RG 1.97 Instrumentation," and Topical Report ULD 0 TOP 05 * Regulatory Guide 1.97, Revision 0,* are provided with identifying marks. The team also noted the following items during the control room walkdown:

.7AR Table 7.11 A is not consistent with the as installed condition with respect to ranges and units for the BWST levelinstruments LIS 1411, LIS.1421, and LR 1411 on panel C14 C16, and C18 (previously discussed in Section E1.3.4.2.2).

The actualinstalled RB Sump levelindicator LI 1405 on panel C14 shows a range of 0-100%. FSAR Table 7.11 A indicates a range of 0 56" for this indicator. The licensee has initiated corrective actions to revise FSAR Table 7.11 The need to correct discrepancies and review the corrective actions related to the accuracy of the FSAR represents a weakness with respect to 10 CFR Part 50.71(e) and is identified as Unresolved item 50 313/97 201-1 During a walkdown of the DH pump rooms and the BWST, the team noted similar weaknesses as in the EFW system on items such as instrument tubing slope, supports, mounting brackets and capillary protection. Because of the generic nature of these observations, CR 197 0074, initially issaed for EFW, has been reclassified by the licensee as a significant condition to also address the DH, LPl. and HPl systems. This item was discussed previously in Section E1.2.5.2.2 of this report. The team observed the monthly surveillance procedure testing of decay heat Channel 2 for proper calibration and operation per 1304.16 The performance of this test satisfied the requirements for calibration and testing of the decay heat system in accordance with TS Table 4.1 1, Half way through the procedure, testing was discontinued because of a defective switch in the test mocule. System operability was not affected since the defective component is not a part of the logic circuitry. The licensee subsequently issued CR 197 0080 to implement corrective action and continue with the testin E1.3.5.2.3 Conclusions The DH/LPIinstrumentation and controls have been designed such that the system will function as describe 6 in the FSAR and the identification of RG 1.97 related instrum the main control room met the intent of the RG. The lack of design basis for instrument

_

._ ___,______. _ _ _....___. - . .__ _ -_____- l

tubing installation is being addressed by the licensee on a generic basis. As indicated by the licensee, the deficiencies that were found in the DH/LPI will be resolved in con; unction with

.

the EFW syste i

! E1.4 Esit Meeting !

2 I

On March 14,1997, the team members conducted a technical debrief with their utility counterparts. On April 3,1997, the team leader conducted a final public exit meeting ;

j where the team's overall conclusions were presented. Upon commencement of the exit meeting, the NRC team leader answered questions from local media representatives. A j

partiallist of persons who attended ti.e exit meeting is contained in Appendix '

i

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3-i t ,

i e

l

i

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i i

L I

i i

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i

4

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j 70 l

!

.

'

, , . _ . . - ...._,..y ..-,,..,m.,_,,m ..,y... _ _ _ . . , ... - -_i,,,,. . , - - ._ , . .y. .,.,,. . ,,,,_m.__..,,.- .. m, . . - , . . . . . . .. ~ , , . .._. _ _ . ,_ . _ - . . ,_ _ __ _ _ .

_ _ _ _ _ _ _ _ _ . . . _ __

,

i Appendix A List Of Open items

,.

NUMBER IYPE SMTJONJ IIILE

50 313/97 201 01 IFl E 1.2.2.2. 2 CST TS Bases Revisions 50 313/97 201 02 URI E 1.2.2. EFW Flow Rates Exceeding B&W Recommendations 50 313/97 201 03 URI E1.2.2. EFW Piping Configuration Differences 50 313/97 201 04 URI E 1. 2.2. ,

Inadequate Piping Pressure And Temperature Specifications 50 313/97 201 05 URI E 1.2. 2. Evaluation of EFW Purnp Room Environment l 50 313/97 201 06 URI E 1.2.2. Drawing and Calculation t

E 1. 2.3. Revisions E 1.2. 3.9. 2 50 313/97 201 07 IFl - E 1.2.3. Modification Work Procedure Revisions 50 313/97 201 08 IFl E 1.2. 3. Additional Vendor Information Necessary for Revised Cable Pulling Calculation 50 313/97 201 09 IFl E 1.2.3.6. 2 Use of Tefzel Cable Ties and PVC Conduit inside Containment

'

50 313/97 201-10 URI E 1.2.3. Maintenance, inspection and '

,

Testing of Cathodic Protection 4 and Grounding Systems l 50 313/97 201 11 IFl E 1.2.3. Verification of Calculation

'

^

Changes Associated with DC Batteries 50 313/97 201 12 IFl E 1.2.3.1 NRC Review of Current Design Basis for Bypassing EDG .

Protective Trips

,

A1

- - . - - - - , . . . - , - - . . . - - - - . . - - , , - - - - --.- ---.

. - . . - . - - . - . - _ . _ - . . . - . - .- .. - - . .- . -- . - - _ - - . ._ -..--.- - -

i j

50 313/97 201 13 IFl E 1.2.3.1 Reconstitution of Tornado Analysis for EDG Exhaust Stacks 50 313/97 201 14 URI E1.2.3.1 Lack of Testing Unit 1 MCCBs- 1

'

50 313/97 201 15 URI E 1.2.3.1 Lack of including all Safety-Related Fuses in Procedure >

2 50 313/97 201 16 URI E 1. 2.3.1 FSAR Discrepancies E1.2.4. E 1. 3.4. ,

E 1.3. 5. , 50 313/97 201 17 URI E 1.2. 5. Ineffective Resolution of Previous Corrective Actions Associated with ,

Solenoid Valve Leakage

50 313/97 201 18 URI E1.2.5. Lack of Seismic Support of

,

OTSG Pressure Transmitters 3 50-313/97 201 19 URI E 1.2. 5. e

Lack of Design Basis for Support of OTSG Instrument Sensing Lines

, 50 313/97 201 20 URI E1.2.5. Inadequate Work Plan for the Control of Post Maintenance Testing 50 313/97 201 21 URI E1.3.2. Inadequate 50.59 review

.

