ML20010B835
| ML20010B835 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 08/10/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Jackie Cook CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 8108180215 | |
| Download: ML20010B835 (18) | |
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l DISTRIBUTION Docket File 50-329/330 LPDR LB #4 r/f PRCPDR DEisenhut ACRS (10)
EAdensam DHood MDuncan Docket Nos.:
330 RT o
RVollmer 0g bg 1 TMurley tir. J. W. Cook RMattson N
Vice President h
RHartfield, MPA cp Consumers Power Conpany OELD 1945 West Parnall Road h
OIE (3)
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Jackson, tiichigan 49201 bcc: WJensen
$6 17 8
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TERA /NSIC/ TIC
Dear !!r. Cook:
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SUBJECT:
TRN4SHITTAL OF PRELIrlINAkY SER DRAFT SECTION 6 f3, N
.tIDLAND PLANT, UHITS 1 AND 2 O
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Enclosed for your review and cor.nent is the preliminary draft section 6.3 of the NRC Staff's Safety Evaluation Report for Hidland Plant, Units 1 ano 2.
Your attention is directed in particular to any open item contained within this draf t section. A principal objective of this tran_.nittal is to provide for timely luentification and resolution of any additional analysis, missing information, clarifications or other work necessary to resolve outstanding issues.
i Please contact tne Staff's Project Manager regarding the need for any meetings and telephone conferences to this end.
l Your corrents, including schedules for conpletion of any further analyses or other work associated with resolution of open items, are requested within two weeks of receipt of this letter.
Sincerely,
" Original signed by:
I Robert L. Tedesco, Assistant Director for Licensin]
Division of Licensing cc: See next page
Enclosure:
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MIDLAND Mr. J. W. Cook Vice President Consumers Power Corpany 1
1945 West Parnall Road Jackson, Michigan 49201 Michael I. Miller, Esq.'
Mr. Don van Farrowe, Chief cc:
Ronald G. Zamarin, Esq.
Division of Radiological Health Alan S. Farnell, Esq.
Department of Public Health 1 sham, Lincoln & Beale P.O. Box 33035 Suite 4200 Lansing, Michigan 48909 1 First National Plaza
- Chicago, Illinois 60603 William J. Scanlon, Esq.
2034 Pauline Boulevard James E. Brunner, Esq.
Ann Arbor, Michigan 48103 Consumers Pooer Company 212 West Michigan Avenue U.S. Nuclear Regulatory Cornission Jackson, Michigan 49201 Resident Inspectors Office Route 7 Myron M. Cherry, Esq.
Midland, Michigan 48640 1 IBM Plaza Chicago, Illinois 60611 Ms. Barbara Stamiris 5795 N. River Ms. Mary Sinclair Freeland, *'.ichigan 48623 E711 Summerset Drive Midland, Michican 4S640 Steaart H. Fr eeman Assistant Attcrney General State cf '<ichigan Environmantal Prctectier, Di vision 720 Law Building Lansing, Michigan 48913 Mr. Wendell Marshall Route 10 Midland, Michigan 48640 Mr. Steve Gadler 2120 Carter Avenue 5t. Paul, Minnesota 55108 5
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Et;CLOSL'RE P.SB OPEN ITEM MIDLAriD SECTION 6.3 Provide modifications to'the HPI system so that operator action to throttle HPI flow in.a line for which the high flow alarm setpoint is reached is not recuired folicming a s.all break LOCA.
This action is necessary for thecberent Midland design in the event of an HPI line break.
o I
o 6.3 Emergency Core Cooling System 6.3.1 Design Bases Criterion 35 of the General Design Criteria and Section 50.46 of 10 CFR Part 50 require that an emergency core cooling system be provided which can perform its safety function assuming a single failure.
.i The Midland emergency core cooling system is desicned to provide 1
eTergency core ccoling daring those postulated accident conditions where it is assumed that mechanical failures occur in the reactor J
coolant system piping resulting in loss of coolant from the reactor f
vessel greater than the available coolant makeup capacity using i
}
normal opera, ting equipment.
The emergency ccre coolin' system is j
also designed to protect against steam line break consecuences.
Evaluation of postulated main steam line breaks is discussed in SER Section 15.1.5.
