ML20027D254
| ML20027D254 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/07/1982 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| IEB-80-04, IEB-80-4, TAC-46844, NUDOCS 8211030447 | |
| Download: ML20027D254 (8) | |
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DC M D ?lb OCT 7 1982 DISTRIBUTION VDocket File Local PDR ORB Rdg D.Eisenhut s
Docket No. 50-336 JHeltemes RAClark PKreutzer OELD J' T' %
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Mr. W. G. Counsil. Vice President E.L. Jordan.
Nuclear Engineering & Operations J.M. Taylor (1)
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Northeast Nuclear Energy Company
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P. O. Box 270 DHWagrer Hartford, Connecticut 06101 Gray File
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Dear Mr. Counsil:
ETourig?y
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N RE: RESOLUTION OF MAIN STEAM LINE BREAK WITH CONTIN $CD FEEDWATER ADDITION EVENT FOR MILLSTONE NUCLEAR POWER STATION., UNIT NO.2,
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7, By letters dated January 25, 1980 and July 6,1982, you responded tb I&E Bulletin 80-04 (issued on February 8, l'80) regardingd4ain Steam Line Brsak with Continued...
Feedwater Addition for Millstone Unit No. 2. '
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In response to this issue, we and our contrhetor, 6.ie Frankil3 Research Center'(FRC{,
have completed a review based on your submittais. Enclosed {'i the Safety Evaldation with attached FRC Technical Evaluation Report documentins.this review. We concl that your analysis is acceptable with respect to our requirenents in I&E ng-80-04, and therefore this issue is closed for Millstone Unit'ifo. 4.
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1 Sincerely.
s OriWnsi mgned h va Robert A. Cldrh., Chief
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Enclosure:
Safety Evaluation.
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MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION
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MILLSTONE NUCLEAR PLANT, UNIT 2 s
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,A In the summer of 1979, a pressurized water reactor (PWR) Licensee
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submitted a report to the NRC'thqt identiffsd a deficiency in its 4
original analysis of the containment pressurization resulting from
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.a postulatec mainSstcam ireak.(M3LB).
A reanalysis of the
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. containment p r e's s u r e,rj s p u n s e following a MSLB was performed, and 7 c-it was dete rmined t ha t, '"i f. t h e auxili a r y f eedwate r (AFW) system 4
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continued to supplf,fepdwater at runout conditions to the steam
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'g ene ra t o r t h a t ' n ad e xp e ri en,c e d t h e s t e a m line break, the
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containmentyd? sign \\,pyessure would N.-
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e exceeded in approximately 10 s-o 3
s ot h e th w o'r d s, the l\\1ng-term blowdown of the water minutes.
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supplied by the AFW system had not been considered in the earlier s
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00n O c t o biir.1', 1,9 79, the f oregoing -inf ormation', was provided to all s
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(s!ders o f.sopergting Licenses and constructjon-permits in IE
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'I n f o r m a t i o\\n Motice 79-Ri C2].
Another licensee performed an s -
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accident" analysis review" pursuant to the,information furnished in
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the above'citednotiYe^anddiscoveredthathwithoffsiteelectrical i
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" p o w'e r-a v aY t a b l e, the condensate pumps would feed the aff.ected steam
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generator at an excessive rate.
This~ excessive feed had not been considered in its analysis of the postulated MSLB accident.
,1 A third Licensee informed the NRC of an Error in the MSLB analysis for their plant.
For a zero or low power condition at the end of core life, the Licensee identified an incorrect postulation that the startup feedwater control valves would remain positioned "as is" during the transient.
In reality, the startup feedwater control valves witL ramp to 80% futt open due to an override signal resulting from the low steam generator pressure reactor trip
' signal.
Reanalysis of the events showed that the rate of feedwater addition to the,affected steam generator associated with.the opening of the startup valve would cause a rapid reactor cooldown and resultant reactor-return-to pnwer response, a condition which is beyond the plant's design basis.
Following the identification of these deficiencies in the original MSLB accident analysis, the NRC issued IE Bulletin 80-04 on February 8, 1980.
This bulletin required all Licensees of PWRs and cert'ain near-term PWR operating License a p p li c'a'n t s to do the
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folLowing:
"1.
Review the containment pressure response analysis to determine if the potential for containment overpressure for M'SLB inside contdinment include *d the impact of runout f lo*w from the auxiliary feedwater system and the impact of other energy sources such as continuation of feedwater or condensate flow.
In your review, consider your ability to
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detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable after extended operation at runout flow.
"2.
Review your analysis of the reactivity increase which results from a MSLB inside or outside containment.
This review should consider the reactor cooldown rate and the potential for the reactor to return to power with the most reactive control rod in the fully withdrawn position.
If your previous analysis did not consider alL potential water sources (such as those
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Listed in 1 above) and if the reactivity increase is greater than previ'ous analysis ind'cated, the report of this' review should include:
a.
The boundary conditions for the analysis, e.g.,
the end of Life shutdown margir, the moderator temperature coefficient, power level and the net effect of the associated steam generato.' water inventory on the reactor system cooling, etc.;
b.
The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of -high concentration boric acid solution to the reactor coolant system; c.
The effect of extended water supply to the affected i
steam generator on the core criticality and return to power; and d.
