ML20113G531
| ML20113G531 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 04/20/1989 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20092F288 | List:
|
| References | |
| CON-IIT05-194-90, CON-IIT05-195-90, CON-IIT5-194-90, CON-IIT5-195-90, RTR-NUREG-1410 13431-1, NUDOCS 9202210446 | |
| Download: ML20113G531 (29) | |
Text
{{#Wiki_filter:J y r ....,,m Vogtla Electric GeneratinD Plant Q' g +%% .i 13431*1 Nw* "o-f) s % w.1M Nutt. EAR OFERAriONS n m.on C 1 a 4 g'g g g Unit. Georgia Power
- e. 1
~~ Q u w,,,,,.3,y ^ .. u u u Dp ; g< w s m \\ u m,./,, m nq-p 4 ~J MOV AC IE VITAL INSTRUMENT _ DISTRIBUTION SYSTEM MANUM. SET' l.0 PjlRPOSE NO. a This procedure provides the necessary instructions for enerhi;ing, operating, and de-energizing the 120V AC lE Vital Instrument Distribution System. Instructions include the following sections 4.1.2 7.5KVA Inverter Startup 4.1.3 10KVA Inverter Startup 4.1.4 Energi::ing A Vital Instrument Panel 4.2.1 T-ansferring Vital Instrument Panel To Alternate Source 4.2.2 h Transferring Vital Instrument Panel To Normal Source 4.2.3 Energi::ing The Sequencer A(B) During System Operation 4.2.4 De-energi::ing The Sequencer A(B) During System Oportion i 4.3.1 De-energi::ing A Vital Instrument Panel 4.3.2 7.5KVA Inverter Shutdown f 4.3.3 10KVA Inverter Shutdown 2.0 PRECAUTIONS AND LIMITATIONS 2.1 PRECAUTIONS 2.1.1 Extreme caution shall be used during the transfer of a vital instrument panel to ensure a reactor trip does occur from the momentary loss of power to the not l Instrutent Panel. N 2.1.2 When Instrument Distribution Panel LAY 2A or 18Y2B is de-energized containment Ventilation Isolation will C) '"*~' ' ^" ^' 1RX-0003 (lBY2B), 9202210446 920116 goa ADOCK0500g4 wm som . _ - - - _ _ - - _ - - _ - - _ - - - - - - - - - - - - - - - - - - - - - - ~ - - ^ - ~ ^ ~ ~ - ^
f VEGP 134h-1 4 2 of 17 2.1.3 When Instrument Discribution Panel LAY 2A or 18Y2A Sequencer Board is energized, Containment Ventilation, Control Room, and Fuel Puilding Isolations may occur due to voltage transients. Control Room and ruel Handling Isolatter, should be blocked por Checklist 2. 2.1.4 t When Instrument Distribution Panel LAY 1A or IBY1B is energized Containment Ventilation isolation may actuate due to sover being restored to TR A(B) SSPS INPUT /O JTPUT CAUISETS. 2.2 LIMITATIONS 2,2.1 The 120V AC Vital Instrument Panela shall be energized in Modes 1, 2, 3, and 4 per Technical Specification 3.8.3.1. 2.2.2 The 120V AC Vital Instrument Panels shall be energized in Modes 5 and 6 per Technical Specification 3.8.3.2. 2.2.3 The 120V AC Vital Instrument Panel Supply Breakers are interlocked so that only one breaker at a time can be ,i closed. 2.2.4 The inverter DC input voltage shall be between 105-140V DC. 2.2.5 The inverter AC input voltage shall be between 414-506V AC. 3.0 PRERECUISITES OR INITIAL CONDITIONS 3.1 - The 125V DC power is availcble to the inverters. 3.2 The 4dOV AC power is available to the regulated i trans fo rmers, I t t l' f f e ~' 1
l VEGP 13431-1 l 4 3 of 17 4.0 INSTRUCTIONS, 4.1 STARTUP 4.1.1 If required. ALIGN the 120V AC IE Vital Instrument Distribution System per the applicable section of 11431-1, "120V AC 1E Vital Instrument Distribution System Alignment". 4.1.2 7.5KVA Inverter Startup CAUTION The 7.5kVA Inverters should not be operated at less than 20 amps of load for longer than 72 hours. 4.1.2.1 ENSURE that the following breakerr on the inverter are
- OPEN, Rectifier AC Input Breaker (ICB),
a. b. Battery Input Breaker (2CB), i Inverter DC Input Breaker (3CB), c. i d. Inverter AC Output Breaker (4CB). 4.1.2.2 CLOSE the Inverter DC Supply Breaker at the associated 12SV DC switchgest as listed in Table 1. 4.1.2.3 CLOSE the Inverter Battery Input Breaker 2CB. 4.1.2.4 DEPRECS and HOLD the Inverter Pre-charge Pushbutton. 4.1.2.5 After 5 seconds, CLOSE the Inverter DC Input Breaker 3CB. 4.1.2.6 RELEASE the Inverter Pre-charge Pushbutton. 4.1.2.7 VERIFY proper operation of the inverter by observing approxiraately 120V AC on the Inverter AC Output Voleneter 2VM. 4.1.2.8 CLOSE the Inverter AC Output Breaker 4CB.
- ~. - - - - - - - - - [ VECP 13431-1 4 4 of 17 - e i 4.1.3 10KVA Inverter Startup i 4.1.3.1 ENSURE the Ireverter AC Input Power Battery Input, and AC Output Breakers on the inverter are OPEN. 4.1.3.2 PLACE the Inverter Transfer Switch to INVERTER. 4.1.3.3 CLOSE the Inverter AC Supply Breaker at the associated 480V AC Motor Control Center (MCC) as listed in ^ Table 1. Jl j'<) PRESENTATION TO REGION 11 NUCl. EAR REGULATORY COMMISSION ON V0GT1.E SITE AREA EMERGENCY KARCil 20, 1990 AGENDA C. K. MCC0Y e OPENING REMARKS G. 30CKHCLD event REVIEW IEAM CRITIQUE e e TRUCK / SWITCHYARD e OFF-SITE NOTIFICATIONS e PERSONNEL ACCOUNTABILITY COMMUNICATIONS CORPORATE / SITE a e MID-LOOP OPERATIONS G bocks 0Lo DIE 3EL IESTINc/0PERABILITY e G, BoCKnoto e Ou,RANTINE COMPONENTS G. BocKHOLD e UNIT 2 i b k 1 '!I -it
ca ec co os ed rc eine,le-t 03a ubas 0 INITIATlHG_EyENT FUELING TRUCK STRUCK IfiSULATOP SUPPORT INSIDE TO THE LOW VOLTAGE SWITCHYARD CAUS!!1G A FAULT DIE 1A RESERvt AUXILI ARY IRAflSFORMER. e DIRECT CAUSE I9UCK DRIVER AftD ESCORT WE9E INATTEff rive 70 SAFE OPERATION OF THE TRUCK. CONI'RIBUTING CAUSES s CcNTROL OF VEHICLES NEAR VULNEPABLE AND SENSITIVE AREAS NOT ESTABLISHED. MAINTEtlANCE EGV!PMENT ST AGED lilAPPPQP91 ATELY. IHE USE OF GROUND-GUIDES INSIDE THE PRCTECTED A9E A WAS ff0T CLEAR. k 2 1
r 'd Gciso osesoero e1%,19 3.g33 ggg,4 Georgia Power interoffice Correspondence DATE: March 21. 1990 RE: Vehicles In Peritmeter Area FROM: G. Bockhold. Jr. T0: Site Personnel Due to the recent plant event of "tarch 20, 1990, the following shall ne implementeo imediately: All vehicles within the Perinneter Area (PA) in 41ch the driver does not have rearvien visibility OR that are larger than a pickup tnick, are required to have a fTaysian at all tim:s when the vehicle is backing up. Adottional policies / procedures on this issue will be fortheeming. m e e 4 3
e G 'd op GO 06/60/r0 ' ~"yg ggqy y.g. 0 3d 008 3 EMERG.E.NC PLAN IMPLEMENTATIQH Y DURING THE EMERGENCY, OFF-SITE fl0TIFICAT10NS WERE LATE AND/OR DELAYED BEYOND THE 15 slNUrE TIME LIMIT. e DIRECT CAUSES POWER TO THE PRIMARY ENN (1E EMERGENCv POWER) WAS LOST. ALL EMERGENCY AGENCIES WERE NOT IfiCLUDED ON THE BACKUP ENN. (BunKE COUNTY AND GEMA ADDED 4/6/90) CONTRIBUTING CAUSES e CONTROL ROOM COMMUNICATORS AND SUPERVISORS WERE NOT FULLY KNOWLEDGEADLE OF THE COMMUN!CAfl0NS (PRIMARY ENN IN TSC SYSTEM CAFABILITIES, HAD POWER FROM THE SECURITY SYSTEM DIESEL.) IHE SERIES METHOD OF NOTIFICATICl4 CONTAINED UNSATISFACTORY DELAYS. EMERGENCY DIRECTOR DID NOT ENSURE PROMPT NOTIFICATION OF OFF-SITE AGEtlCIES. AMPLIFYING INFORMATION WAS 110T PROVIDED TO LOCAL GOVERNMENT OFFICIALS. k
~ ~ 9 'I oriso os+60 ro nggygg. t GeorgiaPower b Interoffice Correspondence OATE: April 4.1990 I Emergency Notification Network (ENN) Ocessunicaticn i RE: FRCH: George Bockhold. Jr. t Emergency Direetors (ED) and Echnunit. TO: To ensure that ENN coruunication is t ieel y. Emergency Directors will ensure that the (ollowing improve *nents are implemented: 1. ! mediately upon the declaration of an emergency, the comn unicator (Shift Clerk) will perform a roll call to determine the operability of the EfiN while the message is being prepared by the ED. Burke County and GEMA is in the process of being added to the backup ENN and this will be installed and tested 2. within the next few days. The E0 will personally ensure notificatinns are t5ely ano problems are resolved. The ED will assign extra l 3. personnel or use TSC facilities to solve comunication problems as necessary. supplies than the Centrol 4. The TSC uses different power Rocm and TSC comunication systems may be operatile when Control Room systems are not. to most emergencies. ED's Since Burke County must respond Quickett will ensure that Burke County receives the highest priority for ENN noti f t:ations. We are investigating improved comunication hardware and techniques. In the meantim, your personal attention to ENN comunicatinns cust ensure that we do not have the problems that we experienced on 3/20/S0. N GB/gww . s 5
1 'd IP150 06.'60ero bitte 1Ae-g 03d WOdd I L EMERGENCY.fLAlliMPLEMENTAT10.8 DURifl0 THE EMERGENCY, $1TE PER$0t1NEL ACCOUNTABILITY 4 NEEDED IMPROVEMEtJT. e DIRECT CAUSE ACCOUNTABILITY PROCEDURES DID NOT Pf:CVIDE FOR THE SITUATION OF NOT EVACUATING THE SITE. (6EtlERAL MAf4AGER'S MEMO OF 4/6/90) e CONTRIBUTING CAUSES IHE Iff!TIAL PAGE AtJNOUNCEMEtJT WAS DELAYED APPROXIMATELY 20 Minutes. PcRs0NNEL WERE ALLOWED TO RE-ENTER THE PROTECTED AREA. PAGE ANNOUtlCEMEllTS ARE DIFFICULT TO HEAR If4 SOME PLANT APEAS. THE COMPUTER ',ENERATED PRINTCUT DID NOT ALLOW QUICK IDENTIFICATION OF PERSONtJEL. THE EMERGr.NCY O! RECTOR FAILED TO PROVIDE GUIDANCE AFTER DECIDING NOT TO EVACUATE PERSONNEL. + k i 6 l l
aT it. iso ogggo,pe ug7 9,g,03u u0dd f; 1 Georgial\\mer b. , Int:roffice Correspondence ~ D/.TE: April 6, 1990 q c Accountability During Emergencies RE: Log: NOV-00426 FROM G. Bockhold Jr. TO: All Emergency Directors and Site Personnel In the event of site emergency conditions, we will implementThese changes w the following reviseo procedures. accountability and safety and ensure better information flow forTh employees. responding to emergency situations. When the Energency Director (ED) makes an emergency classification, he will make the appropriata tone and page announcement on the plant }le will direct site personnel to the appropriate locations. He or she If you can not hear the page, report to your supervisor.Normally non-essential personne PA system. will direct you appropriately. David will report to the Aomin. Building auditorium or parking lot.P l with the ED and control the disposition of non essential person Captain for additional assistance. Emergency Response Organization (ERO) personnel should report Other shift personnel. imediately to the appropriate facility. supervisors, and managers on site should report initially to th OSC. shop area. When directed by the EO, the security department will initiateT who fail to log into the appropriate EAF (e.g., contist roos accountability. s m as possible. Your assistance implementing these instructions will ensurei t we manage emergencies better and provide plant personnel with suffic en information to keep thtm informed of abnomal plant activities. Thankyouforyourassjstance, tbb-> 7 iB/ erd l
- Department iteus NORMS s
ea zriso os co to viim,2e-t one woes a 1 l i i MEEGSiCY Pl.AN IMPl.(PENTAD Dftt CesMuralcATION BETWEEfl COPPCRATE AtiD ISC NEEDS TO BE JP, PROVED. DIRECT CAUSES o IHE STATUS LOOP TELEPH0 lie BRIDGE WAS t10T i OPEWABLE AT THE BEGINf4!f1G OF TWE EMERGErlCY BECAUSE OF THE LOSS OF POWER. t i e in 4 f J ) 4 \\ l 1 I' 1
01*d Cr160 06 60/r0 VAN 9"11e-2*D3d WOdd 4 3 l. ? I 1. 't k MLD-LOOP OPERAT10JS n 1 ol AcT t0 tis TO RESP 0tlD TO LCES OF CORE C00Llf4G AT MID-LOOP SHOULD EE IMP 90VED. lt DIRECT CAUSE ,i e \\ THE " LOSS OF RCSIDUAL HEAT REMOVAL" PROCEDUP I SHOULD PROVIDE IMPROVED GUIDAf1CE FOR A =, LOSP cot 4DITloN, l CONTRIBUTlHG CAllSES e IHE " LOSS OF RHR" FROCEDURES ARE TOO NARROWL i
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FOCUSED FOR MODE 5 8 6 cotmlilotas. DIRECY10tlS FROM THE EMERGEtlCY DIRECTOR l WERE NOT ALHAYS EXPLICIT. I' 1' r s e e O E 9 [ 4 L- - -- - ---- - _ _ __ _ ____ _ _ _
g,, c f DIESELESTING NORMAL 36 MONTH OVERHAUL AND INSPECTION SPECIAL TESTING E )k t l" "^ ?/20 EVENT 5 STr.RTs, TROUeLESH0011NG SEnson CALIBRATION LOGIC IESTING E-RUN BUBBLE IESTING MULTIPLE STARTS (14)- l' UV RUN TEST U 6 MONTH RuN SURVEILLANCE DIESEL OPERABLE UV RuN Test SENSOR' CALIBRATION LoGtc TEST!NG LUBE Oil DCP RUN E-RuN BussLa TESTING DCP UV RuN FUNCTIONAL MULTIPLE STARTS (5) UV RuN TEST 6 MONTH SuavEILLANCT. DIESEL OPERABLE Hi JACKET WATER RUNS (3) DCP UV RUN TEST 19 SUCCESSFUL STARTS li 18 SUCCESSFUL STARTS 10 5 4 m
~' eine,in .038 (403 3 ~ x <*~m 'n 21'd t't'160 06 / E O. t 0 4 91ARANTINElg@jfNTS ,'f 4 TEMPERATURE SWITCilES ) y pro l r. TRIP '. 0,sE o 1AJACKET WATER TEMPERATURE (2/3 Locic) 1 INTERMITTENT } 1 Post CALIBRATION LOW (1F'F 3 VENTING 1A OTHER IEMPERATURE COMPGNENTS 2 1 LUBE OIL IEMPEP ATURE (SLUGGISH) J ) ! 1B IEMPERATURE COMPONENTS 4 JACKET W.eTER TEMP (VENTING) e 2 LusE O!L TEMP (VENTING & CAL!B.) l i PRESSURE SWITCHES i l 1 1.UBE Olt PRESSURE (TR!PPED) 2 LUBE Olt PRESSURE (CONSERVATIV b a j e 1B 2 LosIc (woutD MOT TRIP ENGINE) L, ( 4 ot / ~- - -_ i
..,,0..o-. ux wu } s Il 1 .) l j '. 1 UNIT 2 e UNIT 2 TRIP UNIT 2 RAT B TRIP / PRIMARY DIFFERENTIAL IRIP i i - h TURB1NE TRIP / REACTOR IRIP 3AFETY $YSTEM RESF0 hse PROFER p e CAUSE DIFFERENTIAL RELAY CT SET 3000/5 vicE 2000/5 1 CORRECTIVE ACTIONS F e TEST THE REMAINING RELAYS ON UNtT 2 f UPDATE $WITCHYARD'ORAWING3 BASED ON AUDIT CLARIFY EXISTING POLICIES FOR SWITCHYARD f i s t i S e k I 12 i i _____________._________________m
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.. NUCLEAR REGULATORY COMMISSION UNITED STATES !.i WASHINGTON, D, C 20566
- g July 9, 1987 e...w i
ALL-LICENSEES OF OPERATING PWRS AND HOLDERS OF CONSTRUCTION' . PERMITS FOR PHRS E Gentlemen:
SUBJECT:
LOSS OF RESIDUAL HEAT REMOVAL (RHR) WHILE THE REAC10R COOLANT SYSTEM (RCS) IS PARTIALLY FILLED (GENEGIC LETTER 87-12 ) -Pursuant to 10 CFR 50.54(f), the NRC is requesting information to assess safe operation of pressurized-water reactors (PWRs) when the reactor coolant system (RCS) water level is below the top of the reactor vessel (RV). The principal concerns-are (1) whether the RHR system meets the licensing basis of the plant, such as General Design Criterion 34 (10 CFR Part 50, Appendix A) and Technical. Specifications (TS), in this condition; (2) whether there is a t resultant unanalyzed event that may have an impact upon safety; and (3) whether any threat to safety that warrants further NRC attention exists in this condition. Our concerns regarding this-issue have increased over the-past several years. and lessons learned from the April 10, 1987 Diablo Canyon loss-of-RHR event require an assesssnent of operations and planned operations at all PWR -
- facilities to ensure that these plants meet the licensing basis, Study of the j
Diablo Canyon event has led to identification of unanalyzed conditions that
- are of significance to safety.. Although Diablo Canyon never came close to core damage. and could have withstood the loss-of-RHR condition for more than a day -
(with no operator action, slightly different conditions could have: led to an accident _ involving core damage within several hours. One unanalyzed condition involves boiling within the RCS in the_ presence of air, leaning to-RCS pressurization with the potential-for ejecting RCS water via cold-leg openings, suchascouldexistduringrepairtoareactorcoolantpump(RCP)ortoa-loop isolation valve. The lost water would no-longer be available to cool the core, and if makeup water were-uravailable, the cora could be damaged in_ a - significantly decreased time. The pressurization could also affect the capability to provide makeup waterfo the core. ' Other unanalyzed situations are'also possible, and occurred at blabio Canyon (e.g., boiling in the-core).- -The seriousness of_this situation is exacerbated by the practice of conducting operations with the equipment-hatch removed,-and by the lack of procedures that address prompt containment isolation should the'need arise. Losstof RHR and related topics are not a.new concern to the NRC staff. .nis topic has been addressed in numerous comunications with the licensee. Yet, these< events continue to occur at a rate of-several per year. This condition needs to be fully considered in order to ensure compliance with the licensing 4 H basis. Thereiore, we request that you provide the NRC with a description of the . operation of your plant during the approach to a partially filled RCS condition i and during operation with a partially filled RCS to ensure that you meet the licensing basis. Your description is to include the following: l ' - 6707100112' %