The F.idland emercency core cooling system is similar in desicn, size, and capacity to those of previously licensed Eabccck and
,,ilcc>,177 Fuel Assembly lowered-icop pressured water reactors.
i I
The system design tases are to prevent' fuel ar.d cladding damage that w;uld interfere with adecua.te emergency core ccoling and to rininize the a Dunt of clad-water reaction for any size break up to and a
4 including a double-ended rupture of the largest primary coolant pipe.
These requirements are to be met even with loss of one train of engineered safety features.
l The emergency core cooling system is designed with the required number, diversity, reliability, and redundancy of components such that no sincle failure of active emergency core cooling sys.em equipment during the short term or no single failure of active or passive equipment during the long-term of an accident will' result i
i 1
Instrunentation and controls in inadequate cooling of the reactor core.
The environ-for ECCS systems are discussed in Section 7 of this SER.
mental qualification of the ECCS is discussed in Section 3.11.
6.3.2 Svstem Desicn The Midland emercency core cooling system desicn consists cf two separate ar.d independent subsystem trains.
Each series train consists of a high pressure injection pump, a low pressure inj;; tion pump, and associated valves,in the flow path.
Separate er..gency power sources are supplied to the redundant active compor,ents for actuation of the emergency core cooling system.
- :parate diesel s
The generators prov{de power in the event offsite power is lost.
pumps in each train for the energency core ccoling system injection
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node of operation will initially'take suction from the borated water s torage tar.k and deliver ficw to the reactor throuch the reactor coolant cold lecs and core flood tank connections.
For the recirculation moce of operation, each train will automatically 4
switch to take suction from the contair. rent sura.
Each of the cere flood tants will have a total voluna of 1,210 cetic fest with a r.ominal vclura of borated water of 1,040 cubic feet at a pressure of 600 ppunds ;er scuare inch gauge.
The minirum boric acid ccncentration will be 2,270 parts per millien.
The tanks are connected to the reactor coolant syster via two separate ar.d ccr,pletely independent injection lines.
These injection lines, each containing a normally open, motor-operated gate valve and two check valves, connect to the reactor coolant system at the reactor vertel via two injection no::les.
Adr.inistrative procedures identified in.the technical specifications will require the power to these norma,lly open isolation valves to be disconnected by locking open the circuit breaker to the motor operator during reactor power ope ra ti on.
This procedere provides additional assurance that inadvertent valve closure, would not occur at the tin 2 of a loss-of-coolant accident.
Tne high pressure injection mode of operation will censist of the i
operation of two high pressure in.iection pumps (rated at 500 gallons per minute at 600 pounds per square inch gauge) which provide injection of borated water from the borated water storage tank.
There are three high pressure injection pumps, each sized to deliver 100 percent of the design high oressure injection flow.
Two of these puros are capable of being powered by the emeroency c esels, and ont serves as a spare.
One of the pumps is operating continuously in its dual role es a makeup pump, as part of the plant makeup and purification system (Section 9 of this report).
Each of the two high pressuFe injection pumps discharce to headers which are cross connected to provice ficw to the four cold legs through rotor-oper,ated injection valves which cpen on receipt of an emargency core cooling actuation signal (ECCAS).
The cross connect line! allow any one HP1 oump to inject water through
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all four cold lug ' injection nozzles.
The emergency core cooling actuation signal will be generated,then a low reactor coolant system pressure or high containment pressure condition exists.
Lcw pressure injection will be provihd by two low pressure injection
- umps, which also serve as the decay heat re ci'al pumps caring the shutdc..n cooling mode of cperation.
Each' pump. icill provide a r,ir.imu-ficw of '000 gallons ;er mir.ute at 100 pounds per scuare inch cauge and will take its suction frc the bora'.ed water storage tank and inject into the core ficod injection lir.e.
Upon actuationof the borated water storage tank low level signal from the recirculation i'
actuation system (PAS), tre vaives ccatrt.il Sg ssction frcm the I
containment sump autenatically open and tne low pressure injection valves controlli g suction f' rom the borated water storage tank are auto-ratically shut.
The emergency core cooling system will then provide the long-term cooling requirements by recirculating the spilled reactor coolant collected in the containment' sump back to the reactor vessel through tha core flood nozzles in the vessel.
