ML20067D532

PWR Main Steam Line Break W/Continued Feedwater Addition, Re Ginna Nuclear Power Plant, Technical Evaluation Rept

ML20067D532
Person / Time
Site:	Ginna
Issue date:	12/15/1982
From:	Herrick R, Vosbury F FRANKLIN INSTITUTE
To:	Peter Hearn NRC
Shared Package
ML17256A490	List: ML17256A489 ML17256A491 ML20067D532
References
CON-NRC-03-81-130, CON-NRC-3-81-130, REF-SSINS-6820, TASK-04-03, TASK-06-02.D, TASK-4-3, TASK-6-2.D, TASK-B-69, TASK-OR, TASK-RR IEB-80-04, IEB-80-4, TAC-46835, TER-C5506-146, NUDOCS 8212200054
Download: ML20067D532 (24)

Text

{{#Wiki_filter:. 1 TECHNICAL EVALUATION REPORT PWR MAIN STEAM LINE BREAK.WITH CONTINUED FEEDWATER ADDITION (B-69) ROCHESTER GAS AND ELECTRIC COMPANY ROBERT E. GINNA NUCLEAR POWER PLANT NRC DOCKETNO. 50-244 FRC PROJECT C5506 NRCTAC t'O. 46835 FRC ASSIGNMENT:i NRC CONTRACT NO. NRC-03-81 130 FRCTASK 146 I Preparedby l Franklin Research Center Author: F. W. Vosbury 20th and Race Street Philadelphia, PA 19103 FRC Group Leader: R. C. Herrick 4 Preparedfor Nuclear Regulatory Commission jj Washington, D.C. 20555 Lead NRC En.gineer: P. Hearn t December 15, 1982 l This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the Uni'ed 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, p oduct or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. Prepared by: Reviewed by: Approved by: .S d Ymb W L dr. & flua. Principal AuthC Group Leader de'p'artment D rec r M~~ /3 /#/.FE Date: I U ' N N- / Date: Date: !DHes Been Sene so legt "b

00. Franklin Research Center l

A Division of The Franklin Institute The Benjarnm Franktn Parkway. PMa, Pa. 19103(215) 448-1000 ' '1 % '2.ln 0 00 % N

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.s TER-C5506-14 6 ' t i yI 'I CONTENTS ,l j Section Title ,Page j 1 INTRODUCTION. 1 't l 1.1 Purpose of Review 1 . f 1.2 Generic Background. 1 1.3 Plant-Specific Background 3 j 2 ACCEPTANCE CRITERIA. 4 3 TECHNICAL EVALUATION. 8 3.1 Review of Containment Pressure Response Analysis 8 3.2 Review of Reactivity Increase Analysis. 13 3.3 Review of Corrective Actions 17 19 4 COtCLUSIONS. 5 REFERENCES 20 4 1 1 4 i r. 4 iii A O..d Franklin Research Center 4 e=.on on w rr n,e %. m l

TER-C5506-14 6 FOREWORD This Technical Evaluation Report was prepared by Franklin Research Center l under a contract with the U.S. Nuclear Regulatory Commission (Office of f Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions. The techtsical evaluation was conducted in accordance with criteria established by the NRC. Mr. F. W. Vosbury contributed to the technical preparation of this report through a subcontract with WESTEC Services, Inc. i l' ili l? l l Ag v b'.! Franklin Research Center 4 mma w n. rr.ae www. ' ' ~7 ? ~ T..... ~

I TER-C5506-14 6 ~! l. INTRODUCTION. i y 1.1 PURPOSE OF REVIEW t This Technical Evaluation Report (TER) documents the review of Rochester Gas and Electric Corporation's (RG&E) response to the Nuclear Regulatory e [l Commission's (NBC) IE Bulletin 80-04, " Analysis of a Pressurized Water Reactor Main Steam Line Break with Continued Feedwater Addition" [1], as it pertains e to the Robert E. Ginna Nuclear Power Plant. This evaluation was performed with the following objectives: o to assess the conformr.nce of RG&E's main steam line break (MSLB) analyses with the requirements of IE Bulletin 80-04 .? -l o to assess RG&E's proposed interim and long-range corrective action plans and schedules, if needed, as a result of the MSLB analyses. 1.2 G2NERIC BACKGROUND + t: 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 MSLB. A reanalysis of the containment pressure response following a MSLB 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 experienced the steam line break, containment design pressure would be exceeded in approximately 10 minutes. The long-term blowdown of the water supplied b[ the Am system had not been considered in the earlier analysis. ) on October 1,1979, the foregoing information was provided to all holders of operating licenses and construction permits as IE Information Notice 79-24 [ (2]. Another facility performed an accident analysis review pursuant to receipt of the information in the notice and disevvered that, with offsite j electrical power available, the condensate pumps would feed the affected steam = 1 generator at an excessive rate. This excessive feed was not previously i considered in the plant's analysis of a MSLB accident. 4 l ij . b0 Franklin Research Center .j A Dmsson d The frannbn Insetute ~ Y ~ ~ ~ '~ ~ ^~ ~~ ~~ " ~ ~ ~ ~ ^ '~' '~

