ML18051A856
| ML18051A856 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 04/11/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18051A855 | List: |
| References | |
| IDB-80-04, IDB-80-4, NUDOCS 8404180112 | |
| Download: ML18051A856 (9) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION REPORT OFFICE OF NUCLEAR REACTOR REGULATION MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION PALISADES NUCLEAR PLANT DOCKET NO. 50-255 i.o Introduction In the summer of 1979, a pressurized water reactor CPWR>
licens~e submitted a report to the NRC that identifi~d a deficiency in its original analysis of the containment pressurization resulting from a postulated main steam line break (ASLB>.
A reanalysis of the containment pressure response following a MSLB vas performed, and it was determined that, if the auxiliary feedwater CAFW) system c~ntinued to supply feedwater at runout conditions to the steam generator that had expereinced the steam line break, the containment design pressure ~ould be exceeded* in approximately 10 minutes.
In other words, the long-term blowdown of the water supplied by the AFW systen had not been considered in the earlier snalysis.
On October 1, 1979, the foregoing information vas provided to all holders of operating licenses end construction permits in IE Inf6rmation Notice 79-24 [2J.
Another licensee perfor~ed an
- accident analysis review pursuant to the information furnished in the above cited notice and discovered 'that, with offsite electrical power Gvailable, the condensate pumps.would feed the affected steam generator at 11n excessive rate. This.. exce.isive feed had not been considered in its mnalysis of the postulated ASLB accident.
8404180112 B404ft : '1 *.I
PDR ADOCK 05000255 L '
G PDR
A third licensee infor*ed the NRC of an error in the MSLB analysis for thei~ plant.
For a zero ~r low po~er condition et the.end of core life, the licensee identified an incorrect postulation that the startup feedvater control valves would remain positioned "as is" duri~g the transient.
I~ reality, the startup feedvater control valves will ~~mp to 80% full open due to en 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 st~am generator associated with the opening of the
'\\startup valve would cause a rnpid reactor cooldovn and resultant.
reactor-return-to-power response, a condition which is beyond the plant's design basis.
following the identification of these deficiencies in the origin~l MSLB accident analysis, the NRC issued IE Bulletin 80-04 on February 8, 1980.
This bulletin required all licensees of PWRs and certain near-term PWR operating license ap.plicants to do the
- following:
"1.
Review the containment pressure response analysis to determine if the ~otential for containment overpressure for MSLB inside containment included the impact of runout flow from the auxiliary feedwater system and the i~pact of other energy sources such as continuation of feedwater or condensate flow.
In your review, consider your ability to detect and isol*te the domeged steam generator from ~he~~ sources and the ability of the pumps to remain operable after extended op~~ation at runout flow.
2 -
- 2.
Review your analysis of the reactivity increase which results fro~ a MSLB inside or outside containment.
This review should consider the reactor cooldown rote mnd 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 listed 1n 1 above> and if the reactivity increase is greater than*
previous analysi$ indicated, 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 eost 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 syste~;
- c.
The effect of extended water ~upply to the affected steam generator on the core criticality and return to po~er; and
- d.
The hot channel factors corresponding to the most reactive rod in the fully withdrawn positions at the end of life, mnd the Ainimum De~arture from Nucleate Boiling Ratio C~DNBR) values for the analyzed transient.
- 3.
If the.Potential for containment overpressure exists or the reactor return-to-po~er response worsens, provide a proposed 3 -.
corrective action nnd c schedule for co~pletion of the corr..ec.tive action.
If the unit is operating, provide.n description of any interim Action that will be tnken until the proposed corrective action is completed."
Following the licensee's initial response to IE Bulletin 80-04, a request for additional i~formation was developed to obtain all the information necessary.to evaluate th~ li~ensee's analysis.
T he re s u l t s of o u r. e v a l u at i on for t h e P a l i s a d e s - N u c l° e a r -P L a n t - - -*
( P-a L i s a d e s
) a re p r o vi de d be low
- 2.0.
Eva l_uat ion
.Our consultant, the Franklin Research Center <FRC>, has reviewed
.the submittals made by the licensee in response to IE Bulletin 80-04, and prepared the attached Technical Evaluation Report.
We have reviewed this evaluation and concur in its bases and findings.
3.0 Conclusion B~sed on our review of the enclosed Techriical Evaluation Report, the followi~g f~ndings
~re made regarding the postulated MSLB with :ontinued feedwater *ddition fo~ Pili sades lhere is no potential for contain*ent overpressurization re*ulting from * ~SLB with continued feedwater *ddition
- b*C~y\\. the aain feed~lter System is a l
d 1~0 ate
- nd a~xiliary
'**dw*t*r *. ttYetion systea prevents the affected steam I
c l
[
i
- 2.
