ML19206B384

Summary of 770721 Meeting W/Metropolitan Edison Co & Contractors Re Open Items in NRC Review Including Hydrogen Line Break Containment Sump & Steam Line Break

ML19206B384
Person / Time
Site:	Crane
Issue date:	07/27/1977
From:	Silver H Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 7905090405
Download: ML19206B384 (20)

Text

{{#Wiki_filter:. . n atog C S j UNITED STATES = ~ NUCLEAR REGULATORY COMMISSION 4 f WASHINGTON, D. C. 20655 July 27, 1977 COCKPT NO. 50-320 APPLICANT: Metropolitan Edison Company FACILITY: Three Mile Island Unit 2 SUWARY OF MEETING CN OPEN ITEMS Representatives of the atplicant end his contractors met with memcers of the staff on July 21, 1977 to distuss and resolve cpen items in the staff review. Items discussec are sumari::ed belcw: Hydrocen Line Break The applicant noted that since the existing routing of the hydrogen line includes a corridor in the Auxiliary Building through whicn run caole trays of both trains, the hydrogen line will be rerouted. (This corridor has been identified in the fire hazards analysis.) To avoid any cossible damaging introduction of hydrogen into the auxiliary building, the only feasible solution is installation of a guard pipe on the hydrogen line and addition of a rupture disc. We noted that we would verify if inspection of the line was a proclem. The applicant will inform us of their schedule for this modificaticn. Containment Sumo After some discussion the apolicant agreed to: 1. provide calculations showing availahle pumo NPSH for the as-built configuration 2. add intermediate mes - a eening on the inside of the trash rack 3. add a screened cover on the sumo With these changes, we agreed the stmp design would be acceptable pending performance of pre-cperational verificaticn tests and additional calculations to which the c'.plicant had previously comitted. Steam Line Break In response to our inquires, the apolicant stated that they excect to cccplete the analysis of an unmitigated steam line break within a montn. Preliminary indications are that for cceration oeycnd the first fuel cycle with the recuired assumctions, consequences might be .a gig y, 7905090 @5

. acceptacle. For interim operation for the first fuel cycle, results might be ircroved due to the higher shutdown margin. Discussion of other possible relief in assu::ptions focused on loss of offsite pcwer or elimination of the stuck rod. Although the applicant objected to providing additional justification new for interim operation they agreed to provide an analysis of the steam line break accident without credit for non-safety grade equipnent, including increased first cycle shutdown margin s.2, if required, loss of offsite power, in such a time fr e; as to permit our review before initial criticality. Based on the applicants statements, we new have reasonable assurance that this analysis will support the deter-mination permitting operation for the first fuel cycle. We noted that the present analysis shows the core reaches 22% pcwer at 45 seconds, and requested analyses with and without RC pu:tp operation to show that this dces not result in significant fuel failure or if fuel failure does occur, that coolable gecmetry is maintained. The applicant noted they would inform us within one week as to whether they could perform this analysis and when. Considerable discussion took place on the question of long term cooling (see question 1. of "Attachement #1", handed out for discussion purocses). We noted this question was asked with relation to the existing steam line break analysis. The applicant stated they would ascertain their best schedule for response to this question. Boron Flow Measurement Some additional verbal information was presented by the agolicant, whicn we will consider further. The applicant agreed to provide further information on calibration of this instrument. Qualification of Balance-of-Plant Class IE Ecuitment In response to the applicants statements regarding certification and expected envircnmental conditions for equipment in question, we noted that we require that this equipment be operable in normal, abnormal, and accident envircrJnent. Equip::ent must at all times be available for accident mitigation or other safety-related purposes, even if exposed to unexpected extreme environmental conditions. We provided to the applicant ccpies of our current position on this subject (see attached " Class IE Equipnent Qualification (Outside Containment)"). We noted that for recent plants, resolution of this item was required prior to cg ration. Electrical Site ViJit Plans were made to start the electrical site visit at 1:30 P.M. Monday, August 1, 1977. Our agenda was handed to the apolicant. 971 L3 L \\