l 't TER-CS506-14 6 A third licensee informed the NRC of an error 'in the MSLB analysis for o their plant. During a review of the MSLB 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 80% full open due to an override signal resulting from the low steam generator ~ 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 raturn-to-power response, a condition which is outside the plant design basis. Because of these deficiencies identified in original MSLB accident analyses, the NBC issued IE Bulletin 80-04 on' February 8,1980. This bulletin required all PWRs with cperating licenses and certain near-term PNR operating license applicants to perform the following: 4 4 "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 li auxiliary feedwater system and the impact 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 panps to remain operable after extended operation at runout flow. 2. 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 i 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 include: 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., -i j 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 ,j coolant system,

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j TER-C550 6-14 6 j 'I J c. Se effect of extended water supply to. the affected steam generator on the core criticality and rettir'ii to power, t a d. We hot channel factors corresponding to the most reactive rod in 1 the fully withdrawn pcsition at the end of life, and the Minimum Departure from Nucleate Boiling Ratio (MDNBR) values for the 1 analyzed transient. 3. If the potential for containment overpressure exists or the reactor \\ return-to-power response worseas, provide a proposed corrective i action and a schedule for completion of the corrective action. If the unit is operating, provide a description of any interim action .i that will be taken until the proposed corrective action is completed." '5 1.3 PLANf-SPECIFIC BACKGROUND 1 j Rochester Gas and Electric Corporation responded to IE Bulletin 80-04 in a letter to the NRC dated April 30, 1980 [3]; additional information was obtained from an NRC letter dated Ncvember 3, 1981 [4] and RG&E's response (' dated February 1,1982 [5] concerning the Systematic Evaluation Program (SZP) .] topics VI-2.D, " Mass and Energy Release for Possible Pipe Break Inside Containment," and VI-3, " Containment Pressure and Heat Removal Capability," ] for the Ginna Plant. The information in References 3, 4, and 5 has been- )i evaluated along with pertinent information from the R. E. Ginna Nuclear Power a Plant Final Facility Description and Safety Analysis Report (FFDSAR) [6] to determine the adequacy of the Licensee's compliance with IE Bulletin 80-04. ,1 i 001.) Franklin Research Center 7 A Cansson of The Frannha hsoame

.i* i .k I 4 TER-C5506-14 6 !i 2. ACCEPTANCE CRITERIA .' "e. ? ..j The following criteria against which the Licensee's MSLB respc nse was y evaluated were provided by the NRC [7): ii 1. PRR 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 - 'l containment

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/ a. A discussion of the continuation of flow to the affected steam jj generator, including the impact of runout flow from the AFW j system and the impact of other energy sources, such as continuation of feedwater or condensate flow. AFW system runout ii flow should be determined from the manufacturer's pump curves at Ji no backpressure, unless the system contains reliable anti-runout [I 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 ~ .j 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] a'nd passive devices (e.g., flow -orifices or

j cavitating venturis).~

Lj b. A determination of potential containment overpressure as a result 6, of the impact of runout flow from the AFW system or the impact of j other energy sources such as continuation of feedwater or 2 condensate flow. Where a revised analysis Js submitted or where j reference is made to the existing FSAR analysis, the analysis must show that runout AFW flow was included and that design containment pressure was not exceeded. c. A discussion of the ability to detect and isolate the damaged f steam generator from continued feedwater addition during the MSLB accident. Operator action to isolate AFW flow to the affected 7 steam generator within the first 30 minutes of the start of the J, MSLB should be justified. Where operator action is to be i completed within the first 10 minutes, then the justification should address the indication availtble to the operator and the actions required. Where operator action is required to prevent exceeding a design value, i.e., containment design pressure or ? specified acceptable fuel design limits, then the discussion should include the calculated time when the design value would be N.- exceeded if no operator action were assumed. Where operator actions are to be performed between 10 and 30 minutes after the start of the MSLB, the justification should address the d$ Franklin Research Center A Ce eme Franen inscue . N :. -

i TER-C5506-14 6 I:] indications available to the operator and.the operator actions lj required, noting that for the first 30 minutes, all actions y should be performed from the control room. j d. Where all water sources were not considered in the previous j analysis, an indication should be provided of the core reactivity change which results from the inclusion of additional water 1 sources. A submittal which does not determine the magnitude of H reactivity change from an original analysis is not responsive to O the requirements of IE Bulletin 80-04. !{ 2. If containment overpressure or a worsening of the reactor return-to-j power with a violation of the specified acceptable fuel design limits

