ML17256A491
| ML17256A491 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/09/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17256A490 | List: |
| References | |
| TASK-04-03, TASK-06-02.D, TASK-4-3, TASK-6-2.D, TASK-RR IEB-80-04, IEB-80-4, NUDOCS 8302160266 | |
| Download: ML17256A491 (11) | |
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1 UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGUL'ATION MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION ROCHESTER GAS AND ELECTRIC CORPORATION R.
E.
GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 1.0 'NTRODUCTION In the summer of 1979' pressuri zed water r eactor (PWR) Licensee submitted a report to the NRC that identified a deficiency. in its original analysis of containment pressurization resul.ting from a
postulated main steam l.ine break (NSLB).
A reana Lysis of the containment pressure response foLLowing a MSLB was performed~
and it was determined that~ if the auxiliary feedwater (AFW) system continued to suppLy feedwater at runout conditions,to the steam generator that had experi,enced'the steam Line break~
the containment design pressure would be exceeded in approximately 10 minutes.
In other words~ the long-term blowdown of the water
'upplied by the AFW system had, not been considered in the earlier ana Lysis.
On October 1~ 1979~ the foregoing information was provided to all holders of operating Licenses and construction permits in IE Information Notice 79-24 C23.
Another L,icensee performed an accident analysis review pursuant to the information furnished in the above c'ited notice and discovered that~ with offsite "I
electricaL power available~
the condensate pumps would feed the" affected steam generator at an excessive rate.
This excessive feed had not been considered in the analysis of the postulated MSLB accident.,'D
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A third Li censee informed the NRC of an er ror in the NSLB analysis for their plant.
For a
hiero or low'p>wer 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 tr ansient.
In reality~
the startup feedwater controL valves wilL ramp to 80X fuLL 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"power response~
a condition which is beyond the plant's design basis.
Following the'dentification of these. deficiericies in the originaL NSLB accident analysis~
the NRC issued IE BulLetin"..
80-04 on Febr uary 8~ 1980.
This buL Letin required alL Licensees of PMRs and near"term PMR operating License applicants to,do the following:
"1.
Review the containment pressure response analysis to determine if the potentiaL for containment overpressur e
in the event of a
HSLB inside containment included the impact of runout flow from the auxiliary feedwater
'ystem and the impact of other energy sources such as continuation of feedwater or condensate fLow.
In your review~ consider the ability to detect and isolate the damaged steam generator from these sources and the ability
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of the pumps to remain operable after extended operation at runout fLow.
2.
Revie~ your analysis of the reactivity increase which results from a
HSLB 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.
Xf your previous analysis did not consider aLL potentiaL t
water sources (such as those Listed in I above) and if the reactivity increase is greater than prev'ious analysis indi cated the report of this review should include:
a.
The boundary conditions for the analysis~
e.g.~
the end of life shutdown margin~ the 'moderator temperature coefficient~ power Level and the net effect of the associated steam generator 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 bor ic acid solution to the reactor coolant system; c.
The effect of extended water supply to the affected 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 Ninimum Departure from Nucleate BoiLing Ratio (NDNBR) values for the analyzed, transient.
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I 3., If the potential for containment over pressure exists or the reactor return-to-power response worsens~
provide a
proposed corrective action and a schedule for completion of the corrective action.
If the unit is oper ating~
provide a description of any inter im action that wiLL be taken until the proposed corrective action is completed."
Following the Licensee's initial response to IE BulLetin 80-0'4~
a request for additional information was developed to obtain all E
the information necessary to evaluate the Licensee's analysis.
The results of our evaluation for Ginna Nuclear PLant (Ginna) are provided beLo~.
2.0 Evaluation Our consultant~
the Franklin Resear'ch Center (FRC) ~ has reviewed 9'he, submittals made by the Licensee in: response to IE Bulletin 80"04~, and prepared the attached Technical Evaluation Report.
Me have reviewed this evaluation and concur in its bases and findings.
3.0.
