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| TECHNICAL EVALUATION REPORT
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| . AUXILIARY FEEDWATER SYSTEM AUTOMATIC i
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| INITIATION AND FLOW INDICATION (F-16, F-17)
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| ~ TOLEDO EDISON COMPANY .
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| DAVIS-BESSE UNIT 1 NRC DOCKET NO. 50-346 FRC PROJECT C5257 NRCTACNO. 42964 FRC ASSIGNMENT 9 NRC CONTRACT NO. NRC-03-79-118 FRCTASK 302 Preparedby Franklin Research Center Author: J. E. Kaucher ,
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| The Parkway at Twentieth Street Philadelphia, PA 19103 FRC Group Leader: K. Fertner Preparedfor Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer: R. Kendall May 10, 1982 This rwort was prepared as an account of work sp'onsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the recults of such use, of any Information, apparatus, product or process disclosed in.this report, or represents that its use by such third party would not infringe privately owned rights.
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| - A
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| .. 0. Franklin Research Center Xg A Division of The Franklin Institute gg/ggC , p~~, 13. song,mo r, ,,xto p.,* y. rs. p.. isics caiss4s.ioco. _
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| TER-C5257-302 CONTENTS Title Page Section 1 INTRODUCTION . . . . . . . . . . . . . 1 1.1 Purpose of Review . . . . . . . . . . 1 1.2 Generic Issue Background . . . . . . . . 1 1.3 Plant-Specific Background . . . . . . . . 2 REVIEW CRITERIA . . . 3 2 . . . . . . . . .
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| TECHNICAL EVALUATION . . . . . . 5 3 . . . . .
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| 3.1 General Description of Auxiliary Feedwater Syste:n . . 5 3.2 Automatic Initiation. . . . . . . . . . 5
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| . . 5 3.2.1 Evaluat, ion . . . . . . . .
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| 3.2.2 Conclusion . . . . . . . . . . 9 3.3 Flow Indication . . . . . . . . . . . 9 3.3.1 Evaluation . . . . . . . . . . 9 3.3.2 Conclusion . . . . . . . . . . 11 3.4 Description of Steam Generator Level Indication . . . 11 4 CONCLUSIONS . . . . . . . . . . . . . 15 5 RSFERENCES . . . . . . . . . . . . . 16
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| TER-C5257-302 FOREWORD This Technical Evaluation Report was pre'ared p by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Commission (Office of Nuclear Peactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by the NRC.
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| Mr. J. E. Kaucher contributed to the technical preparation of this report through a subcontract with WESTEC Services, Inc.
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| : 1. INTRODUCTION 1.1 PURPOSE OF REVIEN hnical evaluation of the The purpose of this review is to provide a tecify that both safety-emergency feedwater system design to ver ided at the Davis-Besse initiation circuitry and flow indication are provIn addition, the ste Nuclear Power Plant, Unit 1. t is described to assist subsequent NRC tion available at the Davis-Besse plan staff review.
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| 1.2 GENERIC ISSUE BACKGROUND Regulatory Commission (NRC)
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| A post-accident design review by the NuclearMile Island (TMI) Unit 2 28, 1979 incident at Three after the March (AN) system should be treated as a established that the auxiliary feedwater The designs of (PWR) plant.
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| safety system in a pressurized water reactor i d to meet general design safety systems in a nuclear power plant are requ re specified in Appendix A of 10CFR50 [1].
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| criteria (GDC) design are GDC 13, GDC The relevant design criteria for the AN system tion to GDC 13 sets forth the requirement for instrumenta ted ranges of operation) 20, and GDC 34. ii monitor variables and systems (over their ant c paGDC 20 requi that can affect reactor safety. that acceptable fuel I
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| designed to initiate automatically in order to assuref anticipated operat t
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| cesign limits are not exceeded as a result o f the designed system, occurrences. GDC 34 requires that the safety function othe A N that is, the residual heat removal by the case of a single failure.
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| [2] to each PWR licensee On September 13, 1979, the NRC issued a letter It required ified in NUREG-0578 [3].
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| that defined a set of requirements speci tion and single failure-proof design that the AN system have automatic init a 'In addition, auxiliary I d GDC 34.
