ML17319B580
| ML17319B580 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/05/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17319B579 | List: |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 8210150437 | |
| Download: ML17319B580 (8) | |
Text
SAFETY EVALUATION REPORT CONTAINNENT SYSTENS BRANCH MAIN STEAN lINE BREAK WITH CONTINUED FEEDWATER ADDITION D.
C.
COOK NUCLEAR PLANT Docket No.: 50-315~
-316
1.0 INTRODUCTION
In the summer of 1979~
a pressurized water reactor (PWR) Licensee submitted a report to the NRC that identified a deficiency in its originaL analysis of containment pressurization resulting from a
postULateG main steam Line break (NSLB).
A reanalysis of the containment pressure response following a NSLB 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 experienced the steam Line break~
the containment design pressure would be exceeded in appr oximately 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 information was provided to aLL holders of operating Licenses and construction permits in IE Information Notice 79-24 C23.
Another Licensee performed an accident analysis review pursuant to the information furnished in the above ci ted notice and di scove red that~
wi th,of fsite electrical power available, the condensate pumps would feed the af fected steam generator at an excessive rate.
This excessive feed had not been considered in the analysis of the postulated NSLB accident.
8210i50437 821005 PDR ADOCK OS0003i5 P
e l
A third Licensee informed the NRC of an error in the NSLB analysis for their plant.
For a zero or Low power condition at the end of core Li fe~ the Licensee identified an incorrect postulation that the startup feedwater control valves would remain positioned "as is" during the transient.
In reality~ the star tup feedwater controL valves will ramp to 80K 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 addi tion to the af fected steam gene-rator associated with the opening of the startup valve would cause a
rapped reactor coo ldown and resultant reactor-return-to-power r esponse~
a condition which is beyond the plant's design basis.
Following the identification of these deficiencies in the original NSLB accident analysis~
the NRC issued IE Bulletin 80"04 on February 8, 1980.
This bulletin required all Licensees of PWRs and near-term PWR operating License applicants to do the folLowing:
Revi ew the containment pressure response analysis to determine if the potential for containment overpressure in tpe event of a
NSLB inside, containment included the impact of runout f Low f rom the auxi Liary'eedwater system and the impact of other ene of feedwater or condensate the ability to detect and i rgy sources such as continuation flow.
In your review~ consider so late the damaged steam generator f rom these sources and the abi Lity of the pumps to remain operable after extended operation at runout flow.
Review your analysis of the reactivity increase which results from a
NSLB inside or outside containment.
This review should consider the reactor cooldown rate and the potentia l 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 Listed in 1
above) and if the reactivity increase is greater than previous ana lysis indicated~
the report of this review should include:
a.
The boundary conditions for the analysis, e.g.~
the r
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 boric acid soLution to the reactor coolant system; c.
The effect of extended water supply to the affected steam generator on the core critica lity 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 Boi Ling Ratio (NDNBR) values for the analyzed transient.
3.
If the potential for containment overpressurization 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 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 request for addi tional information was developed to obtain aL L the information necessary to evaluate the Licensee's analysis.
t The results of our evaluation for D.
C.
Cook Nuclear PLant~ Units 1
and 2
(D.
C.
- Cook, 1
and 2). are provided below.
2.0 Eva luat i on Our consultant, the Fr ank lin Research Center (FRC),
has revi ewed the submittals made by the Licensee in response to IE Bulletin 80-04~
and prepared the attached Technical Evaluation Report.
Qe 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 made regarding the postulated NSLB with continued feedwater addition for D.
C.
Cook 1
and 2:
1.
Ther e is no potentiaL for containment overpressurization resulting from a
NSLB with continued feedwater addition because the main feedwater system is isolated and auxiliary feedwater f Low i s restricted to the affected steam generator.
2.
ALthough a single failure of the runout protection system wilL subject one of the auxi liary feedwater (AFM) pumps to possible damage through operation at runout conditions~
the two other pumps wilL remain avai Lable to supply auxi Liary feedwater to the steam generators.
3.
ALL potential water sources were identified.. ALthough a
return to power is predi cted~
there is no violation of the specified acceptable fuel design Limits.
Therefore~
the Final Safety Analysis Report reactivity analysis remains va Lid.
4.
No further action by the Licensee is required regarding IE Bulletin 80-04.
4.0 Ref erences 2.
3.
"Analysis of a
PWR Hain Steam Line Break with Continued Feedwater Addition~" NRC Office of Inspection and Enforcement February 8~ 1980 IE Bulletin 80-04 "Overpressuri zation of the Containment of a
PWR Plant After a Main Steam Line Break~"
NRC Office of Inspection and Enforcement, October 1,
1979~
IE Information Notice 79-24 J.
E. Dolan (INEC)
Letter to J.
P. Keppler (NRC, Region III)
Subject:
IE Bulletin 80-04, Analysis of a
PWR Nain Steam Line Break with Continued Feedwater Addition 8-Apr-80 4 ~
5.
6.
7.
S.
A. Varga (NRC)
Letter to INEC
Subject:
Request for Information~ IE Bulletin 80-04 January 15~
1982 R.
S.
Hunt e r (IHE C)
Letter to H.
R. Denton (NRC)
Subject:
Additional Information Relative to IE Bulletin No. 80-04 Apri l 26 1982 Donald C.
Cook Nuclear Power Plant Units 1
and 2
Final Safety Analysis Report~
through Amendment 83 Indiana and Nichigan Electric Company~
March "1979 Techni ca l Eva luati on Report "PWR flain Steam Line Break with Continued Feedwater Addition - Review of Acceptance Criteria" Frank lin Research Center~
November 17~
1981 TER-C5506-119 8.
"Criteria for Protection Systems for Nuclear Power Generating Stations" Institute of Electrical and Electronic Engineers~
New York NY, 1971.
IEEE Std ?79-119 9.
Standard Review Plan~
Section 4.2 "Fuel System Design NRC, July 1981 NUREG-0800 10.
"Criteria for Accident Monitoring Functions in Light" Water-Cooled Reactors" American Nuclear Society~
Hinsdale~ IL, December 1980 ANS/ANSI-4.5-1980
Instrumentation for Light"Water-Cooled Nuclear Power PLants to Assess PLant and Environs Conditions During and Following an Accident~" Revision 2~
NRC~ December 1980~ Regulatory Guide 1.97.
"Single Failure Criteria for PWR FLuid Systems~"
American Nuclear Society~
Minsdale~
IL~ June 1976~
ANS-51.7/N658-1976 "Quality Group CLassifications and Standards for Water~
Steam~
and Radioactive-Waste-Containing Components of NucLear Power Plants" Revision 3~
NRC, February 1976~
Regu Latory Guide 1.26 "Interim Staff Position on Environmental Qualification of Safety Related ELectricaL Equipment~" Revision 1~
NRC.
July 1981~