o ~ 1 l .p, A detailed description of the circumstances and conditions under which (1) your plant would be entered into and brought through a draindoun process and operated with the RCS partially filled, including any interlocks that could cause a disturbance to the system. Examples of the type of information required are the time between full power operation and reaching a partially filled condition (used to detcrmine decay heat loads); requirements for minimum steam generator (SG) levels; changes in the status of equipm( t for maintenance and testing and coordination of such operations while the RCS is partially filled; restrictions regarding testing, operations, and maintenance that could perturb the nuclear steam supply system (NSSS); ability of the RCS to withstand pressurization if the reactor vcssel head and steem generator manway are in place; requirements pertaining to isolation of containment; the time required to replace the equipment hatch should replacement be necessary; and requirements pertinent to reestablishing the integrity of the RCS pressure boundary. A detailed dcScriptien of the instrumentation and alarms provided to the (2) op'erators for controlling thermal and hydraelic aspects of the NSSS during operation with the RCS partially filled. You should describe temporary connections, piping, and instrumentation used for this RCS condition and the quality control process to ensure proper functioning of such connections, piping, and instrumentation, including assurance that they do not contribute to loss cf RCS inventory or otherwise lead to perturbation of the MS$5 while the OCS is partially filled. You should also provide a description of your ability to monitor RCS pressure, temperature, and level after the RHR function may be lost. Identification of all pumps that can ce ased to control NSSS inventory. (3) (a) pumps you require be operable or capable of operation Include: (include information about such pumps that may be temporarily removed from service for testing or maintenance); (b) other pumps not included in item a (above); and (c) an evaluation of items a and b (above) with respect to applicable TS requirements. A description of the containment closure condition you require for the (4) conduct of operations while the RCS is partially filled. Examples of areas of consideration are the equipment hatch, personnel hatches, containment purge valves, SG secondary-side condition upstream of the alation valves (including the valves), piping penetrations, and electrical penetrations. Referetice to and a summary description of procedures in the control room (5) of your plant which describe operation while the RCS is partially filled. Your response should include the anAtic basis you used for procedures We are particularly incarested in your treatment of development. draindown to the condition where the RCS is partially filled, treatment of minor variations from expected behc"ior such as caused by air entrainment and de-entrainment, treatment of boiling in the core with and without RCS pressure boundary integrity, calculations of approximate time i ( t
[ 3-frem loss of RHR to core damage, level differences in the RCS and the 7 l effect upon instrumentation indications, treatment of air in the RCS/RHR j system, including the impact of air upon NSSS and in:trumentation l response, and treatment of vortexing at the connection of the RHR suction ? line(s) to the RCS. Explain how your analytic basis supports the following as pertaining to your facility: (a) procedural guidance pertinent to timing of operations, required instru;nentation, cautions, and critical parameters; (b) operations control and communications requirements regarding operations that may perturb the NSSS, including restrictions upon testing, maintenance, and coordination of operations that could upset the condition of the NSSS; and (c) response to loss of RHR, including regaining control of RCS heat removal, operations involving the NSSS if RHR cannot be restored, control of effluent from the containment if containment was not in an isolated condition at the time of loss of RHR, and operations to provide containment isolation if containment was not isolated at the time of loss of RHR (guidance pertinent to timing of operations, cautions and warnings, critical parameters, and notifications is to be clearly described). (6) A brief description of training provided to operators and other affected l personnel that is specific to the issue of operation while the RCS is partially filled. We are particularly interested in such areas as i maintenance personnel training regarding avoidance of perturbing the NSSS and response to loss of decay heat removal while the RCS is partially filled. (D Identification of additional rusources provited to the operators while the RCS is partially filled, such as assignnent of additional personnel with cpecialized knowledge involving the phenomena and instrumentation. (8) Comparison of the requirements implemented while the RCS is partially filled and requirements.used in other Mode 5 operation:;. Some requirements and procedures followed while the RCS is partially filled may not appear in the other modes. An example of such differences is operation with a reduced RHR flow rate to minimize the likelihood of vortexing and air it.gestion. (9) As a result of your consideration of these issues, you may have made I changes to your current program related to these issues. If such changes have strengthened your ability to operate safely during a partially filled situation, describe those changes and tell when they were made or are scheduled to be made. j contains insight which experience indicates should be well understood before commencing operation with a partially filled RCS. Your response to this 50.54(f) letter request should encompass the topics ccntained in Enclosure 1. Additional information is contained in the NRC Augmented Inspection Team report, NUREG-1269, " Loss of Residual Heat Removal System, Diablo Canyon Unit 2, April 10,1987." A copy of NUREG-1269 is enclosed.
4 Your response addressing ite m 1 through 9 (ubove) is to be signed under oath or affirmation, as specified in 10 CFR 50.54(f), and will be used to determine We request whether your license should be modified, suspended, or revoked. This information is your response within 60 days of receipt of this letter. required pursuant to 10 CFR 50.54(f) to assess conformance of PWRs with their licensing basis and to determine whether additional ?!RC action is necessary. Our review of information you submit is not subject to fees under the provision If you choose to provide a portion of your response in of 10 CFR 170. association with your owners group, such action is acceptable. i This request for information was approved by the Office of Management and 3150-0011 which expires December 31, 1939. Budget under clearance number Comments on burden and duplication may be directed to the Office of Management ar.d Budget, Reports Management Room 3208, New Executive Office Building, Washington 9.C. 20503. Sincerely, /l auNW ur ,}A Frank J. Miraglia Associate Director or Projects Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
Enclosures:
As stated
. _. ~. 1 1 i t Y ENCLOSURE 1 s j INFORMATION PERTINENT TO LOSS OF RESIDUAL HEAT REMOVAL SYSTEMS WHILE THE RCS IS PARTIALLY FILLED Many maintenance and test activities conducted during an outage require lowering the water level in the reactor coolant system (RCS) to below the top i of the reactor vessel (RV) or (as is done many times) to the renterline 1 elevation of the RV cozzles. This operating regime is sometires known as "mid-loop" operatien. It places unusual demands on plant equipment and t operators because of narrow control margins and limitations associated with equipment, instrumentation, procedures, training, and the ability to isolate j containment. Difficulty in controlling the plant while in this condition often leads to loss of the residual heat removal (RHR) system (lable 1). i Although this issue has been 8.he topic of many communications and investigations, such events continue to occur at a rate of several per year. j Recent knowledge has provided additional insight into these events. Although the full implications of this knowledge remain to be realized, our preliminary assestments have clearly established real and potential inadequacies associated with operation while the RCS is partially filled. These include: not understanding the nuclear steam supply system (NSSS) response to loss of t l RHR, inadequate instrumentation, lack of analyses addressing the issue, lack l of applicable precedures and training, and failure to adequately address the safety impact of loss of decay heat removal capability. The following items are applicable to these conclusions: (1) Plants enter an unanalyzed conditiort if boiling occurs following loss of RHR. For example: f (a) Unexpected RCS pressurization can occur. No pressurization would occur with a water / steam-filled RCS with water on the steam generator (SG) secondary side, because RCS stea, i
- _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ would condence in the SG tubes and the condensate would return to the RV. Air in the RCS can block the flow of steam through passages, such as the entrance portion of SG tubes, so that steam cannot reach cool surfaces. Failure to condense the steam causes pressurization in the RCS until the air compresses enough for steam to reach cooled tube surfaces. This pressurization occurred during the April 10, 1987 event at Diablo Canyon since the RCS contained air. Pressure reached 7 to 10 psig, and would have continued to increase if RHR had { not been restored. The operators began to terminate the event by allowing water to flow from the refueling water storage tank (RWST) inte the RCS. Increasing pressure would have eliminated this option, and would have jeopardized options involving pumps with suction lines aligned (in part) to the RCS. (b) Water that ordinarily would be available to cool the core might be forced out of the RV, thereby reducing the time between loss of RHR and initiation of core damage. This is a potential concern whenever there is an opening in the cold leg, such as may exist for repair of reactor coolant pumps (RCPs) or f' loop isolation valves. Upper vessel / hot-leg pressuritation could force the RV water level down,vith the displaced water lost through the cold-leg opening. A corresponding decrease in level would occur in the SG side of the crossover pipes between the SGs and the RCPs. This occurrence could be particularly serious if the cold-leg ~ opening were large or if makeup water flow to the RCS were small, as from a chargirg pump. Cold-leg injection with elevated pressure in the upper vessel may not provide water to the core. (2) RCS water level instrumentation may provide inaccurate information. There are many facets to this issue. Instrumentation may be indicating
l \\ a level that, differs from level at the RHR suction line, a temporary i instrument may be in use that has no indication or alarms in the control r- } toom, and de<;ign and installation deficiencies may exist. We have observed the following: } (ta) Connections to th3 RCS actually provide a water level indication up-stream of the RCP location. This water level is h-lqher than the water level at the RHR suction connection because of flow from the injection to the suction locations and because of entering water momentum, which increases level on the RCP side of the cold-leg injectionlocation. Ingestion of air at the RHR suction connection will result in transporting air into the cold legs, this can potentially increase pressure in the air space in the cold legs relative to the hot legs, } Level instrumentation may respond to such a pressure change as I though RCS level were changing. In addition, such a pressurization would move cold-leg water into the hot legs and upper RV (or the reverse if a depressurization occurs). Use of long lengths of small-diameter tubing which can lengthen instrument (b) response time and cause perturbations such as RCS pressure changes to appear as level changes, installetion with tubing elevation changes which can trap air bubbles or water droplets, and installation which makes it possible for tubing to be kinked or constricted. (c) Some installations provide no indication in the control room, yet level is important to safety. Some provide one indication. Others e, provide diversity via different instrumentation, but do not provide independence because they share common connections. )
1 (d) Tygon tube installations faintly marked at 1-foot intervals that have no provision for holding the tube in place. (e) Instrumentation in which critical inspections were not performed after the installation. (f) Instrumentatior,in which no provisions were made to ensure a single phase in connection tubing or that tubing was not plugged. (g) Use of instrumentation without performing an evaluation of indicated RCS leve'; bebavior and instrument response. (3) Vortexing and air ingestion from the RCS into the RP9 suction line are not always understood, nor is NSSS response understood for this condition. (a) On April 30, 1987, Diablo Canyon operaters reduced indicated RCS le,el to plant elevation 106' 6" immediately after steam generator tubes drained, and indications of er-atic RilR pump current were observed. Restoring the RCS level to 106' 10" was reported to have eliminated the problem. RHR operation was terminated a few hours later at an indicated level of 107' 4" because the operators observed erratic RHR pun.p current irdications. The licensee later reported that vortexing initiated under those conditions at 107' 5-1/2", and was fully developed at 107' 3-1/2". Procedures in place at the time of the event indicated the minimum allowable level to be 107' 0" (the hot-and cold-leg centerline elevation) or 107' 3". (b) Additional phenomena appear to occur under air ingestion conditions. These include:
RHR pumps at Diablo Canyon were reported to handle several percent air with no discernible flow or pump current change l from that of single phase operation. A postulate is that air ir. the RHR/ reactor coolant system can migrate or redistribute, and thus cause level changes which are at variance with those one would expect. This is a possible explanation for observed behavior in which lowering the RCS water level is followed by a level increase. Water in the RHR appears to be replaceo by air. Similarly, an increase in RCS water level that is fullowed by a decreasing level may be due to voids in the RHR system being replaced by RCS water. Failure to understand such behavior leads operators to mistrust level instrumentation and to perform cperational errors. (c) Operators typically will start another RHK pump if the operating pump is lost. Experience and an understanding of the phenomena clearly show that loss of the second pump should De expected. The cause of loss of the first pump should be well understood und normally should be corrected before attempting to run another RHR pump. (d) Typical operation while the RCS is ncrtially filled provides a high RHR flow rate, which ny be required by TS, but which may be unnecessary under the unique conditions assuciated with the partially filled RCS. Air ingestion problems are less at low +1cw rates. (4) Only limited instrumentation may be available to the operator while the RCS is partially filled. L I \\ l u_
l , (a) Level indication is many times available only in containment via a Tygon tube. Some plants provide one or more level indications in the control room, and additionally provide level alarms. (b) Typically, RHR systen temperature indication is the only temperature provided to the operators. Loss of RHR leaves the operator with no RCS temperature indication. This can result in a TS violation, as occurred at Diablo Canyon on April 10 when the plant entered Mode 4, unknown to the operators, with the containment equipment hatch removad. It also resulted in failure to recognize the seriousness of the heatup rcte, or that boiling had initiatad. (c) RHR pump motor current and flow rata may not be alarmed and scales may not be suitable for operation with a partially filled RCS. (d) RHR suction and discharge pressures may not be alarmed and scales may not be suitable for operation with a partially filled RCS. (5) L'icensees typically conduct operatione, whila the RCS is partially filled, the containment equipment hatch has been removed, and operations are in progress which impact the ability to isolate containment.