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| consistent with the requirements of GDC 20 anld be provided to satisfy feedwater flow indication in the control room shou the requirements set forth in GDC 13 l
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| TER-C5257-302 During the week of September 24, 1979, seminars were held in four regions of the country to discuss the short-term requirements. On October 30, 1979, another letter was issued to each PWR licensee providing additional clarifica-tion of the NRC staff short-term requirements without altering their intent [4].
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| Post-MI analyses of primary system response to feedwater transients and .
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| reliability of installed AN systems also established that, in the long term, the AN system should be upgraded in accordance with safety-grade require-ments. These long-term requirements were clarified in the letter of September 5, 1980 [5]. This letter incorporated in one document, NUREG-0737 [6], all TMI-related items approved by the commission for implementation at this time.
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| Section II.E.1.2 of NUREG-0737 clarifies the requirements for the AM system automatic initiation and flow indication.
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| 1.3 PIANT-SPECIFIC BACKGROUND
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| ~
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| The Toledo Edison Company responded to the NRC requirements in a letter dated September 16, 1981 [7]. This letter provided electrical and piping diagrams as well as other information to supplement the Davis-Besse FSAR.
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| The review of the AN system at the Davis-Besse plant began in November 1981, cased on the criteria described in Section 2 of this report.
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| TER-C5257-302
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| : 2. REVIEW CRITERIA i d licensees To improve the reliability of the AN system, i l the NRC requ reautomatic initi to upgrade the system, where necessary, to ensure t me y ,
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| In the short The system upgrade was to proceed in two phases.
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| when required. i to be used to auto- ,
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| term, as a minimum, control-grade signals and circu ts wereThis c matica11y initiate the AN system.
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| following requirements of NUREG-0578, Section 2.1.7.a [3]:
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| *1.
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| The design shall provide for the automatic initiation of the auxiliary feedwater system.
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| 2.
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| The automatic initiation signals and circuits shall be designed so that a single failure will not result in the I
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| loss of auxiliary feedwater system function.
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| 3.
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| Testability of the initiating signals and circuits shall be -
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| a feature of the design.
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| ' hall be powered from
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| : 4. The initiating signals and circuits the emergency buses.
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| 5.
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| Manual capability to initiate the auxiliary feedwater sys- be tem from the control room shall be retained and
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| ! will not result in the loss of system function.
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| feed-6.
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| The ac motor-driven pumps and valves in the auxiliary tion water system shall be included in the automatic actua (simultaneous and/or sequential) of the loads to the emer-gency buses.
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| : 7. The automatic initiating signals and h circuits shai of manual capability to initiate the AN system from t e Control room." d d in accor-In the long term, the.se signals and circuits were to be upgra e Specifically, in addition to the above dance with safety-grade requirements. i st have indepen-rc iuirements, the automatic initiation signals and circu ts muhave system dent channels, use environmentally qualified components, tion criteria, inoperable status features, and conform to control system interac as stipulated in IEEE Std 279-1971 (8].
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| 4 TER-C5257-302 The capability to ascertain the AFW system performance from the control room must also be provided. In the short term, steam generator level indica-tion and flow measurement were to be used to assist the operator in maintaining the required steam generator level during AEW system operation. This system was to meet the following requirements from NUREG-0578, Section 2.1.7.b:
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| "1.
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| Safety-grade indication of auxiliary feedwater flow to each steam generator shall be provided in the control room.
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| : 2. .
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| The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency power diversity requirements of the auxiliary feedwater system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9 [9] ."
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| The NPC staff has determined that, in the long term, the overall flowrate indication system for Babcock & Wilcox plants must include at least two safety-grade auxiliary feedwater flowrate indicators for each steam generator.
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| The flowrate indication system should conform to the following salient paragraphs of IEEE Std 279-1971:
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| 1 o 4.1 - General Functional Requirements
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| : o 4.2 - Single Failure '
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| o o
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| 4.3 and 4.4 - Qualification 4.6 - Channel Independence t
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| o 4.7 - Control and Protection System Interaction o 4.9 and 4.10 - Capability for Testing.
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| The operator relies on steam generator level instrumentation, in addition to auxiliary feedwater flow indication, to determine AEW system performance.
