ML20010A278
| ML20010A278 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 07/31/1981 |
| From: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM NUDOCS 8108110272 | |
| Download: ML20010A278 (15) | |
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July 31,1981 e g, Director, Nuclear Reactor Regulation C AUG 1 E, g,, Qjggj > _ Att Mr Dennis M Crutchfield, Chief Operating Reactors Branch No 5 coamissum 8 US Nuclear Regulatory Co==ission Q Washington, DC 20555 h/, DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - EVALUATION OF ADDITIONAL WIDE-RANGE LEVEL INSTRUMENTATION USING PRA BASED METHODS; RESPONSE TO NUREG-0737 ITEM II.F.2 Consu=ers Power Cc=pany sub=ittal dated July 9,1981 reaffir=ed our co==itment to ec tinue evaluating the usefulness of vide-range level instru=entation as part of the Probabilistic Risk Assessment (FRA) based Continuing Risk Manage- =ent Progra=. This evaluation was acco=plished by developing and analyzing Operator. Action Event Trees. As co==itted to in our submittal of July 9, 1981, this letter provides Const=ers Power Co=pany's final evaluation of the usefulness of aug=enting the existing instru=entatien at the Big Rock Point Plant with vide-rame level instru=entation. A su==ary of our evaluation, including a description of the develop =ent of Operator Action Event Trees and the results of analyzing these t rees is provided in Attachment 1. Supplementary infor=ation used in this effort is also provided in Attachments 2 through 5 As a result of this evaluation, Consu=ers Power Company has concluded that existing instru=entation at Big Rock Point provides una=biguous indication of inadequate core cooling as well as the approach of inadequate core cooling. Additional vide-range level instru=entation would provide only indirect. indication of failures in core cooling syste=s and would serve as a redundant backup to the existing instru- =entation. Based on these conclusions, Consu=ers Power Co=pany does not intend to install vide-range level instru=entation at the Big Rock Pcint Plant. Accordingly, Consu=ers Pcver Company will not be sub="tting a plan of action for installing additional instru=entation by September 1,1981. This letter completes our response to NURIG-0737 Ite= II.F.2. I f David P Hoffcan l l Nuclear Licensing Administrator $\\I CC Director, Eegion III, USIEC NRC Resident Inspecter - Big Rock Point 8108110272 810731 -PDR ADOCK 05000155 P PDR
e CONSUMERS P0h 3 COMPANY Big Rock Point Plant NUREG-0737, Clarification of 'IMI Action Plan Requirements Supplemental Response to NRC Letter, dated October 31, 1980 I Docket No 50-155 ( License No DPR-ti At the request of the Commission and pursuant to the Atomic. Energy Act of 1954, and the Energy Reorganization Act of 1974, as amended,'and the Commission's Rules and Regulations thereunder, Consumers Power Company submits our supplemental response to NRC letter dated October 31, 1980 ~, (NUREG-0737 " Clarification of TMI Action Plan Requirements", Item II.F.2). Consumers Power Company's supplemental response, entitled " Evaluation of Additional Wide-Range Level Instrumentation Using PRA-Based Methods;. Response to NUREG-0737 Item II.F.2", is dated July 31, 1981 and provides i information in addition to that provided by our responses dated December 1979, 1980 and July 9,1981. CONSUMERS POWER COMPANY / By N b JWeynolds, Executive Vice President Sworn and subscribed to before me this 31st day of July 1981. r. ?J Y N ams-e Helen I Dempski, Notary Edblic Jackson County, Michigan My commission expires Decem*>r 14, 1983
i l ATTACHMENT 1 Summary of Wide-Range Level Instrumentation Evaluation Introduction Item II.F.2 of NUREG-0737 presented the following NRC position with respect to instrumentation for detection of inadequate core cooling. 1 " Licensees shall provide a description of any additional instrumentation or I controls (primary or backup) proposed for the plant to supplement existing instrumentation (including primary coolant saturation monitors) in order to provide an unambiguous, easy-to-interpret indication of inadequate core cooling (ICC). A description of the zunctional design requirements for the system shall also be included. A description of the procedures to be used with the proposed j equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided." i Expanding on this position, the NRC further indicated that this instrumentation must also indicate the approach of inadequate cooling, the existance of inadequate ( cooling from multiple causes, must not erroneously indicate inadequate cooling due to unrelated phenomena, and must indicate full range from normal operating conditions to full