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| TECHNICAL EVALUATION REPORT PWR MAIN STEAM LINE BREAK WITH
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| * CONTINUED FEEDWATER ADDITION (B-69)
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| ARKANSAS POWER AND LIGHT COMPANY ARKANSAS NUCLEAR ONE UNIT 1 NRC DOCKET NO. 50-313 FRC PROJECT C5606 NRCTACNO. 46824 ,
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| FRC ASSIGNMENT 5 NRC CONTRACT NO. NRC43-81 130 FRC TASK 122 Preparedby Franklin Research Center Author: F. W. Vosbury 20th and Race Street Philadelphia, PA 19103 FRC Group Leader: R. C. Herrick Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer: P. Hearn September 28, 1982 This report was prepared as an account of work sponsored 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 results of such use, of any information, appa-ratus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rl0 hts.
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| P.repared by: Reviewed by: Approved by:
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| hcb A- &f '
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| - e$
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| _ $$ -sw Principal Autheb D'ekrtment DirdctoO Date: k c,79 -ff-C. Date: 9/gp/ft Date: 9-21-92.
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| al
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| . 0. Franklin Research Center A Division of The FranWin Institute QQ.{ ,
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| The seni m n Frenidin Perhey, PMe Pa. 19103(215)448-1000
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| TER-C5506-122 CONTENTS Section Title Page 1 INTRODUCTION. . . . . . . . . . . . . . 1 1.1 Purp'se of Review . . . . . . . . . . . 1
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| ; 1.2 Generic Background . . . . . . . . . . . 1 1.3 Plant-Specific Background . . . . . . . . . 3 2 ACCEPTANCE CRITERIA . . . . . . . . . . . . 4 3 TECHNICAL EVALUATION. . . . . . . . . . . . 8 3.1 Review of Containment Pressure Response Analysis . . . 8 3.2 Review of Reactivity Increase Analysis . . . . . . 15 3.3 Review of Corrective Actions . . . . . . . . 17 4 CONCLUSIONS . . . . . . . . . . . . . . 18 5 REFERENCES . . . . . . . . . . . . . . 19 S .
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| . . . , l TER-C5506-122 P.
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| FORIMORD This hchnical Evaluation Report was prepared by Franklin Research Center
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| 'under a contract with the U.S. Rac.'. ear Regulatory Cbanission (Office of Nuclear Reactor 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. F. W. Vosbury contributed to the technical preparation of this report through a subcontract with WESTEC Services, Inc.
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| f BOO ranidin Research Center A Dmman of The Fremen innumme
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| TER-C5506-122
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| : 1. INTRODUCTION 1.1 PURPOSE OF REVIEW Sis Technical Evaluation Report (TER) documents the review of Arkansas Power and Light Company's (APL) response to the Maclear Regulatory Commission's (NRC) IE Bulletin 80-04, " Analysis of a Pressurized Water Reactor Main Steam Line Break with Continued Feedwater Addition" [1], as it pertains to Arkansas Nuclear One Unit 1. This evaluation was performed with the following objectives:
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| o to assess the conformance of APL's main steam line break (MSIA) analyses with the requirements of IE Bulletin 80-04 o to assess APL's proposed interia and long-range corrective action plans and schedules if needed as a result of the MSLB analyses.
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| 1.2 GENERIC BACKGROUND In the summer of 1979, a pressurized water reactor (PWR) licensee submitted a report to the NRC that identified a deficiency in the plant's original analysis of the containment pressurization resulting from a MSIA. A reanalysis of the containment pressure response following a MSIA was performed, and it was determined that, if the auxiliary feedwater (AFW) system continued to supply feedwater at runout conditions to the steam generator that had l experienced the steam line break, containment design pressure would be exceeded in approximately 10 airiutes. Se long-term blowdown of the water supplied by the AFW system had not been considered in the earlier analysis.
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| On October 1, 1979, the foregoing information was provided to all holders of operating licenses and 'constructior. permits as IE Information Notice 79-24
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| [2]. Another facility performed an accident analysis review pursuant to receipt of the infarmation in the notice and discovered that, with offsite ,
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| electrical per available, the condensate pumps would feed the affected steam generator a,t an excessive rate. This excessive feed was not previously considered in the plant's analysis of a MSLB accident.
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| 00bu Franklin Research Center A Demeen of The Frermen humane
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| . . - _ . . . . .-- . - - . _ - - - _ _ - _ . _ _. . ._.- . - ._ _~
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| TER-C5506-122 A third licensee informed the NRC of an error in the MSI2 analysis for thyrplant. During a review of the MSIa analysis, for zero or low power at the end of core life, the licensee identified an incorrect postulation that the startup feedwater control valves would remain positioned "as is" during the transient. In reality, the startup feedwater control valves will ramp to 804 full open due to an override signal resulting from the low steam gener'ator pressure reactor trip signal. Reanalysis of the events showed that opening of the startup valve and associated high feedwater addition to the affected steam generator would cause a rapid reactor cooldown and resultant reactor return-to-power response, a condition which is outside the plant design basis.