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| The requirements for this steam generator level instrumentation are specified in Regulatory Guide 1.97, Revision 2, " Instr mentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" (10]. ,
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| : 3. TECHNICAL EVALUATION 3.1 GENERAL DESCRIPTION OF AUXILIARY FEEDWATER SYSTEM system at the Davis-Besse plant supplies The auxiliary feedwater (AN) for reactor decay heat water to the secondary side of the steam generators to loss of offsite removal when normal feedwater sources are unavailable dueT power or other malfunctions. capable of delivering feedwater to either auxiliary feedwater pumps (1050 gpm)The AN system is part of-the steam a or both steam generators. and, as stated in the FSAR, is designed in rupture control system (SFRCS) 279-1971 [81 accordance with IfEE Std 3.2 AUTOMATIC INITIATION 3.2.1 Evaluation d by the SFRCS, Auxiliary feedwater flow to the steam generators steam andistoprovide automatically which is designed to prevent release of high energy t re or main start the AN system in the event of a main steam line rup u parameters llowing feedwater line rupture or when preset levels of any of-the fo are exceeded:
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| 1 low level in either steam generator rator 2
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| 3.
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| main steam line rupture (low-pressure) main feedwater lin ,
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| differential pressure) 4.
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| loss of all four reactor coolant pumps .
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| : 5. loss of both main feed pumps.
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| The AW system consists of two System valves are controlled by the SFRCS.Each train has two motor-ope turbine-driven pump trains in parallel. lly shut valve in normally shut valves in series and one motor-operated, ' i normavalves on one train a cross-connect line which connects between the two ser esThe cross i and downstream of both series valves on the other tra Ope.n/ n. closed lines allow either pump to feed either or both lsteam room for generators.
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| the metor-valve position indication is provided in the contro CS The A N system can be controlled automatically by the SFR operated valve.
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| TER-C5257-302 l
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| i or manually from the control room or auxiliary shutdown panel. The operation of either pump provides the capacity to remove decay heat from the steam generators at a rate sufficient to prevent overpressurization of the reactor coolant system and to maintain steam generator levels. Therefore, the APW system is capable of automatically initiating appropriate protective action with precision and reliability whenever a condition monitored by the system -
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| reaches a preset level.
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| The primary source of water is the condensate storage tank which is sized so that a total condensate inventory may be available to the pumps sufficient to remove decay heat for 13 hours plus a subsequent cooldown to 280*F. The backup water supplies are the service water system and the fire protection system. Iow-pressure switches are provided on the APW pump suction line which will automatically shift supply to service water system.
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| The AEW system at the Davis-Besse plant is designed as an-engineered safeguards system, the entire system meets Safety Class I criteria, and the automatic initiation signals and circuits comply with the single-failure criterion of IEEE Std 279-1971. A review of initiation logic and wiring diagrams revealed no credible single malfunction that would prevent proper protective action at the system level when required. The diverse signals and redundant channels that provide automatic initiation are physically separated, electrically independent, and powered from emergency buses. In addition, the SFRCS consists of two identical, redundant, and independent channels, and each channel consists of one ac-supplied logic train and one de-supplied logic train. AEW pump 1-1 steam inlet valve (MS-106) and discharge valves (AF3870 and_ AF360) are powered by dc (Class 1E, essential power supplies EllC, EllD, and EllE) and AEW pump 1-2 steam inlet valve (MS-107) and discharge valves (AF3872 and AF388) are powered by ac (Class lE, 480-volt power supplies Fila, F12B, and F12A) . -
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| The AFW system can be manually initiated. The turbine-driven pumps can be started either from the control room or the local control panel. The automatic initiating circuits are designed to be electrically independent from 4 _ . , _
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| TER-C5257-302 4 valves, and initiate the auxiliary feedwater system when the steam generator pressure exceeds the main feedwater pressure by 197.6 psig."
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| In Section 7.4.2.3 of the Davis-Besse FSAR, the Licensee states that the design of the SFRCS is in compliance with the requirement of IEEE Std 279-1971. In addition, the Licensee states that no output signals from the SFRCS are used for control functions.
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| The SFRCS is capable of being tested at power, including system logics, actuation devices, and actuated equipment. A manual testing capability is provided for each input signal to the SFRCS to simulate sensor operation. The system level transmitters, pressure switches, and the differential pressure switches are capable of being independently isolated and, while isolated, simulated process parameters can be applied to check calibration.