core uncovery. Evaluation of this instrur atation was to include reactor level indication. In our initial response to NUREG-0737, dated December 19, 1981, Consumers Power indicated its participation in the work being performed for the General Electric l Owners Group with respect to the need for vide-range level indication. The result of this work was that the existance of adequate core cooling could be verified by assuring that existing vessel water instrumentation indicated water level above the col. or that rated flow was indicated in one of the lov pressure core spray loops. The conclusions of the Owners Group were felt to be applicable at Big Rock Point. Consumers Power also indicated that procedures requiring use of this instrumentation for the detection of inadequate cooling would be imple-mented as dictated by the schedule of Item I.C.1 of NUREG-0737 (Guidance for Development of D:ergency Procedures). Any further consideration of the addition of vide-range level instrumentation would occur pending the completion of the l Probabilistic Risk Assessment (FRA) of the Big Rock Point Plant. Consumers Power Company PRA submittal, dated March 31, 1981, reported that the need for vide-range level instrumentation was not indicated in any of the dominant accident sequences. Addition of this instrumentation would not result in an in-provement of the safety of operation of the Big Rock Point Plant, and continued deferral of this requirement was in order. It was proposed that as a part of the Continuing Risk Management Program at Big Rock, an explicit examination of the benefits of vide-range instrumentation vould be performed by expanding key accident sequences as defined in the PRA into operator action event trees using the methodol-ogy of NUREG-CR/1hh0 (see Attachment 2). To assure that the operator could perform critical functions during accident sequences, the specific needs of the operator with respect to instrumentation could be identified using this methodology. If vide-rcnge level instrumentation proved useful as a result of this evaluation, appropriate instrumentation would be identified and modification proposed.
ATTACID S T I Pa"ge 2 of h Development of Operator Action Event Trees Examination of the benefits of vide-range level instrumentation using the approach defined in NUREG-CR/lkh0 has been completed. Operator Action Event Trees for key dominant accident sequences were developed and the instrumentation required to ascer-tain the existance of inadequate core cooling identified. Selection of accident sequences for this study was based on the following criteria: 1) the sequences result in core uncovery with the potential for significant core damage unless oper-ator n ion is taken, 2) the sequences have the potential for causing mmbiguous _lon to be transmitted to the operator with respect to action necessary to infor respcna to the accident, 3) the sequences must be dominant from a probabilistic identifies and briefly describes the selected sequences. standpoint. These sequences include small break LOCAs below the core, large break LOCAs above and below the core which coincidentally disable portions of, or both of the core spray lines to the vessel, spurious blovdevn of the primary system through the turbine bypass line, loss of instrument air, loss of the main condenser, loss of The selected off-site power including loss of all AC power sources, and ATWS. sequences encompass a vide spectrum of potential transients and accidents at Big Rock Point. Equally important, the Operator Action Event Tree branches and operator actions discussed here are representative of transient and accident sequences other than Consideration was given to the identi-those apecifically developed for this study. fication of inadequate core cooling for both long term decay heat removal situations as well as short term core cooling due to core uncovery. At each of the branch points developed for the operator action event trees, three items were specifically considered; 1) what the actual level of the primary system was at that branch point, 2) whether the operator knew what this level was, and 3) what additional information or benefits vide range level instrumentation into the core region vould provide. Results With few exceptions, a typical sequence of operator actions occurs in all accident sequences depending on the information available from existing steam drum and reactor level indication (see Attachment h):
- Reactor water level indication above Reactor Depressurization System (RDS) actuation setpoint (reactor level > 2'9" above core). This is indication that the core is covered and adequate cooling exists.