Associated with the Removal of

' BWST Vacuum Breaker and Followup Corrective Actions 50 313/97 201 22 URI E1.3.2.2.2

'

Revised Calculations Associated l' with BWST Vortexing and Pump NPSH

50 313/97 201 23 URI E 1.3.2. Inadequate 50.59 Evaluation <

Associated with BWST Releases

.

A2

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Appendix B Exit Meeting Attendees NUCLEAR _ REGULATORY _ COMMISSION NAME ROSITJON R. Pettis Team Leader, NRR .

C. Vandenburgh Chief, Engineering Branch, DRS, RIV K Kennedy Senior Resident inspector, ANO J. Melfi Resident inspector, ANO

,

ENTERGY_ORERATIONS._INC.

NAME POSLTlON C. Hutchinson Vice President Operations R. Lane Director Design Engineering D. Mims Licensing Director C. Zimmerman Unit 1 Plant Manager M. Cooper Licensing Specialist C. Tyrone Design Engineering B. Day Design Engineering Manager ARKANSAS.OEMOCRAT_GAZET.TE A. Moreau Business Writer B1

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l l

Appendix C List uf Documents Reviewed

Safety _AnalysisDocumentation ANO 1 Final Safety Analysis Report, Docket No. 50 31 ANO 1 Technical Specification, Docket No. 50 31 Upa.ar LawaLDecumentation .

Design Configuration Documentation Project, ULD 1 SYS 04, 'ANO 1 Decay Heat Removal / Low Pressure injection System." Rev. Design Configuration Documentation, ULD 1 SYS 08, *ANO 1 Emergency Feedwater Initiation and Control System," Rev. '

Design Configuration Documentation, ULD 1 SYS 09, * Engineered Safeguards Actur, tion System (ESAS),* Rev. O Design Configuration Documentation ULD 1 SYS 10,"ANO.) Service Water System?

Re Design Configuration Documentation, ULD 1 SYS 12. " Emergency Feedwater System,"

Rev.1. January 16,199 Design Configuration Documentation, ULD 0 TOP-04," Control of instrument Set Points Topical," Rev. Design Configuration Documentation, ULD 0 TOP 05," Regulatory Guide 1.97," Rev. Design Configuration Documentation, ULD 0 TOP 07, "HELB/MELB Topical ULD," Rev. O, June 4,199 Design Configuration Documentation, ULD 0 TOP 08, "ANO Missiles Topical," Rev. O, July 15,199 Design Configuration Documentation, ULD-0 TOP-09, " Loss of Decay Heat Removal Topical," Rev. Design Description Document ULD 0-TOP 10, " Electrical Separation," Rev. Design Configuration Documentation, ULD-0 TOP 17, "Ai40 Flooding Topical,".Rev. O, July 30,199 C1

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_

,

Design Configuration Documentation Project 'ULD 0 TOP 18, ANO 1 and 2, * Single Failure Criteria," Rev. Design Configuration Documentation ULD 1 TOP-01, "Small Break LOCA Analysis," Rev.1, February 5,199 Design Configuration Documentation, ULD 1 TOP 02, "INO 1 Main Steam Line Break Accident," Rev. O, December 18,199 ,

Design Configuration Documentation, ULD 1 TOP-07, "ANO 1 Emergency Feedwater Sizing Analysis," Rev. O, January 26,199 Design Configuration Documentation, ULD 1 TOP 11. "ANO 1 Loss of Coolant Flow Accident Analysis.' Rev. O, April 5,199 Design Configuration Documentation, ULD 1 TOP 18, "ANO 1 Single Failure Criterion Topical," Rev. O, January 11,199 Training Manuals STM 105, ANO 1 System Training Menual, * Decay Heat Removal System," Rev. STM 1-66, ANO 1 System Training Manual, * Emergency Feedwater Initiation and Control,"

Re Calculations Calculation 97 E 0012 01, "PSV 1412 FME Cover Air Gap Calculation," February 10,199 Calculation 89 E 0010 26, "LPI Pump NPSH," Rev 5, February 7,199 Calculation 89 E 0010 27, "HPl NPSHA During Piggyback Operation," Rev. March 7,199 Calculation 89 E 0010-28, "P 34A/B and P 35A/B NPSH from BWST," Rev. O,

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February 13,'199 ' Calculation 92 E-0077 03, "ANO 1 LPI System Pump Performance Requirements " Rev.O, November 30,199 ,

Calculation 82 2086140, "ANO 1 Technical Specification Water Level (in feet) for CST T-41 B," Rev. O, May 12,198 Calculation 82 D 2086149, " Volume Lost From Design Basis Tornado Missile Strike,"

Rev. O, October 31,198 Calculation 82 D 2086 01, " Volume of CST T 41B Requiring Tornado Missile Protection "

Rev -1, June 8,1984.

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i C2

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Calculation 88E 0086 01, "NRC8 88-04 Review for P7A and 8 Minimum Flow Evaluation,"

Rev. O, June 2,198 ,

Calculation 80 010838102A, "EFW Pump Capacity," Rev. O April 11,198 Calculation 85 E 007101, " Evaluation of Time Available for ManualInitiation of EFW Following LOFW," Rev. O, March 31,198 Calculation 88 E 0100116, " Evaluation of Emergency Feedwater (EFW) Systems Pressures ano Temperatures in P T Calculation 88 E 010016 " Rev, O. October 2,199 Calculation 88 E 010016. "P T Calculation for Unit 1 Emergency Feedwater System,"

Rev. O, October 13,199 Calculation 82 D 2086-02, "Datermination of Pipe Size and NPSHA for EFW Pumps P 7A and 8 Suction from CST T 418," Rev. 3, September 22,1986.

l

'

Calculation Re E 0077 04 " Unit 1 EFW System Pump Performance Requirements,"

O, December 19,199 Calculation Re O, July 80 D 10838102, "EFW Pump Discharge Pressure Needed at 650 gpm,"

27,198 Calculation 89 E 0040-01, "ANO 1 Rcsponse Time Evaluation," Rev. O, July 6,19R Calculation 86 D 1005 20, " Main Steam Line to EFW Turbine Flow Analysis,"