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described in Section 4.2 of the Standard Review Plan (9] (i.e., increase in core reactivity) can occur by the licensee's analysis, j the licensee shall provide the following additional information: Il p a. the proposed corrective actions to prevent containment over-1 pressure or the violation of fuel design limits and the schedule d for their completion l$ j b. the interim actions that will.be taken until the proposed corrective action is completed, if the unit is operating. 3p ], 3. The acceptable input assumptions used in the licensee's analysis of b the core reactivity changes during a MSLB'are given in Section 15.1.5 L,' of the Standard Review Plan (10]. The following specific assumptions / should be used unless the analysis shows that a different assumption H is more limiting: q d Assumption II.3.b.: Analysis should be performed to determine the i most conservative assumption with respect to a loss of electrical power. A reactivity analysis should be conducted for a normal j power situation as well as a loss of offsite power scenario, unless the licensee has previously conducted a sensitivity analysis which demonstrates that a particular assumption is more conservative. 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 l} concentration boric acid solution to the reactor coolant system, or any other single active failure affecting the plant response, should be considered. I b t l

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,s TER-C5506-14 6 j Assumption II.3.g.: 'Ihe initial core, flow should be chosen such p that the post-MSLB shutdown margin is i minimized (i.e., maximum initial core flow). The acceptable computer codes for the licensee's analysis of core l? reactivity changes are, by nuclear steam supply system (NSSS) vendor, d the followings CESEC (Combustion Engineering), LOPTRAN (Westing-q house), and TRAP (Babcock & Wilcox). Other computer codes may be j used, provided that these codes have previously been reviewod and 1 found to be acceptable by the NRC staff. If a computer code is used M' which has not been reviewed, the licensee must describe the method j employed to verify the code results in sufficient detail to permit

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the code to be reviewed for acceptability. .r j 4. If the AFW pumps can be damaged by extended operation at runout flow, * ] the licensee's action to preclude damage should be reviewed for >j technical merit. Any active features should satisfy the requirements W of IEEE Std 279-1971. Where no corrective action has been proposed, Q this should be indic&ted to the NRC for further action and resolution. 'id 5. Modifications to electrical instrumentation and controls needed to p detect and initiate isolation of the affected steam 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 relies upon to follow the accident and to datermine isolation of th'e affected. steam ,.j generator and feedwater sources should conform to the criteria y contained in ANS/ ANSI-4.5-1980, " Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors" [11], and the regulatory positions in Regulatory Guide 1.97, Rev. 2, " Instrumentation for j Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" [12]. 6. AN 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. 'j 7. The safety-grade requirements (redundancy, seismic and environmental qualifications, etc.) of the equipment that isolates the main '] feedwater (MFW) and AN systems from the affected steam generator j should be specified. The modifications of equipment that is relied 'i upon to isolate the MFW and AW systems from the affected steam generator should satisfy the following criteria to be considered j safety-grade: 9 4

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l bd Frankjin Research Center i A Onasson of The Franun Insetute .-.~

I .E !i .] TER-C5506-14 6 i o Redundancy and power source requirements: n e. isolation valves 1 should be designed to accommodate a' single failure. A failure-d modes-and-effects analysis should demonstrate that the system is e capable of withstanding a single failure without loss of function. W e single failure analysis should be conducted in .s accordance with the appropriate rules of application of .f ANS-51.7/N658-1976, " Single Failure Criteria for PWR Fluid j Systems" [13]. g ad o Seismic requirements: The isolation valves should be designed to .3 Category I as recommended in Regulatory Guide 1.26 [14]. e .:) o Environmental qualification: The isolation valves should satisfy the requirements of NUREG-0588, Rev.1, " Interim Staff Position .l on Environmental Qualification of Safety-Related Electrical J Equipment" [15]. .i., f; o Quality standards: The isolation valves should satisfy Group B i quality standards as recommended in Regulatory Guide 1.26 or j similar quality standards from the plant's licensing bases. 1 i j

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.l .i 1 ' 9 TER-C5506-14 6 4 l .3 3. TECHNICAL EVALUATION

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1 i The scope of work included the followings 'i U l. Review the Licensee's response to IE Bulletin 80-04 against th, f acceptance criteria. '.~ lj 2. a. Evaluate the Licensee's MSLB analyses for the potential of overpressurizing the containment ahd with respect to the core l reactivity increase due to the effect of continued feedwater flow. 4 i b. Evaluate the Licensee's proposed corrective actions and schedule. for implementation if the findings of Task 2a indicate that a ' l potential exists for overpressurizing the containment or I worsening the reactor return-to-power in the event of a MSta accident. t

3., Prepare a TER for each plant based on the evaluation of the information presented for Tasks 1 and 2 above.