Operator Action - Steam drum level below centerline or downscale. Add water via con-1. densate/feedvater, control rod drive, or,if reactor is depressurized, core spray. 2. Steam drum level at or above cer.cerline. Control water injection as necessary via condensate /feedvater, control rod drive, or core spray systems (if reactor depressurized) to maintain drun level instrumentation on scale.
ATTAcm0RC 1 3 Page 3 of 4 Reactor water level indication below RDS actuation setpoint (4,2"9" above core). This is indication that uncovering of the core is imminent or has occurred. Operator Action - 1 1. Insure RDS actuation a. Check depressurization valves open via position indication, check reactor pressure lov, and/or listen for noise associated with rapid depressurization from containment. b. If unsure of depressurization, manually _ actuate RDS. 2. Insure adequate core spray a. Check core spray flow indicators for sufficient flow. b. Check fire pumps (annunciated in control room) on and core spray valves open via position indication, c. If unsure of adequate flow, open redundant flow path to core spray system through post incident systems (M07072). 3. Start condensate /feedvater system to provide contents of hotwell and condensate stcrage to the primary coolant system. a. Crack feedvater flow for water addition. b. Check condensate pump current, feed purp current, feedwater control valve position, and hotvell level as backup to feedvater flow. The benefits of vide-range level instrumentation into the core region are available when reactor level is lov, that is when reactor level drops below the RDS actuation setpoint. The actions identified above to maintain vessel level are the only actions available for the operator to take when reactor inventory drops to this level. The installation of vide-range level instrumentation provides no additional indication that these actions need to be pa-formed nor does it make available any additional systems to provide makeup to the primary coolant system. Indication that there is inadequate cooling is ascertained by the concurrent indication of reactor water below the RDS actuation setpoint and insufficient core spray flow. Additional vide-range level indication can serve as backup to existing level instrumentation and provide indirect indication of-inadequate spray flow in that vessel level is not recovering or recovering slowly. However, this additional instrumentation does not provide direct indication of adequate cooling nor does it provide any additional information with respect to the reason inadequate cooling is occurring. The most beneficial aspect of vide-range level instrumentation is to provide one additional piece of information to the operator that indicates core spray and makeup systems are not working and, even though he has performed all appropriate actions, additional trouble shooting is required. Attachment 5 identifies such accident sequences in which core region level indication provides this information when it would not be available otherwise. The nature and number of failures re-quired to result in such sequences are such that they can be considered probabilis-tically insignificant.
ATTACHMDTf 1 h Page 4 of h Conclusions Existing instrumentation at Big Rock Point provides unambiguous indication of inadequate core cooling and its approach when it occurs. This instrumentation includes reactor vessel level indication and core spray flow. The operator can perform only a few specific actions if this instrumentation indicated the potential for inadequate cooling and procedures require him to perform them any time low reactor water level occurs. This conclusion is consistent with the BWR Owners Group position. Additional vide-range level instrumentation provides indirect inaication of failures in core cooling