. Grid Stability We conraented on the preliminary information the apolicant had recently submitted, noting some oossible problems in the 480 volt equipment. The applicant attenpted to explain their subaittal, which will be trans-mitted formally in Amendment 57, and defended it as adequat We pre-sented our current position on this sub3ect (see attached 't.nclosure 1") which has already been given to cperating plants including TMI-1. We will reconsider their subnittal and, if necessary, transmit our new position formally. Cormittment to this position, but not necessarily physical con-formance, would be required prior to licensing. Centrol Rod Droo Time Testing The applicant presented revised information on shutdown margin at 75% rod insertion, and conmitted to correct minor errors in rod drop time test procedures and to add to the purpose ar4 acceptance criteria in the appropriate test procedure the requirements to evaluate performance of the control rod dashpot. Resconse Time Testing Responses to our recent questions on this subject were reviewed; the applicant agreed to subnit advance.;opies of these responses next week. Other Items The. applicant provided the following schedules: Fuel Densification Report mid-August, 1977 Single Rod Withdrawal Report (Q 22.38) mid-August, 1977 Overpressurization response mid-August, 1977 Instrument qualification for steam line break (Preliminary results appear satisfactory) - late-August, 1977 Advance copies of various sections of future acendments 57 and 58 were handed the staff for our use. n., 1 t Harle' ilver, Project Manger Ligh 'ater Reactors Branch 4 Divi on of Project Managment r)?O

FliCIDSL'FE Attendance List NRC B&R Harley Silver Scott Dam Tom Novak Peter Hearn B&W Sandy Isreal Jim Watt Bill Gray Frank Ashe Lee Pletke Chas. Miller Ren Lovell Jim Shapaker Farouk Eltawila GPUSC Scb McDermott Lou Lanese T. G. Brouchton E. G. Wallace 25 233

Attachment f 1 RequireEents for additional analysis and clarification. 1. Provide a reanalysis of the turbine stop valve failure case with loss of offsite power at turbine trip. Initiation of emergency feedwater and high pressure injection are to follow design actuation. After the unaffected steam generator has been isolated by the MSIV, let it refill and stabilize considering only the secondary system safety valves and the cavitating venturi for control. If operator action is required relative to el'her primary or secondary system control, allow 10 minutes after first indication of a need for action. Include effects of startup and loading of diesels following loss of offsite power. Allow the pressurizer safety valve to relieve primary system pressure unless relief valve operation results in worst case. Principal. consequences of this event to be considered are core cooling, pressure vessel integrity, system stability, and dose release during both the snort-term transient and while achieving long-term cooling. In presenting the results, provide a chronology of all trips, actuations, operator actions, etc., considered in the analysis. Support this chronology with a list of all systems and components utilized and their response or actuation times. Carry the events and analysis out in time until the heat removed by the emergency feedwater system balances the core heat generation rate in a stable condition. This would occur after mass and energy are no longer being released from the pressurizer. LJ Ls '

Provide the folicwing results as a functicn of time: (a) DNBR until minimum is achieved and subcriticality is assured. (b) affected and unaffected steam generator levels. (c) affected and unaffected steam generator pressures. (d) feedwater ficw rate (main and auxiliary). (e) cladding temperature if DNBR is violated. (f) primary system pressure. (g) core inlet and outlet temperatures. (h) pressurizer level. Discuss the results in terms of fuel damage, primary coolat; system pressure limits, and dose releases through failed steam senerator tubes. 2. A return to 22% power is indicated in figure 155-10. Discuss this period in terms of DNBR, fuel damage, and number of reds affected. 3. A minimum DNSR of 1.7 is stated to occur at two seconds. This is not consistent with figure 15B.11. Clari fy. 9 7, cc. J o-LJ

-I Class lE Ecuicment Qualification (Outside Containment) -b With regard to all Class lE equipment located outside the containment building, we require assurance that the environment is maintained within the temperature range for which the equipment is qualified to operate. In those locations where the temperature could exceed that for which the Class lE equipment is qualified, the staff requires that the applicant provide a temperature monitoring system. The system should at a minimum meet the following requirements: The control roem shculd receive an alarm when the temperature a. range has been exceede).- This alarm should be provided by instrumentation which (1) is of a high quality, (2) is checked to verify its functional capability by plant technical specification requirements, and (3) is powered frcm a continuous power source or is redundant with separate channels and power sources. b. The operator should have a method of maintaining a continuous record of the temperature during the time that the temperature range is exceeded. Based on the monitoring system the applicant shall report the occurrence of the temperature exceeding the equipmer.t qualification range as an abnormal occurrence to the NRC. In addition to this, the applicant shall provide results of an analysis to demonstrate that the excess temperature has not degraded the involved Class lE equipment below an acceptable level for continued plant operation. k_ 25 236 s