This report constitutes a TER in satisfaction of Task 3. Sections 3.1 through 3.3 of this report state the requirements of IE Bulletin 80-04 by subsection, summarize the Licensee's" statements and conclusiens regarding 2 these requirements, and present a discussion of the Licensee',s evaluation i followed by conclusions and recommendations. 3.1 REVIEW OF CONTAINMENT PRESSURE RESPONSE ANALYSIS ^ The requirement from IE Bulletin 80-04, Item 1, is as follows: i " Review the containment pressure response analysis to determine if the i potential for containment overpressure for a main steam line break inside containment included the impact of runout flow from the auxiliary 4 feedwater system and the impact 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." i 3.1.1 Summary of Licensee Statements and conclusions l l', In regard to the review of the containment pressure response analysis, the i Licensee stated (3]; i lv .. 2 O kFranklin Research Center .q A o, .e m n. e na. l -

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o TER-C5506-14 6 d '"Although the Ginna post-steamline break containment pressure analysis. in the FSAR did not include the effects of auxiliary feedwater flow to the j affected steam generator, it is important to rece$nize that the evaluation j also did not include the benefits of passive and active heat sinks inside lj containment. Continued feedwater/ condensate addition to the steam l generator will not occur, since the safety injection signal (generated by g a variety of proccss parameters, including high steam line flow, high 'jj containment pressure, and low pressurizer pressure) will close the ] feed 1. Iter control valves and stop the feedwater pumps. The addition of ~ q maximum auxiliary feedwater flow to the broken steam generator will l5 ' eventually require operator action to 1) realign flow to the intact i generator, 2) terminate auxiliary feedwater flow to the broken generators. Positive information is,available to the operator to determine which is 3 the affected steam gencrator. Steam generator level instrumentation is { located inside containment and steam generator pressure is located outside

j containment where it would not be affected by the accident environment

l inside containment.

It is expected that, through proper training and by use of the emergency procedures, the operators will be capable of quickly l.j recognizing the steam line break, and will perform the proper operations. M There is substantial time available for the operator to perform the two lj safety functions noted above. The SEP Safe Shutdown review concluded i}* following their site visit in June 1978 that one steam generator would not boil dry for over thirty minutes. Thus there is substantial time to I align flow to the intact steam generator. The termination of auxiliary feedwater flow to the affe'cted steam generator, under the pessimistic iI circumstances, would require more rapid action (but still easily within i the capability of the operators) to maintain containment pressure below ] design pressure. The analysis presented in Attachment I concludes that,

j assuming minimum safeguards for containment cooling, auxiliary feedwater flow would have to be terminated in about 26 minutes. With maximum safeguards, this time would be extended to about 44 minutes."

In regard to considering AMf runout flow in the containment analysis, the l} Licensee stated [3]: "A potential single f ailure of the flow controller to control flow to 200 gpm is not considered a worst-case single failure in terms of net energy addition to the containment, since the operation of all containment f cooling safeguards (vs. the minimum safeguards assumed in this evalua-tion) would result in a substantial increase in energy removal from i ' containment." In regard to the ability of the Aaf pumps to remain operable during a 'l MSLB, the Licensee stated [3]: i "There is no need to consider the operation of the auxiliary feedwater ij pumps at runout flow. The turbine-driven pumps are controlled by a .i 1 E1 !,1 P+ l L4 N'h) Franklin Research Center A oms.on of The Frerman insonne t .. - - - - ~ -

~ t TER-C550 6-14 6 i governor, and will not exceed about 400 gpm. The mo_ tor driven pump flow is controlled by the AFW control valves, which Egceive an automatic 3 1 throttle signal to 200 gpm from their flow controllers." ij 3.1.2 Evaluation. l The Licensee's submittals (3, 4, 5] concerning the containment pressure response following a MSLB and applicable sections of the Ginna R.iclear Power Plant FFDSAR [6] were reviewed in order to evaluate whether the following j j portions of the acceptance criteria were met: 3 o Criterion 1.a - Continuation of flow to the affected steam' generator 9, m j o Criterion 1.b - Potential for containment overpressure o Criterion 1.c - Ability to detect and isolate the damaged steam j [j generator f 'lj o Criterion 4 - Potential for AN pump damage

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o Criterion 5 - Design of steam and feedwater isolation system 'l 4 o criterion 6 - Decay heat removal capacity 1j' o Oriterion 7 - Safety-grade requirements fort MFW and AFW isolacion i valves. qj The Ginna Nuclear Power Plant is a Westinghouse-designed, 2-loop,1520-MWt 's plant. q [] The following systems provide the necessary protection against a steam 1 >; pipe ruptures I ,o Safety injection system actuation on: a. two out of three pressurizer low pressure signals i Li b. two out of three low pressure signals in any steam line !i Ll c. two out of three high containment pressure signals a d. high steam flow. l3 !) I; o The overpower reactor trips (nuclear flux and differential ,i temperature) and the reactor trip occurring upon actuation of the d,. safety injection system. p. j A b') Franklin Research Center A Diesen of The Franakn insature

s 1 a TER-C550 6-14 6 o Redundant isolation of the MW lines. A, safety, injection signal will ~, rapidly close all MN control valves, trip the main MFW pumps, and j close the MW pump discharge valves. In addition, normal control j action will also signal the MN valves to close. o Trip of the fast-acting steam line isolation valves (designed to close in less than 5 seconds with no flow) on: O2 a. one out of two high steam flow signals in a steam line in ~ coincidence with any safety injection signal. (Dual setpoints are provided, with the lower setpoint used in coincidence with two out of four indications of low reactor coolant average temperature) 1 j b. two out of three high containment pressure signals. 4 Each steam line has a fast-closing isolation valve and a check valve.