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| Operating and channel bypasses are provided as part of the SFRCS and are designed in accordance with IEEE Std 279-1971, as stated by the Licensee. The FSAR describes these operating and channel bypasses in Section 7.4.1.3.4, as follows:
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| "7.4.1.3.4 Bypasses
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| ~
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| SFRCS includes channel bypasses and operating bypasses.
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| : 1. Channel bypass: The only bypasses provided are those on the MCB and ASP for the auxiliary feedwater system. These channel bypasses will l permit operator selection of manual control or ICS automatic l
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| control. The operation of these switches is under administrative l contrcl. The switch position of these bypasses are indicated on the
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| ( MCB and/or ASP and alarmed in the control room.
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| : 2. Operating bypasses: Two out of two logic is provided to allow the cperator to bypass each channel to prevent initiation under normal coc1 down when the main steam line pressure drops below 650 psig.
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| The bypast.es are automatically reset by a one out of two looic when the nain steam line pressure exceeds 650 psig."
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| All active components of the AEW system are accessible for inspection during normal station operation. The AFW pumps are tested monthly during station operation, by discharging via the recirculation piping back to the i condensate storage tanks.
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| l l A _ _ _ _
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| TER-CS257-302.
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| j .
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| 1 3.2.2 conclusion Based on the evaluation in Section 3.2.1, it is concluded that the initiation signali, logic, and associated circuitry of the AFW system at the Davis-Besse plant comply with the long-term. safety-grade requirements of Section 2.1.7.a of NUREG-0578 [3] and the subsequent clarification issued by .
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| the NRC, with the exception of the AEW pump power supply diversity require-ment, i.e., both AFW pumps are turbine-driven.
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| 3.3 FI4W INDICATION ,e 3.3.1 Evaluation The performance of the AFW system at the Davis-Besse plant can be assessed by indication of AEW flow, steam generator startup range level, steam generator operating range level, APW system valve position, and APW pump status.
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| The Davis-Besse plant has one AFW flow channel per steam generator which is a differential pressure device located downstream of any cross connect to The power ensure indication of'.AFW flow delivered to each steam generator.
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| supplies (Class lE) , instrumentation, and displays are designed as safety-grade. The AEW flow indication channels provide no protection function.
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| In Reference 3, the Licensee provided information to justify having only The Licensee's one safety-grade AFW flow channel per steam generator.
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| statement is as follows: ,
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| "As part of our effort to continuously review and upgrade the Auxiliary Feedwater System (AFWS) for reliability and performance, we have completed an extensive reliability analysis of the Davis-Besse AEWS.
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| Results of this analysis, performed by EDS duelear Inc., will be made available to the NRC in the near future. The analysis provides a useful~
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| framework for evaluating the impact of AFWS flow indicators on overall plant safety.
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| Results of the Davis-Besse AEWS reliability analysis indicate that the AFdS flow indicators do not contribute substantially to AEWS reliability. The reasons for this are as follows: .
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| : 1. The actuation of the two trains of AEWS is accomplished automatically through safety-grade signals. The automatic action includes
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| _9_ _ _ _
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| TER-C5257-302 switching to a safety-grade backup water supply (service water supply) should the primary MS water supply (condensai,e storage tanks) fail. Operator action based on control room indication of MS flows, is therefore not required.
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| : 2. Emergency procedures exist to guide operator actions for mitigating faults should one or both trains of the MS fail. The M S flow '
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| indicators provide only one of several measures of M S performance, however. The primary indicator of system performance is steam generator secondary side water level which is monitored and indicated inside the control room by safety grade instrumentation. In addition, separate redundant safety grade level indicators are also available in the cabinet room. Other parameters available to the plant operators for diagnosis of faults and identification of corrective actions include condensate storage tank level, m pump speed, m pump discharge pressure (all non safety grade) and safety grade steam generator pressure.
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| : 3. The steam generator level instrumentation which provides the primary and most important indication of availability of secondary side, heat sink is powered from a safety grade power supply redundant from the one supplying the flow instrumentation. Thus, it is highly improbable to lose both the level and existing flow indication caused by a loss of channel power.