/ / / [h / ENCLOSURE 1 SAFETY EVALUATION AND STATEMENT OF STAFF POSITIONS RELATIVE TO THE E>!ERGENCY POWER SYSTEMS W m?> w M2dA A. INTRODUCTION The onsite emergency power systems of operating nuclear power facilities are being reviewed to assess the susceptibility of their associated redundant safety-related electrical ecuipment to: (a) Sustained degraded voltage conditions at the offsite ;ower source; and (b) Interacticn of the offsite and onsite emergency pcwer systems. We have completed our review of the res:enses to cur generic recuest for additional information1/ relative to the electrical pcwer distribution systems of currentiv c;eratinc nuclear Ocwer facilities. In res:onse to our request,'all licensees have analy:ed their system designs to determine that the voltage levels at the safety-related buses have been optimized for the full lo,ad and minimum load c:ncitiens that are ex:ected thrcugncut the anticipated range of voltage variations for the offsite ;cwer scurces. The transformer voltage tac adjustments that were ne:.essary to optimi:e the voltage levels have been a'c::molished. In additic9 to the above co rective action, we have develcped the folicwing staff posi; ions for use in evaluation Of each of the 0:erating nuclear These Oositi:ns ocwer pl;en :3 with regard to the two items identified abcve. were develeted On tne basis of our review of the licensee rescense to our or v~ cJ J/ 1/ Letters n all licensees, dated August 12 and 13,1975.

/ 3 / 2-requests for additional informatien and of other related information as cited in the text. B. POSITIONS

1), Position 1:

Second Level of Under-or-Over Voltace Protection with a Time Delav 'de require that a sec0nd level of voltage pr0tection for the onsite power system be provided and that this second level of voltage protection sha'.1 satisfy the followino criteria: a) The selection cf voltr.ge and time set points shall be determined frem an analysis of the voltage requirements of the safety-related loads at all onsite system " distribution l evels; b) The voltage trotection shall include coincidence logic to creclude spurious trips of the offsite ;cwer scurce; c) The time delay selected shall be based on the f licwing conditions: (1) The allowable time delay, including margin, shall not exceed the maximuE' time delay that is assumed. in the FSAR accident analyses; (2) The time delay shall minimi:e the effect of snort. duration disturbances from reducing the availability of the offsite power source (s); and (3) The allowable tica eiration of a degradec voltage condition at all distribution system levels 'shall not result in failure of safety systems or cemecnents; c ; c..

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,g / d) The voltage monitors shall automatically initiate the disconnection of offsite power sources whenever the voitage set point and time delay limits have been exceeded; e) The voltage monitors shall be designed to satisfy the requi.ements 'of IEEE Std. 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations"; and f) The Technical Specifications shall include limiting conditions for operation, surveillance requirements, trip set points with minimum and maximum limits, and allowable values for the seccnd-level voltage protection monitors. General Design Criterion 17 (GCC 17) " Electric Power Systems", of Appendix A, " General Design Criteria for Nuclear Power Plants," of 10 CFR Part 50 requires: (a) two physically independent circuits frem the offsite trans-mission network (although one of these circuits may be a delayed access circuit, one circuit must be automatically available within a few seconds following a loss-of-coolant acci, dent); (b) redundant onsite A.C. power supplies; and (c) redundant 0.C. power supplies. GDC-17 further requires that the safety function of each a.c. system (assuming the other system is not functioning) shall be to provide suffic4ent capacity and capability to assure that: (a) specified acceptable fuel design limits and the design conditions for the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences; and (b) the core is cooled and containment integrity and other vital functions are maintained during any of the postulated accidents. 25 2M -f. O

s.. / Existing undervoltage acnitors aut:matically perform the required func-tion of switching frem offsite power, the preferred ower scarce, to the redundant onsite pcwer sources when the monitored voltage degrades to a level of between 50 to 70 percent of the ncminal rated safety bus voltage. This is usually accomplished after a one-half to one second time delay. These undervoltage monitors are designed to function on a ccmolete loss e 'he offsite power source. The offsite pcwer system is the comon source which normally sucolies power to the redundant safety-related buses. Any transient or sustained degradatien of this c mon source will be reflected onto the onsite systen's safety-related buses. ~ A sustained degradation of the offsite ecwer system's voltage c:uld result in the l'oss of capability of the redundant safety loads, their centrol circuitry, and the associated electrical components recuired for performing safety functions. The operating procedures and guidelines utilized by electric utilities and their intercennected coocerative or:anizations minimi:e One pro-bability for the above conditions to oc:ur. However. since decradation of an offsite ocwer system that could lead to or cause the fai' rare of redundant safety-related electrical equipment is unacceptable, we recuire the additional safety margins associated witn imolementation of the protectivi measures detailed abcVe. M c ,O

