ML20062H228

Safety Evaluation Accepting Util Responses to IE Bulletin 80-04, Main Steam Line Break W/Continued Feedwater Addition

ML20062H228
Person / Time
Site:	Beaver Valley
Issue date:	08/04/1982
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20062H226	List: ML20062H221 ML20062H228
References
IEB-80-04, IEB-80-4, NUDOCS 8208130436
Download: ML20062H228 (4)

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• SAFETY EVALUATION REPORT OFFICE OF NUCLEAR REACTOR REGULATION UNITED STATES NUCLEAR REGULATORY COMMISSION MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION BEAVER VALLEY NUCLEAR PLANT, UNIT 1 DOCKET N0. 50-334 Introduction In the summer of 1979, a pressurized water reactor (PWR) licensee submitted a report to the NRC that identified a deficiency in his original analysis of the containment pressurization resulting from a postulated main steam line break (MSLB). A reanalysis 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 experienced 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 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 (2). Another licensee performed an accident analysis review pursuant to the information furnished in the above cited notice and discovered that, with offsite electrical power available, the condensate pumps would feed the affected steam generator at an excessive rate. This excessive feed had not been considered in his analysis of the postulated MSLB accident.

A third licensee informed the NRC of an error in the MSLB analysis for his plant. For a zero or low power 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 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 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.

r C208130436 820804  :

PDR ADOCK 05000334 l C PDR

y Following the identification of these deficiencies in the original 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 applicants to do the following:  :

"1. Review the containment pressure response analysis to determine  !

if the potential for containment overpressure for 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 your ability to detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable after extended operation at runout flow.

2. Review your analysis of the reactivity increase which results l from a MSLB inside or outside containment. This review should

! consider the reactor ccoldown rate and 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 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 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 f 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 t solution to the reactor coolant system l c. The effect of extended water supply to the affected steam i j

generator on the core criticality and return to power; and  !

d. The hot channel facters corresponding to.the most reactive l rod in the fully withdrawn positions at the end of life,  :

and the Minimum Departure trom Nucleate Boiling Ratio  !

(MDNBR) values for the analyzed transient.  ;

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3. If the potential for containment overpressure exists or the i

, reactor return-to-power response worsens, provide a proposed j 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 j j the proposed corrective action is completed."  ;

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Following the licensee's initial response to IE Bulletin 80-04, a request for additional information was developed to obtain all the information necessary to evaluate the licensee's analysis.

The results of our evaluation for Beaver Valley Power Station, Unit 1 (Beaver Valley 1) are provided below.

Evaluation ,

i Our consultant, the Franklin Research Center (FRC), has reviewed the submittals made by the licensee in resoonse to IE Bulletin '

80-04, and prepared the attached Technical Evaluation Report. We  ;

have reviewed this evaluation and concur in its bases and findings. (

i' Conclusion  ;

Based on our review of the enclosed Technical Evaluation Report, I the following conclusions are made regarding the postulated MSLB with i continued feedwater addition f

1. There is no potential for containment overpressurization l resulting from a MSLB with continued auxiliary feedwater '

addition; the steam line break protection system-provides actuation signals which, in fact, isolate the main feedwater i

lines; {

2. The AFW pumps will experience runout conditions; however, based on evaluation of data from a 10-minute run under runout conditions,  !

the licensee concluded that the pump will not be damaged and [

will remain available after extended operation at runout flow '

conditions; i,

3. All potential water sources were identified and no reactor  !

i return-to-power or DNBR violation occurs; therefore, the  ;

FSAR reactivity increase analysis remains valid. z

References

1. IE Bulletin 80-04, " Analysis of a PWR Main Steam Line Break r with Continued Feedwater Addition," NRC Office of Inspection  ;

and Enforcement, February 8,1980  ;

IE Information Notice 79-24, "0verpressurization of the 2.

Containment of a PWR Plant After a Main Steam Line Break," i NRC Office of Inspection and Enforcement, October 1,1979 '

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3. C. N. Dunn (DLC) Letter to B. H. Grier (NRC, Region 1),

Subject:

Beaver Valley Power Station, Unit 1 Response to IE Bulletin No. 80-04, April 28, 1980

4. Beaver Valley Power Station, Unit 1, Final Safety Analysis Report, through; Amendments.9,.Duquesne Light Compar.y, June 1974

5. S. A. Varga (NRC) Letter to C. N. Dunn (DLC),

Subject:

Amendment No. 30 to Facility Operating License No. DPR-66 for the Beaver Valley Nuclear Power Station, Unit No.1, September 10, 1980

6. Technical Evaluation Report, TER-c5506-119, "PWR Main Steam Line Break with Continued Feedwater Addition - Review of Acceptance Criteria," Franklin Research Center, November 17, 1981

7. IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations," Institute of Electrical and Electronics Engineers, New York, NY,1971

8. NUREG-0800, Standard Review Plan, Section 15.1.5, " Steam System Piping Failures Inside and Outside of Containment -

(PWR)," NRC, July 1971

9. ANS/ ANSI-4.5-1980, " Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors," American Nuclear Society, Hinsdale, IL, December 1980

10. Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess P.lant and Environs Conditions During and Following an Accident," Revision 2, NRC, December 1980

11. ANS-51.7/N658-1976,'" Single Failure Criteria for PWR Flu'id Systems," American Nuclear Society, Hinsdale, IL, June 1976

12. Regulatory Guide 1.26, " Quality Group Classifications and Standards for Water, Steam, and Radioactive-Waste-Containing Components of Nuclear Power Plants," Revision 3, NRC, February 1976

13. NUREG-0588, " Interim Staff Position on Environmental Qualifi-cation of Safety-Related Electrical Equipment," Revision 1, NRC, July 1981

14. J. J. Carey (DLC) Letter to S. A. Varga (NRC),

Subject:

Potential for Damage to the AFW Pumps Due to Operation at Runout Flow For 30 minutes, July 21, 1982 Principal Contributor P. Hearn

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