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j These four valves prevent blowdown of more than one steam generator for any j break location even if one valve fails to close. For a break upstream of the j isolation valve in one line, closure of either the check valve in that line or N the isolation valve in the other line will prevent blowdown of the other steam ij generator. 1 U The AN sys, tem consists of two redundant systems, the main AN system and d the standby A W system. The main Am system includes one turbine-driven pump ,o (400 gpm) and two motor-driven pumps (200 gym each) and is automatically actuated on the receipt of the following signals. d 1 Motor-driven A N pumps a 9 o low-low steam generator level (two out of three channels on either i st.eam generator) o trip of both MFW pumps o safety injection signal i Turbine-driven A W pump .;.) ] o low-low steam generator level (two out of three channels on both steam generators) ii q o loss of voltage on both 4-kV buses. fj () 1 .i _11-d JNi! Franklin Resea.rch Center 4 won a n. n. mau. -{ I L _j ~

TER-C5506-145 i 1 j The main moton:-driven A N pumps are normally aligned so that each pump j supplies one steam generator. The turbine-driven A N punp supplies'both steam tj generators. i ] The motor-driven AN pumps are protected against runout conditions by l flow-limiting valves; the turbine-driven pump flow is limited by its governor 1 so it will not achieve runout conditions. Because the main A N system pumps 4 are all located in the same room and subject to a high energy line break 2 e (HELB), a standby AN system was installed independent of the main AN l system. The standby AN system includes two motor-driven pumps (200 gpm each) ] and is manually initiated. Each standby AN pump supplies one steam ~ [! generator. The main A N system complies with safety-grade and IEEE Std !1 1 . 279-1971 [8] requirements. The compliance of the remainder of the above systems with safety-grade and IEEE Std 279-1971 requirements was not reviewed. ,j The environmental qualification of safety-related electrical and mechani- .4] cal. components is being reviewed separately by the NRC and is not within the scope of this review. ~ 1 i The review did not determine whether the instrumentation that the lj operator relies upon to follow the accident and isolate the affected steam s1 generator conforms with the criteria in ANS/ ANSI-4.5-1980 (11] and Regulatory qb Guide 1.97 [12]. il d As part of the Systematic Evaluation Program (SEP), the NRC performed an 1] evaluation of SEP topics VI-2.D, " Mass and Energy.blease for Possible Pipe ,1 Break Inside Containment," and VI-3, " Containment Pressure and Heat Removal Capability," [4] for the Ginna plant. As part of this evaluation, the NRC j performed an analysis to determine the containment pressure cnd temperature 'f response. The analysis determined that the highest peak pressure and f temperature were produced assuming hot standby conditions (zero power), (! failure of a containment spray pump, and AN flow of 200 gpm. This worst-case jj accident produced a calculated peak containment pressure of 71.1 psig at 91 j seconds. The containment design pressure is 60 psig; thus the analysis !j determined the containment design pressure was exceeded. In response to the .s draf t version of the NRC evaluation of SEP topics VI-2.D and VI-3, RG&E 1 i .I j 4 _w. L' OM Franklin Research Center j 4 camn a n. n.n.en m.us. ~ l_, _

4:l ' Lll TER-C5506-14 6 ti l4 identified several conservatisms in the analysis for the-MSLB. RG&E performed

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a reevaluation of the containment conditions following a MSLB by reconstruc- ] ting the worst-casa transient produced by the NRC analysis, evaluating the d assumptions used, and then removing some of the conservatism to produce a more q fi reasonable result. The RG&E reanalysis determined that for the worst case d (hot standby and failure of containment spray pump), a peak containment pressure of 57.7 psig occurred at 128.6 seconds. The mass normally associated ]1 with auxiliary feedwater was not included in the analysis; the effect of.

j auxiliary feedwater on peak containment pressure would be negligible since the

{I mass added during the time frame of interest is a very small fracton of the / secondary side inventory (less than 1%). From the Licensee's response [5], /j the NRC drew the following conclusion in Appendix B of the evaluation report I '4:j on SEP topics VI-2.D and VI-3 [16]: l

j "The calculated peak containment pressure for a MSLB acci3ent is less j

than the containment design pressure." r; y 3.1.3 ' Conclusion The Licensee's responses [3, 5] and the NRC Evaluation Reports [4,16] Ik adequately address the concerns of ' Item 1 of IE Bulletin 80-04. The containment pressure response analysis and the design of the mitigating 7 systems satisfy the NRC's acceptance criteria. Regarding Item 1,'it is q.