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| Because of these deficiencies identified in original MSLB accident analyses, the NRC issued IE Bulletin 80-04 on February 8, 1980. His bulletin required all PWRs with operating licenses and certain near-term PWR operating license applicants to perform the following:
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| "1. Review the containment pressure response analysis to determine if the potential for containment overpressure for a main steam line break inside containment included the impact of runout flow from the auxiliary feedwater system and the impset of other energy sources, such as continuation of feedwater or condensate flow. In your review, consider your ability to detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable after extended operation at runout flow.
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| : 2. Review your analysis of the reactivity increase which results from a main steam line break inside or outside containment. His review should consider the reactor cooldown rate and the potential for the reactor to return to power with the most reactive control rod in the fully withdrawn position. If your previous analysis did not consider all potential water sources (such as those listed in 1 above) and if the reactivity increase is greater than previous analysis indicated, the report of th%s paview should include:
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| ,a. Se boundary conditions for the analysis, e.g. , the end of life shutdown margin, the moderator temperature coefficient, power level and the not effect of the associated steam generator water inventcry on the reactor system cooling, etc.,
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| : b. S e most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution to the reactor coolant system,
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| . nidin Research Center A Osnum of The Fransen hemmae ,
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| - . . . . , - - - . . . - . . - _ - . ~ - . . . _ . . _ . _ . . . . ..
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| TER-C5506-122
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| : c. The effect of extended water supply to the affected steam generator on the core criticality and return to power, F
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| : d. The hot channel factors corresponding to the most reactive rod in the fully withdrawn position at the end of life, and the Minimum Departure from Itacleate Boiling Ratio (MDNBR) values for the analyzed transient.
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| : 3. If the potential for containment overpressure exists or the reactor return-to-power response worsens, provide a proposed corrective action and a schedule for completion of the corrective action. If the unit is operating, provide a description of any interim action
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| . that will be taken until the pecposed corrective action is completed."
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| 1.3 PLANT-SPECIFIC BACKGROUND Arkansas Power and Light Company responded to IE Bulletin 80-04 in letters to the NRC dated May 27, 1980 [3] and July 9, 1980 [4]. Additional information was provided in a letter dated July 30, 1982 [5]. The information in References 3, 4, and 5 has been evaluated along with pertinent information from tne Arkansas Nuclear One Unit 1 Final Safety Analysis Report (FSAR) [6]
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| to determine the adequacy of the Licensee's coipliance with IE Bulletin 80-04.
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| i TER-C5506-122 l
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| : 2. ACCEPTANCE CRITERIA
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| : p. 2 The following criteria against which the Licensee's MS2 response was l 1
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| evaluated were provided by the NitC [7]:
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| : 1. PWR licensees' responses to IE Bulletin 80-04 shall include the following information related to their analysis of containment pressure and core reactivity response to a MSLB within or outside containment:
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| : a. A discussion of the continuation of flow to the affected steam generator, including the impact of runout flow from the AFW system and the 13 pact of other energy sources, such as continuation of feedwater or condensat'e flow. ANW system runout flow should be determined from the manufacturer's pump curves at no backpressure, unless the system contains reliable anti-runout provisions or a more representative backpressure has been conservatively calculated. If a licensee assumes credit for anti-runout provisions, then justification and/or documentation used to determine that the provisions are reliable should be provided. Examples of devices for which provisions are reliable are anti-runout devices that use active components (e.g., automatically throttled valves) which meet the requirements of IEEE Std 279-1971 [8] and passive devices (e.g. ,
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| flow orifices or cavitating venturis) .
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| : b. A determination of potential containment overpressure as a result of the impact of runout flow from the AFW system or the impact of other i energy sources such as continuation of feedwater or condensate l flow. Where a revised analysis is submitted or where reference is made to the existing FSAR analysis, the analysis must show that
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| + runout AFW flow was included and that design containment pressure was not exceeded.
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| : c. A discussion of the ability to detect and isolate the damaged steam generator from continued feedwater addition during the MS 2 accident.
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| Operator action to isolate AFW flow to the affected steam generator within the first 30 minutes of the start of the HS2 should be justified. If operator action is to be completed within the first 10 minutes, then the justification should address the indication available to the operator and the actions required. Where operator
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| . action is required to prevent exceeding a design value, i.e.,
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| containment design pressure or specified acceptable fuel design limits, then the discussion should include the calculated time when the design value would be exceeded if no operator action were assumed. latere operator actions are to be performed between 10 and 30 minutes after the start of the MS 2 , the justification should address the indications available to the operator and the operator actions required, noting that for the first 30 minutes, all actions should be performed from the control room.
