ML17256B022

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Pages 4 & 36-61 to Directors Decision DD-82-3,Attachment 1
ML17256B022
Person / Time
Site: Ginna Constellation icon.png
Issue date: 05/22/1982
From:
Office of Nuclear Reactor Regulation
To:
Shared Package
ML17256B021 List:
References
DD-82-03, DD-82-3, NUDOCS 8206100394
Download: ML17256B022 (28)


Text

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B The major factor in assessing the safety margin of any of the SEP facilities depends upon the ability to provide adequate protection for postulated design basis events (OBEs).

The SEP topics provide, a major input to the OBE review, both from the standpoint of assessing the proba" bility of certain.events and that of determining the consequences of events.

As examples, the safe shutdown topics pertain to the listed OBEs (the extent of applicability will be determined during the SEP OBE review for Ginna):

~TO ic V-10. B V-11.A V-11. B VII"3

. IX-3

~GBE G

YII (Spectrum of Loss-of-Coolant Accidents)

YII (Oefined above)

~

'II (Oefined above)

All (Oefined as a generic topic)"

.III (Steam Line Break Inside Containment)-

(Steam Line Break Outside Containment)

IY (Loss of AC Power to Station Auxiliaries)

(Loss of all AC Power)

V (Loss of Forced Coolant Flow)

(Primary Pump Rotor Seizure)

(Primary Pump Shaft Break)

YII (Oefined above)

Impact Upon Probability Or Conse uences of OBE Consequences Probability Probability Consequences Consequences Consequences Probability Consequences For a listing of OBE groups and generic topics, see Reference 10.