Rev 0,- August 8,198 Calculation April 86 D 1026 07, "EFW Pump Room Temperature Assessment," Rev. 2, 30,199 Calculation 87 E 0061-01, "EFW Pump Room Temperature Test," Rev. O, July 29,1987 Calculation March 87 E 0026 09, "EFW Pump Room Temperature Profiles," Rev. O, 7,199 Calculation 87 E 0059-01, "ANO 1 LOFW Analysis," Rev.1, April 15,198 Calculation 87 E 0062 01, "Cooldown Times to DHRS," July 29,198 Calculation 83 D 1049 05, "ESAS ano RPS String Errors with 18 Month Stability,"

Revision Calculation 85 EO 000319, " Accuracy, Set point and Response Time Analysis for Pressurizer Water Temperature, THOT, TSAT, and HPl/LPI Instrumentation," Revision Calculation 85 EQ 000318, " Error Analysis for FC 2645, FC 2646, FC 2647 and FC 2648,"

Re _ _ ._

C3

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Calculation 85 EO 0003 20, *EFIC LevelIndication Accurar v Analysis," Rev. Calculation 85 EO 0003 04, " Accuracy Analys's for FT 2b45, FT 2646, FT 2647 and FT 2648," Rev. Calculation 85-EO 000312 " Loop Error Analysis for Indicator Accuracy Associated with EFIC SG Level Transmitters," Rev. Calculation 80 D 1083C 01, "EFIC System Error Set point Analys!s," Rev. Calculation 85 EO 000311, " Loop Error Analysis for LE 14058, LE 5645A/LE 56458, LE 5646A and LE 56468," Rev. Calculation 82 D 2086B 41. Error and Set point Analysis for BWST LevelInstrumentation Loop," Rev. Calculation 83 D 1153 01, " Accuracy Analysis for LT 4204, LT 4205,2LT 0727-1,2LT-0727 2 and 2TE 0727A/B," Hev. Calculation 91 R 1018-02, "EOP Set point Basis Document " Rev. Calculation 84 E 0083 001, " Protective Device Coordination Study (Criterla for Protective

.

Device Settings)," Rev. 5 February 8,199 Calculation 84 E 0083-002, " Plant Protection Study, Bkt, 152/301 (480 Volt Loao Center B5 Primary)," Rev. O, November 19,1985, i

Calculation 84 E 0083 006, " Plant Protection Study, Bkr. 152/305 (Decay Heat Pump P34A)," Rev. O, November 18,1P9 Calculation 84 E C083 012, " Plant Protection Study,8kr.152/311 (Emergency Feedwater Pump P78)," Rev. O, November 19,198 Calculstion 84-E-0083 013 * Plant Protection Study, Bkr. 152/401 (480 Volt Load Center B6 Primary)," Rev. O, November 19,198 Calculation 84 E 0083 017,"" Plant Piotection Study, Bkr. 152/405 (Decay Heat Pump P348)," Rev. O, November 19,198 Calculation 84 E 0083 023, " Plant Protection Study, Bkr. 52/512 " (480 Volt Load Center B5 Secondary)," Rev. 2, July 14,198 Calculation 84 E 0083-030, " Plant Protection Study, Bkr. 52/612 (480 Volt Load Center B6 Secondary)," Rev. 2, July 14,198 Calculation 84 E 0083-052, " General Criteria for Nonsafety Buses," Rev. 3, April 21,198 C-4

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Calculation June 27,199 E 0002 01 "ANO 1 Diesel Generator #1 and #2 Load Study," Rev. 8 6 Calculation 90-E-0062 02, " Unit 1 Diesel Generator Loads," Ru. 3, August 18,198 Calculation Rev. O, May 91-E 10 ;99 , (Excerpt), "ANO 1 Battery, DC and Corridor 98 HVAC Evaluatio Calculation 92 E-0003 001, " Emergency Diesel Generator Transient Loading Analysis -

Voltage and Frequency Profiles," Rev. O, June 10,199 Calculation 94 E 0018-01, "GL 8910 MOV Power Cable, Breaker, and Thermal Overloa (TOL) Device Evaluation," Rev.1, June 30,199 Calculation 97-E-0015-01 "E-2080 Conduit Support Evaluation," Rev 0, March 7,199 D 1083A 04, " Voltage Drop Starting Torque," Rev. O, January 24,198 D 1054A September 01, "Derating of Cable Ampacity, Due to Fire Barriers," Rev. ,198 E-0083 50, "DC Load Center D01," Rev. 2, February 2,199 E-0083-51, "DC Lo:.i Center D02," Rev. 2, February 2,199 D-1088 09, " Cable Ampt. city Study for Fan Motor VSFM1E," Re O, August 21,199 D-1012-03, July 26,1988. " Cable Ampacity Study for Chiller Pump VP1 A&VP1B," Rev. O, 89 E 0102 August 26,101. E.,.*~> /minal Voltage Calculation for DC Motor Operated Valves," Rev. 8, 92-E 0009-01, "AC Motor Operator Valve Terminal Voltage," Rev. C 002101, September " Emergency Duty Cycle and Battery Sizing Calculations," Rev. 4, 16,199 E-002102, ' .)01 DC System Short Circuit Study," Rev. 3, January 21,1994, 92 E 0021-03, D02 DC System Short Circuit Study," Rev. 3, January 21,199 E-0021-08, Re O, January 13,199 "ANO Unit 1, Class 1E 125 VDC, Train 1, DC Voltage Drop Study,"