1 concluded that there is no potential for containment overpressurization

!.1 N. resulting from a MSLB with continued feedwater addition. In addition, since ] the AFW pumps are protected from experiencing runout flow conditions, the Ial pumps will be able to carry out their intended function without damage. i N lj I i 3.2 REVIEW OF REACTIVITY INCREASE ANALYSIS d I;I The requirement from IE Bulletin 80-04, Item 2, is as follows: 4 .j " Review your analysis of the reactivity increase which results from a

j main steam line break inside or outside containment. This review should consider the reactor cooldown rate and the potential for the reactor to d

return-to-power with the most reactive control rod in the fully withdrawn 71 position. If your previous analysis did not consider all potential water ..N { j U'b'j Franklin Research Center A Onnseon of The Franan insatute d 1 ~

f TER-C5506-14 6 .i

3 sources (such as those listed in 1 above) and if the reactivity increase I

is greater than previous analysis indicated the report of this review j should include: j a. The boundary conditions for the analysis, e.g., the end of life -) shutdown margin, the moderator temperature coefficient, power level -]J and the net effect of the associated steam generator water ' inventory on the reactor system cooling, etc., i 3 b. The most restrictive single active failure in the safety I injection system and the effect of that failure on delaying the

j delivery of high concentration boric acid solution to the reactor coolant system, a

.) c. The effect of extended water supply to the affected steam generator ~j on the core criticality and return-to-power,

a

'j d. The hot channel factors corresponding to the most reactive rod in the (j fully withdrawn position at the end of life, and the Minimum Departure from Nucleate Boiling Ratio (MDNBR) values for the analyzed d transient." 1 s ,I i 3.2.1 Summary of Licensee Statements and conclusions LA d.4 In regard to the reactivity increase resulting from a MSLB with continued y feedwater addition, the Licensee stated [3]: 'j " Westinghouse Electric Corporation performed the original steam break analysis for Ginna as reported in the FSAR and a reanalysis submitted to the NBC in September 1975. Westinghouse has reviewed the assumptions ,.m ] made for main and auxiliary feedwater flow as they apply to licensing )) basis steam line break transients. Several of the relevant assumptions j used in all core transient analyses follow, and are further explained in the Ginna FSAR. Ej 1. The reactor is assumed initially to be at hot shutdcwn conditions, at i the minimum allowable shutdown margin.

t) 2.

For the Condition IV breaks, i.e., double-ended rupture of a main -j steara pipe, full main fee <' ater is assumed from the beginning of the '7 transient at a very conservative cold temperature. 3 3. All auxiliary feedwater pumps are initially assumed to be operating, d in addition to the main feedwater. The flow is equivalent to the g rated flow of all pumps at the steam generator design pressure. R j] 4. Feedwater is assumed to continue at its initial flow rate until feedwater isolation is complete, approximately 10 seconds after the l1; I ' nklin Research Center j A Cms.on of The Franun :nsotute ~

-? TER-C550 6-14 6 i .1 break occurs, while auxiliary feedwater isi assumed' to continue at its j initial flow rate. N l 5. Main feedwater flow is completely terminated following feedwater i isolation. 'i .j Based on the manner in which the analysis is performed for Westinghouse Jj plants, the core transient results are very insensitive to auxiliary d feedwater flow. The first minute of the transient la dominated entirely y by the steam flow contribution to primary-secondary heat transfer, which j, is the forcing function for both the reactivity and thermal-hydraulic .j transients in the core. The effect of auxiliary feedwater runout (or ] failure of runout protection where applicable) is minimal. Greater f." feedwater flow during the large steamline breaks serves to reduce

1 secondary pressures, accelerating the automatic safeguards actions, i.e.

j steam line isolation, feedwater isolation and safety injection. The j assumptions described above are therefore appropriate and conservative q.j for the short-term aspect of the steamline break transient. pj The auxiliary feedwater flow becomes a dominant factor in determining the -l duration and magnitude of the steam flow trar.sient during later stages in .} the transient. However, the limiting portion of the transient occurs. -i during the first minute, both due to higher steam flows inherently .. ] present early in the transient and due to the introduction of boron to .] the core via the safety injection system. li 'l In conclusion, Westinghouse has evaluated the effect of runout auxilierf feedwater flows in the core transient for steamline break, and based on U this evaluation, has determined that the assumptions presently made are I appropriate for use as a licensing basis. 'Ihe concerns outlined in the introduction to IE Bulletin 80-04 relative to,1) limiting core ] conditions occuring during portions of the transient where auxiliary feedwater flow is a relevant contributor to plant cooldown; and 2) j incomplete isolation of main feedwater flow, are not representative of )d the Westinghouse NSSS designs and associated Balance of Plant requirements. i 4j The most limiting steam line break determined by Westinghouse was analyzed j by Exxon Nuclear Co., Inc. and presented in XN-NF-77-40 Supplement 1, " Plant Transient Analysis for the R. E. Ginna Unit 1 Nuclear Power Plant," -j March 1980. This transient occurs at hot zero power with outside power available and the break occurring at the exit of the steam generator. The Exxon analysis does not specifically account for auxiliary feedwater. I However, the Steam Generator heat transfer model, using constant heat 4 transfer coefficients, continues to calculate heat transfer from the i primary to the secondary side after the broken steam generator has been d estimated to be empty. If auxiliary flow was specifically accounted for, -1 its effect would be negligible during the initial portion of the transient and would have minimal effect during later portions of the i '1,. 4 NJ Franklin Research Center a cws.an er ne rmon mme \\ ~ ~