- P1snning, procedures, and training do not address containment closure in response te loss of RhR or care damage events.
This is inconsistent with the sensitivity associated with partially filled RCS operation and the history of loss of RHR under this operating condition. (6) Licensees typically conduct test and maintenance operations that can perturb the RCS and RHR system while in a partially filled RCS condition. The sensitivity of the operation and the historical record indicate this is not prudent. 1 I
l Table 1 37 LOSS-OF-OHR* EVENTS ATTRIBUTED TO INADEQUATE RCS LEVEL Docket Plant Date Ourstion Heatup 344 Trojan 05/21/77 55 min. Unknown 03/25/78 10 min. Unknown 10 min. Unknown 04/17/78 Unknown Unknown 334 Beaver Valley 1 09/04/78 60 min. 145 175*F 366 Millstone 2 03/04/79 Unknown 150-208*f 272 Salem 1 06/30/79 34 min. Unknown 334 Beaver Valley 1 01/17/80 Unknown Unknown 04/08/80 35 min. None 04/11/60 70 min. 101-108*F 03/05/81 54 min. 102 163*F 344 Trojan-06/26/81 75 min. 140-150 F 369 McGuire 1 03/02/82 50 min. 105-130 F l 339 North Anna 2 07/30/82 46 min. Unknown 338 North Anna 1 10/19/82 36 min. Unknown 10/20/82 33 min. Unknown 369 McGuire 1 04/05/83 Unknown Unknown 339 North Anna 2 05/03/83 Unknown Unknown 05/20/82 8 min. Unknown 26 min. Unknown 60 min. Unknown 280 Surry 1 05/17/83 Unknown Unknown 328 Sequoyah 2 08/06/83 77 min. 103-19 W 370 McGuire 2 12/31/83 43 min. Unknowr 01/09/84 62 min. Unknowt. 344 Trojan 05/04/84 40 min. 105-201*F 316 DC Cock 2 05/21/84 25 min. Unknbwn r 368 ANO-2 08/29/84 35 min. 140-205 F 295 Zion-1 09/14/84 45 min < 110-147 F 339 North Anna 2 10/16/84 120 min. Unknown 413 Catawba 1 04/22/85 81 min. 140-175 F 327 Sequoyah 1 10/09/85 43 min. <1 F 296 Zion 2 12/14/85 75 min.
- 15*
361-San Onofre 2 03/26/86 49 min. 114-210 F 382 Waterford 3 07/14/86 221 min. 138-175 F 327 .Sequoyah 1 01/28/87 90 min. 95-115 F 323 Diablo Canyon 2 04/10/87 85 min. 100-220 F
- Decay heat removal a
1 l L- --- - --- _- -
L ist et RECtNTLv Itsuto stNtalc ;tTTEas $7asetE Date of Letter No, Swelect Itasaats leswet to l 16 87-12 50.2414) Lttitt RE. LC$5 0F 07/C9/07 ALL LICENBEtt REttDUAL NEAT RtmovAs t 4%) Or CPERATING DLRtN3 MID-600* CPERATICN PwRS AND HOLCass et CCNSTRVOTION PERMITS FOR PWRS 3. 87-14 htLARAT!DN IN 149tTRARY CA/23/87 ALL t*ERATIN3 INTimmt01 ATE PIPE RvPvuht LICENSEtt. nEgytREMENTa CONstRutt!ON PERMIT KOLDraw, AND APPLICANTS FCR CCNSTRUCY1CN PEmmtT5 SL 57-10 !MPLEMEtOATION OF 10 CFR 04/12/87 AL. 80wt4 73.57, NEQu! REM 64To FOR Ft1 REACTOR CRIMINAL MISTORY CHECKS LICENSEE 5 et p5 09 SECfl0NS 3.0 ANP 4.0 0F Twt 04/04/87 ALL L10HT STANDARD TECHNICA. wATEM REACTOR SPECIFICAf!Cns ON THE LICEN$tES AND (_ APPLICASILITD OF LCO AND APPLICANTS SURVEILLANCE REQUIREMENTS GL 87-00 tMPL1=ENTAT30N OF 10 CFR 73.53 C5/11/97 ALL power MISCELLANECUS AMEN 0 FELTS 4440 REACTOR 14 ARCH REQUIREMCN15 LICENSEES DL 67-07 INF ORMAT ION ThANSMIftAL OF 03/19/87 ALL FACILtTY F lf*AL 8"JLEMA8(ING FOR REV!610N9 LICENSEES TO O*ERATOR a ICENstNO-10CFPS3 AND CohFORMjhde AMENOMENTS GL 37-04 ftSTING OF PRESSURE IS0teTION 03/13/87 Aut. OPuMATING VALVES REACYOR LICENSEES 46 87-05 REQiXST #N ADDITIONAL 03/12/07 LICENSEE 5 0' 1M'ORMAth0N-ASSESSMENT OF OR' $. LICENSEE MEA $vHED TO MtTIGATE APPLICANTS FLM AND/OR IDENTIFv POTENTIAL OL'5. Ano DE*mADATION MK! HOLDERS OF CP's FOR Bwn MARM ! CONTAINMENTS OL 87*04 TdMPORARY EltMPTION
- ROM 03/04/07 ALi, POwLR PRov!SIONS OF THE FBI CRIMINAL REACTOR HISTORY RULE P084 TE PtPOR ARY LICEW.ES m ERs UNITED STATES NUCLEAR REGULATORY COMMISSION Po7TUSD Ulio US*":
WASHINGTON. D.C 20555 W A SM DC PsawiT N. om OFFiflAL BUSINESS PEN ALTV FOR PFUVATE USE. 5300 (
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- g"o UNITED STATES
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- g NUCLEAR REGULATORY COMMIS310N g
wasmoTom.o.c. osss y ,k oct.ober 17, 1998 L TO ALL HOLDERS OF OPERATING LICENSES OR CONSTRUCT 10h PERM 115 FOR PRESSUR! ZED WATER REACTORS (PWRs)
SUBJECT:
LOSS OF DECAY HEAT REMOVAL (GENERIC LETTER N04 88-17) 10 CFR 50.54(f) Loss of decay heat removal (DHR) during nonpower operation and the consequences of such a loss have been of increasing concern for years. Numerous industry and NRC publications have addressed the subject. The Diablo Canyon event of April 10, 1987, and ensuing work by both the staff and industry organizations have provided additional insight. Yet the problems continue, as illustrated by t (1) the inadequacies demonstrated by many licensees in their response to Generic Letter (GL) 87-12; (2) the event at Waterford on May 12, 19E8; (3) the event at Seouoyah on May 23, 1988; (4) the D4R perturbations due to inadequate level at San Onofre on July 7, 198A; and (5) the apparent lack of a complete i industry understanding of the potential seriousness of such events. The report of the Diablo Canyon event, NUREG-1269, stated that operating a plant with a reduced reactor cc,4 nt system (RCS) inventory was a particularly sensitive condition and identifm many generir weaknesses in DHR. GL B7-12, which requested information from all PWR licensees, provided additional in-sight, and NUREG-1269 was transmitted with the generic letter to ensare that licensees had the latest information. Despite this, many of the responders to GL 87-12 demonstrated that they did r.ot understand the identified problems. H Deficiencies exist in procedt;res, hardware, and training in the dreas of (1) prevention of accident,nitiation., (2) mitigation of accidents before they potertially progress to core damage, and (3) control of radicactive material if a core damage accident should occur. Although deficiencias exist in all PWRs, certain design features make initiation and the time available for mitication in the Westinghouse and Combustion Engineering designs of more concern than in the nuclear steam supply systems (NSSSs) designed by Babcock and Wilcox. Nevertheless, we believe expeditious actions are necessary at all PWRs to rectify these deficiencies. These should be paralleled by programmed enhance-ments which supplement, add to, or replace the expeditious actions to accom-plish a more comprehensive improvement. Recommendations covering these iteas are summarized in the attachment, and additional information and guidance are provided in the three enclosures. u 8810130350*
4 2 Pursuant to 10 CFR 50.54(f), we request your response regarding your plans with respect to each of the recommendations as related to operation following placement of the HSSS on shutdown cooling, or following the attainment of NSSS 1 conditions under which shutdown cooling would normal'y be initiated. Your [ response is to include the following: (1) A description of th0 actions you have taken to implenent each of the eight recommended exceditious actinns identified in the attachment. Your rep'y shall be submitted to us within 60 days of recaipt of this letter. (2) A description of enhancements, specific plans, and o schedule for implo-nentatior, for ea:h of the six programned enhancement recommenda tions identified in the attachment. Your reply shall be provided to us within 90 days of receipt of this letter. Individual deviations from the recommendations will be ccasidered en a case by case basis provided compensatory measures are provided which will achieve a compareble level of protectica. No further responses are required to GL 87-12 and licensees or construction permit holders need not provide any supplemental information in a response to GL 87-1? to which they previously committed. We will-accept documents such as technical reports, action plans, and schedules prepared by industry groups when accompanied of commitments f rom participating licensees in lieu of individual documents from those licensees. Alternatively, such industry group documents may be incorporated by reference in licensee documentation. We ericourage your participation in cocoerative efforts to effecitvely resolve these issues. Your written response shall te submitted under oath or affirmation under the provisions of Section 182a, Atomic Energy Act of 1954, as amended. Your written response is needed to determine whether actions to modify, suspend, or revoke your license are necessary. An analysis as required by 19 CFR 50.109 s has been performed regaraing this request. The original copy of your written responsa shall be transmU ted to the U. S. Nuclear Regulatory Concission, Dccument Control Desk, Washington, D.C. 20555 for reproduction and distribution. This request is covered by Office of Management and Budget Clearance !! umber 3150-0011 which expires Decenter 31, 1989. The estimated average burden hours is 200 person-hours per licensee response, including assessment of the new requirements, searching data sources, gathering and analyzing the data, and peeparino the required reports. Cornents on the accuracy of this estimate and suggestions to reduce the burden may be directed to the Office of Management and Budget, Room 3208, Ncw Executive Office Euilding, Washington, D.C.
- 20503, ar.d to the U. S. Nuclear Regulatory Commission _ Records and Reports Managenent Branch, Office of Administration and Resources Management, bashington, D.C.
