ML17213B016
| ML17213B016 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 01/17/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17213B015 | List: |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 8302010018 | |
| Download: ML17213B016 (8) | |
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UNITED STATES NUCLEAR REGULATORY COiViVilSSION 7'ASHIVGTON D. C. 20SGS SAFETY EVALUATION YrAI W STEAN L I WE BREAK WITH CONTI WUED FEEDHATER ADDITION ST.
LUCIE NUCLEAR PLANT UNIT 1
DOCKET WO. 50-335 1.0 Introduction
~ In the summer of 1979~
a pressurized water reactor (PWR)
Licensee submitteo a report to the WRC that identified a deficiency in its originaL ana lysis of the containment pressurization resulting from a postuLated main steam Line break (NSLB).
A reanalysis of the containnent pressure response fo llowing a
HSLB was performed~
and it was deternined that, if the auxiLiary feedwater (AFW) systen continued to supply feedwater at runout 'conditions,to the steam generator that had experienced the steam Line break~
the containnent design pressure would be exceeded in approximately 10 minutes.
In other words~
the Long-term blowdown of the water supplied by the AFW system had not been considered in the earlier anaLysis.
On October 1,
1979~
the foregoing in ormat'ion was provided to alL holders of operating Licenses and construction permits in IE Inf'ormat ion Not accident analys the above cited i%a power avai Lab le e
generator at an considered in ice 79-24 L23.
Another Licensee performed an I
is revi ew pursuant to the in'formation furnished in notice and discovered that with offsite electrical the conde'nsate pumps would feed the affected steam excessive rate.
This excessive feed had not been ts anaLysis of the po'stulated 7lSLB accident.
83020i00i8 830ii7 PDR ADOCK 05000335 P
PDR I
ti third Licensee infor"ed the f'RC of cn error 1n the RSLB analysis for their plant.
For c =ero or Lou poser condition at the end of core Life~ the Licensee identified an incorroct postulation that the otartup feedvater control valves mould roaa5n positioned "as 5s" during the transient.
Xn reality~ the startup feedarater control valves vkLL ramp to 80X fulL open due to an override signal resuLting from the Los steam gener ator pressure reactor trip 'signaL ~
4 Reanalysis of the events shoved tha't the rate of feedvater addition to the affected steam generator associated with the opening of the startup vaLve mould cause a rapid reactor cooldovn and resultant reactor-return-to-poser response~
a.condition thaT is beyond the plant's design basis.
'Following the identification of these deficiencies in the originaL NSLB ac c3 dent anaLys 3 s+ the HRC issued IF. Bulletin 80"04 on February 8~ 1980.
This buLletin required aLL Licensees of PMRs and certain ne'ar-term PMR operating Licensee appLicants to do.the foLLoving:
"1.
Review the containment pressur e response analysis to determine if the potential for containaent overpressure for HSLB.inside containment included the impact of runout f LoM from the auxiliary feedeater system and the impact of other energy sour ces such as.continuation of feed~ater or condensate f Los.
Zn your revie~~ consider your abi lity to detect and isoLate the damaged steam generator from these sources and the ability of the pumps to remain operable after extended operation nt runout fLos.
Review your cnalycis of the reactivity increase which results fron a
RSLB inside or outside containaent.
This review should consider the reactor cooldovn rate and the potential for the
~
reactor to return to poser arith the nost reactive controL rod in the ful.ly vhthdraMn position. If your previous anaLysis did not consider aLL potentiaL eater sources (such as those Listed.
in 1 above) and E f the reactivity increaie is greater than previous analysis indicated~ the report of this review should include:
I a
The boundary conditions for the analysis~
o g.~ the end of.
life shutdown margin~ the aoderator temperature coef ficient~ power Level and the ne't ef fect of the associated steam generator stater inventory on. the reactor'ycte~
cooling~ etc ~,
b.
The nost restrictive singLe active faiLure in the safety injection systea and the effect of that faiLure on delaying the del.ivery of'igh concentration boric acid soLution to the reactor coolant cystea;
/
c.
The effect of extended Mater supply to the affected stean generator on the core criticality and return to poMer; and d.
The hot channel factors corresponding to the most reactive rod in the fuLLy withdrawn positions at the end of Life~
and the Riniaua'Depar ture'r oa Nucleate Boiling Ratio (NDNBR) values for the analyred transient.