92 E 002104, " Battery D03 and D07 Recharge Time," Rev. O, March 5,199 C-5

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93 D 100B 04, " Battery Load and DC Voltage Dion Study P.avisions Due to Larger Motors installed on CV 2620 and CV 2627," Rev. O, Arg.P 1 199 D 1010-01, "D01 Short circuit Study For DCP 9.'i 1010, Rev. O, September 27,199 D 1010-02, *D02 Short Circuit Study for DCP 931010," Rev. O, September 27,1994, 95-D 1010-04, September 7,199 " Recharge Time for D06 and 007 with changes per DCP 9310?O," Rev. O, 93 D 1010 08, " Protection Device Coordination Study for D01 per DCP 931010, Re October 10,199 D 6012-03, "2C-75A&B Protective Device Sizing," Rev. O October 7,199 E-0007-01, "RB Penetration Overcurrent Protection Study," Rev. O, October 14,199 R-2026-02, January 17,199 "ANO Unit 1 GL 89-10 MOV Cable Ampacity Report," Rev. O, 95-E-0001-05, "ANO Unit 1. Millstone Study-Startup No.1 Cases," Rev. O, March12,199 PM EE-64, Rev. 7 - Low Voltage Circuit Breaker " Breaker Maintenance," NP-7410, Volume 3, September,199 " Molded-Case Circuit Breaker". EPR Drawings M-204, "P&lD Emergency Feedwater," Sh. 3, Rev. 2 M-204, "P&lD Emergency Feedwater Storage," Sh. 5, Rev.1 M 204, "P&lD Emergency EFW Pump Turbine," Sh. 6, Rev.1 M-205, "P&lD Extraction Steam Heater Vents and Drains," Sh.1, Rev. 7 M-206, "P&lD Steam Generator Secondary System," Sh.1, Rev.11 M-206, "P&lD MSIV Operator Controls," Sh. 2, Rev.1 M-200, "P&lD instrumentation and Component Symbols," Sh.1, Rev.1 EFW-1, "Large Pipe Isometric Emergency Feedwater to Steam Generator E-24B South,"

Sh.1,Rev.1 C-6

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3 EFW 2, "Large Pipe isometric Emergency Feedwater from Containment Penetration P-17 to Steam Generator E 24A," Sh.1, Rev.11, 3 EFW 2, "Large Pipe !sometric Emergency Feedwater from Containment Penetrntian P 17 to Steam Generator E 24A, " Sh. 2, Rev. EFW 108, "Large Pipe isometric Emergency Feedwater Piping to Containment Penetration P-6 5, " Sh.1, I'.ev EFW" 108, Piping, Sh. 2,"Large Rev. 4. Pipe Iso'netric Emergency Feedwater Pump-7A Discharge and lest Loo 3-EFW 109, "Large Pipe Isometric Emergency Discharge Feedwater Pump 7A to isolation Valve," Sh.1, Rev. EFW-109, " Emergency Feedwater System Piping Isometric," Sh. 2, Rev. EFW 110, "Large Pipe isometric Emergency Feedy .ner from P-78 to CV 2626,"

Sh.1 Re EFW-110, "Large Pipe isometric Emergency Feedwater from P-78 to CV-2626,"

Sh.2,Re EFW-111. "Large Pipe isometric EFW from P 78 to CV-2670,"

Sh.1, Rev. ,

3-EFW S ,

2, Rev. 2. "Large Pipe isometric Emergency Feedwater from P-78 to CV-2670, "

3 EFW Sh.1, 11 Re . "Large Pipe isometric Emergency Feedwater Pump P-7A & P-78 inlet Piping,"

3 EFW-113, "Large Pipe isometric Emergency Feedwater Pump P 7A & P-78 Inlet Piping, Sh.2,Re EFW-114, Rev.16, "Large Pipe isometric Emergency Feedwater Pump Inlet Header Piping," Sh.1, 3 EFW 114. "Large Pipe isometric Emergency Feedwater Pump Inlet Header Piping," Sh. 2 Re EFW-114, "Large Pipe isometric Emergency Feedwater Pump Inlet Header Piping," Sh. 3, Re EFW-115, "Large Pipe isometric Emergency Feedwater Pump P-7B Piping to CV-2670,"

Sh.1,Re C-7

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3 EFW-115, Piping, " Sh. 2, "Large Rev. 2. Pipe isometric Emergency Feedwater Pump P 7A & P-7B Discharge 3 EFW 116, "Large Pipe isometric Emergency Feedwater Penetration to Steam Generator,"

Sh.1,Re EFW-116. "Large Pipe (sometric Emergency Feedwater Pump 7A Piping to CV 2627,"

Sh.2.Re EFW 116 "Large Pipe isometric Emergency Feedwater P 78 to CV 2670," Sh. 3. Rev. EFW 118, "Large Pipe isometric Emergency Feedwater Pump.7A to CV 2627," Sh.1, Re EFW-119, Piping," Sh.1, Re "Large2, Pipe isometric Emergency Feedwater Discharge P 7A and Test Loop l,

12-CON 8," Sh.1, Re ,2,"Large Pipe isometric Emergency Feedwater from T-41B to EFW Pumps P 7A &

12 CON 142, "Large Pipe isometric Emergency Feedwater from T-418 to P-7A and P-78,"

Sh.1,Rev.2, 12 CON-143, "Large Pipe isometric Emergency Feedwater from T-418 to P-7A & P-78",

Sheet 1. Rev. 2 12 CON 143, "Large Pipe isometric Emergency Feedwater from T-41B to P 7A and P 78,"

Sh.2.Re '

!

P1278," CON S , Rev."Large 2. Pipe Isometric Emergency Feedwater Piping from~T 418 to P 7A and P&lD Rev, Sh.1, M 209, 89. " Circulating Water, Service Water & Fire Water intake Structure Equipment,"

P&lD Sh.2,Rev.3 M 209, " Condenser Vacuum, Circulating Water & Discharge Structure Equipment,"

P&lD M 210, " Service Water," Sh.1. Rev.12 _

E 258, " Wiring Sh.5,Re Block Diagram, Emergency Feedwater initiation and Control (EFIC),"

M 212, " Piping and instrument Diagram Demineralize Water Distribution," Sh. 2 Rev. 5 M 230, " Piping and instrument Diagram Reactor Coolant System," Sh.1 Rev. 9 M-232, " Piping and instrument Diagram Decay Heat Removal System," Sh.1. Rev 8 C-8

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. . . . . . . . ,

M 402, ' Functional Description and Logic Diagram, Condensate Feedwater System,"

Sh. 3, Rev. 2 M-402, " Logic Diagram, EFIC Annunciation," Sh. 5, Rev. M 418, " Functional Description & Logic Diagram. Decay Heat Removal System,"

Sh.1 Rev.16, M 418, " Functional Description & Logic Diagram, Decay Heat Removal System,"

Sh. 2, Rev.1 M 418, " Logic Diagram, Decay Heat Removal System," Sh. 3, Re M 587, " Console C09 Arrangement Emergency Feedwater," Rev.-1 M-590, " Panels C14 Arrangement Primary Coolant," Rev. 2 M 591, " Panels C16 & C18 Arrangement Engineered Safeguards Panels," Sh.1, Rev. 39.