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| e TER-C5506-122 J. Where all water sources were not considered in the previous analysis, an indication should be provided of the core reactivity change which results from the inclusion of additional water sources. A submittal P- which does not determine the magnitude of reactivity change fro's an original analysis is not responsive to the requirements of IE Bulletin 80-04. ,
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| : 2. If containment overpressure or a worsening of the reactor return-to- ,
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| power with a violation of the specified acceptable fuel design limits described in Section 4.2 of the Standard Review Plan [9] (i.e.,
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| increase in core reactivity) can occur by the licensee's analysis, the licensee shall provide the following additional informations
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| : a. The proposed corrective actions to prevent containment overpressure or the violation of fuel design limits, and the schedule for their completion.
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| : b. The interim actions that will be taken until the proposed corrective action is completed, if the unit is operating.
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| : 3. The acceptable input assumptions used in the licensee's analysis of the core reactivity changes during a MSLB are given in Section 15.1.5 of the Standard Review Plan [10]. The following specific assumptions should be used unless the analysis shows that a different assumption' is more limiting:
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| Assumption II.3.b. : Analysis should be performed to determine the most conservative assumption with respect to a loss of electrical power. A reactivity analysis should be conducted for a normal power situation as well as a loss of offsite power scenario, unless the licenses has previously conducted a sensitivity analysis which demonstrates that a particular assumption is more conservative.
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| Assumption II.3.d.: The most restrictive single active failure in the safety injection system which has the effect of delaying the delivery of high ,
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| . - concentration boric acid solution to the reactor coolant system, or any other single .
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| . active failure affecting the plant response, ,
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| should be considered. :
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| Assumption II.3.g.: The initial core flow should be chosen such ,
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| that the postHMSLB shutdown margin is ,
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| 3 minimized (i.e., maximum initial core flow) .
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| I TER-C5506-122 l l
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| R e acceptable computer codes for the licensee's analysis of core I reactivity changes are, by nuclear steam supply system (NSSS) vendor, y the following: CESEC (Combustion Engineering), INTRAN (Westing-house), and TRAP (Baccock (i Wilcox) . Other cetr.puter codes may be used, provided that these codes have previously been reviewed and found to be acceptable by the NRC staff. If a computer code is used which has not been reviewed, the licensee must describe the method employed to verify the code results in sufficient detail to permit the code to be reviewed for acceptability.
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| : 4. It the APW pumps can be damaged by extended operation at runout flow, the licensee's action to preclude damage should be reviewed for technical merit. Any active features should satisfy the requirements of IEEE Std 279-1971. Where no corrective action has been proposed, ;
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| this should be indicated to the NBC for further action and resolution.
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| ~
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| J 5. Modifications to the electrical instrumentation and controls needed to. detect and initiate isolation of the affected eteam generator and feedwater sources in order to prevent containment overpressure and/or unacceptable core reactivity increases must satisfy safety-grade requirements. Instrumentation that the operator relias upon to follow the accident and to determine isolation of the affected steam generator and feedwater sources should conform to the criteria contained in ANS/ ANSI-4.5-1980, " Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors" [ll),' and the regulatory positions in Regulatory Guide 1.97, Rev. 2, " Instrumentation for Light-Water-Cooled Ikiclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" [12].
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| : 6. AFW system status should be reviewed to ensure that system heat removal capacity does not decrease below the minimum required level as a result of isolation of the affected steam generator and also that recent changes have not been made in the system which adversely affect vital assumptions of the containment pressure and core reactivity response analyses.
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| : 7. H e safety-grade requirements (redundancy, seismic and environmental qualifications, etc.) of the equipment that isolates the main feedwater (MFW) and AFW systems from the affected steam generator should be specifitd. He modifications of equipment that are relied upon to isolate the MFW and AFW systems from the affected steam
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| * generator should satisfy the following criteria to be considered safety-grade ,
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| o Redundancy and power source requirements: S e isolation valves should .be designed to acce==adate a single failure. A failure-modes-and-effects analysis should demonstrate that the system is capable of withstanding a single failure without loss of function.
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| *% v s TER-C5506-122 1 The single failure analysis should be conducted in accordanct, l.En -
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| E the appropriate rules of application of ANS-51.7/NG58%976, '
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| " Single Failure Criteria for MfR Fluid Sys,tems" [13]d ' s
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| ,' . s o seismic requirements: Se isolation, valves should be designed to .
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| Category I as recommended in Regulatory Guide 1.26 [14). , - 3 p.
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| o Environmental qualification: The isolation valves should satisfy .