Should the break size be
small enough to maintain reactor coolant system pressure higher than the low pressure injection pump head, the required flow will l
be delivered by connecting the low pressure injection pumps t
j-to the suction of the b*gh pressure pumps. This "pincy-back" alignment will be atoomplished nanually from the control room prior to the borated water storage tank low level signal.
6.3.3
_Desien Evaluation As discussed in the previcus section, the e >ergency core cooling l
system includes the piping, valves, pumps, heat exchangers, instrumentation and controls used to transfer heat from the core following a loss-of-coolant accident.
The scope of review of tne emergency core cooli,ng system for the Midland units included piping and instrumentatien diagrams, equipment layout drawings, failure redes and effects, analyses, and design specifications for essential conponents.
The review has included the applicant's proposed cesign criteria and design bases for the enargency core cooling system and the manner in which he design confoms to these criteria and bases.
Specifically, we evaluated the system's ability to l
withstand a single active failure during the short term or a single active or passive failure durine the long term following j
a pcs tulated loss-cf-coolant accident.
i f
The t.;plicent's emergsncy core coolinc system failure nodes and cf fects tnalysis provide the basis for shewing that the ",idland energenc;. core cooling system can prcvide sufficient core cooling f
assuming a single failure.
The applicant has identified the j
folitwing mc or-operated valves which do not receive an ECCS ectuatien l
signal and which must be properly aligned in the event of a postulated loss'-of-coolant accident:
Function (Positior) valve j
Low Pressure Injecticn Block Yalve (0 pen) l 1120 A,B 1114 A,B Decay Heat Removal Heat Exchanger Bypass (Shut) 1101I,B Core Ficod Tank Discharge Line Isolation Valve (0 pen) 1105 A,B Core Flood Tank Vent Valve (Shut) l L'e have verified that the above valves and the associated positions are listed in the Midland technical specifications.
Le have reviewed
- L: bars given are Unit 2.
The same valves are on Unit 1.
o the Midland emergency core cooling system piping and identification diagrams in the Final Safety Analysis Report and conclude that 4
the applicant's failure modes and effects analysis is complete, and that the single failure assumptions of the ECCS performance evaluation are valid.
The energency core cooling system capability to perform its function assuming single electrical failures is address-1 ed in Section 7 and 8 of this safety evaluation report.
The potential for valve rotors and associated valve control /
actuation systems to became submarged within the containment following l
a loss-of-coolant stci'de'nt was reviewed. At the request of the staff, the applicant provided a list of all emergency core cooling system valves inside containment and their position with respect to the
. maximum calculated water level in the containme u following a loss-of-caol ant a.ccident. All valves are ic:ated abcVe the maximum cal-culated level.
We reviewed the applicant's assumptions for this calculation, and the function and the location of the ecuipment in-side containment es stated in the Midland' safety analysis reptet.
The naxis;m calculated containment water level considers the displacement of water by submerged structures and ecuipment and is based on con-servative assumptions related to tne folio.cing scurces of water:
i j
1.
The entire reacter coolant system, inventory; 2.
The ccntents of two core flood tanks; i
3.
Full torated water stor. age tank containing 500,000 call ons ;
a.
Licaid inventory of nonseismic Category I con;;nent cooling i
cnd service water within containment; and, 5.
Additional two feet level margin to provide for uncertainties.
the calculation assumptions are conservative and all active equipment inside containmet,- wnich functiers as part of the emergency core cooling We therefore system in the long and short term does not become submerged.
find' the Midland units acceptable in this regard.
With regard to long-term cooling, we requested that the applicant evaluate the emergency core ccoling system performance by applying a single failure analysis; which includes consicering passive as well as active f ailures in the emergency core cociing system
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4 (assuming no prior failure during the short-term phase).
As part of the passive failure evaluation, the applicant addressed the concern of post-loss-of-coolant accident water leakage from emergency core cooling system components such as a failed pump seal or valve stem packing which could degrade more thin one sub-system. The maximum postulated leakace rate from passive failures originally proposed by the applicant was 500 milliliters per minute from a failed low pressure injection pump bushing.