systems and can serve as redundant backup to existing instrumentation'. Wide-range level indication provides no additional useful information'to the operator in trouble shooting the cause of inadequate cooling. Additional vide-range level indication does provide some benefit to the operator in defining the best course of action for the operator to take during certain transient sequences which have been determined to be insignificant from a probabilistic standpoint. Based 'on this study, Consumers Power Company concludes that little benefit can be derived from the addition of vide-range level instrumentation in the Big Rock Point primary coolant system and plans no further work activity in this area. 1 -,-y ,w -.p ~,- -,,- ym, --9 -.. = - -, ,y.-
._. ~ _ -. ATTACHMENT 2 Description of Operato'r Action Event Tree Methodology HUREG/CH-lh40 entitled " Light Water Reactor Status Monitoring During Accident Conditions" presents a systematic approach for addressing operator informational needs during reactor accidents. Particular Accident Sequences are obtained or e developed in the form of event trees as were derived for the Reactor Safety Study (WASH-lh00). The selected sequences are generally those which are signi-ficant from a risk standpoint. The heading of each branch of a particular event tree sequence identifies a 1 unique ~ plant state or equipment function. Operator action event tree methodol-ogy begins by identifying a set of reactor conditions for the plant states associated with the event tree headings. An appropriate operator response to this set of conditions is identified. These operator responses make up the headings of the operator action event tree for the sequence of interest. With knowledge of the plant state and appropriate operator response, the instru-mentation required to supply the operator with the necessary and sufficient information to respond appropriately can be identified. Success or failure of the operator to act is dependent on the availability of this instrumentation. i i l The source of the dominant accident sequences for studying the benefits of wide._ - range level instrumentation was Appendix I of the Big Rock Point PRA. Selected I sequences for the wide-range. instrumentation study are identified in Attachment 3 An example event tree and operator action event tree for the PE F C sequence vs follows. i e 4 1 e i i
. continued LOSS OF 0FFSITE PG'.'ER EVEi!T TP.EE Fower Power Frimary Festored Inve - Festore Long n imerg. System trerg. During .ory RDS Durtne 1eru LOSP FP5 AE Power Isolation Cond. Short Terr g raseup 5.0 5.E C5 tong term Cochng P A Q I E F, 7 J K C F I a 5 t- ?> I 4 PI,f t g 9, FEL 14 13 PIF L E 5 15, 37 Pi 3 y g PET,L I P!F,J 17 Fit 19 r j PIF t g 16 PlyC 23, 22 PIT t i 24, ?! PIF,F t g 26 PIf TL g 25 28, PlF,TF t g PIF,YC ??, 31 PQL Me 33 i PQF t g 30 35, PQit 40, 34 39 PQiF,L 33 41, PO!F P t gg PQIF,C E FQEF,r. PQEF,J 44r' 43 ICIL 461 45 i PQIYL 42 PQITC 491 43 PQlF,L 47 64 PQlF,F t g PQlF,C ATW5
- continued - OPERATOR ACTION EVEllT TREE FOR THE PE F C SEQUENCE ys LOSP RPS Emergency Primary Emergency Operator Power RDS/CS Operator Long Term Power System Condenser Manually Opens Restored Manually Cooling with Isolation EC Valves and Be fore Depressur-Emergency Establishes Safety izes Reactor Condenser, Heat Sink Valves Vessel and/ Shutdown System, Lift or Provides or Core Spray / 5 Alternate PIS cooling to i Core 4 10a Core Cooled 6a I .10b Passible Failure Core Damage 3 5 Assumed Failure = Assumed 6b = 2 7 10c Co_r.e Cooled 1 9a Suc: cess Assumed 8 10d Possible Core Damage Success Assumed 9b Core Damage ~~ Success Assur.ed