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| the requirements of NUREG-0588, aav.1, " Interim Staff Positioh on ,
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| Environmental Qualification of Safety-Related "'
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| . Equipment" [15]. -
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| o Quality atandards: Se isolation valves should satisfy. Group's '* ,
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| quality standards as recommended in Regulatory Guide 1.26 or *I similar quality standards frora the plant's . licensing bases. .N 3
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| : 3. 'fEC5MICAL EVALUATION s, % -
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| ^ U i 'Ibe'srope of Wosk Arcluded the followjngs m 3' l y8 s . ,
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| ' ' , ' f., RevA 7 w;che Licensec,'s respons; to IE Bulletin 80-04 against the accoppace criteria. - "
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| k i i 2. a. - Eval ste the Licensee's MSL3 analyses for thy potential of overpressurizing the containhent and ,with respect to the core t 'r'asctivity increase due to the effect of continued feedwater flow.
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| ^
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| ' b'.' Evalunte the Licensee's pcupoded corrective actions and schedule i forsi:splementation if the yndings o! Task 2a indicate that a potential existe for overpressurizing the containment or worseting the reactor return-to-power in the event of a MSLB S ' -
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| 'S 7 accidant.
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| + <
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| , x
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| : 3. Prepare a TER for each plant based on.ihe evaluation of the
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| . information presented for Tasks 1 and 2 above. .
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| This report" constitutes a TER in satisfaction of Task 3. Sections 3.1 ithrough 3.3 of this r eport state the requirements of IE Bulletin 80-04 by s@section, summarist the Lic<ensee's statements and conclusions regarding these requirements,' and present a discussion of the Licensee's. evaluation
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| \
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| 'followed by conclusions and recommendations.
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| . s '
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| s ,
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| <. 3.1 REVIEW OF CONTAIleGMT PRESSURE RESPONSE. ANALYSIS
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| , . ~ ^
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| l l The requirement from IE Bulletin 80-04, item'1, is as follows:
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| s &
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| " Review the containment prosyure \r\esponse analysis to determine if the L{ potential for containment ovecocesaufe for a main steam line break inside containment included the impact of, runout flow from the auxiliary feedwater system and the impact of other energy sources, such as continuation of feedwater ,
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| or cctdensate flow. In your review, consider N your ability to detect and isolate the damaged steam generator from these i
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| sources and the ability of the pumps to remain operable after extended operation at runout flow."
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| 3.1.1 Sammary of Licensee Statements and (bnclusions s
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| In regard to the review of the containment pressure response analysis, the Licensees stated .(3):
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| f TER- C5506-122 "As.a restilt of a main steam line break inside the reactor building of ANO-1, the steam pressure ir. both Once Through Steam Generators (OTSG)
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| P would decrease quite rapidly. The rate of depressurization would of I course depend upon the break size. For a steam line break accident, the reactor power would increase with the decreasing average reactor coolant temperature as a result of a negative moderator coefficient. The ICS will cause insertion of control rods in an attempt to limit the reactor power,to 102 percent.. If the break were large, the reactor power increa'se suld not be limited sufficiently by the ICS and a reactor trip would occur due to high neutron flux and/or low reactor coolant pressure.
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| 4 Following the reactor trip, the turbine will trip and the ICS will run back the feedwater flow. Due to the low OTSG pressure in the affected OTSG, the safety grade Steam Line Break Instrumentation and Control System (SIBIC) would actuate, isolating the affected OTSG by closing the respectivs feedwater isolation valve and both main steam block valves. A SLBIC signal also opens the steam supply to the turbine driven emergency feedwater pump. As the affected OTSG boils dry, the emergency feedwater actuation and control system will actuate the emergency feedwater system when it receives a OTSG level of less than 18 inches in either generator.
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| This signal will actuate the motor driven emergency feedwater pump (the turbine driven pump has already been actuated by SLBIC) and align the emergency feedwater valves in both trains.
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| Upon realizing he has a steam line break accident, the operator, using Emergency Operating Procedure 1202.24, will determine the affected OTSG by observing the OTSG 1evels and pressures. ITpon identifying the affected OTSG, the operator will close the affected OTSG's emergency feedwater system steam supply and feed valves, and open, if not presently open, the corresponding steam supply valve on the unaffected OTSG. The operator would then commence cooldown to cold shutdown utilizing the unaffected OTSG.
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| If for some unlikely reason the operator fails to isolate the emergency fecdwater to the affected Cm)G, it has been shown through analysis using the assumptions in Attachment A that the reactor building pressure would not reach the design pressure of 59 psig until approximately 3 hours and 45 minutes into the accident allowing more than sufficient time for tne operators to take corrective action."
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| In regard to a request for additional information concerning the possibility and consequences of continued main feedwater addition to the affected steam generator af ter a MSta, the Licensee stated [5]:
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| "The analysis provided you in our original response [3] did consider the l' effects of concern. Although not explicitly stated in that response, an assumption in the 'no operator action' case was the failure of SLBIC.