The basis for the proposal was tests performad by the pump seal manufacturer which demonstrated a maximum leakage rate of 29 milliliters per miHute. While the test and associated test report indicate that seai integrity was maintained for the conditions
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under which they were tested, we did not concur wita the propcsal
-of 500 millilithrs per minute to serve as the bounding leak rate for a passive failure following a less-of-coolant accident, since operating data indicate that leaks in excess of-this rate have occurred.
Subsec,uently, the applicant assumed 30 gallons per minute fcr the purposes of evaluating emergency core cooling system performance.
The applicant described the capability of the "idiand design in the safety ant jsis repcrt and in response to car recuest for additicnal info-maticn.
Cetat:icn of passive failures is through the use of rco
..ner level switches and airborne radiation monitors.
Emergency core cooling system ecuip.ent and cc ponents are lobated in s..i:":ipt s=partments of the auniiary buildinc.
Se;arate watertight cc ;artments are' provided for each ECC5 train so that fic; ding of one cc partment will not jeo;ardize operation of the redundant train.
Each compartment is monitored for ficoding using redundant Class lE room water level switches placed at two levels i
in each room.
The airborne radicac:ivity tc-itors will also alert the operater to lea'. ige so that the leak can be isolated.
Each ECCS co.ictment is also provided with drain lines which c;n te opened remotely.
We have concluded that the Midland design for protection against passive failures is acceptable.
5 Ve have also reviewed the system design for preventing excessive b ric acid buildup in the reactor vessel during the long-term j
cooling period after a loss-of-coolar.t accider.t.
The Midland system which perforts this function will consist of redundant flow l
paths from the decay heat removal system drop line (from one reactor coolant locp hot leg) directly to the reactor building enargency I
sunp.
Each " dump to sump" line contains a relief valve and two l
. solenoid-operated valves which are placed at an elevation to l
l prevent the possibility of flooding.
The applicant has calculated t
that 40 gallons per minute flow through one " dump to sump" line initiated 7 days after the posteiated less-of-coolant accident will prevent excessive borjc.' acid buildup in the core during the post-los s-of-coolant accident period.
Calculatiens on similar Babcock and Wilcox plants by applicants and the staff shcw that l
relatively long periods are available before " dump to sump" would be requ. red due to th.e naturcl circulation between the reactor 2
vessel upper ple'num and downtonar provided by the internal vent valves which are unique to Babcock and Wilcox plant's.
The " dump to sump" made was previously found acceptable.on other Babcock and Wilcox piants.
Also, for additional conservatism, the applicant will pro. sdurally open the " dump to sump" valves at approxir.ately i
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter a loss-of-coolant accident.
At the request of the l
s t,aff, in order to verify adequate dilution flow, the applicant w li instail a class IE fica-reasuring devjce and associated indication in each line so that the control r:c cperator can
(
r verify ade;; ate ficw when the system is initiated.
The applicant l
has also co7:ittcd to a preoperational test to verify the system design basis.
On the basis of the similarity of this design to evicusly reviewed Sabcock and Wilcox plants and the commitment to conduct a preoperational test, we find the system design acceptable.
We have revie. ed the Midland enercency core c:ciing system with respect to differ.ences from previously licensed 177 fuel assembly j
plants cc.ergency, ore cooling system.
The following differences c
were identified:
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1.
Midland has two emargency core cooling sysic valves located in series in the makeup tank outlet line.
These valves are closed automatically by an emercency core cooling actuation signal.
This action eliminates any pottntial for the gas in the makeup tank to reach the high pressure injection pudp suctio'n in the emergency core cooling system operating mode.
These valves eliminate an administrative control on makeup tank fluid level and gas pressure during normal operation used by sore previously licensed 177 fuel assembly plants.
Midland has proteckive interlocks for the high pressure 2.
injection / makeup. pumps and low pressure injection pumps for pump protection agaii.st misaligned valves during normal operatitn.
Thpse normal operation interlocks a., overridden by an emergency core cooling actuation signal.
7 3.
Midland has two ECCS valves in the normal maEcbp line to the reactor coolant system which are automati.cally closed by an ECCS actuation signal.
This ensures automatic closure of the line even wit 5 a single failure and ensures adequate ficw distribution for ECCS.
4.
Midland has two valves in series in the letdtv.n line upstream cf the letdo,n cooiers.
These valves close autonatically when rich fic.erate is ser.sa: u;;;r:a c' tha letco,en coolers, which minimizes the amount of coolant released for a letdown line break outside of containment.