ATfACHIGiT 3 Wide-Range Level Instrumentation - Operator Action-Event Trees Probabigty(1) dequence(g (yr ) Description -6 HZ 1.1 x 10 High energy line break in the pipe. tunnel y outside containment which disables both core spray lines. H 3.9 x 10 High energy line break in the recire pump -7 2 room which disables both core spray lines or a portion of the core spray system. BB ZY G 2.0 x 10 Spurious opening.of the turbine bypass g valve, failure to isolate the bypass line or MSIV, failure to makeup with. feedvater, followed by RDS/ core spray failure. TAY L 1.4 x 10-ATWS with loss of feedvater. f 3 1 x 10-6 Loss of off-site power, failure of emergency PE F C condenser outlet vt.1ves, failure to restore off-site power quickly, RDS/ core spray failure. Loss of off-site power, failure of diesel PQE F C 2.5 x 10 generator to start, failure to makeup to emergency condenser, failure to restore off-site power quickly, RDS/ core' spray failure. -6 Loss of main condenser, failure of emergency - ME NL 1.7 x 10 condenser outlet valves, failure to restore main condenser, long term cooling failure. via post incident system. 1.9 x 10 ' Loss of instrument air, failure to makeup UE UL to emergency condenser shell, failure to restore instrument air, ~ long term cooling failure via-post incident system. -6 Small break LOCA below the core, RDS/ core spray SEC h.0 x 10 failure. -5 Small break LOCA below the core, long term SEL 3 7 x 10 cooling failure via post incident system. Sequences and probabilities from. Appeudix I of Big Rock Point PRA. (1) s
ATTACHMENT ls Identification of Existing Instrumentation at Big Rock Point used to Indicate Inadequate Core Cooling A Lr3!b7 e LT.3 t BS h (f-7 p.h ~ ttse LT3f B5 j .i. t13 e4 LT3 t e t. LT3 g4 s L12E!9 8 l LIREl9 A LT P.E73 B LT Re 2c A i ~ -d L b ', S RGob A LsReo6 B STEAM DRUM (g gg,g LE REOB l [A ~ [ i Ah n n FS252'3 TT 2t/,3 h -- 4 NG'!C F/6tQ p7 7 V5IQ1A TL",' l l -- u,q-.A>(H,; ~~'., CE2$\\1 re?*rs f y - m,r,4 s ~ I t FI2335 -gd ggggag r-- I LEgc9 4,., gg g c4 W ] 4 v"- FT2fk? A B l C D FS2520
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100 gpm annunciates FS2523 2500 gpm FE2815 Frimary Core Spray Flow-
'600"'H O 2
'T2162 0-600".H O 2
F.2335 0-900 gpm FS2522
> 100 gpm annunciates
> 300 gpm LERE08A Steam Drum Level East 60" H O 2
LTRE20A 60" H O 2
LIRE 19A
-30" to +30" St drum g.
LT3186 RDS actuation 0-39 1" H O annunc. 17" 2
below Q, LI3386
-30" to +30" St drum E LT3187 RDS actuation 0-39 1" H O annunc. 17" 2
below [.
ll LI3387
-30" to +30" St drum %
4 LERE08B Steam Drum Level West 60" Hj0 LTRE20B 60" H O 2
LIRE 19B
-30" to-+30" St drum E LT318h RDS actuation 39 1" H O :annunc. 17" 2
below {.
LI333h
-30" 2 +30" - St drum g 4
LT3185 RDS-actustion 0-39.1" H O annunc. 17" 2
below Q, i
LI3335
-30" to +30" St Drum %.
LERE09A
-Reactor Water Level 18" H O 2
i LSRE09A ECCS/RFS actuation 33" above Core ISRE09E B.U. ECCS actuation 33" above core LT3181 RDS actuation 0-27.h" H O annune..' 33" I
2 above Core LI3381 2h" to h2" above-Core 9
. -.. _, _ _ -,. _.. _, _ _. _.. ~
Attacament h page 2 of 3 Instrument Description / Function Range /Setpoint LERE09B Reactor Water Level 18" H O 2
LSRE09B ECCS/RPS actuation 33" above core LSRE09F B.U. ECCS actuation 33" above Core LT3180 RDS actuation 0-27.h" HO annunc. 33" 2
above Core LI3300 2h" to h2" above Core LERE09C Reactor Water Level 18 H O 2
LSRE09C ECCS/RPS actuation 33" LSRE09G U. ECCS actuation 33" above Core LT3183
.DS actuation 0-22.h" H O annunc. 33" 2
above Core LI3383 24" to 42" above Core LERE09D
'.leactor Water Level 18" H O 2
LSRE09D ECCS/RPS actuation 33" above Core LSEE09G B.U. ECCS actuation 33" above Core LT3182 RDS actuation 0-27.h" H O annunc. 33" 2
above Core LI3382 24" to 42" above Cora Abbreviations:
ECCS - Energency Core Coolir.., System RFS - Reactor Protection System B.U.
- Back Up RDS - Reactor Depressurization System
ATTACHMENT 5 Discussion of Accident and Transient Sequences in which Wide-Range Level Instrumentation May be Beneficial 1.