,

M 592, " Panels C19 Arrangement Plant Auxiliary Systet .s," Rev. 28.

l-L 35720 002-0, " General Plan for Condensate Storage Tank," Re", 7, 12 CON 146, " Largo Pipe isometric LT-4205 and T41B Spare Connection," Re CON 145, "Large Pipe isometric Spare Connection for Condensatt Storage Tank T 41B," Re B-003 G 1025 001-0, "BWST (T 3) Level Sensing Line," Rev. O, i

C-46, " Field Erected Tanks Sheet No. 2," Rev.1 , " Initiate Logic Module (l) Functional Logic Diagram B and W EFIC, Rev. ..

58526 242, "EFIC Connection Diagram, System Organization Layout & Index Sh,1",Re , "EFIC Connection Diagrams, Legend Symbols and Notes Sh. 2," Rev. , "EFIC Connection Diagrams, Module SEA (CH. A&B) Sh. A," Rev. , "EFIC Connection Diagrams, Module SEA (CH. C&D) Sh. L.."

Re , "EFIC Connection Diagrams, Initiate Module Channel A&B Sh. 4A," Rev. , "EFIC Connection Diagrams, Initiate Module Channel A&B Sh. 4B " Rev. C-9

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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..

58526 253, "EFIC Connection Diagrams, initiate Module Channel C&D Sh. 4C." Rev. , "EFIC Connection Diagrams, Initiate Module Channel C&D Sh. 4D " Rev. , "EFIC Cennection Diagrams, Vector Module Logic A&B Sh. SA," Rev. , "EFIC Connection Diagrams, Vector Module Logic C&D Sh. SB," Rev. , "EFIC Connection Diagrams, Trip Module EFW Logic Sh. 6A." Rev. O, 58526 255, "EFIC Connection Diagrams, Trip Module SEA MSLI Logic Sh. 68," Rev. , "EFIC Connection Diagrams, Trip Module SEA MSLI Logic Sh. 6C," Rev. , "EFIC Connection Diagrams, Trip Module SGB MSLI Logic Sh. 6D." Rev. , "EFIC Connection Diagrams, Trip Module SGB MSLI Logic Sh. 6E " Rev. , "EFIC Connection Diagrams, Control Module Organization Sh. 8A," Rev. M 516 Sh.153, " Data Sheet, Pressure Transmitters," Rev. M 516 Sh.159, " Data Sheet, Panel Mounted Indicators," Rev. M 516 Sh.160, " Data Sheet, Panel Mounted Indicators," Re M 516 Sh.164, " Data Sheet, hecorders," Re M-516 Sh.165, " Data Sheet, Indicating Controllers," Rev. M 516 Sh.169, " Data Sheet, Difterential Pressure Transmitters," Re M 516 Sh.173, " Data Sheet, Differential Pressure Transmitters," Re M-516 Sh.201, " Data Sheet, Pressure Transmitters," Rev. M 516 Sh.209, " Data Sheet, Recorders" Rev. M 516 Sh.234, " Data Sheet, Indicating Switches", Rev. 2 M 516 Sh.235, " Data Sheet, Differential Pressure Transmitters", Rev. 2 M 516 Sh.306, " Data Sheet, Thermocouple Sensors", Rev. O E-1, " Station Single Line Diagram", Sh.1, Rev. 37 E-1, " Single Line Diagram. 500kV Switchyard Auxiliary Power", Sh 2, Rev. 2 C-10 y

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E 5, " Single Revisicq 23. Line Meter & Relay Diagram,4160 Vrilt System, Engineered Safeguard,"

-

E 7, " Re Sh.1, Single 2 Line Meter & Relay Diagram,480 Volt Load Centers, Main Supply",

E 8, " Single Line Meter & Relay Diagram,480 Volt Load Centers, Engineered Safeguard &

Main Supply," Sh.1. Rev. 2 E 15, " Single Line Meter & Relay Diagrara,480 Volt Motor Control Centers B51 &

B52," Sh.1, Rev. 5 E 16, " Single Line Diagram,480 Volt Motor Control Centers 855 B56," Sh.1, Rev. 5 E 18, " Single Line Diagram,480 Volt Motor Control Centers B61 - 862," Sh.1, Rev. 6 E 19, " Single Line Diagram,480 Volt Motor Control Centers B53 - 863," Sh.1.Rev. 3 E-19, " Single Line Diagram, 480 Volt Motor Control Centers B57 - B65." Sh. 2, Rev.11.

E E 55, " Grounding Notes & Details," Sh.1, Rev. E 59, " Conduit & Cable Notes & Details," Sh.1 - 80, Rev. E 62, " Cathodic Protection System, Notes & Details", Sh. A - 15. Rev. E-96, " Schematic Starter," Rev. Diagram.125 V.D.C. Motor Control Center Full Voltage Reversing E 181, " Schematic Diagram, Decay Heat Removal P.:mp P34A," Sh.1, Rev. 2 E-181, " Schematic Diagram, Decay Heat Removal Pump P348," Sh.1 A, Rev. E-183, S A,"Re Schematic Diagram, Decay Heat Cooler E358 Isolation Valve CV-1400," '

E 265, S , Re " Schematic / Wiring Block Diagram, Plant Auxiliary Control System, DH/LPI,"

E 265, Sh. 4, Re " Schematic / Wiring Block Diagram, Plant Auxiliary Control System, DH/LPI,"

E 293, " Schematic Diagram, Emergency Feedwater to SEA From P78 Isolation Valve CV-2670," Sh. 2, Rev. E 294, " Schematic Diagram, Emergency Feed Water Pump P78," Sh.1, Rev.1 E-331, " Schematic Diagram, Miscellaneous Instrumentation," Sh. 2, Rev.1 C-11

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E 559, " Connection Diagram, Main Control Panel C09," Sh, 2 Rev.1 E 567, " Pan Layout, Main Control Panel C18," Sh.1 Rev 2 E 567, S " Connection 2, Rev, 6 Diagram, Front (Rear View) - Left Section, Main Control Panel C18,"