5.
The Midland design includts two fioW control valves in each 1 05: pres.ure injection line.
The second valve is downstream 1
of the low pressure injection-to-lme pressure inje c:icn cross-
.l ccnnect and is used af ter a loss-of-cociant accicent to split the flow from one Icw pressure injection pump through botn low pre'ssure injection lines should one low pressure injection pump fail.
The flow split can be accomplished frem the control room.
This codification enhances ECCS effectiseness for certain break 4
5 locations.
6.
Midland has automatic switchover of the suction for the emergency core cooling system pumps from the borated water storage tank to the containment sunp.
In some previously licensed plants this switchover is acconolisheo by the control room operator.
7.
The Midland design has two direct dump-to-sump lines to prevent boron precipitation in the core during the long-term cooling period af ter a loss-of-coolant accident.
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8.
Midland has croes connected high pressure injection lines so that one HpI pump can inject in all four cold leg locations.
This design redu'ces the amount of ECCS water that could be lost from a postviated small cold leg break.
Each of these differences vias included in our ~ review to assure that the requirements of single failure requirements of 10 CFR 50.46 and GDC 35 are met in all respects.
1:e ccnclude that they provide an overail inprovement in the emergency core cooling system design and are aer.eptable.
6. 3..'
Ferfor ance Evaluation Acceptance criteria for Emargancy Core Ceoling System for Light Urter ^eactcrs are mvided ir. iC CFr EL.'E c' tFc Co-ission's re gul ati cr.s.
The recuirerent for acce-tance is that in the ever.t of a postulated loss-of-coolant accident the calculated cooling perfom.ance of the ECCS will limit ccre darage below specified criteria for (1) peak cladding terperature (2) maximum cladding oxidation, (3) maximum hydrogen generation, (4) coolable geometry and (5) long te r.n ccolinc.
Tne requireants for calculational m.:his are rovided in Appendix K to 10 CFR 50.
The l'.idlend Fina Safety Evaluation report incorporates by reference the B&W ECCS Evaluation model.
This mcdel is described in the topical report bah'-10104, Rev. 4, Ref.1.
Final NRC approval
was issued in Reference 2 in wnich the model was found to be in rompliance with Appendix K.
Calculations of core damage for a spectrum of postulated break sizes are presented in~ EAW-10103A Ref. 3 and additional analyses of small breaks are presented in Reference 4.
These analyses utilize input assumptions designed to be conservative for all S&W designed plants with lowered loops and therefore utilize decay heat assumptions which are 8% higher than those which would be required for the Midland plants.
The analyses ident,ified the worst break as a postulated 8.55 square The calculated foot rupture at-the discharge of a reactor coolant pump.
consequences of the entire range of break sizes is below the core i
damage criteria lof 10 CFR 50.46.
The table below sumr.arizes the results of the analysis and provides
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the allowable linear heat generation rate as a function of core elevation.
Linear Heat Peak Cladding Maximum Local Elevation Generation E-te Temperature Oxidation (feet)
Limit (kilowatts (degrees Fahrenheit)
(Fercent) cer foot) 2 15.5 2002 3.92 4
16.6 2135 4.59 5
1E.0 2146 5.45 8
i7.0 2i10 5.19 10 16.0 19'7 3.02 The maxinum core-wide tetal-water reaction was calculated to be 0.647 As percent, a value which is below the allowaoie limit of one percent.
shown in the tabulation, the calculated values for the peak clad temper-ature and local metal-water reaction were telow the alia.able limits specified in 10 CFR Part 50.45 of 220L^F and 17 ;erc;nt respectively.
E AU-10103 has,. also shown. that the core gecmetry remains amenable to cooiing and that long-term core ccoling can be established.
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l y
Following the March 1979 event at TMI-2, a number of break si:es were Unile analyzed by B&W for the purposes of operator training, Reference 7.
the ccre conditions calculated for these break sizes were less limiting than those break sir?s previously analyzed, operation of the auxiliary feedwater system was found to be required for break sizes of.01 ft and less to adequately remove decay heat during the recovery from the small break.
The, staff evaluation of the auxiliary feedwater system for the Midland units is described in Section 10.4.