The AFW pumps will not experience runout conditions and will, therefore, be able to carry out their intended function during a MSLB without incurring damage.
- 3.
All potential water sources were identi1ied, no return-to-power is predicted, and there is no vi6lation of the specified acceptable fuel design limits.
Therefore, the current analysis remains valid.
4~
No further action by the licensee is required regarding IE Bulletin 80-04.
However, there is a need for a licensee submittal and staff review of the following proposed modifications (8), to be made during the 1985 refueling outage, that will provide safety grade single-failure proof protection against the continued addition of main feedwater during a MSLB:
- a.
The circ~itry of the MFW stop valves and MFW bypass valves will be modified to close on a low steam generator pressure trip signal.
- b.
Additional control val~es will b.e installed in series with the MFW stop valves and bypass valves which also close on a Low steam generator pressure signal.
4.0 Acknowledgement This safety evaluation was prepared by P. Hearn.
- 4.
REFERENCES
- 1.
"A~alysis of a PWR Main Steam Line Break with Continued Feedwater Addition" NRC Office of Inspection and Enforcement, February 8, 1980 IE Bulletin 80-04
- 2.
"OVerpressurization of the Containment of a PWR Plant after a Main Line Steam Break" NRC, Off ice of Inspection and Enforcement, October 1, 1979 IE Information Notice 79-24
- 3.
S. R. Frost (CPC)
Letter to J. G. Keppler (NRC, Region III)
Subject:
Response to IE Bulletin 80-04 9-May-80
- 4.
D. M. Crutchfield (NRC)
Letter to Consumers Power Co.
Subject:
IE Bulletin 80-04, Request for Additional Information March 8, 1982
- 5.
D. M. Crutchfield (NRC)
Letter to Consumers Power Co.
Subject:
IE Bulletin 80-04, Request for Additional Information
- Se.ptember 7*, 1982
- 6.
B. D. Johnson (CPC)
Letter to D. M. Crutchfield (NRC)
Subject:
Response to IE Bulletin 80-04 Additional Information April 26, 1982
- 7.
B. D. Johnson (CPC)
Letter to D. M. Crutchfield (NRC)
Subject:
Analysis of Main Steam Line Break with Continued Feedwater Addition October 26, 1982
- 8.
B. D. Johnson (CPC)
Letter to J. G. Keppler (NRC IE Region III)
Subject:
Analysis of Main Steam Line Break with Continued Feedwater Addition July 22, 1983
- 9.
~- D. Johnson (CPC)
Letter to D. M. Crutchfield (NRC)
Subject:
Analysis of 1'1ain Steam Line Break with Continued Feedwater Addition September 8, 1983
- 10.
Palisades Plant Final Safety Analysis Report Consumers Power Company
- 11.
Integrated Plant Safety Assessment Pa~isades Plant NUREG-0820 October 1982
- 12.
B. D. Johnson (CPC)*
Letter to D. M. Crutchfield (NRR)
Subject:
NUREG-0737, Item II.E.l.l - Additional Information May 29, 1981
- 13.
Technical Evaluation Report, *pwR Main Steam Line Break with Continued Feedwater Addition - Review o*f Acceptance Criteria" Franklin Research Center, November 17, 1981 TER-C5506-119
- 14.
"Criteria for Protection Systems for Nuclear Power Generating_
Stations*
Institute of Electrical and Electronics Engineers, New York, NY, 1971 IEEE Std 279-1971
- 15.
Standard Review Plan, Section 4.2 "Fuel System Design" NRC, July 1981 NUREG-0800
- 16.
Standard Review Plan, Section 15.l.5 "Steam System Piping Failures Inside and Outside of Containment (PWR)
- NRC, July 1981 NUREG-0800
- 17.
- criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors*
American Nuclear Society, Hinsdale, IL, December 1980 ANS/ANSI-4.5-1980
- 18.
- Instrumentation for Light-Water-cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident,* Rev. 2 NRC, December 1980 Regulatory Guide 1. 9 7
- 19.
"Single Failure Criteria for F\\oiR Fluid Systems" American Nuclear Society, Hinsdale, IL, June 1976 ANS-51.7/N658-1976
- .. 7 -
.~.* y
- 20.
"Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants"
. Rev. 3 NRC, February 1976 Regulatory Guide 1.26
- 21.
- interim Staff Position on Environmental Qualification of Safety~Related Electrical Equipment,* Rev. l NRC, July 1981 NUREG-0588
- 22.
Technical Evaluation Report, *Auxiliary Feedwater System Automatic Initiation and Flow Indication*
Franklin Research Center, April 1982 TER-5257-298
- 23.
- plant Transient Analysis of the Palisades Reactor for Operation at 2530 MWt*
Exxon Nuclear Corporation XN-NF-77-18
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