20555. l
If you have technical questions regarding this matter please contact Wayne j i Hodges at 301-492-0895. Other questions reay he directed to the NRR Project Manager assigned to this issue, Charles M. Trannell (301492-3121) or to the Project Manager assigned to your plant. p5/ lb Yrf Dennis M. Crutc tle Acting Associate 01 ector for Projects Office of Nuclear Reacto F.egulation
Attachment:
Recoreended Actions
Enclosures:
.g, 1. Overview and Background Information Pertinent to Generic Letter 88-17 2. Guidance for tieeting Generic Letter 88-17 3. Abbreviations and Definitions
'rk
s 1 1 LIST OF RECENTLY ISSUED GENERIC LETTERS Generic Date of Letter No. Sub,iect Issuance issued Te 88-16 REMOVAL OF CYCLE-SPECIFIC 10/04/88 ALL POWER REACTOR DARAMETER LIMITS FROM LICENSEES AND TECHNICAL SPECIFICATIONS APPLICANTS 88-15 ELECTRIC POWER SYSTEMS - 09/12/88 ALL POWER REACTOR INADEQUATE CONTROL OVER LICENSEES AND DESIGN PROCESSES APPLICANTS 88-14 INSTRUMENT AIR SUPPLY 08/08/88 ALL HOLDERS OF SYSTEM PROBLEMS AFFECTING OPERATING LICENSES SAFETY-RELATED EQUIPMENT OR CONSTRUCTION PERMITS FOR NUCLEAR POWER REACTORS 88-13 OPERATOR LICENSING 08/08/88 ALL POWER REACTOR EXAMINATIONS LICENSEES AND APPLICANTS FOR AN CPERATING LICENSE. 88-12 REMOVAL OF FIRE PROTECTION 08/02/88 ALL POWER REACTOR REQUIREMENTS FROM TECHNICAL LICENSEES AND SPECIFICATIONS APPLICANTS 88-11 NRC POSITION ON RADIATION 07/12/88 ALL LICENSEES OF EMBRITTLEMENT OF REACTOR OPERATING REACTORS VESSEL MATERIALS AND ITS AND HOLDERS OF IMPACT ON PLANT OPERATIONS CONSTRUCTION PEPMITS 88-10 PURCHASE OF GSA APPPOVED 07/01/88 ALL POWER REACTOR SECURITY CONTAINERS LICENSEES AND HOLDERS OF PART 95 APPROVALS 88-09 PILOT TESTING 0F FUNDAMENTALS 05/17/88 ALL LICENSEES OF ALL EXAMINATION BOILING WATER REACTORS AND APPLICANTS FOR A BOILING WATER REACTOR OPERATOR'S LJCENSE UNDER 10 CFR PART 55 l 88-08 MAIL SENT OR CELIVERED TO 05/03/88 ALL LICENSEES FOR POWER THE OFFICE OF NUCLEAR REACTOR AND NON-POWER REACTORS REGULAT. ION AND HOLDERS OF 1 CONSTRUCTION PERMITS FOR NUCLEAR POWER REACTORS i
"i ATTACHMENT TO GENERIC LETTER RECOMidENDEO ACTIONS Expeditious actions and progranned enhancements are recommended concerning operation of the NSSS during shutdown coolino or during conditions where such cooling would normally be provided. The reconmendations apply whenever there is irradiated fuel in the reactor vessel (RV). These recommendations are summarized below and discussed further in enclosure 2: Expeditious actions: The following expeditious actions should be 1mplementeo prior to operat-ing in a reduced inventory condition *: -P (1) Discuss the Diablo Canyon event, related events, lessons learned, and iniplicatiers with appropriate plant personnel. Provide training shortly before entering a reduced inventory condition. (2) Implement procedures and administration controls that reasonably assure that containment closure ** will be achieved prior to the time at which a core uncovery could result from a loss of DHR coupled with an inacility to initiate alternate cooling or addition of water to the RCS inventory. Containment closure procedures should include consideration of potential steam and radioactive material release from the RCS should closure activities extend into the time boiling takes place within the RCS. These proteaures and administrative controls should be active and in use: (a) prior to entering a reduced RCS inventory condition for NSSSs supplied by Combustion Engineerine or Westinghouse, and (b) prior to entering an RCS condition wherein the water level is lower than four inches below the too of the flow area of the hot legs at the junction of the hot legs to the RV for NSSSs supplied by Babcock and Wilcox, and should apply whenever operating in those conditions. If such procedures and administrative controls are not operational, then either do not enter the applicable condition or maintain a closed containment. A reduced inventory condition exists whenever RV water level is lower than three feet below the RV flange. Containment closure is defined as a containment condition where at least one integral barrter to the rele2se of radioactive material is orovided. Further discussion and qualifications which the intearal barrier must meet are provided ir, enclosure 2 and in the definitinns provided in enclosure 3. i-em = - - m
e .+. ( ' ? I (3) Provide at least two independent, csntinuous temperature indications thet are representathe of the core exit conditions whenever the RCS is in a mid-loop condition
- and the reactor vessel head is located on top of the reactor vessel. Tcmperature indice.tions should be periodically checked and recorded by an operator or automatically and continuously monitored and alarned. Temperature monitoring should be performed either:
(a) by an operator in the control room (CR), or (b) from a location outside of the containment building with b. provision for providing immediate temperature values to an operator in the CR if significant changes occur. Observations should be recorded at an interval no greater than 15 minutes during normal conditions.** (4) Provide at leaeJ two independent, continuous RCS water level indica-tions wheneve n he RCS is in a reduced irventory ccndition. Water level indications shoulc be periodically checked and recorded by an operator or automatically and continuously monitored and alarmed. Water level monitoring should be capable of being performed either: (a) by an operator in the CR, or (b) from a location other than the CR with provision for providing irn:ediate water level values to an operator in the CR if significant changes occur. Observations should be recorded at an interval no greater than 15 minutes during normal condi-tions.** (5) Implement procedures and administrative controls that generally avoid operations that deliberately or knowincly lead to perturba-tions to the RCS and/or to systems tnat are necessary to maintain the RCS in a stable ard controlled conditien while the RCS is in a reduced inventory condition. If operations that could perturb the RCS or systems supporting the RCS must be conducted while in a reduced inventory condition, then ddditional measures should be taken to assure that the PCS will y. remain in a stable and controlled condition. Sach additional measures include both preventien of a loss of DHR and enhanced monitoring recuirements to ensure timely response to a loss of DMR should such a loss occur. A mid-loop condition exists whenever RCS water level is below the top of the flow area of the hot legs at the junction with the PV. Guidunre should be developed and provided to operatnrs that covers evacuation of the monitoring post. The cuidance should properly balance reactor zad personnel safety. (
3 w 3 t (6} Provide at least two available* or operable means of adding inventory to the RCS that are in addition to pumps that are a part of the normal DliR systems. These should include at least one high pressure injection pump. The water addition rate capable of being provided by i each of the means should be at least sufficient to keep the core covered. Procedures for use of tnese syste;ns during loss of DHR events should be provided. The path of water addition must be specified to assure the flow does not bypass the reactor vessel before exiting any opening in the RCS. (7) (applicable to Westinohouse an( Contustian Engineering nuclear steam supply system (NSSS) designs) Inplement procedures and administra-tive controls that reasonaoly assure that all hot legs are not blocked simultaneously by nozzle dams unless a vent path is provided that is large enough to prevent pressurization of the upper plenum of the RV. See references 1 and 2. (8) (applicable to NSSSs with loop stop valves) Implement prncedures and administrative controls that reasonably assure that all hot legs are not blocked simultaneously by closed stop valves unless a vent path is provided that is large enough to prevent pressurization of the RV upptr plenum or unless the RCS configuration prevents RV water loss if RV pressurization should occur. Closino ccid legs by noz?le dans does not meet this condition. Programmed enhancements: Programmed enhancements should be developed in parallel with the exredi-tious actions and they may replace, supplement, or add to the expediticus actions. For example, programmed enhancements may be used to change expeditious actions as a result of better understanding or improved This uy lessen the initial impact of expeditious actions procedures. such as the speed with which containment closure must be achieved and nay include consideration of such factors 8s the decay heat rate. Additional guidance is provided in enclosure 2. For example the first paragraph of ~ section 2.2.2 and the first paragraph of section 3.3.2 illustrate the flexibility we have in mind as 1cng as safety is adsquately addressed. We intend that prcgrammed enhancements be incorporated into plant opera-tions as they are developed when this results in significant safety improvement or enhancement of plant operations with no decrease in sa fe ty. Procedural and har>Jware 'rodifications may be implenented without prior staff approval where the criteria of 10 CFR 50.59 are -met, although it is our intent to review and/or audit such changes. Frograamed enhancements should be implemented as scon as is practical, but no later than the following schedule:
- Available means ready for use quickly enough to meet the intended functional need.
t e 4 I (1) Programmed enhancements consisting of hardware installation and/or modification, and programed enhancements that depend upon hardware installation and/or modification, should be implemented: (a) by the end of the first refueling outage that is initiated 18 l months or later following receipt of this letter, or (b) by the end of the second refueling outage frilowing receipt of this letter, whichever occurs first. If a shutdown for refueling has been initiated as of the date of receipt of this letter, that is tn he counted as the first refueling cutage. (2) Programed enhancements that do not depend upon hardware changes shculd be implemented within 18 months of receipt of this letter. We recommend you implement the following six proorammed enhancements: 1 (1) Instrumentation ) Provide reliable indication of parameters that dest. ribe the st6tc of the RCS and the performance of systems normally used to cool the RCS l for both normal and accident conditions. At a minimum, provide the following in the CR: (a) two independent RCS level indications (b) at least two independent temperature measurerents representa-tive of the core exit whenever the RV head is located on top of the RV (We suggest that temperature indications be previded at all times.) (c) the capability of continuous 1v monitoring DHR system perfor-mance whenever a DHR systen is being used for cooling the RCS (d) visible and audible indications of abnormal conditions in temperature, level, and DHR system performance (2) Procedures Develor and implement procedures that cover reduced inventory operation and that provide an adequate basis for entry into a reduced inventory condition. These include: -(a) procedures that cover normal operation of the NSSS, the con-tainment, and supporting tystems under renditions for which cooling would normally be provided by DPR systems.
t a c (b) procedures that cover emergency, abnormal, off-normal, or the equivalent operation of the NSSS,. the containment, and support-ing systems if an off-normal condition occurs while operating under conditions for which cooling would normally be provided by DHR systems. (c) administrative controls that support and supplement the proce-dures in items (a), (b), and all other actions identified in this communication, as appropriate. (3) Equipment (a, Assure that adeovate operating, operable, and/or available 1 equipment of high reliebt11ty* is provided for cooling the RCS and for avoiding a loss of RCS cooling. ~ (b) Maintain sufficient existing equipment in an operable or available status so as to mitigate loss of CHR or loss of RCS inventory should they occur. This should include at least one high pressure injection pump and one other system. The water addition rate capable of being provided by each equipment item should be at least sufficient to keep the core covered. f (c) Provide adequate equipment for personnel communications that i i involve activities related to the RCS or systems necessary to ~ maintain the RCC in a stable and controlled conditicn. (4) Analyses Conduct analyses to supplement existing information and develop a basis fnr procedures, instrumentation installation and response, and equipment /flSSS interactions and response. The analyses should j enecmpass thermodynamic and physical (configuration) states to which the hardware can be sub.iected and should provide sufficient depth that the ba~ sis is developed. Emphasis should be placed upon obtain-ing a complete understanding of NSSS behavior under ncnpower opera-tion. (5) Technical Specifications Technical specifications (TSs) that restrict or limit the safety benefit of the actions identified in this letter should be identi-fied and appropriate changes should be submitted. l
- Reliable equipment is equipment that can be reasonably expected to perform the intended function. See Enclosure 2 for additional information.
l l
r -.n. 6 (6),P.CS perturbations Item (5) of the expeditious actions should be reexamined and opera-tions refined as necessary to reasonably minimize the luelihood of loss of DHR, Additional information and guidcnce are given in enclosure 2. REFERENCES (1) C E. Rossi, "Possible sudden loss of RCS Inventory during Low Coolant Level Operation," NRC Information Notice 88-36, June 8, 1988, (2) R. A. Newton, " Westinghouse Owners Group Early Notification of Mid-loop Operation Concerns," Letter from Chairman of Westinghouse Owners Group to Westinghouse Owners Group Prinary Representatives (IL, IA), OG-88-21, May 27, 1988. -.v-
[ -{ o L i ENCLOSURE 1 TO GENERIC LETTER i OVERVIEW AND BACKGROUND INF0F.NAT10N PERTINENT TO GEttERIC LETTER 88-17 l 1 i 4
? + 5 C0hTENTS Pace 3 1.0 THE IS?UE........................... 4 2.0 PERSPECTIVE.......................... 4 2.1 Phenomena and Impact................... 5 2.1.1 Pressurirstion.................. 5 2.1.2 Vortexing..................... 6 2.1.3 SG tube draining i jn plants with "U" tube SGs 6 2.1.4 RCS level differences............... 7 2.1.5 DHR systen effects................ 7 2.1.6 Instrumentation 8 2.2 Time available for mitigation.............. 9 2.3 Generic Letter 87-12 review............... 3.0 NEEDED RESPONSE 10 10
4.0 REFERENCES
i l
t s \\* 3 l.0 Tile IS5tJE Concern has been increasing for some time that an event involving the loss of decay heat removal (DHR)* while there is substantial core decay heat may pose a i ( significant likelihood of a release due to a severe core damage accident. Recently obtained probabilistic risk information and a survey of industry operations substantiate this concern, independent engineering evaluation of plant operation while cooling is provided by DHR systcms lands to a similar conclusion. Consideration of plant behavior points out several phenomena that had previously gone unrecognized and that potentially could lead to severe core damage in approximately one hour rather than in the previously believed conservative time of snore than tour hours. Plants are operating under conditions that have not been analyzed and in which plant response is not L understood. ? Evaltation of plant dats $ bows that an unacceptably large number of events have occurred and continue to occur. If not mr igated, such events lead to core damage. Many of these events have involvN : lost of DHR for one or more hours. A number of events have resulted N hos. mg in the core, a condition that has not been -analyzed at most plants. Often, plant personnel were unaware of the real difficulty for some time during or after the event. Experience t clearly substantiates that a problen exists. Information obtainec :,ir.ce the Diablo Canyon event of April 10, 1987 shows that many previously unrecognized r"echanisms exist that eracerbate the prc'- a lem, and that are rot. represented in tne FRA results. Some of these can realistically cause core uncovery or complete core voiding in less than half an hour; significantly less than the previously believed " conservative' boil down of water with uncovery of the top of the core in four hours. Our review of lisensee responses to Generic Letter (GL) 87-12" and plant experience has clearly established that few procedures exist to avoid these scenarios. Many licensees demonstrated in the GL 87-12 responses that they were not even aware that such scenarios exist. Review of industry responses to GL 07-12 shows that most licensees are poorly d_ prepa.'ed for reduced PCS inventory operation. Procedures are incomplete, ir. correct, er r.onexistent. Little effective thr.unht has been given to avoiding the initiation of accidents or to nitigatine an eccident once it has begen, i l
- Enclosure 3 provides a list of abbreviations and definitions.