E 567, S " Connection 3. Rev, 5 Diagram, Front (Rear View) Middle Section, Main Control Panel C18,"

E 567, Sh " Connection Diagram, Front (Rear View) Right Section, Main Contiol Panel C18,"

4.Rev,1 E 567,1 ,Rev " Connection Diagram, Pan "A" - Right Side (Rear View), Main Control Panel C18," S E 567, " Connection Diagram, Pan "B" - Left Side, Main Control Par ><.a C18," Sh. 8, Rev.1 E 567, " Connection Diagram, Pan "C," Main Control Panel C18," Sh. 9. Rev, E 567, " Connection Diagram, Pan "D" Bottom Main Control Panel C18c' Sh,12, Rev.1 E-601, " Electrical Plot Plan Outdoor Area," Sh.1 Rev. 2 E-602, " Embedded Conduit & Grounding, Turbine & Auxiliary Bldg. E '0," 335'-0"

& 354'-0," Rev.1 E 605, " Embedded Conduit, Auxiliary Building Area No 4, Plan at EL. 335' 0," Rev.10.

l E-613, " Underground Conduit & Grounding, Transformer Yard Area," Rev.1 E-618, " Cathodic Protection System," Rev,1 E 661, " Partial Plan - Conduit & Tray Layout - EFW Pump Room," Sh. 2 Rev.1 E 666, " Conduit & Tray Layout, Auxiliary Bldg. Area 4 & 6, Plan EL. 317'-0," Rev. 43, E-667, " Conduit & Tray Layout, Auxiliary Bldg. Area 4, Plan EL. 335' 0," Sh,1. Rev. 5 E 667, " Conduit & Tray Layout, Auxiliary Bldg. Area 4. Plan EL. 335'-0," Sh. 2, Rev.1 E 667, " Conduit & Tray Layout, Auxiliary Bldg. Area 4, Plan EL. 335'-0," Sh. 4, Rev 2 E-668, " Conduit & Tray Layout, Auxiliary Bldg. Area 4, Plan EL. 354'-0," Sh.1, Rev. 5 E-668, " Conduit & Tray Layout, Auxiliary Bldg. Area 4, Plan EL. 354'-0," Sh. 3. Rev. 2 E-668, " Conduit & Tray Layout, Auxiliary Bldg. Area 4 Plan EL. 354'-0," Sh. 4, Rev.1 C-12 l

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E-669, " Conduit & Tray Layout, Auxiliary Bldg. Area 4, Plan EL. 372'-0," Sh.1 Rev. 7 E 669, * Conduit & Tray Layout, Auxiliary Sidg. Area 4, Plan EL. 372' 0," Sh. 2 Rev. 2 E-669, " Conduit & Tray Layout, Auxiliary Bldg. Area 4, Plan EL. 372'-0," Sh. 4 Rev. 3 E-673, " Conduit & Tray Layout, Auxiliary Bldg. Area 6, Plan EL. 35 4'-0" & EL. 360'-0,"

Sh.1, Rev. 2 E-676, " Conduit & Tray Layout Section & Details," Rev. 4 E-684, " Sections and Details, Cable Spreading Room," Rev.1 E-685, " Conduit & Tray Layout, Cable Spreading Room," Rev. 2 E 686, " Conduit & Tray Layout, Reactor Bldg. Penetration Area 1, Part Plans EL. 373' 6"

& 386'-0," Rev 5 E-690, " Conduit Layout, Cable Spreading Room," Rev. 6 E 721, " Connection Diagram, Auxiliary Instrumentation Panel C543," Sh. 5, Rev. 5,.

,

!

l M 263, " Piping & Instrument Diagram, Aux. Bldg. HVAC Elevation 372'-0," Sh. 2, Rev.1 E-280, Sh 3, Pev 6, Schematic Diagram Service Water Syste E-295, Sh 4. Rev. 7. Emergency Feedwater Turbine MOVs, E 331. Sh 31, Rev. 6, Misc. Instrumentatio E-331, Sh 57 Rev.1, Alternate Heating Stem for Feedtvater Heaters E2A and E2B, E-258, Sh 6A, Rev 2, Wire Block Diagram EFI E 293, Sh 1, Rev.17 Schematic Diagram EFW Steam Generator Isolation Valv E 295, Sh 4A, Rev 3, Schematic Diagram EFW Turbine MOV E-293, Sh 2. Rev 8, Schematic Diagram EFW to SEA from P78 Iso. Valve C267 E-295, Sh 1, Rev. 33, Schematic D;agram EFW Turbine MO E-294, Sh 1 Rev.15, Schematic Diameter EFW Pump P7 E-296, Sh 1, Rev.17, Schematic Diagram EFW Pump P78 Condensate Water Valve CV280 E 296, Sh 1 A, Rev. 2, Schematic Diagram EFW Seivice Water Valve CV280 C-13

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E 296, Sh 1B, Rev. 2, Schematic Diagram EFW P7A Condensate Water Valve CV280 E 296, Sh 1C, Rev. 3, Schematic Diagram EFW P7A Service Water CV280 E 295, Sh 3, Rev.16, Schematic Diagram EFW Turb. MOV E 295, Sh 4, Rev. 7, Schernetic Diagram EFW Turb MOV E 295, Sh 1 A, Rev. 3. Schematic Diagram EFW Turbine MO E 290, Sh 1D, Rev. 3. Schematic Diagram Pump P7B Service Water Valve CV385 E 296, Sh 1E, Rev. 2. Schematic Diagram Pump P7A Service Water Valve CV385 E 318, Sh 1, Rev. 27 Schematic Diagram Recirculation Test isolation Valv E 318, Sh 2, Rev.- 8 Schematic Diagram Recirculation Test Isolation Valv E 17, Eh 1, Rev. 38, Vital AC and 125VD E 17, Sh 1 A, Rev. 2, Green Train, o

E-17, Sh 2, Rev.1, Singfeline Meter, Diagram 125 VDC Syste E-24, Sh 1, Rev.10, Singfeline Diagram 125VDC MCCs D15& D2 E-4, Sh 1, Rev. 24, Singleline Meter Diagram 4160 V Eng. Safeguard Main Suppl Ensinaaring. Reports ER 92-R-1019-01, " Evaluation of ANO-1 EFW Per SPIP-216 EFW," Rev. O, June 24,199 ER 93-R-102912. " System Review Report for the ANO-1 Emergency Feedwater System,"

Rev. O, February 28,199 ER 93 R-1040-01, "ANO-1 AB Limiting Component Qualification Temperatures," Rev. O, March 4,199 ER 89 E 0054 01, "ANO-1EFW Delay Justification", Rev. O May 10,198 ER 963378E101, " Equivalency Evaluation for Material Change," P7A/B Impellers."