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| l
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| : 4. Other plant design and procedures modifications have been implemented at Davis-Besse subsequent to the TiI-2 event. These modifications have addressed the most significant contributions to APWS unavailability and have reduced the system unavailability by over an order of magnitude. These modifications include:
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| diverse electric power sources (DC-power supplies) for motor operated valves in one train of the APWS.
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| ~ redundant and seismically qualified turbine exhausts.
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| administrative procedures to lock in position all manual valves and local control stations and handwheels for motor operated valves in the AFWS.
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| automatic steam generator dual level setpoint control.
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| an emergency procedure to supply water to the stean. generator through the main feedwater startup pump should both trains of the M S fail.
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| These modifications have greatly reduced the probability for failure to achieve the AEWS safety function, and have thereby diminished the relative significance of the APWS flow indication.
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| 5.
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| The inclusion of one APWS flow indicator per train has little impact on the reliability analysis results. The inclusion of a second flow The only application for indicator per train has even less impact.the flow indicator in the r Generally, operator in recovering from a faulted system condition.th instrumentation performance. A second flow indicator would not address the more dominant human factors, and would therefore not .
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| significantly improve system reliability.
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| : 6. The failure of a flow indicator, in itself, will not lead to adve consequences.
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| pressure indications provide adequate informationEven for operator in the to ensure availability of adequate secondary heat steam sink. even unaffected steam generator, the availability of (unaffected) genearator has been previously shown to be acceptable for decay hea removal purposes. There are no credible occurrences in which automatic or manual actions would be taken to terminate AEW based on a faulty indication of that flow.
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| l The insensitivity of the overall AEWS reliability to a second AEWS f ow indicator and the inclusion of truly diverse indication of the primary d performance parameter (steam generator level) make unnecessary a seco flow indicator in each train."
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| Even though, as the Licensee has indicated, other indication and auto-i f two AEW flow matic features exist to assisti the operator, the availabil ty o The single channels for each Babcock & Wilcox steam generator is essential. i ts of AEW flow channel proposed by Toledo Edison does not meet the requ remen NUREG-0737, in particular the single failure requirements.
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| 3.3.2 Conclusion _
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| d The present AEW flow indicat ion does not meet the long-term, ification safety-gr requirements of NUREG-0578, Section 2.1.7.b, and the subsequent clar In order to comply with these requirements, a second
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| ' issued by the NRC.
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| safety-grade AEW flow indication channel per steam generator should be installed. .
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| ==3.4 DESCRIPTION==
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| OF STEAM GENERA'IOR LEVEL INDICATION There are three level channels for monitoring steam generator level at
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| 'Ihese include (1) startup range, (2) operate range, the Davis-Besse plant.
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| , TER-C5257-302 and (3) full range level instrumentation.- Following is a grouping of these level instrumentations and an indication of their power supply. The instrumentation used for control or automatic initiation of auxiliary feedwater is indicated by an asterisk (*) . Under the power source column, some of the instruments are shown to be powered from uninterruptible buses YAU (and YBU) or vice versa. At present, these instruments are powered from power supplies shown outside the parenthesis. Following the modifications scheduled for implementation in the 1982 refueling outage, these instruments will be able to be powered from both sources.