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    ne ol o~ing shaii 'e prcvidea in
    ",a suc n side f e ~HR sysm~ -" isolate 'urs Ne RCS. (a) soiation -<<ari >e provided by't s- -~o peer coer tact valves in sar.as.
    ne vaive positions shall '.e inaiczt a fn
    ~aa cantro 1 rocrrre {b) one valves shall have '.ncaoendant "'.ver.a.ntariocks ta pr vent 'e valves fxrrr aing:oened unless -.e RCS presser fs helm.he M system cas"gn prassu.. F fiure ef s pmer suopiy shall not c use my valve to -Wanga pcsi,cn. (c) The valves shaii have fndapenC nt "'.ve.se <ntariocxs pratac-against cne or ct!r valves '."eing ooan dur'.ng xn RCS fncreasa aaove "e "asign pressure of 'le QiR sys~. 2. One of the.aliavfng snail "e "r.v'"ad on -"e ".scnarge sfoe
    t'e M sys~ w isolate i ra ae RCS II
    {a).na !alves cs ~ an...dl Qrs .d '" err c'.cs ascr'd ~ n K%1{k) ( ) ,b) One:r ~or ="'.act valves in sar'.as i ". *.".crnaiiy closed pe er operated valve.:ne pc~er sacra-4 valve.:osf:!on ~snail =e '.ndicztart in .".e ="nt~i..ca..r. '.e ?HR sys= ~ oi sonar e i ine ~ 3 usact ~ Qr sn ~ ur~ ~ cn tha olet caaratac valve is.a "e maenad con "acaio= "f a safety 'naction signai =nca .e."eac-"r ="alert =. ssur nas acr>>asaa caid ".a = "~ casign pr assur ~ ~ ~ q ~ -4T (c) Three check valves in series, or (d) Tw) check valves fn series, provided chat ""ere ar desfgn provisions o permit periodfc astfng oi'he "".eck ralves for leaktfghtness and he:asting is per,ormed st least annually." The RHR suc-,on md dfsc.".arge calves c"nnec-ng =afs systan o =~e pr'.mary coolant system zr s".o~n "n."-igure 9.3-l of:he R. "=. Ginna;~R. ~i.e r acwr coolant system suc '.on supp1y to -Ne RHR oumos is, rom ~",e.-.ot leg of l.op A >rough zoto~peratad valves NOV 700 and .".OV 701 ',n series. The RHR umo discnarge r turn w Ne leap 8 cold 1 g of che reac=-r c"olsnt system 's "%rough .~o series zocor coeracad
    valves, vOV 729 ann 8)V 721.:nere ar no ~".eck valves in sar.'es
    <<i~a ."OY 7ZQ snd ".OV:Zl. Permfssf ra inter'.ocks ." quiM to =pen ~he 'our RHR austen !sol-tion 0 valves are '.istad cela. .".OV 744 NOV 701
    ".OV 7ZO
    'RlV 7Z1 (1) Reactor c"olant sys 85 pressur 1us -e 'ass ~Man 410 psig (Z) RHR suc"fon valves M)V.3cQA ala uQV acGg,r m t~e conminment sump must e c'.osad (1) RHR suc-fon ralves .'OV "-MA and.'SV 3508.~m ~ae c"ntafnment sumo must 'e =losed (Z~ The 'reive is operated t'J a <ay P4i+wr1 ('i) .'Io inter'.ocks axis" ':ut ",e valve is ce. scao "v <ey F41 ~? (1) Reacwr coalwt systam or ssure ~ust:e 'ass ~lan !10 sfg Ho interlocks are associated with valve closure. There are no autamatiC functians which close the valves and na alarms generated by the valves (Reference 5). The valves fail "as is" upon loss of power supply and have renote position indication in the control room. The RHR system discharge line is not used for an ECCS unction that would require HOV 720 or MOV 721 to open;
    however, a branch of the RHR discharge line provides low pressure safety injection (LPSl) to the reactar vessel via parallel lines with. one normally closed motor-aper atad valve and one check valve in each line.
    Tne check valves are periodically tasted. The motar operated valve position '.ndicat.'on is provided in the control roam and these valves receive an open signal.caincident with the safety injection (Sl) signal. Based on the above description, the RHR system deviates fram these BTP provisions (a) The power operated valves in the LPSZ lines open on an SL signal before RCS pressure drops below RHR design pressure. (b) The ZHR discharge and suc ion isolation valves do not have independent diverse interlocks ta prevent opening the salves until RCS pressure is below 410 psig. Only the inboard valves (700, 721) have his interlock. The outboard valves (701, 720) are manually controlled with key-lacked switches. By procedure, HOV 701 and NPl 720 ar not opened until RCS pressure is less than 410 psig. -49" (c) The RHR isolation valves have no interlock feature to close them when RCS pressure increases above the design RHR pressure. The staff has concluded that the deviation regarding the independent, diverse interlocks to prevent opening of the RHR isolation valves until pressure is below 410 psig is acceptable. The RHR isolation valves are designed such that they are physically unable to open against a dif" ferential pressure of greater than 500 psi. The'nboard isolation valves are provided with a pressure interlock. By administrative procedure, the RHR valves are key-locked closed, with power removed. In addition, a relief valve (RY203), set at 600 psig>is available. The staff therefore has concluded that the probability of an intersystem LOCA is acceptably low. The deviation regarding the