Loss of Off-Site Power, no AC power (Emergency Diesel Generator, EDG, failure)
During loss of off-sice power, core spray flow indication derives its power from the emer6ency bus. Failure of on-site emergency AC power equipment will therefore disable thi; important indication of inadequate core cooling. Tb place the reactor in this situaticn in which inadequate cooling results, all of the following must occur:
a.
LOSP - loss of off-site power b.
Energency diesel failure to start or run c.
Failure of the emergency condenser as a heat sink either by failure of the outlet valves to open or failure of makeup to the shell.
d.
1411ure to repair the EDG by the time RDS actuation setpoint is reached.
e.
tailure to place the standby diesel on the energency bus by the time the RDS actuatica setpoint is reached f.
Failure to restore off-site power by the +ime the RDS actuation set-point is reached g.
Failure of the RDS/ core spray systemt for reasons not noticeable from the control room (ie, other than core spray salves failure to open or diesel fire pump failure to start)
Probabilities of occurrence of items a through g are fcund in Appendices I, II, mad III of the PRA.
The total probability of occurrence of sequences involving loss of all AC power sources and the RDS/CS system (in a manner not noticeable from the control room) is estimated to be less than 1.6x10~I/yr.
2.
Sequences involving failure of RDS/CS and condensate /feedwater systems On recognition of failure of the RDS/ core spray systems, the operator will begin injection of water to the primary system with the condensate / feed-water systems. Condensate /feedwater can be used in an attempt to reflood the vessel for all sequences in which off-site power is available and feedwater has not been disabled. Irimary indication of feedwater addition to the primary system is feedwater flow. Backup indication includes con-densate and feed pump motor currents, hotwell level, and feedwater reg valve nosition indication.
From Appendix I of the PRA dominant accident.
sequences ending in RDS/ core spray failure which could potentially be nitg/yr.
gated by rapid addition of feedwater occur with a frequency of 5.5 x 10-Wide-range-level instrumentation would serve as backup to feedvater system indication for these sequences in that it verifics the-water addition to the vessel. A lack of rising water level during feedwater addition indi-cates to the operator that the system is not adequate to keep up with primary coolant losses or additional trouble shooting of the feedwater
Q The most likely system is required to satisfactorily reflood the core.
cause of unsuccessful reflood of the vessel with feedvater while all feedvater instrumentation indicate normal is diversion of water through 2
a rupture 'in piping downstream of the feedvater flow instrumentation.
The likelihood of rupture of this particular section of piping sinul-taneous with the failure of RDS/ core gpray reduces the probability of these sequences to much less than 10- /yr.
3.
Partial core spray failure as a result of a high energy line break below the core Appendix I of the PRA identifies hi',h energy line. breaks witb!< the con-t tainment which could potentially disable portions of the core -pray sy9 em piping.
The probability of occurrence of these sequences is 3.9 x 10 /yr.
Assuming these seauences are a result of breaks below the core and esult in dr ;raded core spray flow to the vessel, vide-range level instrunentation coul be ceneficial in assisting the opert'. tor in determining whether to keep the et re cooled with core spray or feedvater. ' A range of break sizes exist in which feedvater is capable of keeping up 5ith break flow and maintaining the core at least partially covered, but for which the degraded core spray system is unable. Wide-range level instrumet.tation would indicate that the level in the vessel would fall after feedvate ? tripped due to a lack of inventory in the hotwell. With the informat: on that prior to the feedvater trip vessel level was near satisfactory, the op
'or could preferentially route fire system flow to the hotwell and rainti issel level with this.
high volume system rather than attempting to coc
.th a degraded spray system.
The range of break sizen which fall in this category are believed to be in This is a very narrow range of breaks and comprises the 400-1000 gpm range.
only a small fr9etion of the spectrum of medium and large breaks which make up to 3.9 x lo /yr. high energy line break sequences.
The probability of oc.urrence of those LOCA sequences which vould potentially benefit from vide-range level indication is considered to be much less then 10-8/yr.
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