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| SIBIC automatically isolates main feedwater thus its failure assumed the isolation did not take place. The assumption of 'no operator action' was ranklin Research Center A Denman of The Feensen huumme
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| TER-C5506122 l
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| made to preclude manual isolation of main feedwater by.the operators.
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| (it should be noted that not all B4W plants were designed with SLBIC y system thus manual isolation of main feedwater was the only isolation means available. As such, B&W maintained the 'no operator action' assumption in all analysis to generically bound all plants.)"
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| In regard to the ability of t' e AFW pumps to remain operable during a MSta, the Licensee stated [4]:
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| " Analyses were performed by the Architect Engineer, using plant specific data and input from the pumps manufacturer, to determine if the emergency feedwater pumps would remain operable after*possible runout flow conditions following a main steam line break (MSta) . These analyses demonstrated, even assuming no operator action, that the emergency feedwater pumps will remain operable during and following a MSIa accident considering runout flow conditions."
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| i r 3.1.2 Evaluation The Licensee's submittals [3, 4, 5] concerning the containment pressure response following a MSLB and applicable sections of the Arkansas Nuclear One Unit 1 FSAR [6] were reviewed in order to evaluate whether the following portions of the acceptance criteria were mets o Criterion 1.a - Continuation of flow to the affected steam generator o Criterion 1.b - Potential for containment overpressure o Criterion 1.c - Ability to detect and isolate the damaged steam generator o Criterion 4 - Potential for AFW pump damage o criterion 5 - Design of steam and feedwater isolation system o criterion 6 - Decay heat removal capacity o Cr'iterion 7 -
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| Safety-grade requirements for MFW and AFW isolation I
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| valves.
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| Arkansas Nuclear One Unit 1 is af Babcock and Wilcox-designed, two-loop, 25684 eft plant.
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| i In the event of a MSIB, the following systems actuate to provide ne:essary protection:
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| .f'
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| TER-C5506-122 o The engineered safeguards actuation system (ESAS) initiates the high pressure and lo.# presr.ure injection systems on receipt of the
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| : p. following:
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| : a. two out of three (2/3) low reactor coolant pressure signals (1500 psig)
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| : b. 2/3 high reactor building pressure signals (4 psig) o The reactor protection system (RPS) trips the reactor to protect aganst fuel damage on recaipt of the followings
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| : a. 2/4 overpower signals (105.5%)
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| : b. 2/4 low reactor coolant system pressure signals (1800 psig)
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| : c. 2/4 high reactor building pressure signals (4 psig) o Reactor building cooling system (4 units at 60x10 6Btu /hr) is actuated on receipt of 2/3 high reactor building alge-13 (4 psig) o Reactor building spray system (2 trains at 120x106 Btu /br) is actuated on receipt of 2/3 high reactor building pressure signals (30 -
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| Psig) o The steam line break instrumentation and control (SIJIIC) is designed to isolate each steam generator by closing the main steam block valve and/or the feedwater isolation valve on each line upon receipt of 2/4 low steam generator pressure signals.
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| The emergency feedwater (EFW) system includes one motor-driven pump (672 gpm) and one turbine-driven pump (105 gpm) which are aligned so that either pump can supply both steam generators. The flow from either pump will ensure that the heat removal capacity exceeds the minimum level required for decay heat removal after a r.SLB. The EPW syften is automatically initiated on the following:
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| Motor-driven Pump o loss of both main feedwater pumps o loss of all four reactor coolant pumps o low steam generator level .
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| Turbine-driven Pumps __
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| o loss of both main feedwater pumps 4
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| bd Franklin Research Center
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| l TER-C5506-122 o loss of all four reactor coolant pumps
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| : p. o low steam generator level o SLBIC signal The SLBIC system is designed to meet safety-grade and IEEE Std 279-1971
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| [8] requirements; the ESAS and RPS are designed to safety-grade and IEEE Std 279-1968 requirements. ,
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| The environmental qualification of safety-related electrical and mechanical components is being reviewed separately by thh NRC and is not within the scope of this review.
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| The review did not determine if the instrumentation that the operator -
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| relies upon to follow the accident and isolate the affected steam generator conforms with the criteria in ANS/ ANSI-4.5-1980 [11] and Regulatory Guide 1.97
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| [12].
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| The Licensee's analysis assumed that the SIBIC fails to isolate main feedwater and that no operator action is taken to isolate main feedwater or emergency feedwater. The ICS is then assumed to control both main and emergency feedwater flow to maintain a minimum level in the steam generators. The Licensee's analysis determined that the containment design pressure of 59 psig 1
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| would not be exceeded for 3 hours and 45 minutes. This is ample time for the operator to analyre the accident and take the appropriate actions to prevent exceeding the design pressure.
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| An analysis of the EFW pumps determined that the pumps would remain operable without operator action, when subject to runout flow conditions during a MSIB accident.