j i ~ ,j TER-C5506-14 6 !$j transient since by the time the broken steam generator empties, the total system reactivity is negative and core power.,is decreasing. The ,l additional reactivity addition associated with the slight cooldown due to d runout flow is more than negated by the boron reactivity inserted by d safety injection. Therefore, the severity of the transient is not j increased." 1:: .1 .j 3.2.2 Evaluation ') j,,. The Licensee's analysis of the core reactivity increase resulting from a ..( d. MSLB with continued feedwater addition was reviewed in order to evaluate ij] whether the following acceptance criteria were met: 1

,j o Criterion 1.c - Ability to detect and isolate the damaged steam j

generator N 4 o criterion 1.d - changes in core reactivity increase i) o Criterion 3 Analysis assumptions. .4 i! The Exxon Nuclear analysis [17] of the reactivity increase resulting from .] ] a MSLB and Reference 3 were reviewed. From that review, it was determined M 'that the analysis is conservative in its assumptions and that the assumptions are in accordance with those in Acceptance Criterion 3, with the exception l/ 4 that runout AN flow was not considered. ]1 In the worst-case MSLB, which assumes no load conditions, a double-ended E rupture at the steam generator exit, with offsite power available, a peak 1q reactivity of about +0.004, occurs at 20 seconds which produces a maximum core ] power of 39% at 91.4 seconds. At this time, 20,000 ppm boron solution reaches ] the core, rapidly shutting down the reactor. The calculated return-to-power i does not result in a violation of the specified acceptable fuel design limits. I The Licensee's conclusion that the core transient for the MSLB is f insensitive to AFW flow is valid for the following reasons: 1 o Early in the transient, the primary-to-secondary heat transfer rate (from the blowdown of the initial steam generator mass) is several ,j orderg of magnitude greater than that contributed by the additional AFW flow due to runout. I$ 1'Ii l1 t ?? l, I b ' ddu Franklin Research Center / A o-won at ne Frenen in.come ~~

i 2 1 1 i TER-C5506-14 6 l o Later in the transient (when the majority of the initial mass has 'j blown down), AFW flow becomes a dominant factor in determining the cj magnitude and duration of the transient. .I o The limiting core conditionc will occur within the first minute due to j the initial high cooldown rate contributing to the reactivity addition j which is terminated by the introduction of 20,000 ppm boron solution et into the care region. j4 Since the limiting ccra conditions occur before the AFW flow beccmes a ] major contributing factor, it can be concluded that the core transient is ins,ensitive to the contribution of AFW flow, and therefore the assumptions of 3 the FSAR analysis remain valid. d 3d 3.2.3 Conclusion Il 'i The Licensee's response and reactivity response analysis (17] adequately r} l address the concerns of Item 2 of IE Bulletin 80-04. All potential sources of ) "1 water were identified, and although a reactor return-to-power is predicted, 4 there is no violation of the specified acceptable fuel design limits. Therefore, the reactivity increase analysis [17] of the reactivity increase resulting from a MSLB remains valid. t,y 3.3 REVIEW OF CORRECTIVE ACTIONS The requirement from IE Bulletin 80-04, Item 3, is as follows: in j "If the pctential for containment overpressure exists or the reactor 'y 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 Lj until the proposed corrective action is completed." i ,j 3.3.1 Summary of Licensee Statements and Conclusions e:: L.! The Licensee stated [3]: l} "Since neither the potential for containment overpressurization nor the e, reactor-return-to-power response worsens no corrective action is i required." 1 5 . ?[. 'i i t Udd Franklin Research Center a o~, on or n. r,.wn m.ou. y . - - *? . L _._, 1 ^7

4-6 4 -i i j TER-C5506-14 6 4. 3.3.2 Evaluat. ion and Conclusion i j The Licensee's analysis determined that neither a containment over-l pressurization nor a reactor return-to-power with a violation of the specified 1 'j acceptable fuel design limits would occur from a MSLB. Therefore, it is I concluded that no further action regarding IE Bulletin 80-04 is required of RG&E for the Ginna Nuclear Power Plant. 1 i 1 1 !a b, .j .'j .-) J1 s n a]