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| Instrument Number Servica Description Power Source Startup Rance Instrumentation N ES-SP9B2 SG 1 SU Level Selector YAUandhBU LT-SP9B3 SG 1 SU Level Transmitter
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| * Y1 LY-SP9B3 SG 1 SU Level for AFPT 1* Y1 LC-SP9B3 SG 1 SU Ievel for AFPT 1* '
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| Y1 LI-SP9B3 SG 1 SU I4 vel for Indicator Y1 LT-SP9B4 SG 1 SU Level Transmitter
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| * Y2 LY-SP9B4 SG 1 SU Level for AFPT 2*
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| LC-SP984 Y2 SG 1 SU Level for AFPT 2* Y2 LY-SP9B6 SG 1 SU Level for Buffer Module Y2 LS-SP98 SG 1 SU Imvel for Iow Alarm LI-SP9B1 YBU and YAU SG 1 SU I4 vel Indicator Y1 ES-SP9A2 SG 2 Level Selector YAU and YBU LT-SP9A3 SG 2 SU I4 vel Transmitter
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| * Y2 LY-SP9A3 SG 2 SU Level for AFPT 2* Y2 LC-SP9A3 SG 2 SU Level for AFPT 2* Y2 LI-SP9A3 SG 2 SU Level Indicator Y2 LT-SP9A4 SG 2 SU .I4 vel Transmitter
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| * Y1 LY-SP9A4 SG 2 SU I4 vel for AFPT 1*
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| LC-SP9A4 Y1 SG 2 SU.Ievel for AFPT 1* Y1 LY-SP9A6 SG 2 SU Level Buffer Module LS-SP9A Y1 SG 2 SU Level Icw Alarm YBU and YAU
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| ; LI-SP9Al SG 2 SU Level Indicator Y2
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| _ ~ . - - _ _ _ _ _ _ _ _ - _ . _ _ , . _ _ _ _ . _ . _ . _ _ . , _ _ _ _ _ _ . _ _ , . _ , _
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| TER-C5257-302 Instrument Number Service Description Power Source LT-SP9B6 SG 1 Level Transmitter for SFRCS (CH 2)* Y2 LI-SP9B6 SG 1 Level Indicator for SFRCS (CH 2) Y2 LSLL-SP986 ,
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| SG 1 Low Ievel Trip for SFRCS (CH 2)
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| * Y2 LT-SP9B7 SG 1 Level Transmitter for SFRCS (CH 2)* D2P ,
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| LI-SP987 SG 1 Level Indicator for SFRCS (CH 2) D2P LSLL-SP9B7 SG 1 Low Level Trip for SFRCS (CH 2)
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| * D2P LT-SP9B8 SG 1 Level Transmitter for SFRCS (CH 1)* Y1 LI-SP9B8 SG 1 Level Indicator for SFRCS (CH 1) Y1 LSLL-SPSB8 SG 1 Low Level Trip for SFRCS (CH 1)
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| * Y1 LT-SP9B9 SG 1 Level Transmitter for SFRCS (CH 1)* DlP LI-SP9B9 SG 1 Level Indicator for SFRCS (CH 1) D1P LSLL-SP9B9 SG 1 Iow -Level Trip for SFRCS (CH 1)
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| * DlP LT-SP9A6 SG 2 Level Transmitter for SFRCS (CH 1)
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| * Y1 LI-SP9A6 SG 2 Level Indicator for SFRCS (CH 1) Y1 .
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| LSLL-SP9A6 SG 2 Low Level Trip for SFRCS (CH 1)
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| * Y1 LT-SP9A7 SG 2 Level Transmitter for SFRCS (CH 1)* DlP LI-SP9A7 SG 2 Level Indicator for SFRCS (CH 1) DlP LSLL-SP9A7 SG 2 Low Level Trip for SFRCS (CH 1)
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| * DlP LT-SP9A8 SG 2 Level Transmitter for SFRCS (CH 2)
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| * Y2 LI-SP9A8 SG 2 Level Indicator for SFRCS (CH 2) Y2 LSLL-SP9A8 SG 2 Low' Level Trip for SFRCS (CH 2)
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| * Y2 LT-SP9A9 SG 2 Level Transmitter for SFRCS (CH 2)
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| * D2P
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| ; LI-SP9A9 SG 2 Level Indicator for SFRCS (CH 2) D2P
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| ! LSLL-SP9A9 SG 2 Low Level Trip for SFRCS (CH 2)
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| * D2P l
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| Ooerate Range Level Instrumentation ,
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| HS-SP9B1 SG 1 Operate Level Selector YAU (and YBU)
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| LT-SP9B1 SG 1 Operate Level Transmitter YBU (and YAU)
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| LY-SP9B1 SG 1 Operate Level Temperature Compensator YBU (and YAU) l LS-SP9B1 SG 1 Operate Level High Level Alarm YBU and YAU l
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| l LT-SP982 SG 1 Operate Level Transmitter YAU (and YBU)
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| LY-SP982 SG 1 Operate Level Temperature Compensator YAU (and YBU) l LRS-SP9B SG 1 Operate Level Recorder YBU (and YAU)
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| HS-SP9Al SG 2 Operate Level Selector YAU (and YBU) l l
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| l 4 _. _
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| bd Franklin Research Center A Dmm ori of The Fransert kwetuta
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| . . _ . . . . , , _ . . . _ , . . . _ . . . . . . . . . _ . I... ._ . . _ . . ..