LPSI isolation valve is considered acceptable since the check valve testing provides sufficient assurance that these valves will perform their isolation function until RCS pressure decreases below RHR pressure. The staff's position on these deviations is given in Section 5.2. The deviation regarding lack of automatic closure for the RHR isolatioh valves is acceptable based on the administrative controls which the licensee provides for the operation of these valves, coupled with the RHR system high pressure alarm at 550 psig and the RCS interlock pressure alarm at 410 psig (Reference 5). These alarms provide adequate assurance that the operator action required by procedure will be taken to shut the isolation valves when RCS pressure is increasing towards the RHR design pressure. (See the following discussion of BTP provision C.l, "Pressure Relief Requirements.") 4.2 "C. Pressure Relief Re uirements The RHR system shall satisfy the pressure relief requirements listed below. I. To protect the RHR system against accidental overpressurization when it is in operation (not isolated from the RCS), pressure relief in the RHR system shall be provided with relieving capacity in accordance with the ASME Boiler and Pressure Vessel Code. The most limiting pressure transient during the plant operating condition when the RHR system is not isolated from the RCS shall be considered when selecting the pressure relieving capacity of the RHR system. For example, during shutdown cooling in a PMR with no steam bubble in the pressurizer, inadvertent operation of an additional charging pump ot inadver tent opening of an ECCS accumulator valve should be considered in selection of the design basis. The RHR relief valve has a setpoint of 600 psig and a capacity of ~g 70,000 lb/hr. The RHR system is provided with a 550 psig high pressure alarm and a reactor coolant system interloc" pressure alarm at 410 psig. The RHR system is connected to the loop A hot leg on the suction side and the loop 8 cold leg on the discharge side. The design pressure and temperature of the RHRS are 600 psig and 400~F. The design basis with regard to overpressure protection for Qinna Station's RHRS is to prevent opening of the RHR isolation valves when RCS pressure exceeds 450 psig and to provide relief capacity sufficient to accommodate thermal expansion of water in the RHR and/or leakage past the system isolation valves. An analysis of incidents which might lead to overpressurizing the RHR system was performed (Reference 5). Three events were considered in the analysis: ~ ~ / 4 (a} Nth RCS in solid condition and RHR and charging pumps operating, the letdown line from the RCS is isolated. (b} Ouring cooldown using two RHR trains, one RHR train suffers a failure at a time when the core heat generation rata exceeds the heat removal capability of one train. (c) Pressurizer heaters are energized with RHR in operation and RCS sol id. The results of these analyses show that the RHR system is provided adequate / relief capacity provided certain procedural changes are implemented. / These changes have been implemented in the licensee's operating procedures. ~g Overpr ssure transients more severe than the three listed above have been analyzed by the licensee in conjunction with the reactor vessel overpressuri" zation protection sys~ (OPS) (Reference 4). To successfully mitigate these ~orst case transients, the licensee has modified the pressurizer power operated relief valve (PQRVs) to provide a low pressure relief setpoint of 435 psig during plant cold shutdown condit ons and has imple-mented several administrative controls changes. The PORVs also provide overpressure protac ion for the RHR sys~ when the RHR ',s aligned "o the RCS for shutdown cooling. I ~.. ~ ~ ~ The staft has evaluated the effects of the worst casa mass and heat input events to establish the capabi1ity of the OPS and RHR relief to prevent RHR overprassurization. For the mass input case presented in Reference 4, the OPS alone prevents pressure from exceeding the RHR design pressure. For the heat input case, the Reference 4 data was extrapolated to include a 504F steam generator to RCS temperature difference at an RCS tempera-tur of 300 F. (The data in Reference 4 only applied to heat input transients at RCS temperatures from ISOoF to 250~F.) 300~F was chosen
    because, this is the maximum temperature for which re steam generator to RCS temperature dif,eranca is "O~F based on RHR initiation at 3 0 F Tha staff determined that pressure transients, at an RCS temperature of 300'F which would result from heat addition, would not exceed 110 of RHR design pressure even a5suming the failure of one PORY.