- GL 87-1? Iref. 1) requested licensees to describe operation of their plants under conditions where sore of the water inventory bad been rencved from the reactor conlant system (ECS). We have completed our review of these retoonses with assistance from the Idaho National Engineering Laboratory, and a NUP.EG/CR docur.ent describing that review is beira prepared. Further infor-mation is provided in Section 2,2 of this erclosure.
F F
t 4 The inability of containment to mitigate an accident is seldom addressed at any le<el of operating procedures, administrative cont *ols, or training. Instrumentation is often of low quality or inat. curate, and little provision is mode for using equipment effectively. The responses establish that the problem is extensive, many disciplines are involved, many licensees are not adequately responding, and information is not being effectively shared within the industry. 2.0 PERSPECTIVE 2.1 Phenomeaa and Impact A number of phenomena have been recognized as affecting nuclear power plant operatton wner ' m e plants are operating in a nonpower condition. Sorne of these pheno % 4 c-cause the tine between loss of DHR &nd severe cord damage to be as shv c approximately one hour. Such phenorrena also cause instru-mentation errors, loss of DHR, and unstable operation. These phenornena are of particular concern at operating conditions where the water level is below the top of the hot and cold legs. This level permits air to be distributed throughout the RCS. This complicates interpretation of the event, in addi-tion, the al bwable operating band for water level is often only a few inches (too low, and DHR is lost; too high, and steam generator (SG) tubes do not drain or water floods the SGs and containment). This is a challenging environment for the operstors, and one with a high probabili,ty of failure. For example: (1) The actual state of the RCS may differ from the analyzed state, and phenonena may occur that have been neither recognized nor analyzed. This can lead to RCS behavior that operators and advista do not anticipate. Of serious concern is the discovery of accident sequences that can cause core uncovery or complete core voiding in 15 or ?O minutes and severe core damage in approximately an hour from the time DHR is lost. (0) Operators and advisnrs may not recognize the potential seriousness of the situation until unanticipated pher.omena becwe obvious. Corrective action may be further delayed because operators and advisors disbelieve the symptoms as indicateo by available instrumentation. '(3) Changes in FCS state may cause y Sble mitigation paths to be unavailable. l (4) Failure to recognize the potential seriousness of the situation and lack cf clear, appropriate procedures can lead to significant delay in obtain-ing resources needed to cope with the event. We discuss a number of phenomena and reltted concerns in the subsections that follow. Although incomplete, these distassinrs will help to illustrate the j L magnitude and breadth of the issue. We will discuss: l (1) pressurization (2) vortexing l I3) SG tube draining in plants with U-tube SGs I t
s 5 (4) RCS level differences (F) DHR system effects (6) instrumentation 2.1.1 Presserization The principal concern is that a snall pressurization can occur as a result of conditions unique to operation with a reduced RCS inventory - and this pres-sure increase can seriously affect plant safety. Previously at least four hours are believed to be available between loss of DHR and core uncovery. lie now know that these newly appreciated phenomena can cause core uncovery or complete core voiding in 15 or 20 'ninutes and severe core damace in approri-mately an hour following loss of OM. A number of considerations are applicable (refs. 1 - 4), including: (1) Inappropriate use of SG nozzle dains can had to complete core voidino within 15 or 20 minutes of loss of DHR. A similar phenomenon can occur wwhen loop stop valves are inappropriately used. (2) Cold leg openings can allow water to be ejected from the vessel 'ollowing loss of DHR until suf ficient water is lost that steam is relieved by clearing t,f the crossover pipes. (3) Phenomena associated with pressure differences within the RCS may prevent injection water f rom reaching the reactor vessel (RV). (4) Rapid RCS pressurhotion may prevent gravity feed of water f rom tanks that are a: icipated to be available. I (5) Rapid pressuri/stion may cause instruments to malfunction or provide misleading indications. (6) Papid pressurization may cause the RCS to respond in unanticipated ways. (7) St all RCS pressure boundary openings at various lccations (vents and drains both above and below the water level) may lead to instrument malfunctions or unanticipated RCS responses. (8) Large RCS pressure boundary openings at various locations (SG manways, reactor coolant pump (RCP) bowl, loop stop valves, pressurizer manways) nay lead to instrument nalfunctions or unanticipated RCS respcoses. (9) Steam generator secondary side inventory and openinas tray influer.ce RCS behavior. 2.1.? Vortey i rig Vortexino at the,iunction of the DHR systen suction line and the RCS will occur if water level is too low, a situation to be avoided cince this may introduce air into the DHR pumo suction. Small amour.ts of air nay lead to subtle changes that occur over a tirre of minutes to an hour or more, and may I i = 7
r o! 6 propagate to loss of DHR. Large amounts of air may cause immediate loss of pump suction and hence loss of DHR. Vortexing inay occur at levels higher than i anticipated. For example, vortexing may initiate at the level required *to drain SG tubes or if initiated, may continue while at a level where vortexing may not ordinarily initiate. This can lead to operation with unrecognized ~uch vortexing and air vortexing and suction of air into the DHR system. entrainment may not be reflected by pump curre.t and flow rate instrumentation i until it is sufficiently severe that continued operation of the DHR system is jeopardized. As discussed in reference 4 even when vortexing is insufficient to perturb OhR system operation, it may upset the RCS level and level indica. tions and lead to inappropriate operator actions. For :txample, the operators were controlling RCS level at Diablo Canyon to the range of 107'0" to 107'8" i.nmediately before-the April 10, 1987 event, and the3 had drained the RCS to 107'0" before the event to stay within this band. D @ was lost when the instrureentation registered about 107'4". The Diablo Canyon-licensee later reported to us that vortexing begins to occur at 107'5.5" and is fully developed at 107'3.5" with an RHR flow rate of 3000 gpm (the technical specification (TS) requirement at Diablo Canyon at the time of the event). This vortexing behavior was not understood on April 10. 2.1.3 Steam Generator Tube Draining in plants Equipped With Il-tube Steam Generators Operators frequently drain the RCS to the vicinity of vortexing to drain SG tubes.- For example, the RCS was drained to an elevation below 107'5.5" Itop of the pressurizer surge line) to drain SG tubes at Diablo Canyon before the ~ April 10 event. Vortexing was later reported to initiate at 107'S.5". (See Appendix C Sf reference 4 for additional information.) Alternate' approaches exist to draining of SG tubes. These include: (1) Introduce nitrogen via instrument connections located below the SG plena. e This may allow draining of SG tubes with most of tne remainder of the RCS full. (2) Provide nitrogen directly into the SG plena. This may also allow drain-ira of SG tubes with most of the remainder of the RCS full. (3) Use-nitrogen from the RV to drain SG tubes. This often can be done at a higher RCS level than required to drain with nitrogen from the pressuri:- er. 7.1.4 Reactor Coolant System t.evel Differences When operating under mid-loop conditions, the critical level parameter is water level in-the het leo essentially at the junction with the DPR-systen i suction line. The significance of-this is often unrecoonized in connectivo level instrumentation anc in operation. Yet a change in level of only a few inches can cause loss of DHR, and unrecognized and/or unanalyzed pheromena-are more than' sufficient to provide such a change. For example. dif f erences exist between actual level at the suction line and the indicated level because of-such effects as: ---)
4 7 (l) Flow from the injection point to the suction connection will cause a level change between these locations because a d*iving force is necessary to accomplish the flow. The level difference will not be discovered if instrumentation is not independent nor will it be found by calibration between shutdown level instrumentation and the pres:urizer level instru. mentation. (2) RHR return water momentum will result in a level bulldup. 'This will not be found by cross checks between the shutdown lcvel instruments and pressurizer level instrurnentation. Additional information is provided in references 3 and 4. i 2.1.5 Decay Heat Eenoval System Effects DHR systems in a plant are seldom identical. Even changeover from one DHR system to another may result in loss of DHR due to minor differences in the systems, Changeover from one DHR system to the other can also cause a lost o' DHR if it is improperly performed. For example, starting one DHR systen wh'le the other is running will increase flow rate, and can lead to entrainment of sufficient air to cause both DHR systems to be lost. The erfect can occu? as a result of: (1). The increased DHR systen flow rate can cause an increr - ortexing at single drop line plants. (2) Tfie increased flow rate can lead to a decreased level in the upper vessel . and hot legs in plants equipped with one or more drop lines. This can occur because most of the pressure drop occurs between waur injection locations and the hot legs, most of which is a conaon flow path and hence is affected by total flow rate; and by moving RCS inventory into a DHR systen that was initially only partially filled. Arother problem axists with operator response to loss Of a DHR system, if the loss were due to RCS conditions, the conditions may be such that it is likely other DHR system pumps also will be lost if they are started without correct-ing the cause of the initial loss. Shutting off or starting a DHR systen n'ay be followed by a chance in RCS inventory (1) if DHR piping drains into the RCS, (2) if air in the DHR systen is displaced by weter from the RCS, or (3) if air in the RCS is ditplaced by Wdter from the DHR systems. Similar behavior occurs when air ingestion is occurring and there is an increase or decrease in vortexing. Such a vortexing effect may oc;ur when RHR flow rate changes, when FCS inventory is chanced, or when inventory is transferred between systens as a result of the identified effect. 2.1.6 Instrumentation Instrunentation used for level indication needs careful analysis, installatien, %d prottction from damage or changes which may influerre instrumentation indication. Level indications mcy easily be in er ror by half a foot or more. Curther, conrection schemes, flow dyn vics. entrappeo air, or pressurization m.. .1..m..m.i am
P 9-8 may significantly ed simultaneously affect all level instrumentation during operation with a lowered RCS inventory. These contribute to the mis-diagnosis of events and inappropriate operator response, which may ext 'trbate the problem. Inaccerate level indication has often led to or contributed to loss of i. DHR. Many phenomena affect the instrumentation and should be considered in instru-ment design and installation as well as during plant operation. Feilure to do so can lead to r:isunderstood level instrumentation response, operator mistrust of instrumentatier., and inappropriate operator actions. Another instrument related problem is the limiting of operator information by the common practice of disconnecting instrumentation in preparation for removing the RV head and for other cperations commonly conducter' during a refueling outagn. Frequently, thermoccJples in the RV will be disconnected well before the RV head is lifted. Remaining resistance temperature device (PTD) instrumentation in the mcnifolds (typical of many plants) or the hot and cold 17js will not reflect vessel temperatures in a loss of CHR s3 sten flow situation even if they are available, and DHR system temperature indicatico is meaningless if the DHR system pumps are inoperative. 2.2 Time Available for Hitication The traditional approach to determining system response has been to ceaserve-tively calculate the time to uncover the core by assuming that RCS inventory beats to the boiling point and that the inventory is then boiled away. This typically has been calculated to take.four hours. This traditional approach is nonconservative. Boiling initiated at Diablo Canyon in 30 to 45 minutes following loss of DHR in the April 10, 1967 event. More importantly, this boiling caused RCS presserization, an unanticipated condition. A differert RCS configuration, such as blocked hot legs and an opening in the cold legs, could have quickly led to core uncovery following initiation of hoiling, an unantidpated situation. Further, the loss of DHR at Diablo Cenyon occurred at a low initial RCS temperature and with a decay heat generation rate less than half of that which could occur during 1 css of DDR accidents, Clearly, core uncovery can occur auch faster than oreviously believed, an occurrence the Westinohouse Owners Group (WOG) recently reported to West'ng. house owners (ref. 3), (The WOG report identifies boiling in less than 10 minutes.) " vere core damage can follow as soon as adiabatic heatup of the core reaches the point of rapid chemical reaction. There are two i.vertant conclesions: (1) The tir'e available for operators to respond to a loss of DHR can be far less than was previously believed. Irrmediate actions are necessary to reasonably assure an adecuate operator response curing tuch conditions. (2) This situation rcestituteJ a previously unanalyzed plant condition that can realistically be encountered. Generic Letter 88-17 provides guidance in rcrrecting this situation. l
p A 1 l *) l' 9 G,eneric letter 87-12 Review 2,3 e GL 8 M 2 (ref. 1) eas transmitted tn'all licensees and holders of construction permits for PWRs. It requesttd information pertinent to operation of nuclear power plants when the RCS inventory is below that required for normal opera-tion. Licensee responses were evaluated with respect to the following topics (1 interlocks (2 draindown operations (3 DHR cperations (4 SG considerations 5 test and maintenance operations i 6 RCS pressurization consiteratinns
- 7) containment considerations
- 8) irstrumentation and alarms (9) backup Rr5 cooling and makeup (10 analytic basis (11 training (12 Resources available to operator and the evaluations were conducted with consideration of such subjects as:
(1 understanding of issue (2 approach (3. adequacy. (4) procedurer and training (f) malfunction miticative response i The evaluation clearly established that most licensees did not t emonstrate i adequate preparation for reduced FCS inventory operation. The situation may be summarized as follows: (1) Accident initiation. The ma,ior reasons for such accidents is that industry has f ailed to adequately address the issue of operating the piants under ennditions of reduced RCS inventory. Plants are not well designed fnr reduced RCS inventory operation, plant behailor has not been adequately analyzed or understood, instrumentation is inadequate, and procedures sometimes are of poor quality or provide inadequate coverage. (2) Prrgression to core demace. Operators have been ill prepared fnr friti-cating an accicent once it has initiated. Operators are expected to recover the normal DPP system or to provide alternate cooling before the t cendition becomes serious. Yet, operators have not been given the tools to achieve this objective. (3) Ccnsequences. While the plant is in a reduced RCS inventury condition, Ticensees generally have their containtrent open, of ten with the equipment hatch removed. Many licenseet have given little thcught to closing the containment or to t ding other actions to mitigate the consequences of a core damage accident. ,<,,,,,,,.,,,,a
10 Son,e utilities have achieved a signifirant improvement in the past year, and are continuing to work on this issue. Those licensees best oualified to deal with loss of DHR during. lowered RCS inytntory conditions have active improvement programs. Further information on the review criteria, licensee responses, and rev;ew of licensee responses will be reported in a HUREG document within the next few months. 3.0 NEEDED RESPONSE Direct loss of DHR is an important initiator of accidents and its loss could cause a release of radioactive material due to a core damaue accident. The problem is exacerbated by weakness in procedures for restoration of core cooling, weakness in administrative controls, and by a large likelihood of failure to mitigate a release should the core be damaged. Actions to mirimize the initiation and consequences of loss of DHR take two forms: + (1) Expeditious or inmediate actions, which can be implemented quickly and at little direct cost, but thich may affect plant operations under some circumstances and cause an operational cost. These actions will signifi-cantly reduce the likelihood of a significant release of radioactive material for the potential core damage accidents of concern here. (2) Programmeo enhancements or longer term actions, which involve development of understhnding, procedures, training, and min.imal additional ' instrumentation. When implemented, these will modify some inmediate actions and may reduce impact on plant operations caused by the immediate actions, although other impacts may result in some plants. Expeditious actions will reduce the likelihood of a release due to a core damage accident. They will essentially assure the containment will be closed prior to the time significant core damage could occur if DHR is lost. Addi-tional benefits will ensue because the frequency of loss of DHR accidents will be reduced and operator response to such accidents will be improved. The longer tern programmed enhancements attack the root cause of accident initiation and provide enhanced mitigative response.