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| - ,_ , u _
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| TER-C5257-302 Instrument Number Service Description Power Source LT-SP9Al SG 2 Operate Level Transmitter YBU (and YAU)
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| LY-SP9Al SG 2 Operate Ievel Temperature Compensator YBU (and YAU)
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| LS-SP9Al SG 2 Operate Level Hi,gh Level Alarm YBU and YAU LT-SP9A2 SG 2 Operate Level Transmitter YAU (and YBU) -
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| LY-SP9A2 SG 2 Operate Level Temperature Compensator YAU (and YBU)
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| LRS-SP9A SG 2 Operate Level Recorder YBU (and YAU)
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| Full Range Level Instrumentation LI-SP9B2 SG 1 Full Range Level Indicator YBU (and YAU)
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| LT-SP9B5 SG 1 Full Range Level Transmitter YBU (and YAU LY-SP9B5 SG 1 Full Range Level E/I YAU Converter LS-SP9B5 SG 1 Full Range I4 vel Alarm YAU LI-SP9A2 SG 2 Full Range Level Indicator YBU (and YAU)
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| LT-SP9AS SG 2 Full Range Level Transmitter YBU (and YAU)
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| LY-SP9AS SG 2 Full Range Level E/I YBU Converter LS-SP9AS SG 2 Full Range Level Alarm YBU l
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| l I
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| 4 -- -
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| 0003 Frankin Research Center A CNn.an of The Frannan enseue
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| TER-C5257-302-
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| : 4. CONCLUSIONS Based on the evaluation in Section 3.2.1, it is concluded that the initiation signals, logic, and associated circuitry of the A N system at the Davis-Besse plant comply with the long-term safety-grade requirements of ,
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| Section 2.1.7.a of NUREG-0578 [3] and the subsequent clarification issued by the NRC, with the exception of the AN pump power supply diversity require-ment, i.e., both AN pursps are turbine-driven.
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| The present AN flow indication does not meet the long-term, safety-grade requirements of NUREG-0578, Section 2.1.7.b, and the subsequent clarification issued by the NRC, because only one safety-grade channel of APW flow indication is installed. In order to meet the NRC requirements, a second safety-grade AN flow indication channel per steam generator should be .
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| installed.
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| l l
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| t l .
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| (
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| i l
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| i i _
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| l l
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| A Nbil Frank!!n Research Center A Opnen,c of The Frammkn W
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| *- ygm p ;g- y 9 , t .,y=g , 5 -w- - - --e, - ---, . - p ,3,- g7 ,, , , pre.
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| o TER-C5257-302
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| : 5. REFERENCES
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| : 1. Code of Federal Regulations, Title 10, Office of the Federal Register, National Archives and Records Service, General Services Adminis tration -
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| Revised January 1, 1980 .
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| : 2. NRC, Generic letter to all PWR licensees regarding requirements resulting from Three Mile Island Accident September 13, 1979
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| : 3. NUREG-0578, "'IMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" USNRC, July 1979
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| : 4. NRC, Generic letter to all PWR licensees clarifying lessons learned short-term requirements October 30, 1979
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| : 5. NRC, Generic letter to 'all PWR licensees regarding short-term requirement resulting from Three Mile Island accident Septemoer 5, 1980
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| : 6. NUREG-0737, " Clarification of StI Action Plan Requirements"-
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| USNRC, November 1980 .
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| : 7. R. P. Crousa (Tbledo Edison)
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| Letter to J. F. Stolz (NRC)
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| | |
| ==Subject:==
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| Additional Information on AFW System Automatic Initiation Toledo Edison Company,16-Sep-81
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| : 8. IEEE Std 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations," Institute of Electrical and Electronics Engineers, Inc., New York, New York ,
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| : 9. NUREG-75/087, Standard Review Plan, Section 10.4.9, Rev. 1, USNRC,_
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| no date
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| : 10. Regulatory Guide 1.97 (Task RS 917-4) , " Instrumentation for Lighc-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following and Accident," Rev. 2,
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| ; USNRC, December 1980 4
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| ~ ~ ~
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| bbranldin Research Center Atw nonn.r na.. w u, .
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