    No credit is taken for action of relief valve RY-203. The swff then considered tha potential for initiating a heat input transient at Ginna when RCS tamper V ature is between 300 F and 350'F.. For a heat input transient to occur, Da heat from the steam generators must ba rapidly transferred to a cooler, water solid RCS. The means of rapid heat transfer is forced convection caused by a reactor coolant pump start. In im review of ovarprassuri=ation transients, the staff considered stean generator to RCS tamoaratura dif arancas in axc ss of 50~F to ba unlikely occurrences. The administrative measures proposed by he Iic nsae to duce the proba-bil',ty of, heat input transient >era to (I) require an accaptab1e RCS temperature profi1e prior to reactor coolant pump s artup with a water solid RCS, (2) require one coolant pump to be r"'n until RCS tamoeratura -63 ~ 0 is less than or equal to 150'F, and (3) minimize plant operation fn a water solid condition. Although items (1) and (3), above, would not necessarily preclude a heat addition event, item (2) wauld. Also, the staff examined Ne potential for initiat'.ng a heat input event during plant cooldawn, which is the time that steam generator temaerature may exceed RCS temperature with RCS temperature above 300OF. The I icansee initiates RHR cooling at 350oF after cooling down to that point with the steam generators. Cont'inuing the caoldown with the RHR system and with the reactor coolant pumps secured (in violatian of procedures), would result in the 50~F difference being fully developed at an RCS ~mperature of 300'F. As inoted befor, a heat input event at this temperature would not r suit in RHR overpressurization even with an assumed single failure. 8ased an Ne abave discussian we conclude that the OPS and RHR relief provide sufficient RHR overpr ssure protac~Ãon for RCS temperatures of 300~F or less and that the licensee's prac dures acceptably minimize 4e likelihood of a Peat addition overpressure transient at RCS temperatures above 300'F. Therefore, the OPS and De RHR relief meet the pressure relief requirements of the BTP. The OPS and reIatad Technical Specifications were approved by the staf in Refer nce 17. '~ Sy procedure, the OPS is enabled at the same time as RHR c"aling is init.'ated during plant caoldawn, so he RHR sys am is ai orded the addi-tional averpressure prstac ion of the OPS. The lic nsee will be requir d to incorporate, into the plant; Technica'I Speciiications, a requ-cerement -54" for enabling of the OPS whenever RHR cooling is in progress to assure this safety marginis maintained for the life of the plant. The licensee has agreed to incorporate this change (Reference 20). 4.2. 1 "2. Fluid discharged through the RHR system pressure relief valves must 'e collected and contained such that a stuck open relief valve will not: "(a) Result in flooding of any safety-related equipment. "(b) Reduce the capability of the ECCS below that needed to mitigate the consequences of a postulated LOCA. "(c) Result in a nonisolatable situation in which the water provided to the RCS to maintain the core in a safe condition is discharged outside of the containment." Fluid discharged through the 2"inch RHR relief valve (RV203} is directed to the pressure relief tank (PRT) inside the reactor containment. The PRT has a rupture disc which is designed to rupture at 100 psig and allow the contents of the tank to overflow to the containment
    sump, where it would be available for recirculai'.ion.
    Should flow'from a stuck RHR J'relief valve cause the rupture disc to rupture, the consequences to safety-related equipment would be less severe than the consequences of post-LOCA containment flooding which has been previously analyzed and. found acceptable (Reference 6). If RV203 were to stick open in a post"LOCA scenario, RHR flow to the RCS for both low head recirculation and low head safety injection modes would be affected. This is because a flow path would exist from the RHR system to RV203 via valves HCV-133 and 703 in either of these RHR operating modes. HCV-133 fails shut following loss of instrument air on containment ) ~ ~ ~ ~ ~ ~ isolation following a LOCA, but a flaw path would still exist to RV203 via the 3/4-inch locked open manual valve 703. The effect of this flaw diversion would not reduce the capability of the ECCS below that needed to mitigate the consequences of a pos ulated lOCA. This is because the design flow rate through RV203 (70,000 Ib/hr, which is a conservative number in this case since HC'l-133 is shut) is much less than the flow rate of an RHR pump in the low pressure safety injection (LPSI) mode (776,000 lb/hr). Each RHR pump has the capacity'o provide 100 af the required LPSI flow. Therefare, the leakage through RV203 would not be as severe an event as the loss of an RHR pump which has been postulated as a single failure in the ECCS analysis. 