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j I i' i ] TER-C5506-14 6 1 t I 4. CONCLUSIONS. 1

j With respect to the R. E. Ginna Nuclear Power Plant, conclusions

] regarding Rochester G.as and Electric Corporation's response to IE Bulletin 80-04 are as follows: [8, o There is no potential for containment overpressurization resulting ?! from a MSLB with continued feedwater addition. 'l e d5 o The AEW pumps are protected from the effects of runout flow and .f therefore can be expected to carry out their intended function during ~3 the MSLB event.

n 3

d o All potential water sources were identified and,- although a reactor Q return-to-power is predicted, there is no violation of the specified j acceptable fuel design limits. Therefore, the Exxon Nuclear [17] MSLB 4 reactivity increase analysis remains valid. 4 o No further action regarding IE Bulletin 80-04 is required. ..] a 33da a 4

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_1 ') i. e Research Center MU Franklin 6 ,j 4 9.on or m re mu, _m

j I- '!'j TER-C550 6-14 6 .); I ~~ s 5. REFERENCES i -j 1. " Analysis of a PWR Main Steam Line Break with Continued Feedwater J Addition" il NRC Office of Inspection and Enforcement, February 8,1980 i IE Bulletin 80-04 ad-: 2. "Overpressurization of the Containment of a PWR Plant after a Main ] Line Steam Break" .j NRC Office of Inspection and Enforcement, October 1,1979 q IE Information Notice 79-24 Y 3. L. D. White, Jr. (RC&E) Letter to B. H. Grier (NRC, Region I) ] Subject Response to IE Bulletin 80-04 7 April 30,1980 4. D. M. Crutchfield (NRR, ORB #5) 'J. Letter to J. E. Maier (RG&E) j

Subject:

" Systematic Evaluation Program (SEP) for the R. E. Ginna j Nuclear Power Plant - Evaluation Report on Topics VI-2.D and VI-3" j November 3, 1981

1 5.

J. E. Maier (RG:E) ) Letter to D. M. Crutchfield (NRR, ORB #5)

Subject:

SEP Topics VI-2.D and VI-3 11 February 1,1982 3 5 6. Robert Emmett Ginna Nuclear Power Plant Final Facility Description and Safety Analysis Report, through Supplement 11 ] Rochester Gas and Electric Corporation, March 1969 j 7. 'I4chnical Evaluation Report _j "PWR Main Steam Line Break with Continued Feedwater j j Addition - Review of Acceptance Criteria" 3 Franklin Research Center, November 17, 1981 i TER-C5506-il9 i 8. " Criteria for Protection Systems for Nuclear Power Generating Staticas" ~' Institute of Electrical and Electronics Engineers, New York, NY,1971 IEEE Std 279-1971 9. Standard Review Plan, Section 4.2 " Fuel System Design" NRC, July 1981 [. NUREG-0800 4 t, Udil Franklin Research Center A Des.on of The Frar*hn Insciute

t s. TER-C5506-146 f 10. Standard Review Plan, Section 15.1.5 ~ l " Steam System Piping Failures Inside and Outside of Containment (PWR)" ] NRC, July 1981 j NUREG-0800 M d 11. " Criteria for Accident Monitoring' Functions in Light-Water-Cooled h Reactors" ] American Nuclear Society, Hinsdale, IL, Decewoer 1980 0] ANS/ ANSI-4.$-1980 12. " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess d Pir.nt and Environs Conditions During and Ibliowing an Accident" j Rev. 2 M NBC, recember 1980 j Regulatory Guide 1.97 ] 13. " Single Failure Criteria fcr PWR Fluid Systems" American Nuclear Society, Hinsdale, IL, June 1976 ANS-51.7/N658-1976 1 ..] j ' 14. " Quality Group Classifications and Standards for Water, Steam, and Radioactivo-Waste-Containing Components od Nuclear Power Plants" "l Rev. 3 jj NRC, February 1976 p] Regulatory Guide 1.26 ~I p]j 15. " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment" l-Rev. 1 NBC, July 1981 l.. 4 NUREG-0588 g i j 16. T. Speis (NRR, Assistant Director for Reactor Safety) .) Mettorandum to G. Lainas (NRR, Assistant Director for Safety j

Subject:

Systematic Evaluation Program (SEP) for the R. E. Girina Assessment) j 1j Nuclear Power Plant - Evaluation Report on Topics VI-2.D and VI-3 (Docket No. 50-244) '] March 31, 1902 I 17. " Plant Transient Analysis for the R. E. Ginna Unit I Nuclear Power I

• Plant" Exxon Nuclear Company, Inc.

i XN-NF-77-40, Supplement 1 1 March 3, 1980 I 4 nklin Research Center i A Onesson of The Franen tascade - - - - - ~ - '-}}