The auxiliary feedvater system for Midland is designed to safety grade criteria and we have concluded that reliance en the auxiliary feedwater system for LOCA mitigation to te consistent with the Commiss.on's regulations.
r Following the sfaff review of the additional small break LOCA analyses performed subsequent to the March 1979 Ti.:-2 event, several sections of thed&W snall break evaluation model were identified r.s This review is documented requiring further experimental verification.
in l'UREG-05E5, (Reference 8).
These recommendations have been incorpor-ated in the TMI-2 Action Plan Item II.K.3.30 as described in tiU:.EG-0737 The resolution of the f?UREG-0737 requirements is discussed Refer'nce 9.
elsewhere in this Safety Evaluation Report.
The staff cpinicn continues to te that the ELW LOCA evaluation model is in conformance with Appendix K pending completion of the code verification program.
To support an 0:erating ccnfiguration witn less than four reactor crolcr '. p.:mps en the lirs (pc.rtiti loop), the staff requires an analysis of the predicted c'onsequences of a less-of-coolant accident occurring during the propcsed partial loop operating mode (s).
The i
applicant submitted an analysis for partial loop operation with one idle reactor coolant pump (three pumps operating) by reference to Using a reduced power level cf 77 percent of rated E AW-10103.
power, Babcock and Wilcox performd this analysis assuming the
= 1) and maximum LH2R worst case break (8.55 senre feet DE, CD The worst (18.0 Lilowatt'.s per foot) from the 4-pump analysis.
was located in the active leg of the partially breat selected Piacing the break at the discharge of the pump in an idle icop.
active cold leg of the partially idie loop (instead.of at the discharge of the pump in an active cold leg of the fully active loop) yields the rest degraded positive fice through the core during the first half of the blowdown and results in higher cladding temperatures.
Additional conservatism was demonstrated by B&W in revised calculations 2
for the postulated 8.55 ft cold leg break Ref. 5 using recalculated inlct nozzle and reactor coolant system pressure distributions.
These j
analyses resulted in a decrease in the calculated peak cladding temperature of 86*F.
The additional calculations of postulated small break sizes presented in Reference 4 demonstrated that for certain small break sizes operator action w'ould be necessary for currently operating Bla.'
designed plants to manually balance HPI fhw so that no more than 30% of the HPI flow could be lost from a cold leg break adjacent to an HPI nozzle.
This manual action was required'no later than ten minutes after occurrence of the postulated break.
The staff took the pcsition that prompt n
ope ator actionyas not allowed for LOCA mitigation under 10 CFR 50.45 and consequently found it necessary to issue license exemptions to the operating piants designed by B&W.
Reference 6 is an example.
The HPI system for Midland, however, has been designed with cross connect lines between the HPI train:: so that prompt operator action will no longer be necessary to adjust flow in the event of a small cold leg break.
In the event th'e breaks were in an HFI line, calcula-tions sutnitted by the applicant indicate that the cperator would have tc inic:e the CI train. 'th its brcken ii-e t.ithin 20 ninutes to cer.ain within the limits of 10 CFR 50.45.
The existence of a broken EP: line t.culd be indicated,to the operator by a high HPI flew alarm.
He do not believe the Midland design teets the requirements of 10 CFR 50.46 regarding the continued reliance en prompt operatcr action since it requires that the operator terminate HPI flow in the line foc which the flow is the greatest before long terr ccoling cf the core is c.c compl i s hed.
Considering the nature cf ti.is action (termination of HPI flow), he believe that the HPI system for Midland should be modified so that prcnpt operator action is no icnger required for LOCA mitigation.
Such a design, modification has been accepted on TMI-l, a BSW designed plant similar to Midland.
He will report on resolution of this issue in a SER supplement.
The maximum cladding temperature for the one-idle-pump mode of op,eratinc.
is 1784 degrees Fahrenheit, which reets the requiremnts of 10 CFR 50.46. We have requested additional informtion on other transients and accidents which could occur during partial loop operation. These evaluations are discussed in Section 15 of this SER.
6.3.5 Tests and Inspections The applicant will ' demonstrate the operability of the emergency core cooling system by subjecting all components to preoperational The tests will in caneral, be performd in conformance with tests.