4.2.2 "3. If interlocks are provided to automatically close the isolation valves ~hen the RCS pressure exceeds the RHR system design pressure, adequate relief capacity. shall be provided during the time period whiTe the valves are closing." As noted above, these interlocks are not provided. However, the procedures for cgordinat;an of the overpressure prataction and RHR systems as desc".ibed abave provide adequate r lief capacity to prevent the RCS pressure fram exceeding RHR design pressure. 4.3 "Q. P~ ma Protec.ion Reauir ments "The design and operating procedures af any RHR system sha11 have pravisions ta pr vent damage to the RHR system pumos due to overheating, cavitation or loss of adequate pump suc ion fluid.4 The features designed into tNe Ginna RHR system to prevent damage M the sys~ centrifugal pumps are pravision for pump cooling, a puma mini"flow 4 t ~ ~ ~ ~ 4 recirculation flow path, and system design to prevent1oss of net positive 4 suction head (HPSH). The CCN system provides coolfng for the RHK pumps to prevent damage fram overheating. The RHR pumps are pravfded wi h a recirculation line to recycle a portion of the pump discharge fluid to the pump suction. This prevents overheating caused by operating he pumps under na flow conditions. HPSH calculations were perl'armed for the RHK pumps by the licensee. The RHR operating modes evaluated were normal plant shutdown cooling, Iow pressure safety injection, and post-LOCA recirculation. Recirculatfon 'peration develaoed the most limiting NPSH requfr~aents, but the calcu-lations indicated a 4~ NPSH margin is available during recirculatfan (Reference 7, page 6.2<<37). The RHR NPSH requirements wi11 be re valuated during the SP under Topic 'lI.-7.E, "KCS Sumo Oesign and Test for Recfrculat'.on Node Kffectfveness." The above protec"fon features provide adequate protac ion to prevent RHR pump damage. 4.4 "E; Test Requirements "The isolation valve operability and interlock circuits must ba designed so as to permit online tes ing ~hen operating in the RHR 'ode. Tesmflf y shall meet the requirements of iE="- Swdard 338 and Regulatory Guide I.22.'he preaperatfanal and initial smrtup test program shall be in conformance with ReguIatory Guide 1.68. The prograns for PMRs shall include tests with supporting analysis to (a) confirm that adequate mixing of borated ~ater added prior.o or during cooldawn can be achieved under natural circulation conditions and permit es fmation of the times r quired to achieve such mixfng, and (b) confirm that the caaldown under natural circulation ~ ~ 0 t~ conditions can be achieved within the limits specified in the emergency operating procedures. Comparison with performance of previously tested plants of similar design may be substituted for these tests.'he RHR isolation valve operability and interlocks cannot be tasted during the RHR cooling mode of operation. This test requirement is nct applicable to the Ginna facility, since the installed interlocks unction only when the RHR isolation valves are shut. Regulatory Guide I.68 was not in existence when the Ginna preoperaticnal and initial s artup tasting was acccomplished. However, tests have been performed to confirm that cooldown under natural circulation can be achieved (Reference 8). The care flow rates achieved under natural ci=ulation war more than adequate for decay heat removal. The calculated cora flow at approximately 2 r actor power was 4.2 of nominal fuII power flow. At approximately 4~ reactor power, calculated core flow was 5.2 cf nominal. F1ow rates of this magnitude should orovide adequate mixing of boron added to the RCS during cooldowa. An incident at Ginna Station on July 5, 1970, provides further indication that natural circa-Iaticn wi11 provide uniform mixing of boron in the RCS (Reference 9). Ouring that incident while steam system maintenance was in progress with no RC. s operating, natural circulation was indicated by incore .hermocouple readings. While the RC7s were secured, 136o gallons of water wer added to the RCS to dilute the boron concentration. When an RC7 was r sorted reactor power, which was being maintained at a <ow power level corresponding'o 10 amps on the intermediate range channel, did not, change. This "58-indicates that the natural circulation flow had uniformly mixed the boron throughout the RCS. 