ML20116C983

Responds to SER Confirmatory Issue 13 Re Commitment to Provide Piping cross-tie on Feedwater Line Fill Network. Adequate Justification Provided for Taking Credit for Water Seal in Feedwater Sys Piping During Initial LOCA

ML20116C983
Person / Time
Site:	Hope Creek
Issue date:	04/22/1985
From:	Mittl R Public Service Enterprise Group
To:	Schwencer A Office of Nuclear Reactor Regulation
References
NUDOCS 8504290239
Download: ML20116C983 (5)

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Pubhc Servce E!ectnc and Gas Cornpany 80 Pa'k Plaza, Newark, NJ 07101/ 201430 8217 MAILING ADDRESS / P.O. Box 670, Newark, NJ 07101 Robert L. Mitti General Manager Nuclear Assurance and Regulation April 22, 1985 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20814 Attention: Mr. Albert Schwencer, Chief Licensing Branch 2 Division of Licensing Gentlemen:

RESPONSE TO HCGS SER CONFIRMATORY ISSUE 13 HOPE CREEK GENERATING STATION DOCKET NO. 50-354 The Hope Creek Generating Station Safety Evaluation Report (SER) Section 6.2.3 acknowledges the following Public Service Electric and Gas Company (PSE&G) commitments in sup-port of issuance of an operating license for Hope Creek:

"The applicant has recently committed to provide a pip-ing cross-tie on the feedwater line fill network. This cross-tie will permit the fill network to perform its intended safety function following a single active failure and will ensure the sealing function of this system for at least 30 days following a LOCA. The applicant has also committed to perform a confirmatory analysis to demonstrate that during the initial portion of a LOCA, water in the feedwater system piping down-stream of the No. 3 feedwater heater will flash to steam and continue to flow toward the reactor pressure vessel until the feedwater line decreases to the con-tainment pressure, at which time the isolation valves will be manually closed. The intent of this analysis will be to verify that pressure in the feedwater system piping will be sufficient to prevent the outward leakage of radioactive contaminants through the isola-tion valves during the approximately one-hour period after the accident, until the water seal is re-established between the isolation valves via the fill system. Thus, no bypass leakage of the feedwater sys-tem is expected to occur. The staff finds this accept-able, pending receipt of the upplicant's short-term analysis for the feedwater system. This is a confirma-tory item."

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5 Dir. of Nuclear Reactor Regulation 2 4/22/85 I

Table 1.3 of the HCGS SER identifies the above commitments 1

as confirmatory issue 13. They will be referred to herein as "C-13"~ commitments.

In.accordance with the C-13 commitments, PSE&G has' installed I i

the cross-tie on the feedwater line. fill network, and has 1 3

performed a preliminary evaluation of the likelihood of successfully maintaining a steam seal in the.feedwater pip-i ing, until a water seal.i:s established by the fill system. l This preliminary evaluation, however, has not shown suffi-F ciently positive results to warrant performing a more  !

detailed. analysis. l In light of the above, it is no longer PSE&G's intent to demonstrate prevention of leakage past the feedwater inola-tion valves by means of a steam seal. It is the position of PSE&G that a water seal will be maintained upstream of, the third point feedwater heater due to the following conditions:

1. -According to the Heat Balance (HCGS FSAR Figure 10.1-1), the maximum water temperature immediately upstream of the third point feedwater heater is 211. 9 ' F .~

2. Since feedwater will be in a no flow condition, the water seal will be maintained by the water leg in i the piping immediately upstream of the third point i feedwater heater.

This water seal will prevent bypass leakage through.the )

feedwater system until the feedwater fill system establishes a long term water seal between the containment isolation valves. Although feedwater piping.outside of primary con-tainment is not constructed to seismic Category I standards, taking credit for the presence of a water seal in this pip- i ing is a realistic and justifiable assumption in the event of a LOCA, as explained below.

The issue of containment bypass leakage is predicted on very conservative assumptions, including the following:

1. An instantaneous, double-ended guillotine break in the safety related, seismic Category I reactor coolant pressure boundary is assumed. This is not a mechanistic event. ,

2. A sudden loss of all off-site sources of electric power is assumed.

' Director of Nuclear Reactor Regulation 3 4/22/85

3. No credit is allowed for non-seismic Category I piping which implies an earthquake concurrent with a LOCA, and that the earthquake causes a failure in the non-seismic Category I piping. This is an extremely rare combination of events.

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An event sequence, which does not include the simultaneous occurrence of the three virtually independent events listed above, does not result in bypass leakage through the feed-water system for the following respective reasons:

1. Relatively small pipe breaks would result in a much

slower reactor depressurization and consequently a l

much lower amount of feedwater system flashing, resulting in a much higher probability of maintain-ing a water seal inside the feedwater isolation valves.

2. Continuity of off-site power availability would allow operation of the condensate pumps to maintain L

a water seal in the feedwater system throughout the event sequence.

3. In the event of a LOCA, the non-seismic Category I l

l feedwater piping would remain intact to form a

! water seal, thus preventing bypass leakage.

In the unlikely event that the above three conditions did

' occur simultaneously, the off-site radiological consequences would not approach the 10 CFR 100 limits regardless of the existence of a water seal in the feedwater system unless a large degree of core damage with an instantaneous core release is also assumed.

Concerning the seismic issue discussed in Item 3 above, NRC-sponsored research l of power plants and similar facili- l ties which have been subjected to large earthquakes has shown that the earthquakes have not been found to cause significant damage to process equipment such as piping, valves and pumps. Therefore, even though an earthquake and a LOCA are not realistically considered to occur concur-rently, there is a high degree of probability that there would be no damage to the feedwater piping if an earthquake were to occur concurrent with a LOCA.

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1. NUREG/CR-1665, " Equipment Response at the El Centro Steam Plant During the October 15, 197, Imperial Valley Earthquake," USNRC, October 1980.

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Director of Nuclear Reactor Regulation 4 4/22/85 The feedwater piping from the reactor pressure vessel back to the in-line anchor outside containment is designed to seismic Category 1 standards. The piping fran the in-line anchor to the condenser has been designed for de ad we ight, operating and design pressures, and system thermal operating modes. This piping has also been seismically analyzed to the Uniform Building Code, which is above and beyond the j requirements - for typical non-seismic piping .

The primary and secondary condensate pumps, the reactor feed pumps, and the condenser have been designed for horizontal seismic accelerations of 0.lg; the feedwater heaters and drain cooler are designed for 0.3g horizontal and 0.2g ve rt ical . This is beyond the typical requirements for non-seismic equipment, and gives added confidence that the sys tem is not likely to fail under seismic conditions.

In addition, the feedwate.r system equipment and piping are deuigned to withstand the high pressures observed during normal operation. Prior to the postulated LOCA, the feed-water system will be functioning at its operating pressure of 1000 psig . No system leakage , the refo re , is realis-tically expected at the relatively low post-LOCA containment pressures.

It is the position of PSE&G that the above considerations provide adequate justification for taking credit for a water seal in the feedwater system piping during the initial por-tion of a LOCA, even though the feedwater piping outside cont ainment is not seismic Category I. Based on this posi-tion, PSE&G still expects no containment bypass leakage to occur during LOCA. NRC evaluation of this position is hereby requested to enable prompt resolution of confirmatory issue 13.

Should there be any question or concerns in this matter, please feel free to contact us at your earliest conve nie nce .

Very truly yours ,

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Director of Nuclear Reactor

. Regulation 5 4/22/85

. C ' D. H.. Wagner -

USNRC Licensing Project Manager 1

A. R. Blough USNRC Senior Resident Inspector 4

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