4.0 REFERENCES
" Loss of Residual Heat Penoval (PHR) while the Reactor (1) F. J. Miraglia,(RCS) is Partially Filled (Generic Letter 87-12)," Letter Coolent System to all licensees of operating PWRs and holders of construction permits for PURs, July 9,1987 (2) C. E. Rossi. "Possible Sudden Loss of RCS Inventory Durina low Coolant Level Operation," NRC Information Notice No. 88-30, June 8,1988.
5 0 11 's. (3) R. A. Newton, " Westinghouse Owners Group Mid-loop Operations Concerns," Letter to W. Hodges, NRC, from Chairman oI~ Westinghouse Owners Group, OG-88-24, June 20, 1988. Letter transmits R. A. Newton, " Westinghouse Owners Group, Early Notil'ication of hid-Loop Operation Concerns," Letter to Westinchouse Owners Group Primary Pepresentatives (IL,1A) from Chairman, Westinghouse Owners Group, OG-88-21, Nay 27, 1988 (4) U. S. Nuclear Regulatory Comission, " Loss of Residual Heat Removal System, Diablo Canyon, Unit 2, April 10, 1987," NUREG-1269, June 1987, i i s l l
1 e' . y ENCLOSURE 2 TO CENERIC LETTER GUIDANCE FOR HEETING GENERIC LETTER 88-17 f 0 1 1 r.- .-m m
o C 2 t CONTENTS Page 1.0 OVERVIEW 3 1.1 Introduction.......,,................. 3 1.2 Approach................... 4 3 2.0 GUIDANCE AND STAFF POSITION INFORMATION - EXPEDITIOUS ACTIONS... 5 2.1 Diablo Canyon' Event 5 - 2.2 Containment Closure 5 2.3 RCS Temperature 7 2.4 RCS Water. Level 8 2.5 RCS Perturbation....................... 9 2.6 RCS Inventory Addition............... 9 2.7 Nozzle Dams ..................10 l 2.8 Loop S top Va lv es...................... 10 3.0 PROGRtJHED ENHANCEMENTS.,................ 11 \\ 3.1 Instrumentation 11 - 3.2 Procedures......................... 14 L 3.3 Equipment ..................15 3.4 Ana lyses.......... ................, 16 3.5 Technical Specifications.................. 17 3.6 RCS Perturbations ....-..............-.17 t
4.0 REFERENCES
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m-i N 3 1.0 OVERVIEW 1.1 Introduction Events have occurred for years that jeopardite core cooling during nonpower operation. These events often have not been taken seriously because of the impression that the low heat generation rate associated with nonpower opera-tion allows considerable tine to restore core coolinn before core damage begins, and there is a wide range of means aveilable to the operators to restore core cooling. The general industry position seems to have been that the likelihood of a release of radioactive materiai due to a core damage 4 accident during nonpower cperation was so low as *.o be negligible when com-pared with the likelihood associated with full power operation. Significant new informatien has been generated within the past year, notably as a result of the Diablo Canyon event of April 10, 1987, the licensee's efforts following that event, and work conducted by the Festinghouse Owners s Group (WOG). ($ee, for example, refs. 1 + 7.) We now know that several previously unrccognized phenomena need to be addiessed. An ime(dlate response is necessary to deal with this new information. Generic Letter 88-17 requests information from each licensee of a pressurized water reactor (PWR) regarding the licensee response to this need. This enclosure provides irformation relative to the actions identified in the
- letter, lhe information it r.ot intended to cover all topics, nor does it represent the only selbit:ns we will accept in response to actions iden ified in the letter.
It should be used for cuidance. if better solutions ar< found than illustrated in the enclosure, they should be considered and discussed with us. Our initial objective is to obtain reasonable solutions ouickly. The next objective is to develop a more comprehensive solution which nay take longer to develop. Portions of the latter solution may already exist for some plants, and it may thus be feasible to implement some progranmed enhancements 1 on a schedule that meets the expeditious actions identified in GL 88-17. A number of terms are used in the material that follows that are unicue to this issur. Other terms will be more familiar, but the meaning may be more precise as applied to the DHR issue. We suggest you review tha definitions provided in Enclosur e 3 to avoid misunderstandings. 1.2 Approach We 6re using an approach that couples immediate response and a development program to achieve: (1) an irrediate reduction in the likelihood of a release of radioactive matcrial due to a core damage accident - which we call expeditious actions, and (2) a longer term reduction in core darace likelihood - defined as prograreed enhancements. ,-_.,,.-.~--,.m.-
4 The approach addresses the three key aspects which influence this issue: (1) Prevent accident initiators frora occurring. Although some aspects have been incorpo-This addresses the root cause. rated into expeditious actions when the effect on core damage likelihood is immediate and plant implications are understood, effective initiation Conse-rate reduction will require an extended effort at many plants. quently, initiation rate reduction is addressed in the programmed en-hancement recommendations. (2) If an acc1 dent initiates, provide inydepth mitigation capability to prevent core damage. Comprehensive mitigation planning is also a longer term tubject, and is addressed in the prograrped enhancement recommendations with some consid-eration provided in the expeditious actions. (3) Provide a closed containment before the core uncovers if a loss of DHR occurs. This is the primary rvraditious action because it con be implemented s effective protection agair.st a release. immediately and it pt ..e Control of accident initiation, mitigation of an initiated accident to prevent core-damage, and prevention of the release of radioactive material involve the following five topics which are important to safety: instrumentation procedures maintenance and testing l (4) equipment (5) analyses A sixth topic, technical specifications (T9s), will be affected by certain changes in the above. We have carefelly considered the unique aspects of nonpower oneration and We believe their implications using various methods of addressing the issues. that' flexibility in equipment celection and operation will be highly ef fective ender the less demanding physical conditions that exist during nonpower Consequently, with respect to the issue as addressed in GL 88-17, operation. we will accept the following for resolving the items identified in the letter: (1) Containment cinsure in lieu of the comparable power cperatier reouirement of containment isolation. (2) ' Relirble equipment in lieu of the comparable safety crade classification. (3). Realistic thermal-hydraulic and mechanica l analysis methods (with suit-able safety factors in a few situations) rather than the evaluation model-methods and multiple conservatisms that are otten used for evaluation of p;wer operation. I e y r r m---w -,.,.m e v-,- ,w ,,.,.,-,,,,.-g.,,,,-r...,..y,.- ,,,-y.m r t*-*-*'"rT-F-'#CM'F*'N1N"P'-'C-WW'-*FM'*
i 1 4 \\ 5 ~ (4) Peelistic equip nent response (with suitable safety f actors in a few s'.tuations) in lieu of conservativt assumptions. Various aspects of these approaches are discussed in the remainder of this i enclosure. P.0 GUIDANCE Al4D SYAFF POSITION INFORMATION - EXPEDI11005 ACT10fl5 P.1 Diablo Canyon Eve.nt 2.1.1 Recc mendation ] Discuss the Diablo Caryon event, related events. lessons learned, and impli-cations with appropriate plant perscnnel-Provide training shortly before entering a reduced inventory condition. 2.1.2 Discussion We celieve the lestons learned from the Diablo Canyon event are irrportant, and that all persennel involved in plant operations during DHP system operation conditions should be aware of the cvent and more importantly the significance, with emphasis upon knowledge and inright develcped as a result o' the event, j for example, Pow many plant personnel 3re aware that cold leg injection may be l ineffective under 50.9 '.futenwn condit hns, and that they shculd use hot leg injecticn to ef fective ly provide core cun*1ing undar tho e conditions? (See ref. 6.) .iany licensees accorrplished this recorr:endation within d f ew tr.ont h s of the Diablo Canyon event. However, recently developed insight is irportant and warrants coverage, ar.d was not covered during the early implementation of the i recomtrenda tion. The above illustration concerning effective watt.r injection is a good eXdmple - the knowledge was only reecntly disseminated on an industry-wide basis. 2.P Containment Closure 2.?.1 Recormende tion implement procedures and administrative centrols that reason.ibly essure that containment clo!ure will be achieved prior to the time at which a core uncovery i co+21d result fror a loss of DPR. These procedures anc edministrative controls should be active and in use: (a) prior to entering a reduced RCS inventr.ry conriition for nuclear steam supply systers (t:S$5s) supplied by Certtust:on Engineerino er vestinghouse, and (b) prior to entering an RCS condition wherein the water level is lower Lhan four iriches below the top of the. f low area of the hot legs at the ju1ctier of the bot leg', to the RV for f:FFS; supplita by Babcock and Wilcnx, w
t t b 6 and should apply whenever operating in those conditions. If such procedures and administrative centrols are not operational, then either do not enter the applicable condition or maintain a closed containment. 2.2.7 Discussion The expeditious action 4 tem addressing containment closure is a preliminary action that immediately and effectively reducts the likelihood of a release while providing the fic,1bility to have the containment building open under A wide range of times is available in which to close appropriate conditions. the containment building depending upon the state and configuration of the RCS. The expeditious action that we will accept in lieu of analytically determined times includes pretcribed times that reasonably assure containment closure in contpliance with the reconeendation. These times may be modified es soon as suitable analyses provide better estimates of the tire between loss of DHR and core uncovery. Although relaxation of times and other programmed enhancement developments nay relax containment closure actions, and may be implemented without staf f approval sub,iect to the provisions of 10 CFR 50.59, it is not our intention that containment closure provisions be eliminated. We recommend that containment closure considerations remain in effect whenever irradiated fuel is located in the RV unless the decay heat rate is so low that the fuel cannot overheat if completely voided of water. We will accept containment closure actions which include all of the following: (1) Containment c1csure is not necessary if tte reactor vessel (RV) and surrounding pool contain no irradiated fuel. Containment penetrations, including the eouipment hatch, nay remain open (?) provided closure is reasonably assured within 2.5 nours of initial 1055 of DHP - but see the time modifications which are discussed below for some configurations. Freraency procedures which require initiation cf closure activities.shculd be operational. Once initiated, closure activities may not be terminated until controlled and stable DHR has been restored ano the RCS bas been returred to a centrolled and stable-cordition. The followinc modifications should be met for nuclear stean supply systems (3) (NSSSs) supplied by Westingtouse (W) and Cctbustion Encineering (CE): (a) The ?.5 hour requirement in item ? is replaced by 30 minutes (W) or 45 minutes (CE} if openings totalino creater.than ene souare inch exist in the cold legs, reactor coolant pumps (RCps) (connectino into the cold leg water space) and crossover pipes of the RCS. This 30 or AC minute tinde requirement may be increased to two hours if a vent path from the upper RV is provided which is sufficiently large (with a suitable safety factor) that cnre uncovery cannot occur due to pressuri7ation resulting from boiling in the core. (4) As soon as suitable procedures and instrumentation are available and implemented, completion of containment closure followinn initiation of closure activities may be delayed, This may be done on the basis of reliable temperature information obtained durina a trarsient event provided the containment is closed prinr to reaching an RCS temperature , = = -.
7 of 200 F as displayed by the larger of two valid indications of temperature at the top of the core or imrediately above the core. The location of such temperature measurementJ should be at the approximate highest tenperature regions expected as a result of measure.nents obtained during normal power operation or should be representative of those locations. Reasonable assurance of containment closure should include consideration of activities which must be conducted in a harsh environment. For example, once boiling initiates in the RCS, a large volume of steam may be entering containment, potengially leading to high :ontainment temperature and increased i pressure. The 200 F temperature ddentif*.ed above provides assurance that containment is closed prior to the existet.:e of such conditions. There are several differences in the recommendations for different vendor designed NSS$s. These have been developed from differentec in operational history involving loss of DHR and f rom our appraisal of the implications of loss of CHR. For example, the B&W design is not sensitive to phenomena which can cause a pressure difference to develop between the hot and cold legs in the CE and W designs. Therefore, water is not forced from the RV due to a pressure i difference in the P&W design and the allcwable tires for containr:ent closure reflect this difference. Similarly, the specified water level at which containment closure procedures must be operational is lower in the B&W design than in the other two vendor designs because B&W does not encounter the draining dif ficulties, and the B&W operational history reflects less linelihood I of losing DHP systems. There are a number of other considerations which apply as well, including that B&W designs seldom involve.lowerino level to a value commonly used in the other designs, and there is lit',le question whether injection water will reach the core in the B&W design. 2.3 RCS Temperature 2.3.1 Recommendation Provide at least two independent, continuous temperature indications that are representative of the core exit conditions whenever the RCS is in a mid-loop condition and'the reactor vessel head is located on top of the reactor vessel. Temperature indications should be periodicall.y checked and recorded by an operator or automatically and continuously monitored end alarmed. Temperature monitoring should be perforred either: (a) by an operator in the control room (CR), or (b) from a location outside of the containment buildine with provision for providing irrediate temperature values to an operator in the CR if significant changes occur. Observations should be recorded at an interval no greater than 15 minutes under normal conditions.** i'GuidanEe should be developed and provided to operators that covers evacuation of the ronitoring post. The cuidanca should properly talance reactor and personnel safety. i . ~. ,.. _ _ -. _ _ _, _ _, _ _ _ _. _.. _.. _ _.. -. ~.. _ _ _ _. -, _... - - - - _.. _
m 8 e 2.3.2 Discussion The near tenn concerns are that boiling may force water from the RV and significantly decrease the time available between Inss of DHR and initiation of core damage, that operators should have a direct indication of the conditton of the RCS, and that operaters should be able to determine the effectiveress of actions taken in response to a loss of DHR. Temperature is the only variabic that can be measured that will directly track Although level can be used as an the approach to boiling in the CV. indication of the adequacy of cure coverage, often the avaliable range of 4 level indication does not correspond to the range for which inf ormation is necessary. lemperature can assist in bridging that gap. Temperature is also useful as an aid in determining the response necessary to a loss of DHr. Conscouently, we intend that temperature be provided to the operators over as wide a rar.ge of plant conditions as is feasible and for which its indication is valuable in. guiding operator actions. The region of most concern is when the RCS is in condition where inventory is liinor preturbations in RCS level may cause loss of DHR and temperature
- low, increase rate m 'th a low inventory will be f astcr than uncer other conditions.
coverage with respect to expeditious actions while the Consecuently, x minimun: RV head is located on top of the RV, we recommend that operations te conducted io minimize unavailability of temperature indication durino reduced RCS inventory nperation and that tennerature indication be provided whenever operating in a mid-loop cnnditier. 2.4 RCS Water Level 2.4.1 Recommendation Provide at least two independent, continuous RCS water level inoications Water level indications whenever the RCS is in a reduced inventory condition. should be perindically checked and rec 6rded by an operator or automatically and continuously monitored and alarmed. Water level monitoring should be capable of being performed either: (a) b,v an operator in the CR, nr from a location other than the CR with provision for providing fritediate (b) water level values to an operator in the CP if significant changes necur. Observations should be recorded at an interval no greater than 15 mir.utes during normal conditions.** 2.4.2 Discussion We believe reliable,-accurate RCS water level inforr'ation must be provided to H e operators whenever approacHng or operating in a, condition where a loss of leve; can-lead to loss of DHR. Level information is necessury under loss of
- Guidance should be developed and provided to operators that covers evacuattow of the monitoring post, iht guidance sFould-properly balance reactor and personomi safety.