'/ j 4 5-

2) Position 2:

Interaction of Onsite Power Sources with load Shed Feature We require that the current system designs automatically prevent Joad shedding of the emergency buses once the onsite sources are The supplying power to all sequenced loads on the emergency buses. des k shall also include the capability of the load shedding feature to be autcmatically reinstated if the ensite source supply breakers are tripped. The automatic bypass and reinstatement feature shall be verified during the periodic testing identified in Position 3. In the event an adequate basis can be provided for retaining the load shed feature when loads are energi::ed by.ne onsite power system, we will require that the setpoint value in the Technical Specifications, which is currently specified as "... equal to or greater than..." be amended to specify a value having maximum and minimum limits. The licensees' bases for the setpoints and limits selected must be documented. GDC 17 requires that provisions be included to minimi::e the prcbability of losing electric power from any of the remaining supplies as a result ^ of or coincident with the loss of power generated by the nuclear power unit, the loss of power frcm the transmission network, or ihe loss of power from the onsite electric power supplies. e 25 211

./ i / s / /. The functional safety requirement of the " loss-of-offsite power monitors" is to detect the loss of voltage on the offsite (preferred) power system and to initiate the necessary actions required to trans-fer the safety-related buses to the onsite system. The load shedding fenture, which is required to function prior to connecting the onsite power sources to their respective buses can adversely interact with the onsite power sources if the load shedding feature is not bypassed after it has perfomed its required function. The load shed feature should also be reinstated to allow itito perform its function if the onsite sources are interrupted and are subsequently required to be reconnected to their respective buses.

3) Position 3:

Onsite Power Source Testina We require that the Technical Specifications include a test requirement to demonstrate the full functional operability and independence of the onsite power sources at least,once per 18 months during shutdown. The Technical Specifications shall include a requirement for tests: (1) simulating loss of offsite power in conjunction with a safety injection actuation signal; and (2) simulating interruption and subsequent reconnection of onsite power sources to their respective buses. Proper operation shall be determined by: a) Verifying that on loss of offsite power the emergency buses have been de-energized and that the loads have been shed frem the emergency buses in accordance with design requirements. 2.2 6 ro) c.

-,/ ~ ~ b) Verifying that on loss of offsite power the diesel generators start from ambient conditicn on the autostart signal, the emergency buses are energi:ed with permanently connected loads, the auto-connected emergency loads are energized through the load sequencer, and the system operates for five minutes while the generators are loaded with the emergency loads. c) Verifying that on interruption of the onsite sources the loads are shed from the emergency buses in accordance with design requirements and that subsequent loading of the onsite sources is through the load sequencer. GDC 17 requires that provisions be included to minimize the probability of losing electric power from any one of the remaining supplies as a result of or coincident with the loss of power generated by the reactor power unit, the loss of power generated by the nuclear power unit, the loss of power frem the transmission network, or the loss of pcwer from the onsite electric power supplies. The testing requirements identified in position 3 will demonstrate the capability of the onsite power system to perform its required function. The tests will also identify undesirable interaction between the offsite and onsite emergency power systems. Wa Lq.-

\\ y. N TABLE 3.3-3 (Continued) EtiGitiEEllED SAFETY FEATtillE ACTUAT10ft SYSTEH lilSTiltJMEtiTATI0ft_ HittlHl)H TOTAL 110. CilAfitiELS CilAtitlELS APPLICABLE FutiCT10!iAL Utill 0F CilAftriELS TO TitiP __ OPEllABLE_ OPERATit1G MODES ** ACT I0tt *** I g LOSS OF POWEll a. 4.16 kv Emergency Bus Undervoltage (Loss of Voltage) 4(3)/ Bus 2/ Bus 3(2)/ Bus 1, 2, 3 A or B g M r-y b. 4.16 kv Emergency Bus N Voltage) 4(3)/ Bus 2/ Bus 3(2)/ Bus 1, 2, 3 A or B E y Undervoltage (Degraded C E '~ E 'O A E m i a I

• (Entries in parenthesis are applicable for

{% 2 out of 3 coincidence logic)