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| On October 15, 1980 (16] APL provided details of a safety-grade,
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| , automatically initiated, AFW system designed to feed only the unaffected steam generator in the event of a MSIa. The Licensee committed to install this system during the Arkansas Itaclear One Unit 1 fif th refueling outage currently scheduled for January 1983. The final design of this AFW initiation system was provided in Reference 17.
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| lJ'UUU Frenidin Research Center A Osummen of The Fieseen buuuuse
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| .- - . _ - . . . . - . .... = - - - . ~ - . - . .-- -. .
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| . ~ . . . . . .
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| TER-C5506-122 A review of References 16 and 17 determined that the proposed emergency -
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| fey initiation and control system (EFIC) is an instrumentation system j designed to provide the followings o initiation of emergency feedwater (EN) o control of EN at appropriate setpoints (approximately 3, 20, and 31.5 feet) o level rate control when required to minimize overcooling o isolation of the main steam and main feedwater lines of a depressurized steam generator o the selection of the appropriate steam generator (s) under conditions of steamline break or main feedwater or emergency feedwater line break downstream of the last check valve o termination of main feedwater to a steam generator on approach to overfill conditions o termination of EN to a steam generator on approach to overfill conditions o control of atmospheric dump valves to predetermined setpoint.
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| The EFIC logic issues a call for EN auto-initiation when:
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| o all four reactor coolant pumps are tripped o both main feedwater pumps are tripped o the level of either steam generator is low .
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| o either steam generator pressure is low I o flux to MN flow ratio trip is present.
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| Other functions of the htFIC logic area o Issues iE' call for steam generator (SG) A main feedwater and main steamline isolation when SG A pressure is low o Issues a call for SG B main feedwater and main steamline isolation when SG B pressure is low o Signals approach to SG A overfill when SG A level exceeds a high level setpoint
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| ! renidin Research Center A Ohnmen af The hand $n bushet
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| TER-C5506-122 o Signals approach to SG B overfill when SG B level exceeds a high level setpoint P.
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| o Provides for manually initiated individual shutdown bypassing of reactor coolant pumps, main feedwater pumps, and SG pressure initiation of EN as a function of permissive conditions. The bypass (es) is automatically removed when the permissive condition terminates.
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| o Provides for maintenance bypassing cf an EFIC initiate logic.
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| In the event of a steam line break or feed line break, he EFIC system is designed to isolate the steam and feedwater lines and to provide emergency
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| (-
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| feedwater to the intact steam generator. The system is designed so that no single active failure will either prevent emergency feedwater from being supplied to the intact steam generator or allow emergency feedwater to be supplied to the broken steam generator.
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| 'ib meet the requirements for steam line or feed line break protection, the following design was implemented:
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| o Isolation - low steam pressure (below approximately 600 psig) in either SG will isolate the main steamlines and main feedwater lirie to the affected SG.
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| o SG selection
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| : a. If both SGC are above 600 psig, E N is supplied to both SGs.
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| : b. If one SG is below 600 psig, EFW is supplied to the other SG.
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| : c. If both SGs,are below 600 psig, but the pressure difference between the' two SGs exceeds a fixed setpoint (approximately 100 psig), EN 'is supplied only to the SG with the higher pressure,
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| : d. If both SGs are below 600 psig and the pressure difference is less than the fixed,setpoint, E N is supplied to both SGs.
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| The EFIC system was designed to safety-grade and IEEE Std 279-1971 requirements.
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| ?
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| 3.1.3 Conclusion The Licensee's responses [3, 4, 5} and the Arkansas Nuclear One Unit 1 FSAR [6] adequately address the concerns of Item 1 of IE Bulletin 80-04. The containment pressure response analysis and the design of the mitigating d Franklin Research Center A Dnemen of The hensen suunne
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| . , ~ . . - , - , , .. . . . - . - . _ - . - - - - . . . . . _ . . ..- . -
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| l..
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| TER-C5506-122 systems satisfy the NRC acceptance criteria. The proposed EFIC system will proyide safety-grade protection against a MSIa and eliminate the need for operator action to isolate emergency feedwater flow to the ruptured steam
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| ! generator. The EFW pumps will remain operable when subject to runout flow conditions during a MSIB. -
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| n.
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| 3.2 REVIEN OF REACTIVITY INCREASE ANALYSIS The requirement from IE Bulletin 80-04, Item 2, is as follows:
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| " Review your analysis of the reactivity increase which results from a main steam line break inside or outside containment. This review should consider the reactor cooldown rate and the potential for the reactor to return-to-power with the most reactive control rod in the fully withdrawn position. If your previous analysis did not consider all potential water sources (such as those listed in 1 above) and if the reactivity increase is greater than previous analysis indicated, the report of this review should includes
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| : a. The boundary conditions for the analysis, e.g., the end of life shutdown margin, the moderator temperature coefficient, power level, and the net effect of the associated steam generator water inventory on the reactor system cooling, etc.,
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| : b. 'The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution to the reactor coolant system,
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| : c. The effect of extended water supply to the affected steam generator on the core criticality and return-to-power,
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| : d. The hot channel factors corresponding to the most reactive rod in the fully withdrawn position at the end of life, and the. Minimum
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| . Departure from Nucleate Boiling Ratio (MDNBR) va3ues'for the analyzed transient."