_m.._- ' t d( 9 DHR ccoditions since it provides an indication of core coverage arid, if sufficient venting capacity exists, of the time to core uncovery, it is tiso useful in mitigation of a loss of PHR accident, n At a 4rlufmum, the low limit of the range of level indication must 53 belcw the level necessary for operation of DHR systems. Desirable is a low tirait that indicates level to the botton of the core. Where provision of two independent levtl indi;ations is not practical in the short term, we will accept d sinole irdicatica, flowever, these conditions are r unacceptable in the longer term, where we believe at least two b1 dependent 6 indications must be provided it' the CR. 2.5 RCS Perturbation 2.5.1 Recoceendation l Implement procedures andy/ administrative controls that generally avoid operations that deliberate'y or knowingly ? cad to perturbations to the RCS and/or to systems that are necessary to maintain the RCS in a stable and controlled conditier. while the RCS is in a reductd inventory condition. If operations that could perturb the ECS or systens supporting the RCS must be conducted while in a reduced inventory condition, then additirr.a1 reasures should be taken to assure that the RCS will remain in a staule and-controlled condition. Such additional measures include both prevention of a loss of DilR and e,nhanced monitorino requirements to ens'ure TIEely response ~ to a loss of DPE should such a loss occur. 2.5.2 Discussion This expeditious action item should eliminate a major cause of accident initia-tion during reduced RCS inventory operation, Preliminary procedures end/or administrative controls will be accepted as an expeditious action rerponse. We believe complete considerction of this issue is necessary in tre longer tern. 1 2.6 RCS Inventory Addition '?.6.1 Recommendation Previde at least two available or operable means of ndding inventory to the RCS that are in addition to pumps that are a part of the normal pFR systems. These should include at least one high pressure injection pump, The water addition rate capable of being provided by each of the means should be at least sufficient to keep the core covered. Procedures for use of these systems during loss o' DPR events should be provided. The path of water addition-must be specified to assure the flow does not bypass the reactor vessel tefore exiting any openiog in the PCS. P 6.2 Discussion Sufficient eauipment should exist in most plants, but there is little assurance it is evailable or provided for in the procedures and/or administrative controls. 1he expeditious action reconrendation increases asrurance of sufficient accident mitigation capability.
f 10 N,ozl e Dams l P.7 2.7.1 Recommendation (applicabletoWesting)hcuseandCombustionEngineetingnuclearsteamsup systern (NSSS) designs reasonably assure that all hot legs are not blocked '.,imultaneously by nozzle dams unless a vent path is provided that is large ercugh to prevent pressuri-zation of the upper plenum of the RV. See references 5 and (s. 2.7.2 Discussion Addressing closure of PCS legs addresses a major contributor to short tern core damage. The prohibited confiauration, if it existeu, could force water out of the RV within half an hour of loss of DHR. We reconnend that licensees consider removing.: pressurizar manway (if analy-sis shows this to provide a suf ficient vent path) or ctherwise create a i sui'.able openitffif a pressurization potential exists so as to lirnit the pressurization which could follow loss of DHR while nozzle dains and the RV ( l head are in place. l Simihrly, hot leg nozzle dems should be renuved before retroving cold leg nozzle dans or hot leg oczzle dams shoeld be renoved before, or es quictly as l l is practical following, closure of the open vent path from the upper Rl. the concern is that nozzle cams ray not have suf ficient strennth to A part of l Loss of a withstand the pressure that may result under accident conditions. nuzzle dam while prcssurized under hu of DHR conditions could cause rapid PV ( j voiding. 2.8 gen Stop Vahes T.8.1 Recomtre.r,da t lon (applicable tc NSS$s with 1000 stop valves) frplement proredures and acministratise controls that reaserably assure that all het legs are not closed stop valves unless a vent path is provided blocked simultanecusly b) large er.ough to prevent pressurization of the RV upper plenum or unless th t it loss if RV pressur ization shouTB occur. the RCS configuration prevents M water Closing cold legs by nozzle dons does net meet this condition. 2.8,P Discussion Hot leg stop vahes should be opened before opening rold leo stop valves or hot leg step salves shculd be opened before, or as civickly as is practical following, closure of the rpen vent path from thr upper RV. Loop stoo valves may be user 1 in combirntions sufficient to prevent loss of water thrt.unh cold lecs uncer pastulated conditions of RV pressurization und, when th1s conficuration is in place, the timing recuirements of item T of I Sectirr ?.P.? may be appliec,
~- 'R I 11 3.0 PROGRAfit1ED ENHANCEMENTS 3.1 Instrurnentation 3.1.1 Reconsnendation Provide reliable indication of parameters that describe the state of the RCS and the perfornance of systems normally used to cool the PCS for both normal and accident conditions. At a minimum, provide the following in the CR: (a) two independent RCS level indications (b) at least two independent tempert.ture me6sureraents rcpraentative of the core exit whenever the RV head is located on top of the RV (We suggest that ten.perature indications be provided at all times. ) (c) the capability of continuously inonitoring OPR system performance whenever a DHR system is being used for cooling the RCS (d) visible and audible indications of abnormal cer&ditions in tenperaturc, level, and DHR system performance 3.1.2 Discussior 3.1.2.1 RCS level Inadequate determination of RCS level has been involved in many potentially i serious events. This situation must be corrected. We strongly believe independence is important. This includes the conr.ections to RCS, where difficulties with blockage have been encountered in bott the licuid and reference connections. We recognize that it may be difficult to provide independence in isolated instances. Consequently, if the reconrendatiet. for indepundence results in an ennecessary hardship, we will censider compensatory neans. For example, if a cor.ron tap is used for the liquid leg. a neans of. period.fc. draining or flushirg capable of detecting blockage might be proposed as a means of diminishing the potential impact of the dependency. Introducirg a small flow into the sersing line at the Instrument and checking whether this perturbs the level indication is another way of checktrip, tinfortunately such technioues ray have the potential cf causing erroneous level indications. Similarly, a careful investigation of th* implications of determining level at a single location should be perf'>rmed, and a contrast obtained with the information obtained if more than one location were used. Phenomena and instrurrentation hebavior that are of concern include: (1) response time (2) instrument level inadequacies that ray not be iderstified by static inst rumentation ca libr ations ~
12 (3) DHR air entrairment influence (4) DHR flow rate influence (5) RCS drain location and dr ain rate impact influence (6) RCS level, such a: the potential for error because a high water level blocks the pretaurizer surge line connection to the RCS, the inability of air spaces to communicate if tF' legs are full, or erroneous level indication because a portion of the RCS fails to drain as anticirsted (7) thc measured water level at one location may differ f rom.that at the suction line (8) level may be af fected by pressure difference between the RCS ar.d the contain. tent building atmosphere. These phenomena ray be addressed by such actions as: (1) instrumentation error analysir (2) complete review of the instrumentation design (3) quality control and followup review of the installation (4) maintenance, including calibrations and operational checking. We also rote that ordinary plastic tubing does not meet our concept of reliable instrumentation, and its use may not be accepted as a component in instrumcntation systems. 3.1.2.2 RV Temperature liary plants have no indication of RCS state if DHR is inst because teaperature is determined by sensors located in the DHR system. Numerous licensees have demonstrated they do not understand that most RCS temperaturo indicators are inoperative under the conditions of concern. As a result, there have been occurrences of unrecognized boiling in the RCS. This is unacceptable because under some nonpower operation configurations boiling may force water out of the RCS and cause core uncovery in a short time. There are other implications as well. These include: (1) Boiling involves a mode change. A licensee encountering boiling in the mannet discussed here is often in violation of TSs. (2) Temperature is valuable in guiding DHR restoration actions dnd in moni-toring the effectiver(ss of recovery actions. (3) Knowledge of the RCS is necessary to guide actions such as containt.1ert closure and ceclaration of emergency levels. (4) Knowledge of temperature may allow operational flexibility, such as the ability to remove CHR systems f rom operation, j l
i e y 13 Accurate temperature indication is valuable evan if the p,V head is removed, and we prefer this be provided to the operators. Consequently, we suggest that licensees investigate ways to provide ternperature even if the head is rernoved, particularly if a lowered RCS invent:ry condition exists because of the short time that may occur between loss of DHR and iMtiaticn of boiling, an,1 the netd for operator guidance which a knowledge of tenperature can make possible. 3.1.2.3 OHF. System Performance i Pany CR displays provide only limited DHR system perfor* nance inforrhation to the operators. Flow rate is generslly provided. DHR pump motor current often is provided, although the indication rnay be on a back panel and not in the operator's norrnal range of vision. Motor current trend information is seldom i provided. Also rare is pump noise monitoring and a sensitive pump suctior j pressure indication, both of which could provide early indication of an j approach to loss of DHR due to air ingestion and inadequate RCS level, i Our recomtr.codation is broadly stated as a continuous monitoring of the DHR sy s tem ( s). We expect each licensee to consider the individual plant configuration and instrurrentation, and to provioe sufficient information to the operators that an approaching melfunction is clearly indicated. In some cases, available instrumentation may be sufficient, in others, new instrumentation may be necessary. i Provision of pump motor current is a coed example of useful intormation. A simple indication of instantaneous moter current can be useful, but a display which shows a historical trace is more valuable since " noise" due to air ingestion is readily seen, and may be one of the earliest indications of an approach to in60 equate pump suction conditions. Noise monitoring at the DHR pump and sensitive pressure determination in the punp suction pipe ure additional examples of potentially sensitive indications. Also of interest is a performance monitor that senser several parareters and provides an integrated DHR system performance indication (we are nut r. ware of the existence of such a tronitor, although we have seen indications of its consideration as a development instrurent). 3.1.2.4 Visible and Audible Abnormal Condition Indication Alarms are sometimes provided, although they tray be inappropriate for the application - such as an alarm on high flow rote or high pumo motor current, neither of which directly addresses loss of DHR. Alarms are seldom provided which indicate an approach to d loss of Dit cendition. We expect both audible alarns and a parel indication when conditions exist which jeepardire continued operation cf a DHR system, as well as when OHR it lost. For exanple, pump rotor current could be rnonitored continuously and an 61erm set at the time steariy operation is obtained which would provide an abnormal indication if motor current dropped by of the order of 100 (a smaller pertantage might be selected if sufficient to exclude extranenus alarms). A similar prevision could be made with a sensitive pump suction pressure indication ir the DHR drop line. We have provided general guidance in this recommendation. We expect licensees to select existing instrumentation dnd abnormal indicetions
14 ar.d, if necessary, to add instrumentation based upon a practicel approach for their plant configuration. 3..' Procedures 3.2.1 Recommendation Tevelop and implement procedures that cover reduced inventory operttion and tha* provide an adequate basis for entry into a reduced inventory condition. These include: (a) procedures that cover normal operation of the NSSS, the containment, and supporting systerns under conditions for which cooling woulo nortnally be provided by DHR systeras (b) procedures that cover emergency, abnornal, off-normal, or the equivalent operation of the NSSS, the containment, and supporting systems if an off-normal condition occurs while operating under conditions for which cooling would nomally be provided by DHR systems. (r) administrative centrols that support and supplement the procedures in itens (a), (b), and all other actions identified in this cormunication, as eppropriate l 3.2.2 Discussion We note that procedures that adequately cover operation urder all shutdown corditions for wnich cooling would normally be provided by DHR systenis will covt:r both entry into and operation in a reduced insentory condition. 3.2.2.1 Entry !nto Emergency Procedures 1 We define rorrnal and ernergenry precedures in Enclosure 3 to be consistent with power operation procedures, tionpower operation involves unique conditions that co not exist in power operation, and conditions f or entry into errergency procedures need to be defined. The usual entry conditier during power I operation is reactor trip or existence of conditions which sneuld have resulted in reactor trip. Several apropriate conditicns eyht for nonpower operation. We expect entry criteria to include consideration of all of the followina: (1) Accidental loss of a system that is operating to cool the RCS (?) l'nsuccessful atternpt to start a system when the system was to be used for RCS cooling d the RCS was not being actively cooled by another DHR system (3) Uncontrolled and sianificant loss of RCS inventory (4) Uncentrolled and significant break in the RCS coolant boundary (5) Ar,y valid syrrptom nf loss of control of the state of the RCS, such as uncontrolled terperature increase, uncontrolled pressurization, or the a tteintnent of values of these parareters which are suf ficiert ly hiah that action is required that is not contained within normal procedures. l
s il 15 (6) Significant core damage expected (7) Any valid symptom of significant core damage observed 3.3 Equipmcnt 3.3.1 Recomendation (a) Assure that adequate operating, operable, and/or availabic equipment of high reliability is provided for cooling the RCS and for avoiding a loss of RCS cooling. l (b) Maintain sufficient existing equipment in an operable or available c.tatus so as to raitigate loss of DhR or loss of RCS inventory sFould they occur, j This should include at leest one high pressure iniection pump and one i other system. The water addition rate capable of being provided t.y each i equipment iten should be at least sufficient to keep the core covered. (c) Provide adequate equipment for personnel communications that involve activities related to the RCS or systems necessary to taaintain the RCS in d stable and controlled Condition. 3.3.2 Discussion We ba u been prescriptive in the expediticus action recommendation. We will accept more flexibility in the longer term, includino considering such options as linking heatup rate and RCS configuration to both the 09R operational requirements and the operability and availability of backup cooling equipment. For example, if heatup rate permits and other considerations such as horon concentration are satisfactorily addressed, licensees may coniider not operating normal DHR systems for long times, or may consider using other means of cooling tne RCS if suitable precautions are taken while nornal DHR systems are not available. Such an opproach would require TS charoes, j Phere appropriate, licensees should develop procedures for gravity
- makeup from storage tanks and for the use of SGs to provide cooling. Recoqnited drt3S Where it wculd be inappropriate are where RCS pressure is too high for gravity feed from storage tanks, where other means of nakeup are not required to exist, or where the pressure nect.ssary to force steam into ccr, tact with SG tubes to initiate cooling also causes significent loss of RCS inventory.