• allequired when ESF equipment is N

required to be operable 4.> n

• • • Action A for 2 out of 4 logic Action B for 2 out of 3 logic t

• ~,. -

/ ~ ,/ .^ l TABLE 3.3-3 (Continued) ACTION STATEMENTS ACTION A - With the number of OPERABLE channels one less than the Total Number of Channels operation may proceed provided both of the following conditions arr. satisfied: a. The inoperable channel is placad in the tripped condition within one hour, b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours for surveillance testing per Specification (4.3.2.1.1). ACTION B - With the number of OPERABLE Channels one less than the Total Number of Channels operation may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition within i hour. l 1 \\ 4 O a [l[J ~ qr u

l I s i s ? y y a a l l e e d d ,s a a e e hm hm ti ti it i t E w w LS d d BE sn sn AU t o t o WL l c l c OA oe oe LV vs vs LA )) )) S E + [] +1 U LA (( (( V P s I R T N y y O a a I l l T e e A d d T a a N e e E hm hm M ti ti U i t i t R w w ) T d d d S E sn sn e N U t o t o u I L l c l c n A oe oe i M V vs vs e t E n T P )) )) o S I C Y R ( S T +[ +T 4 N O (( (( 3 I T 3 AU E T L C B A AT E e e R g g U a a T t t A l l E o o F v v r r Y e e T d d E n !!n F U A 6 S s s u u D B B E ) R y) ye E c e cg E ng na N e a et I gt gl G rl ro N T eo eV mV m E I N R E Ed U E f e W vo vd L O k k a A P s r N 6 s 6g O F 1 o 1 e I O .L .D T 4( 4( s ;, 7 t C S ] N S w 'J U O f r F L a b L

N', TARLE 4.3-2 (Continued) s EllGillEERED SAFETY FEATURE ACTUATI0fl SYSTEM INSTRUMEllTATION SURVEILLANCE REQ I OPERATING CllANNEL MODES IN WilICil CllANNEL CllANNEL FUNCTI0llAL SURVEILLANCE l FUNCTIONAL UtilT CitrCK CALIBRATION TEST fiEQUIRED i LOSS OF POWER a. 4.16 kv Emergency Bus g. Undervoltage (Loss of Voltage) S R H 1, 2, 3 b. 4.16 kv Emergency Bus Undervoltage (Degraded Voltage) S R M 1, 2, 3 O b, S = at least once per 12 hours R = at least once per 18 months g i M = at least once per 31 days -{ g

}/'f s -~ ELEcrRICAL POWER SYSTEMS e _ SURVEILLANCE RECUIREMENTS 4.8.1.1.X Each diesel generator shall be demonstrated OPERABLE: a. At least once per 18 months during shutdown by: 1. Simulating a loss of offsite power in conjunctic, with a safety injeccion actuation test signal, and: a) Verifying de-energization of the emergency busses and load shedding from the emergency busses, b) Verifying the diesel starts from ambient condition on the auto-start signal, energizes the emergency busses with permanently connected loads, energizes the auto-cennected emergency leads through the load sequencer and operates for > 5 minutes while its gener,ator is loaded with the emergency loads, c) Verifying that on diesel generator trip, the loads are shed from the emergency busses and the diesel re-starts on the auto-start signal, the emergency busses are energized with permanently connected loads, the auto-connected emergency loads are energized through the load sequencer and the diesel operates for > 5 minutes while its generator is loadad with the eEergency loads. 4 e e

' O- ~ W' MEETING SCORRY j f 19d

• C Cd eCl Q-g Docket File J. Knignt NRC PDR D. Ross Local PDR R. Tedesco TIC R. Bosnas tiRR Reading S. Pawlic<i Branen File I. Sinweil B. C. Rusche P. Check E. G. Case

/} T. Novak R. S. Soyd Z. Rosztoczy R. C. DeYoung IE (3) A D. B. Vassallo G. Lainas D. Skovnolt V. Benarova J. Stolz M T. Ipoolito K. Kniel V. Mcore O'. Parr R. Voll.T.er M S. Varga M. Ernst R. Denise A F. Rosa R. Clark W. Gam.ill T. Speis EP Branca Chief P. Collins D. Bur.cn C. Helte.mes J. Collins R. Houston ti. Kreger L. Crocker R. Ballard J. Miller B. Youngolood F. J. Williams J. Steen

7.._i-- S.% %Md L. Hul.hin H. Denton L. Drener D. Muller Branch Chief, CR A

Project Manager War 1ev 5iiver Project Manager, CR Attorney, ELD ACRS (16) Licensing Assistant-M. Service NRC Participants k O*) b A ws! A /mk A A1.lla 25 23.9}}