Engineering Standard CES-14, " Seismic Class 1 Conduit and Cable Tray Support Design,"

Rev. O and Draft for Re ER 93 R-1002-01, "ANO-1 BWST Outlet Vortex Suppressor," Rev. O, February 5,199 ER 96 R-1003-01, " Revised DHR Flow Requirements for ECCS at ANO-1," (Framatome C-14 l

l l

Technologies, Inc . Document 51 1239323 00), April 1,199 ER 92 R 1017-16. " Unit 1 Set point Documentation' Packages for The Decay Heat System,"

Re EAR 91 177, " Set point Documentation Packages for Unit 1 EFW System," Re IDG 001-0, " Instrument Loop Error Analysis and Set point Methodology," Rev. January 16,199 ER 6010.005, * Plant Set point Control," Rev. .

- EE 84181 Engineering Evaluation Station Batteries Operability Cente NRC_Reporta and Documentation Nuclear Regulatory Commission, "Information Report by the Office of Nuclear Reactor Regulation on the Single Failure Criteria," SECY 77-439, August 17,1977.

,

Regulatory Guide 1.82, " Water Sources for Long Term Recircuiation Cooling Following a Loss of-Coolant Accident."

,

NRC Presentation, "ANO 1 Emergency Feedwater System Review," Rev. November 14,198 U.S. NRC Inspection Reports 50 313/94-20 and 50-368/94-20, " Service Water inspection 7/25/94 through 8/12/94," September 28,1994.

l ,

i U.S. NRC Inspection November 9,1994. Reports 50 313/94-20 and 50-368/94 20 (Notice of Violation),

U.S. Nuclear Regulatory Commission, A.W. Serkiz, " Containment Emergency Sump

. Re Performance Octobc,r(Technical 198 Findings Related to Unresolved Safety issue A 43)," NUREG-0897, NRC Information Notice 91-56, " Potential Radioactive Leakage to Tank Vented to Atmosphere," September 19,199 Joh_Drders Job Order 00952943, August 16,199 Job Order 00957859, December 11,199 Job Order 00957910, December.12,199 Job Order 00940673 "P4C Service Water Pump Feeder Cable Replacement,"

November 13,199 C-15

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Job Order, Repetitive Task Number 018389, " Unit #1 Cathodic Protection Preventive Maintenance."

Conditlan.Raports CR C-96-0135, June 11,199 CR 197 0039, February 6,199 CR 1-97 0015, January 16,199 CR 1-96 0663, Derember 5,199 CR 1-97-0019, January 22,199 CR 1-97 0031, February 3,199 CR-197 0080, " Defective Test Switch on DH RC Pressure Test Module,"

February 27,199 CR 197-0017. " Set point Calculation 91-R 01018 02 Error, January 21,199 ,

CR-1-97 0075, "CV 1050 Leakage," February 25,199 CR-197-0040, " inadequate EFW Pump Caction Pressure Alarm Set point," February 6,199 CR 1-97-0074, " Inadequate Instrument Tubing Support, PT-2811/PI-2811 A,"

February 25,199 .

CR 197-0058, " Missing Clips on EDG Instrument Tubing," February 14,199 CR-107-0087, " inadequate instrument Tubing Support, PT-2811/PI-2811 A, March 9,199 Root Cause Analysis Report, Non Outage Testing of PSV-1412 Without Adequate Controls to Ensure BWST Operability, CR-1-97-0019,1/22/97," February 26,199 Placaduras Procedure 1306.034 " Testing of U1 Press VAC Relief Valve PSV-1617,2423, & 1412, Rev. 2, December 4,199 Procedure 1104.029, " Service Water and Auxiliary Cooling System."

Procedure 1203.030, * Loss of Service Water."

Procedure 1106.00d, " Emergency Feedwater Pump Operation," Rev. 54.

'

Procedure 1305.028, -Reg Guide 1.97 Instrumentation Verification," Rev. C-16

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_ _-_- -_ - _- ___ -

Procedure 1104.004, " Decay Heat Removal Operating Procedure " Rev. 6 Procedure 1203.012K, " Annunciator K12 Corrective Action," Rev. 3 Procedure 1203.012H, " Annunciator K09 Corrective Action," Rev. 2 Procedure 1304.162, " Unit 1 Decay Heat Channel 1 Test," Rev. Procedure 1304.145, " Unit 1 EFIC Channel A Monthly Test," Rev.1 Procedure 1304.208, " Unit 1 EFIC Channel D Monthly Test, SG Pressure Greater 1 than 750 psig, Re Procedure 1304.185 " Unit 1 Green Channel Low Pressure injection Flow 18 Month Surveillance " Re l Procedure 1304.197, " Unit 1 Green Train BWST Level Instrument Test," Rev. Procedure 1304.012, " Unit 1 Red Train BWST Level Instrument Test," Rev.17.