It would be appropriate to consider SG cooling if the RCS pressure which thereby resulted was sufficiently low that gravity nakeup remained viable but would not be viable if SG coolinn did not exist, Loss of DHR due to unplanned activation of the autoclosure interlock function is not consistent with provision of reliable ecuipment. You should investi-aate this feature if installed in your plant and should consider chances to obtain a reliable heat removal syster, consistent with other requirements. We encourage removal of this feature nn the basis of our review of operating experience provided suitable con;pensatory reasures are taktn. At present, we recommend the Diablo Canyon approach as a model #0r guidance (refs, 3 and 41 He have received a report funded by the Pestinghouse cwners group that addresses this topic (ref 81, but we have not yet reviewed the document.
,f 16 Equipment (such as a DHR system) is reliable only if its Tupport requirements are reliably met (electrical power, cooling). Support requireme'nts necessary for reliable operation should be considered in rnecting the prograwed enhancement recommendations of this letter. Operation of equipment iri a raanner that would increase the likelihood of its rnalfunction should be addressed. For example, aany TSs require a high CHP system flow rate when core cooling requirernents can be met at a lower rate. The high rate contributes to the likelihood that air will be ingested and cause a loss of DHR. Such operating techniques are inccnsistent with reliable operation and should be adaressed in meeting the longer term recommendations of this letter. 3.4 Analyses 3.4.1 Recommenoation Conduct analytes to supplenent exutirg information and develop a basis for procedures, instrumentation installation and response, and equiprent/NSSS interactions and response. The analyses should encompass thermodyr aric and physical (configuration) states to which the hardware can be subjected and should provide sufficient depth that the basis is developed. Emphasis should be placed upon obtaining a complete understariding of NSSS behavior urder nonpower operation. 3.4.2 Discussion The Westinghouse owners group has funded an analysis program which we consider an excellent start toward rceting this recomnendation. That program covers areas such as: (1) thcreal/ hydraulic modeling with consideratior, of noncondensibles for 2, 3, and a loop plants (2) heatup rate, time to saturation, naximum prestori7ation, effect of water in SGs, vapor ventirg, liquid venting, and tine to core uncovery (3) influence of SG nozzle dams (4) mitigation actions including gravity m keup to the RCS, forced nakeup to the RCS, use of SGs, safety it,jection, and bleed and feed. luportant results are already being achieved in the Westinghcuse progran, and in: pact ori safety. are being factored into plant operations, with a significant Of note is the independent discovery-of the potential impact of improper use of nozzle dams, which is discussed in reference 6, and the increased uncerstanding of plant behavf ar during nonpower iperation. Another area that should be considered in reaching a complete understar' ding of Areas thut behavior durina nonpower operation involves level instrumentation. RHR fler shruid be considered include response tires, PHP air entrainment, rate, draining location and rate, range (PCS cnnnectier: locatior trd irnpact upon instrumentation indication), and RCS level (such as potential for error
i$ 17 y due to a hot leg level high enough to block the pressurizer surce' line connection to the RC5 or thc influence of a full hot leg due to inability of air spaces to tommunicate). See enclosure 1, Section 2.1 for additional information. 3.5 Technical Speci'icationss 3.5.1 Recommendation Technical specifications (TSs) that restrict or limit the saft.ty benefit of the actions identified in this letter should be identified and appropriate chances should be submitted. 3.5.2 Discussion ~ Typical potential im-nclude 1Ss that control containment: DHR syttern flow rate; the autoclosiS ,rlock; equipment operability, operation, and availability; and instrumentation. One objective we wish to achieve is a simplification of TSs at ronpower operation is investigated. Consequently, we will consider alternatives to li placing require.ments in TSs when such alternatives achieve the same purpose. for exartple, procedures reoviring certain DliR couipment to be availalile before an operation it initiated may be sufficient, and tuch specifications ther would not appear in TSs. 3.6 RCS Perturbations 3.6.1 Recornmenda tion item 2.5 of the expeditious actions should be reexaniced and operations refined as necessary to ressonably minimize the likelihood of loss of DHR. 3.6.2 Discussion Where systems or components require lowered RCS inventory for maintenance or testing, reasonable attempts should be made to conduct such activities when decay heat is low, other activities have a low likelihood of interfering, and extra prtcautions are available to mitigate transients should any occur. Extra precautions include such items as additioral equipment to maintain RCS inventory, a closed containment. and an enhanceo ability to close containment should loss of DHR occur. Activities which industry experience shows to have a potential impact on operation, such as electrical tests that could lead to closure of DHR systein suction valves, are not to be conducted during lowered inventory operation if they can be reasonably conducted at another tinte. If such testing must occur, then additional precautions snould be taken to respond if an impact to CHR or to the RCS occurs. Activities that could perturb the p.CS inventory or could lead to a loss of DHR given a single malfunction, such as the partially open velve ahich initiated
f 18 inventory loss at Diablo Canyon on April 10,.1987, should not be conducted during lowered inventory operation unless the symptoms of such a sinole failure have been considered and precautions are previded to compensate if the symptoms occur, for example, the syrptoms of the open valve at Diablo Canyon were an increase in water level in the tank that received the draining water and a de-crease in wcter level in the cnemical and volume control system (CVCS) tank. Precautions would have included identification of the expected response of those tank levels, specifically observing those tank levels during and following initiation of the operation, and assuring that additional independent ways of adding makeup water to the RCS were readily available. Control room personnel should be informed imnediately before initiating an operation which could perturb tbc RCS or a system which is necessary to maintain the RCS in a stable and controlled condition while a reduced RCS inventory condition erists. They should also be immediately informed of any unanticipated activity or symptom associated with the operation which could affect the RCS or DHR, and should be informed when the operation is ended. We note that recent plant dif ficulties have occurred when licensees were improving instrumentation. Typically, there may be more temporary connections than usual, tubing runs may not be well located and controlled, and operators may not be f amiliar with the new instruments and may discount the results, in part because the instruments may not have beer. declared operational. We also ncte that maintenance personnel may not be sensitive to the use of tubing or of openings into the RCS, We believe it important that licensees recognize the potential for perturbation of instrcrent indications. These thould be addressed as part of the overall issue of perturbation of the RCS. N
A \\ 19
4.0 REFERENCES
(1) U.S. Nuclear Regulatory Cornission, " Loss of Residual Heat Penoval System, Otablo Canyon, Unit 2, April 10,1987," NUREG-1269, June 1987 (2) F. J. Mit aglia, "l.oss of Residual Heat Removal (RHR) while the Reactor Coolant System (RCS) is Partially filled (Cencric Letter 87-12) " Letter to all licensees of operating PWRs and holders of construction permits for PWRs, July 9, 1987, (3) J. Shiffer, " Removal of RHR System Autoclosure Interlock Function," Letter to NRC from PG8E, Aug, a, 1987 (4) J. Shiffer, " Removal of RHR Suctinr. Valve Autoclosure Interlock func. tion," Letter to NRC from PGAE, Jan. 19, 1988. (5) C. E. Rossi, "Possible Sudden Loss of RCS Inventory Durino Low Coolant Level Oper@on," NRC Information Notier No. 88-16. June 8,1988. (6) R. A. Newton, " Westinghouse Owners Group Early Notification of Mid-Loop Operation Concerns " Letter from Chairman of Westin Westinghouse Owners Group Primary Repre.tentatives (ghouse Owners Group to { IL, IA), OG 83-21, May 27, 1988. j (7) Pacific Gas and Electric Co., " April 10, 1087 RHR Event, linit 2, Diablo Chnyon Power P1' ant," Volores 1 and 2. April and May,1987 (8) N. L. Burns et al., " Residual Heat Removal System Autoclosure Interlock Removal Report for the Westinghouse Owners Group," WCAP-11736, Revision 0.0, Feb. 1988. I
4-3 2; s 8 l CNCLOSURE 3 TO GENERIC LETTER 88-17 ABBREYIATIONS AND DEFINIT 16hs 6 v e f
T 2 1.0 ABBREVIATIONS CHL Brookhaven National Laboratory B&W Babcock and Wilcox Combustion Engineerirg (also used to describe in-core thermocouples) CE core exit thermocouple CET CFR Code of Federal Regulations CR control room DHR decay heat removal (used in a generz.I sense to describe the process or system) FR Federal Register FSAR final safety analysis report GI generic issue GL generic letter gpm gallcns per minute ICC inadequate core cooling LER licensee event report NRC Nuclear Regulatory Commission NSAC Nuclear Saf 9ty Analysis Center NSSS nuclear steam supply system NUREG Huclear Regulatory Commission document designation NUREG/CR NUREG prepared by a contractor PORY pressure-or power operated relief valve located on the pressurtzer PRA probabilistic risk as?cssment or probabilistic risk analysis PRT pressurizer relief tank psi pounds per square inch PWR pressurized water re. actor RCP reactor coolant pump RCS reactor coolant system rem reentgen eauiva D7t man RHR residual heat removal (used in the specific sense of the DHR system used in Westinghouse plants) RTD resistance temperature device RV reactor vessel SG steam generator TS technical specification USI unresolved safety issue W Westinghouse
o 3 2.0 DEFINITIONS For p:Jrposes of this letter, the following definitions apply: (1) Action - Responsive acts which are recommended in the letter. There are two types of actions. (a) Expeditious action - An action recormnended in tne letter that should be implemented prior to operating in a reduced inventory condition. (b). Programed enhancement - An action which is to be implemented at a later date. Generally, such actions can only be implernented af ter development work has been done, he anticipate implementation 4 concurrent with development and outage availability. (2) Available - Ready for use within a short enough time to meet the intended need, but not necessarily operable because phy"ical manipulations may be needed to realize an operable status. (3) Closed containment - A cortainment that provides at least one integral barrier to the release of qdicactive material. Sufficient sepa* dion of the containment atmosphere from the outside enviror. ment is t6 be provided such that a barrier to the escape of radioactive material is reasonably expected to remain in place tellowing a core melt accident. This can be accomplished by providing' reasonable assurance thPt the following conditions are inet: (1) The equipment hatch door is closed and held in place by a sufficient number of bolts :uch that no gaps exist in the sealing surface, (2) A mintwm of one door in each airlock is closed, and (3) Each penetration providing access from the containment atmosphere to the outside atmosphere shall be closed by a valve or blind flange. Closure by a valve or blind flange used for containment isolation during power operation meets this specification. Closure by other = valves or blind flanges may be used if they are similar in capability to those provided for containment isoletion. These may be constructed of standard materials and may be justified on th< asis of either normal analysis methods or reasonable engineering judoement. ~ (4) Containment - See Closed containment. (5) Emergency Procedures - That set of emergency, abnormal, off-normal, or the equivalent procedures that cover operation of the nuclear steam supply system (NSSS), the containment, and supporting systems if an off-normal condition occurs whtle operating under conditions where heat would normally be removed by DHR systems. These procedures provide 2 coverage using a symptom tased philosophy and organization similar to that used for response tc off-normal,nnditions originating durina power a ^m 1 I m - - - - - - - - - - - ~ - - - - " - ' - - - - - - - " - - ' - - ^ ^ ' - - " ' -
T 4 They cover all aspects of operation where responsibility operation. rests with the operators, including provision of a closed containmer.t. restoration of decay heat ternoval (DHR) by a broad range of meTns and maintenance or replenishment of reactor coolant systern (RCS) inventory. Plant specific features are considered, such as relative elevations of water sources (for gravity drain to the RCS) and presence of high eleva-tions in DHR suction pipes (which may affect attempts to restart DhR systems, particularly if the RCS has reached a boiling condition). (6) Independent - Not vulnerable to the same factors as another entity that For example, if a common tap is used for the has the same purpose. liquid leo of two liquid level instrumo.nts, then they are dependent if the common tap can be plugged by debris or if unrecognized phenomena can influence the indicated water levei so that it is not representative of level at the location of interest. (7) Inventorv - See Reduced RCS inventory. Normal procedures - The set of procedures that provide guidance and (8) instructions to the operators which cover normal operation of the NSSS, the containment, and supporting systems under conditions during which heat may be removed by DHR systems. These procedures cover operation with a water-solid RCS (if this is a normally allowed mode of operation), with a level in the pressurizer, RCS drain down, operation when RCS level is below the pressurizer instiJmentation range, operation under reduced inventory conditions, operation while at mid-loop, and r'efill of the RCS, Containment, RCS state, and equipment criteria that must be satisfied before entering the conditions where these procedures apply and during the existence of these conditions are %:luded, either as entries in the procedures, as administrative controls ce ferenced in the procedures, or by other suitable means which provide reasonable assurance that the entry conditions are met. Mid-loop - The condition that exists whenever the RCS water level is (9) lower than the top of the flow area at the. function of the hot legs with the RV. (10) Procedures - See Norma _, procedures or Emeroency procedures. l (11) g inventory - See Reduced inventory or Mid-loop. (12) Reduced inventory or Reduced RCS inventnry - An RCS 'm entory that results in a reactor vessel water level lower than three feet below the RV flange. (13) Reliable - The condition of having a hich, but reasonable, expectation of being able to perform the intended function. Ordinary plastic tubing oces not meet our concept of reliable instrumentation nor does a DHR system in which inadvertent operation of the "'oclosure interlock is likely to occur.
a u. [' UMITED STATES nm cuss man NUCLEAR REGULATonY COMMISSION
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