Regulatory Guides 1.68, "Preoperational and Initial Startup Test Programs for U ght-Water-Cooled Power Reactors," and 1.79, Revision 1, "Preoperational Testing of Emcrgercy Core Cooling Systems for Pressurized Water Reactors."
Certain features of the Midland design prevent full compliance wil1 the Regulatory Guides.
The degree of compliance is addressed in Appendix 3A to the FSAR.
The NRC staff has reviewed the exceptions to the requirements of Regulatory Guides 1.68 and 1.79 identified by the applicant and concludes that in each case that the essence of the Regulatory Guide recommendations are met.
The more significant exceptions
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are as f'ollows:
a)
Etercency diesel and ECCS pump tests will be conducted separately to avoid the pumping of larce quantities of water.
This is acceptable since the electric recuirenents of the' ECCS pumps is independent of the power scurce.
b)
In accordance with Regulaiory Guide 1.79 the capability of the HPIs to deliver <Eter to the reactor system under accident ccnditions will be verified by analysis based on as-built pump curves rather than by a test to avoid subjecting the reactor system to cold water shock.
Operation of the check valves will be verified in a separate test.
c)
In-plant recirculation tests from the sump to the primary system willnotbeder'ormed.
Instead, the applicant has demonstrated that adequate NFSH will be available analytically.
The calculations i
)
utilize minimJm BWST water level and sump water temperature to min-inize available head and maximum piping losses.
The method will be further checked by comparison of caiculated and r.easured losses in the lines connecting the Decay Heat pumps to the BWST.
Vortex con-trol and sump screen losses have been verified utilizinc a full-scale
model test performed by Western ' Canada Hydraulic Laboratories Ltd.,
F.eference 10.
Sump screen blockages up to 95% were tested.
The flow straightening ability of the sump screen and grating cage pre-vented vortex formation in all cases.
The head loss.across the in-take screen was determined to be only 0.009 feet, which will not prevent adequate NPSH for the LPI pumps.
After the plant is brought into operation, periodic tests and inspections of the emergency core cooling system components are performed in accordance with the unit technical specifications and tre Fump and Valve Inservice Test-ing Program required Ey the American Society of Mechanical Engineers Code,Section XI.
With, respect to testing of valves in the emergency core cooling cystem which serve as a precsure boundary between the reactor coolant system' i
and low pressure Systems, the applicant has committed to test the check valves in the core ficod"ta'nks and decay heat removal injection lines as well as
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the auxiliary pressuricer spray line.
Compliance of these tests to accept-l l
able leakage criteria is addressed in Section 3.9.6.
The emergency sump will be visually examined during ecch refueling.
6.3.6 Conclusions l
Eased on cur review, exceot as noted above, we have deternined that the Midland erergency core ccolir.g syste: will meet the acceptance criteria ar.d ccr form to he Com:nissicn's rec,uirements as set forth in the General Design Criteria, regulatory gu, ides, and staff technical pcsitions.
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SECTION .3 REFERENCES 1.
"E." s ECCS Evaluation Model," E AW-101401, Fev.4, l'ay,1978.
2.
Letter from S. A. Yarga (NRC) to J. H. Taylor (E&W), September 5,1978.
3.
"ECCS Analysis of' B&W's 177-FA Loucered-Loop NSSS Rev.
3," EAW-10103A, Rev. 3, July,1977.
4.
Letter, J. F. Taylor (B&W) to S. A. Yarga (NRC), July 18, 1978.
5.
Letter from J. H. Taylor (B&W) to R. L. Baer (NRC), July 8,1977.
6.
Letter from R. kT. Reid (NRC) to J. G. Henbein (Mad Ed) furch 16, 1978.
7.
" Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the'If7 Fuel Assembly Plant, S&W Report, ley 7,1979.
8.
" Generic Evaluation of Small Break Loss-of _ Coolant Accident Eehavior in Babcock & Wilcox Designed 177-FA Operating Plants," NUREG-0565, January,1980.
9.
" Clarification of TMI Action Fian Requirements," NUREG-0737, November, 1950.
1 10.
" Consumers Power Co. Midland' Nuclear Flant Units 1 & 2, Madel Testing Of Centainment Sump ECCS Recirculation Intales fcr Eechtel Poveer Company" by Western Canada Hydraulic Ltd., April,19S0.
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