Conclusion Based on our review of the enclosed Technical EvaLuation Report~
the foLLowing conclusions are nade regarding the postulated HSLB with continued feedwater addition for Ginna:
Th'ere is no potentiaL for containment overpressurization resulting from a
NSLB with continued feedwater addition because the main feedwater system is automaticaLLy iso-Lated and the auxiLiary feedwater system Limits flow to the affected steam generator;
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2., The AFM pumps are protected from the ef feces of runout fLow and therefore can be expected to carry out their intended function during the NSLB event.
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ALL potential water sour ces were identified and~ although a reactor return-to-power is predicted~
there is no
'ioLation of the specified acceptable fuel design limits.
Therefore, the EXXON Nuc Lear (17) t1SLB react ivity increase
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analys i s remains valid.
No further, action regarding IE Bulletin 80-04 is required.
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4.0 References 1 ~
2 ~
3 0 4.
5.
6.
7 ~
"Analysis of a
PMR Hain Steam Line Break with Continued Feedwater Addition~" NRC Office of Inspection and Enforcement~
February 8~
1980 IE Bulletin 80-04 "Overpressurization of the Containment of a
PMR Plant After a Hain Steam Line Break~"
NRC Office of Inspection and Enforcement~
October 1~ 1979~
IE Information Notice 79>>24 L.
D.
Mhite (RGGE)
Letter to B. J.
Gr ier (NRC~ Region I)
Subject:
Response
to IE Bulletin 80-04 April 30, 1980 I
D.'. Crutchfield (NRR~
ORB P5)
Letter to J.
A. Haier (EGSG)
Subject:
"Systematic Evaluation Program (SEP) for the R.
E. Ginna Nuclear Power Plant - Evaluation Report on Top-i cs VI-2. D and VI-3" November 3~
1981 J.
E. Haier (RGRE)
Letter to D.
H. Crutchfield (NRR~
ORB 05)
Subject~
SEP Topics VI-2D and VI"3 February 1~ 1982 Robert Emmett Ginna Nuclear Power Plant Final Facility Description and Safety Analysis Report~
through Supplement 11 Rochester Gas and Electric Corporation~
Harch 1969 Technical Evat.uation Re'port "PMR Hain Steam Line Break with Continued Feedwater Addition " Revie~ of Acceptance Criteria" Franklin Research Center~
November 17~
1981 TER-C5506-119 8.
"Criteria for Protection Systems for Nuclear Power Generating Stations" Institute of Electrical and Electronics Engineers~
New York NY 1971~
Standard Review Plan~
Section 15.1.5 "Steam System Piping Failures Inside and Outside of Containment (PMR)"
NRC~ July 1981 NUREG-0800
I I
Standard Review PLan~ Section 15.1.5 "Steam System Piping Fai Lures Inside.
and Outside of Containment (PMR)"'RC~
July 1981 NUREG-0800 Criteria for Accident Monitoring Functions in Light-Water"Cooled Reactors" American Nuc lear Society~
Hinsdale~ IL. December 1980 ANS/ANSI-4 "5-1980 "Instrumentation for Light-Mater "Cooled Nuclear Power Plants to Assess PLant and Environs Conditions During and Following an Accident" Rev.
2 NRC~
December 1980 Regulatory Guide 1.97 "Single Fai lur e Criteria for PWR F luid Systems" American Nuclear Society~
Hinsdale~
IL~ June 1976 ANS-51.7/N658-1976 "Quality Group C L'assi fi cat i ons and Standards for Mater~
Steam~
and Radioactive-Waste~
Containing Components of.
Nuclear Power PLants" Rev.. 3 NRC~ February 1976 Regulatory Guide 1.26 "Interim Staff Position on Environmental Qualification of Sa fety"Related Electrical Equipment" Rev.
1 NRC~ July 1981 NUREG-0588 T. Speis (NRR~ Assistant Director for Reactor Safety)
Memorandum to G. Lainas~
(NRR~ Assistant Director for Safety Assessment)
Subject:
Systematic Evaluation Program (SEP) for the R.
E.
Ginna Nuclear Power PLant - Evaluation Report on Topics VI-2.D and VI-3 (Docket No. 50-244)
Narch 31~
1982
."Plant Transient Analysis for the R.
E.. Ginna Unit I Nuc Lear Power P lant" Exxon Nuclear Company~ Inc.
XN"NF-77-40 Supp lement 1
Nar ch 3~ 1980
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