The hot channel factors corresponding to the most reactive rod in the fully withdrawn positions.at the end - - _. -
of life, and the Minimum Departure from Nucleate Boiling Ratio (MDNBR) values for the analyzed transient.
"3 If the potential for containment ove rpr,etsu r e exi s t s or t he reactor return-to power response worsens, provide a proposed corrective action and a schedule for completion of the corrective action.
If the unit is operating, provide a description of any interim action that will be taken until the proposed corrective action is completed."
Following the licensee's initial response to IE Bulletin 80-04, a
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request for additional information was developed to obtain all i
the information necessary to evaluate the licensee's analysis.
The results of our evaluation for Millstone Nuclear Plant, Unit 2 (Millstone 2) are provided below.
2.0 Evaluation Our consultant, the Franklin Research Center (FRC), has reviewed i
the submittals made by the licenseein response to IE Bulletin 80-04, and prepared the attached Technical Evaluat on Report (TER).
We have reviewed this evaluation and concur in its bases and findings.
3.0 Conclusion Based on our review of the attached TER, the following conclusions are made regarding the postulated MSLB with continued feedwater
-. addit _ ion for Millstone 2:
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There is no potentia'l for containment overpressurization resulting from a MSLB with continued feedwater addition because the main feedwater system is isolated and the initiation of auxiliary feedwater flow'to the affected steam
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generator is delayed; 2.
No damage would be incurred by the AFW pumps since the calculated runout flow rate is written design capability of the pumps; 3.
All potential water sources were identified and, although a reactor return-to power is predicted, there is no violation of the specified acceptable fuel design li m.i t s. ;
4, J{o further action regarding IE Bulletin 80-04 is r e q'u i r e d.
4.0 Refe ences 1.
" Analysis of a PWR Main Steam Line Break with Continued Feedwater Addition," NRC Office of Inspection and Enfore,ement, February 8, 1980, IE Bulletin 80-04 l
2.
"Overpressurization of the Containment of a PWR Plant after a Mai,n Steam Line Break," NRC Office of Inspection and.
Enforcement, Octobe r 1,1979, IE Information Notice 79-24 3.
W.
G. Counsfl (NNECO) letter't'o B.
H. Grier (NRC, Region I)
Subject:
Haddam Neck Plant, Millstone Nuclear Power Station U. nit 2, " Analysis of a PWR Main. Steam Line Break with Continued Feedwater Addition," March 5,1960 4.
W.
G.
Counsit (NNECO) letter to R.
Reid (NRC, ORB)
Subject:
Millstone Nuclear Power Station, Unit 2, " Automatic Initiation of Auxiliary Feedwater," January 25, 1980
l 5.
W.
G.
Counsil and J. P.
Cagnetts (NNECO) letter to R. A.
Subject:
IE Bulletin 80-04, Response to i
Request for Additional Information, July 6, 1982 6.
Millstone Nuclear Power Station, Unit 2, Final Saf ety Analysis Report, Northeast Nuclear Energy Company 7.
Technical Evaluation Report, TER-C5506-119, "PWR Main Steam Line Break with Continued Feedwater Addition - Review of Acceptance Criteria," F rank lin Resea rch Cent e r, November 17, 1981 8.
" Criteria f or Protection Systems for Nuclear Power
. Generating Stations," Institute of Electrical and E le ct roni cp ' Engine e rs, New Yo rk, NY,1971 9'.'
'SYanda rd Revi ew P lan, Section 4.2, " Fuel System Design," NRC, J uly 1981, NUREG-0800 10.
Standard Review Plan, Section 15.1.5, " Steam System Piping Failures Inside and Outside of Containment (PWR)," NRC, July, 1981, NUREG-0800 11.
" Criteria for Accident Monitoring Functions in Light-Water Cooled Reactors," American Nuclear Society, Hinsdale, IL, D e c'e mb e r 19 80, 'AN S / A N S I-4.5-19'80
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12.
" Instrumentation f o r Li ght-Wa,t e r-Cooled Nuclea r Powe r P lant s to Assess Plant and Environs Conditions During and Following an Accident," Rev.
2, NRC, December 1980, Regulatory Guide 1*.97 13.
" Single Failure criteria for PWR Fluid Systems," American Nuclear Society, Hinsdale, IL, June 1976, ANS-51.7/N658-1976 i -
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" Quality Group Classifications a'nd Standards for Water, Steam, and Radioactive-Waste-Containing Components of Nuclear Power Plants," Rev. 3, NRC, February 1976, Regulatory Guide 1.26 15.
" Interim Staff Position on Environmental Qualification of Sa f ety-Re lat ed Ele ct ri ca l Equipment," Rev. 1, NR C, J u ly 1981, NUREG-0588 16.
R.
Reid (NRC, ORB) letter to W. G. Counsfl (NNECO)
December 21, 1979 17.
W.
G.
Counsil (NNECO) letter to J. Hendrie (NRC)
Subject:
Haddam Neck Plant', Millstone Nuclear Power Station, Unit 2, Automatic Initiation of Auxiliary Feedwater, November 30, 1979
- 18. 'W! G.
Counsil (NNECO) letter to R. Reid (NRC, ORB)
Subject:
Millstone Nuclear Power Station, Unit 2, Proposed License Amendment, Power Uprating, February 12, 1979
Attachment:
FRC Technical Evaluation Report l
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