'ji'>1 ' 4.5 "F. 0 erational Procedures "The operational procedures for bringing the plant from normal operating power to cold shutdown shall be in conformance with Regulatory Guide 1.33. For pressurized water reactors, the opera-tional procedures shall include specific procedures and information required for cooldown under natural circulation conditions." Operational.procedures reviewed in this comparison of the Ginna Station to BTP RSB 5"I are discussed in Section 2.0. All of the procedures required the use of nonsafety-grade equipment for portions of the shutdown operation. The licensee performed a review of a plant shutdown utilizing safety-grade equipment only; this procedure would require remote hand operation of certain air operated valves because the control air system is not safety-grade. The procedures for shutdown and cooldown ~y should provide instructions as to how safety-grade equipment could be used to perform the cooldown. Ho procedure exists fol proceeding to cold shutdown conditions from outside the control room. The need for procedures for these evolutions stems from the provisions of BTP RS8 S-I and SEP Topic VII-3 to provide assurance that the capability for decay heat removal with safety-grade equipment exists. The staff'ill consider requiring the licensee to develop these procedures during the integrated SEP assessment of the Ginna plant.'e conclude that the procedures for safe shutdown and cooldown at Ginna are in conformance with Regulatory Guide T.33. The plant operating procedures also include a procedure for cooldown using natural circulation.
    4. 6 "G.
    Auxilia Feedwater Sunni "The seismic Category I water supply for the auxi1iary feedwater system for a PW shall have sufficient inventory to. permit operation at hot shutdown for at least four hours, followed.b~ cooldown to tha conditions permitting operation of the kHR system. The inventory needed for cooldown shall be based on the longest cooldown time needed with either only onsita or only offsite power available with an assumed single failure." The Category l ~ater supply for the auxiliary eed system (AFS) is the service water system (SMS). The SMS, which must be manually aligned to the AFS system, receives its water supply from Lake Cntario via the seismic Class E screen house. This source of water, wnich has never be n interrupted in the nine years of plant operation, provides sufficient AFS water supply with an assumed single failure regardless of he loss of oifsite or onsite power.'he SP wiII reexamine the adequacy of the semen house to provide water during emergency shutdown and maintenance of safe shutdown during oesaioaion of goo =anics an seisnic design snd Tending. The SP has reevaluated the capability of the Ginna plant to achieve cold shutdown conditions within a reasonable period of time in ADpend'ix Ae 5.0 RESOLUTION OF SEP TOPICS The SEP topics associated with safe shutdown have been identified in the INTROOUCTION to this assessment. The following is a discussion of how the Ginna Station meets the safety objectives of these topics.
    5. 1 Topic V-10.8 RHR S stem Reliabilit The safety objective for this topic is to ensure reliable plant shutdown capability using safety"grade equipment using the guidelines of SRP Section 5.4.7, Regulatory Guide 1.139, and BTP RSB 5-1.
    The Ginna Station systems have been compared with these criteria, and the results of these comparisons are discussed in Sections 3.0 and 4.0 of this assessment. Based on these discussions, we have concluded that the Ginna systems fulfillthe topic safety objectives except for the requirement for procedures to shutdown and cooldown using safety-grade systems. The licensee will be required to ensure that their operating procedures contain sufficient information to enable plant operators to perform V required functions, such as decay heat removal, with safety-grade systems. 5.2 To ic V-ll.A Re uirements for Isolation of High and Low Pressure ~Sstems The safety objective of this topic is to assure adequate measures are taken to protect low pressure systems connected to the primary system from being subJected to excessive pressure which could cause failures and in some cases potentially cause a LOCA outside of containment. Thss topic ss assessed in this report only with regard to the isolation requirements of the pHR system from the RCS. As discussed in Sec-tions 4.1 and and 4.2, adequate overpressure protection for the RHR system w)11 ex>st when the plant technical specifications are modified to require enabling the overpressure protection system whenever RHR cooling ss sn progress. The licensee agreed to this change in a letter dated January 13, l981. 5.3 Topic V-ll.B RHR Interlock Requirements The safety objective of this topic is identical to that of Topic V-ll.A. The staff conclusion regarding the Ginna RHR interlocks, as discussed in Section 4. 1, ss that adequate interlocks exist subject to completion of the above modification. J s 4 f