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| 3.2.1 Summary of Licensee Statements and Conclusions v
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| In regard to the reactivity increase resulting from a MSLB with continued feedwater addition, the Licensee stated (3} :
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| "'Rie steam line break accident has been analyzed in the Unit 1 FSAR in
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| . Section 14.2.2 considering no operator action. In this analysis, the affected OTSG is assumed to blow dry after the rupture at which time the minimum level control opens feedwater valves such that the OTSG maintains low-level. Assuming a minimum tripped rod worth with the maximum rod ranklin Research Center A Dhaman of The Fearmen buenae
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| , , _ . . _ _ . - . - . . _ _ _ - _ _ _ _ _ _ . . _ _ . -. - _ _ _ , _ - - - . _ . - . _ . . , _ < _ . _ , . _ _ . - - . - - - . , . _ . _ - _ . . - - - ~ _ _ . - - - - _ _
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| TER-C5506-122 stuck out, the reactor will return to a maximum neutron power level of 2.64 at 44.5 seconds and return to subcriticality at 47.5 seconds. With
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| : p. the low level control valves maintaining a 30-inch minimum downconer level in the affected OTSG, the average coolant temperature will remain below 475 degrees F until feedwater isolation on the affected OTSG is achieved."
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| 3.2.2 Evaluation The Licensee's analysis of the core reactivity increase resulting from a MSLB with continued feedwater addition was reviewed in order to evaluatg whether the follo* ring acceptance cr'iteria were mets o Criterion 1.c - Ability to detect and isolate the damaged steam generator o Criterion 1.d - Oianges in core reactivity increase ,
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| o Criterion 3 -
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| Analysis assumptions.
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| The FSAR analysis of the reactivity increase resulting from a MSLB and Reference 3 were reviewed. From that review, it was determined that the analysis is conservative in its assumptions and that the assumptions are in accordance with those in Acceptance Criterion 3.
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| In the worst case MST.3, which assumes full power conditions, a double-ended rupture at the steam generator exit, and no operator action, a peak power of 106%
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| occurs at 6 seconds, at which time a high flux reactor trip occurs, inserting the control rods. After the reactor trip, the core returns to criticality at 43.5 seconds, reaches a maximum neutron power of 2.6% at 44.5 seconds, and returns to subcriticality at 47.5 seconds. The predicted return-to-power does not result in a violation of the specified acceptable fuel design limits.
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| 3.2.3 Conclusion ,
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| For the current plant design, the Licensee's responses ;[3, 4, 5] and FSAR adequately address the concerns of Item 2 of IE Bulletin 80-D4. All potential sources of water were identified, and although a reactor retiarn-to-power is predicted, there is no violation of the specified acceptable' fuel desian limits, and the FSAR analysis of the reactivity increase resulting from a MSLB I ,
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| remains valid. '
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| ULbd Frenidin Research Center l A Onamen af The Fraseen tunnae
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| l l
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| l TER-C5506-122 I
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| .3 ' REVIEW OF CDRRECTIVE ACTIONS l y The' requirement from IE Bulletin 80-04, Item 3, is as follows:
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| "If the potential for containment overpressure exists or the reactor return-to-power response worsens, provide a proposed corrective action and a schedule for completion of the corrective action. If the unit is operating, provide a description of any interim action that will be taken until the proposed corrective action is completed."
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| 3.3.1 Sumnary of Licensee Statements and conclusions The Licensee stated [3]:
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| ~
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| "Although the potential exists on ANO-1 for reactor building over pressurization, this event will not take place until 3 hours and 45 minutes into the steam line break accident. It is our position that -
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| there is more than sufficient time for the operator to isolate the affected orSG and terminate the event. Thus no corrective action is proposed."
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| 3.3.2 Evaluation and Conclusion The Licensee's analysis determined that neither a containment overpressu-rization nor a reactor return-to-rower with a resultar.t violation of the specified acceptable fuel design limits m uld occur from a MSIa. Therefore, it is concluded that no further ca. tion regarding IE Bulletin 80-04 is required of APL for Arkansas Nuclear One Unit 1.