P
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Xf the potential for containment overpressurc exists or the reactor return-to-poser response worsens~
provide a proposed
0
0 corrective action and a
scheduLe f'r completion of the cor,rective action.
If the unit is operating, provide a
description of any interim action that wiLL be taken untiL the proposed corrective action is conp leted.."
FoLlowing the Licensee's initiaL response to IE BulLetin 80"04~
a request for additional information was developed to obtain aLL the infornation necessary to evaluate the The resuLts of our evaLuation for St.
Luci licensee's'nalysis.
e Nuclear PLant, Unit 1 (St.
Lucie 1) are provided below.
2.0 Eva lua t i on The staff's consultants the Franklin Research Center (FRC)s has re-viewed the submitta ls made by the Licensee i'n response to IE Bulletin 80-04'nd prepared the attached Technical Evaluation Report.
The staff has reviewed this=evaLuation and concurs, in its bases and findings.~
3.0 Conc lusion Based on the staff's review of the attached TechnicaL Evaluation Reports the foLlowing concLusions are made regarding the postulated NSLB with continued. feedwater addition-for St. Lucie 1:
1.
There is no potential for containment overpressurization r suLt ing f rom a
i'1SLB wi th continued feedwa ter addition because the main feedwater system is iso lated and auxiLiary I
feedwater actuation system (AFAS) prevents the affected steam generator from being fed during the time interval when peak containnent pressure occurs;
2.
ALL po
~ ential water sources were identified and although a
reactor return-to"power is predicted, there is no violation of the specified acceptable fueL design Limits..Therefore~
the Updated Final Safety Analysis Report HSLB reactivity increase ana'lys i s rema ins va lid for the control'-grade AFAS.
.3
~
The controL"grade AFAS does not provide runout protection for the auxi liary feedwater pumps.
Installation of the safety" grade AFAS wi Ll provide runout protection
~
4.
The.Licensee's commitment to provide an analysis of the reac-,
tivity resoonse to a
l1SLB prior to installation of the safety" grade AFAS is acceptable.
Attachment:
FRC Technical EvaLuation Report
References "Analysis of a
PWR Hain Steam Line Break with Continued Feedwater Addition" NRC Of fice of Inspection and Enforcement, February 8~
1980 IE Bulletin 80-04 "Overpressurization of the Containment of a
PWR PLant after a Main Stean line Break"~
NRC Offic'e of Inspection and Enforcement~
October 1~
1979 IE Information Notice 79-24 Technical EvaLuation Report "PWR tlain Stean Line Break wi th Continued Feedwa ter Addition - Review of Acceptance Criteria" Fr.ankLin Research
- Center, Novenber 17~
1981 TER-C5506-119 "Criteria for Protection Systems for Nuclear Power Generating Stations" Institute of Electrica l and ELectronics Engineers~
New York~'NYg 1971 IEEE Std 279-1971 Standard Revi ew PLan, Section 4.2 "Fuel System Design" NRC~ July 1981 NUREG-0800 Standard Review PLan Section 15.1.5 "Steam Systen Piping Failures Inside and Outside of Containment PWR" NRCi 'July 1981 NUREG-0800 "Criteria for Accident Honitoring Functions in Light-Water-,Cooled Reactors" American Nuclear Society Hinsdale~
IL~ December 1980
- ANS/AN'SI-4. 5-1980 I
"Instrumentation for Light-Mater-Cooled Nuclear Power PLants to Assess PLant and Environs Co'nditions During and FoLLowing an Accident" Rev.
2
- NRC, December 1980 RegulatOry Guide 1.97 "Single Failure Criteria for PWR Fluid Systems" American Nuclear Society HinMale~ IL~
~ June 1976 ANS-51.7/N658"1976
"Quality Group C lassifications and Standards for Water-Steam-~
and Radioactive-Waste"Containing Components of Nuclear Power PLants~"
Rev.
3~
NRC~ February 1976/
Regulatory. Guide 1.26 "Interim Staff Position on Environnenta l QuaLification of Safety-Re Lated E lect ri cat Equipment~"
Rev.
1 NR C~ July 1981 NUREG-0588 "Clarification of THI Action PLan Requirements"
'NRC, November 1980 NUREG"0737