'

Procedure 1304.098, " Unit 1 EFIC Channel A Calibration," Rev.1 Procedure 1413.134, " Unit 1 CST T41B Level Instrument Loop Calibration, Red Channel," Re Procedure 1413.157, ' Unit 1 CST T418 Level Instrument Loop Calibratio Green Channel," Re Procedure / Work Plan No. 1403.093, " Installation of Raychem Kits," Rev. Procedure / Work Plan No. 6030.005, " Control of Modification Work," Rev. Procedure / Work Plan No. 6030.109, " Installation of Electrical Cable and Wire," Rev. Procedure / Work Plan No. 6030.110, " Termination, Splicing and Soldering of Cable and Wire," Rev. 4,

- Procedure / Work Plan No. 6030.111, " Installation of Electrical Grounding Systems " Rev. Procedure / Work Plan No. 6030.112. " Installation of Raceway Systems," Rev. Procedure / Work Plan No. 6030.113, " Seismic Raceway Supports," Re Procedure / Work Plan No. 6030.116, " Installation & Removal Instruction for EGS Grayboot Connectors," Re Procedure / Work Plan No. 6030.200, " Administration of Post Modification Testina," Rev. C-17

___. . _ - _ -

I Precedure/ Work Plan No. 6030.212, " Insulation Resistance Testing," Rev Maintenance 1025.003, Rev. 42, 9/15/96, Conduct of Maintenanc .056, Rev. O,1/15/93, Fuse Contro .

Work Plan 1307.016, Rev.10. Unit 1, 006, D07 & D17 (switchyard) Pilot Cell Tes Ensinaaring_Operatingfroceduras 1203.002, Rev,13, Alternate Shutdow .007. Rev. 4, Decoded Powe .008, Rev. 5 Blackou .012, Rev. O, Repetitive Task Plant Modifications LCP 92 5008, Rev. O, "CPT Change Out."

DCP 9013, Rev 4," Cable Replacement for DC Valve Operators."

DCP 82-1-5-A, Rev.10. "EFW Upgrade Turbine Replacement."

Miscellaneous Temporary Alteration TAP #971001, *BWST (T-3) PSV 1412 (BWST Vacuum / Relief Valve)

and the BWST Vacuum Relief Capacity, February 5 -199 Design Change Package (DCP) #90-1043 " Upgrade LPl and RBS Flow Indication Loops to RG 1.97, Cat.1, Type A," Rev. 3, July 30,199 Varec Catalogue, "481/482 Series Pressure and Vacuum Relief Valves."

.

ANO SWS Self-Assessment Summary, " Task Description of SWSOPl Support Team."

ANO Self Assessment Followup Action Assignments, April 4,199 Service Water System Independent Review for ANO Unit 1, Devonrue,28,199 March C-18 1/

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.

Appendix D List of Acronyms AB Auxiliary Building AC Alternating Current ADV Atmospheric Dump Valve AE Architect Engineer ANO 1 Arkansas Nuclear One, Unit 1 ANSI American National Standards Institute AOV Air-Operated Valve ASME American Society of Mechanical Engineers ,

B&PV Coiler and Pressure Vessel Code B&W Babcock & Wilcox C BTP Branch Teclinical Position BWST Borated Water Storage Tank CARB Corrective Action Review Board CR Condition Report CST Condensate Storage Tank CV Control Valve CWP Controlled Work Package DBA Design Basis Accident DC Direct Current l DCD Design Configuration Document l DH Decay Heat l DHR Decay Heat Removal D/P Differential Pressure ECCS Emergency Core Cooling System ECP Emergency Cooling Pond 2DG Emergency Diesel Generator EDUP Electrical Drawing Upgrade Project EFIC Emergency Feedwater Initiation and Control EFW Emergericy Fe9dwater EO Environmental Qualification ER Engineering Report ES Emergency Safeguards ESAS Engineered Safeguards Actuation System ESF Engineered Safety Feature FLC Full-Load Current FME Foreign Material Exclusion FTl Framatome Technologies incorporated (formerly B&W)

FSAR Final Safety Analysis Report GDC Genera! Design Criterion GL Generic Letter (NRC)

HELB High Energy Line Break HPl High Pressure injection HVAC Heating, Ventilating and Air Conditioning HZ Hortz (cycles per second)

k I&C Instrumentation and Controls

} D1

1

!!

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_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ -

. .

inspection ar.d Enforcement (NRCl IE inspector Followup item IFI'

IN information Notice (NRC)

institute of Electrical and Electronics Engineers IEEE Insulated Power Cable Engineers Association IPCEA IPE Individual Plant Evaluation Inservice inspection ISI Inservice Testing IST Licensing information Request LIR LOCA Loss of Coolant Accident LOFA Loss-of Flow Accident LOOP Loss-of Offsite Power Low Pressure injection LPl LRC Locked Rotor Current MCCB Molded Case Circuit Breaker MCC Motor Control Center Motor Operated Valve MOV MSLB Main Steamline Break Main Steam Safety Valve MSSV NPSH Net Positive Suction Head Nuclear Regulatory Commission

,

NRC Office of Nuclear Reactor Regulation (NRC)

NRR Nuclear Steam Supply System NSSS Once Through Steam Generator OTSG Plant Data Management System l

PDMS Piping and Instrumentation Diagram

,

P&lD j Pl Pressure Indicator Post Maintenance Testing PMT Pressure and Temperature PT Polyvinyl Chloride PVC Reactor Building RB Reactor Coolant Pump RCP Reactor Coolant Pressure Boundary RCPB Reactor Coolant System RCS Regulatory Guide (NRC)

RG Region II (NRC)

Ril RPV Reactor Pressure Vessel Resistance Temperature Detector RTD Sargent & Lundy S&L SBLOCA Small-Break Loss of-Coolant Accident SBO Station Blackout Sub-Cooling Margin SCM Spent Fuel Pool SFP Safety issues Management System SIMS Solenoid-Operated Valve SOV Safety Systems Functional Inspection SSFI

'

SW Service Water Service Water Integrity Prugram SWlP D-2

'94-ys. ,g .i

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SWSOPl Service Water System Operational Performance Inspection TBS Turbine Bypass System T/C Thermo Couple TDH Total Developed Head TDOC Time Delay Ovetcurrent TM Temporary Modification TMl Three Mile Island TS Technical Specifications

. TOP Topical Report i'*x , ULD Upper Level Document (may be Design Configuration Document, DCD, or

' \, ,

Topical Report, TOP)

URI Unresolved item

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v t e Inspection Report - Arkansas Nuclear - 1997201 (Security)
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1997 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 10(0) 11(0) 13(0) 14(0) 15(0) 16(0) 17(0) 18(0) 19(0) 20(0) 21(0) 22(0) Security 201(0) Operator Exams Emergency Planning Other

IR 05000313/1997201

Text

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