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| l l
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| s
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| ! _17 l 000)A omFrenidina.an= rm n Research C. enter
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| : 1. - -
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| . < ~~
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| TER-C5506-122
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| : 4. CONCLUSIONS 9-With respect to Arkansas Nuclear One Unit 1, conclusions regarding Arkansas Power and Light Company's response to IE Bulletin 80-04 are as follows:
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| o Sere is no potential for containment overpressurization resulting from a main steam ling break with continued feedwater addition.
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| o h e emergency feedwater pumps will remain operable when subject to effects of runout flow and therefore can be expected to carry out their intended function during the MSLB event.
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| o All potential water sources were identified and, although a reactor return-to-power is predicted, there is no violation of the specified acceptable fuel design limits. Merefore, the Final Safety Analysis Report MSI2 reactivity increase analysis remains valid.
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| o No further action regarding IE Bulletin 80-04 is required.
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| i Uuuu Franklin Research Center A Obamen af The Fm humane
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| j . ,
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| e TER-C5506-122
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| : 5. REFERENCES P-
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| : 1. " Analysis of a PWR Main Steam Line Break with Continued Feedwater Addition" NBC Office of Inspection and Enforcement, February 8,1980 IE Bulletin'80-04
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| : 2. "Overpressurization of the Containment of a PWR Plant after a Main Line Steam Break" NBC Office of Inspection and Enforcement, October 1,1979 IE Information Notice 79-24
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| : 3. D. C. Trimble (APL)
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| .Intter to K. V. Seyfrit (NRC, Region IV)
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| | |
| ==Subject:==
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| IE Bulletin 80-04 May 27, 1980
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| : 4. D. C. Trimble (APL)
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| Intter to K. V. Seyfrit (NRC, Region IV)
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| | |
| ==Subject:==
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| IE Bulletin 80-04, Emergency Feedwater Pump Analysis July 9, 1980
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| : 5. J. P. Marshall (APL)
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| * Intter to J. F. Stolz (NRC, ORB No. 4)
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| | |
| ==Subject:==
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| Additional Information Regarding IE Bulletin 80-04 July 30, 1982
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| : 6. Arkansas Nuclear One Unit 1 Final Safety Analysis Report, through Amendwnt 49 Arkansas Power & Light Company, September 1bi3
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| : 7. Technical Evaluation Report "PWR Main Steam Line Break with Continued Feedwater Addition - Review of Acceptance Criteria" Franklin Research Center, November 17, 1981 TER-C5506-119
| |
| : 8. " Criteria for Protection *
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| * Systens for Nuclear Power Generating]
| |
| Stations" Institute of ' Electrical and Electronics Engineers, New York, NY,1971
| |
| * IEEE Std 279-1971 i t
| |
| l
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| : 9. Standard Review Plan, Section 4.2 ,
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| " Fuel System Design", !
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| NRC, July 1981 i NUREG-0800 I
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| NO Franklin Research Center l A Dhimon of The Frumen kuunne
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| ~o_..,- .. . . _ , _ _ . . . . _ _ . _ - _ . . ~ . . . _ _ _ _ . . . _ _ _ _ . . - . . . .
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| TER-C5506-122
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| : 10. Standard Review Plan, Section 15.1.5
| |
| " Steam System Piping Failures Inside and Outside of Containment (PWR)"
| |
| NRC, July 1981
| |
| : p. NUREG-0800
| |
| : 11. " Criteria for Accident Monitoring Functions in Light-Water-cooled Reactors" ,
| |
| American Nuclear Society, Hinsdale, IL, December 1980 ANS/ ANSI-4.5-1980
| |
| : 12. " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During 'and .Pollowing an Accident" Rev. 2 ..
| |
| NBC, December 1980 g.
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| Regulatory Guide 1.97
| |
| : 13. " Single Failure Criteria for PWR Fluid Systems" American Nuclear Society, Hinsdale, IL, June 1976 ANS-51.7/N658-1976 .
| |
| : 14. " Quality Group Classifications and Standards for Water , Steam , and Radioactive-Waste-Containing Components of Nuclear Power Plants" Rev. 3 NBC, February 1976 Regulatory Guide 1.26
| |
| : 15. " Interim Staff Position on Environmental Qualification of Safety-P. elated Electrical Equipment" Rev. 1 NRC, July 1981 NUREG-0588
| |
| : 16. D. C. Trimble (APL)
| |
| Intter to R. W. Reid (NRR)
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| | |
| ==Subject:==
| |
| Submittal of EFW Upgrade' Proposal D6 sign Information October 15, 1980
| |
| : 17. D. C. Trimble (APL)
| |
| Intter to T. M. Novak (NRR ORB)
| |
| | |
| ==Subject:==
| |
| Final Submittal of EFW Upgrade Design Information December 1, 1981*
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| * I .
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| N
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| ~' -
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| 1EL R _,c A Ohemen af The Frumen buumme
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| . , - - _ . - - ., . - - . _ , , . , _ _ _ , _, -.~ . ..__ ._}}
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