ML18087A556
| ML18087A556 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 10/20/1982 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML18087A554 | List: |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 8211060588 | |
| Download: ML18087A556 (7) | |
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.. i SAFETY EVALUATION MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER AD~ITION SALEM NUCLEAR PLANT UNITS 1. AND 2
- Docket No.: 50-272, -311
1.0 INTRODUCTION
In the summer of 1979, a pressurized water reactor CPWR) licensee submitted a report to the NRC that identified a d~fici-ency in its original analysis of containment pressurization resulting from a l
postulated main steam. line break CMSLB)~
A reanalysis of the
."I containment pressure resporise following a MSLB was performed, and it was determined that, if the auxiliary feedwater CAFW> system continued to supply feedwater at runout conditions to the steam generator that had experienced the ~team Line break, -the containment design pressure would be exceeded in approximately 10 minutes.
In other ~ords, the long-term blowdown of *the water supplied by the AFW system had not been considered in the e~rlier 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 t2J.
Another licensee performed an accident analysis review pursuant to the information furni~hed in the above cited notice and discovered that, with offsite electrical
- power available, th~ 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.
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.e 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 sta~tup feedwater control valves would remain positioned "as is" during the transient.
In reality, the startup feedwater control valves will ramp to 80% 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 gene-rator associated with the opening of the startup valve would cause a rapid reacto*r cooldown and resultant reactor-return-to-power response, a condition which is beyond the plant's design basis.
Following the i.dentification of these deficiencies in the or*iginal MSLB 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 ~o do the* following:
- 1.
Review the containment pressure response analysis to*
determine if the potential for containment overpressure in the event of a MSLB inside containment included the impact of runout flow from the auxiliary*feedwater system and the impact of other energy sources* such as continuation of feedwater. or condensate flow.
In your review, consider the ability to ~etect and isolate the damaged steam generator from these sources and the ability of th~ pumps to remain operable after extended operation at runout flow.
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- 2.
Review your analysis of the reactivity increa*e which results from a MSLB inside or outside containment.
This review should consider the reactor coo-ldown rate and the potential for the reactor to return to powe~ 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* analysis indicated, the report of this review should include:
- a.
The boundary conditions fo~ the analysis, e.g., the end of life shutdown margin, the moderator temperature coefficient, power level and the net eff~ct of the associated steam generator water inventory on the reactor system coolfng, etc; b.,
The most *restrictive single active fai Lu.re 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 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 CMDNBR) values for the ~nalyzed transient.
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- 3.
If *the potential for containment overpressurization exists or the reactor return-to-power response worsens, provide a proposed correctiv~ action and.a schedule for completion of the corrective action.
If the unit is operating, previde 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 additio~al information was develDped to obtain all the information necessary to evaluate the licensee's analysis.
The results of our evaluation for Salem Nuclear Plant, Units 1 2 (Salem 1 and 2) are provided below.
2.0 Evaluation Our consultant, the Franklin Research Center CFRC>, has reviewed the submittals made
~Y the license~ 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 Based on our review of-the enclosed Technical Evaluation Report, the following conclusions are made regarding the postulated MSLB with continued feedwater addition for Salem 1 and 2:
1*.
There is no pote.ntial for containment overpressurization resulting from a MSLB with.continued feedwater addition because the main feedwater system is isolated and auxiliary feedwater flow to the affected steam generator is restricted
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- 2.
The CAFW) pumps are protected fr-0~ the effects of runout flow and therefore can be expected to carry out their intended function during a MSLB.
- 3.
All potential water sources were identified and, although a
- 4.
re.actor return-to-power is predicted, there is no violation of the specified acceptable fuel design limitsG Therefore, the FSAR reactivity incr,ase analysis remains valid
- No further action regarding IE Bulletin 80-04 is required.
5.,.
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4.0 REFERENCES
- 1.
"Analysis of a PWR Main Steam Line Break with Continued Feedw~ter 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 Office of Inspection and Enforcement, October 1, 1979 IE Information Notice 79-24
- 3.
PSE&G Gilberts CNSP)
Letter to B. Grier CNRC, Region I)
Subject:
Response to IE Bulletin 80-04 April 17, 1980
- 4.
E. A. Landers CPSE&G>
Letter to S~ AQ Varga CNRR)
Subject:
Additional Information Related to NRC Bulletin 80-04 July 26, 1982
- 5.
Salem Nuclear Generating Station Units 1 and 2 Final Safety Analysis Report, through Amendment 37 Public Servic~ Gas and Electric Company
.6.
"PWR Main Steam Line Break with Continued Feedwater Addition -
Review of A~ceptance Criteria"
- 7.
Franklin Resear~h Center, November 17R 1981 TER-C5506-119 "Criteria for Protection Systems for Nuclear Power ~enerating Stations" Institute of Electrical and Electronics Engineers, New York, NY, 1971.
IEEE Std 279-1971 85 Standard Review Plan, Section 4.2 "Fuel System Design" NRC, July 1981 NUREG-0800
- 9.
Standard Review Plan, Section 15.155 "Steam System Piping Failures Inside and Outside of Containment CPWR) 11 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 6
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- 11.
Regulatory Guide 1.97 "Instrumentation for Light-Water-Cooled Nuclear Power Power Plants to Assess Plant and Environs Conditions *During and Following an Accident" Rev. 2 NRC, December 1980
- 12.
"Single Failure Criteria for PWR Fluid Systems" American Nuclear Society, Hinsdale, IL, June 1976 ANS-51.7/N658-1976
- 13.
- 14.
- "Quality Group Classifications and Standards for Water-,
Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants" Rev.. 3
.NRC, February 1976 "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment" Rev. 1 NRC. July 1981 NUREG-0588
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