ML20024A411

Interim Deficiency Rept Re Potential Excessive Offsite Dose Concerning Feedwater Containment Penetrations Not Maintaining Water Seal for 30 Days post-LOCA.Also Reportable Under Part 21

ML20024A411
Person / Time
Site:	Susquehanna
Issue date:	05/31/1983
From:	Curtis N PENNSYLVANIA POWER & LIGHT CO.
To:	Allan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
REF-PT21-83 10CFR-050.55E, 10CFR-50.55E, PLA-1687, NUDOCS 8306170250
Download: ML20024A411 (4)

Text

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. . PP&L Pennsylvania Power & Light Company Two North Ninth Street

• Allentown, PA 18101

• 215 1 770 5151 Norman W. Curtis Vice President-Engineering & Construction-Nuclear 215/770-7501 May 31, 1983 Mr. J. M. Allan Acting Regional Administrator, Region I U. S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 SUSQUEHANNA STEAM ELECTRIC STATION-UNIT 2 INTERIM REPORT OF A DEFICIENCY INVOLVING FEEDWATER BYPASS LEAKAGE - POTENTIAL EXCESSIVE OFF-SITE DOSE ER 100508 FILE 821-10.

PLA-1687 Docket No. 50-388 Dear Mr. Allan This letter serves to provide the Commission with an interim report on a deficiency involving potential excessive off-site dose related to the feedwater containment penetrations not maintaining a water seal for 30-days post-LOCA.

This deficiency was originally reported by telephone to Mr. D. Johnson of NRC Region I on April 27, 1983 by Mr. J. Saranga of PP&L as potentially reportable under the requirements of 10CFR50.55(e) for SSES Unit 2. Further evaluation of the deficiency by PP&L resulted in the conclusion that the condition should now be classified as reportable as a significant deficiency in final design, as defined in 10CFR50.55(c) .

The attachment to this letter contains a description of the deficiency, its cause and extent, and the safety -implications. Alternative plans for corrective action are currently being assessed. PP&L anticipates providing the Commission with a final report in July, 1983, including the corrective action taken or planned. This information is furnished for Unit 2 pursuant to the provisions of 10CFR50.55 (e) .

Since the details of this report provide information relevant to the reporting requirements of 10CFR21 for Unit 2, this correspondence is considered to also discharge any formal responsibility PP&L may have in compliance thereto.

This information was reported to the Commission for Unit 1 in a letter (PLA-1662) dated 5/12/83 which transmitted Licensee Event Report No.83-057.

We trust the Commission will find this report to be satisfactory.

Very truly yours, 8306170250 830531

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PDR ADOCK 05000388

/ S PDR N. W. Curtis Vice President-Engineering & Construction-Nuclear Attachment \l

c SSES PLA-1687 ER 100508 File 821-10 Mr. J. M. Allan cca Mr. Richard C. DeYoung (15)

Director-Office of Inspection & Enforcement U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Mr. G. Mcdonald, Director Office of Management Information & Program Control U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Mr.-Gary Rhoads U.S. Nuclear Regulatory Commission P.O. Box 52 Shickshinny, PA 18655 Records Center Institute of Nuclear Power Operations 1100 Circle 75 Parkway, Suite 1500 Atlanta, Georgia 30339 1

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. Attachmant PLA-1687 Page 1 of 2 Interim Report on Feedwater Bypass Leakage (8.1.2A #83-12) r,

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S- PROBLEM:

e The feedwater containment penetrations at Susquehanna were designed to remain water filled for 30 days post-LOCA. Review of the design has yielded the conclusion that the water seal.may not exist under certain design basis conditions.

CAUSE:

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The Susquehanna FSAR states that the feedwater containment penetrations are designed to remain water sealed for 30-days post-LOCA. The water scal is relied upon to prevent the primary containment atmosphere from bypassing secondary containmen.t and the Standby Gas Treatment System through the feedwater lines during a LOCA. During the Technical Specification review, NRC questioned whether the penetrations were truly water sealed and suggested that local leak rate testing be performed using air as the working fluid.

PP&L agreed to use air, but retained the contention that the lines will remain water filled. Bechtel Power Corporation provided an analysis of post-a'ecident flashing in,feedwater lines and concluded that enough water remained above the top of the valve seats to seal the penetration for 30-days. This analysis ass.umed valve leakage would be less than or equal to the leak test criteria in Manufacturer's Stand'rda MSS-SP-61. The calculated allowable leak rate was transmitted to the Susquehanna Integrated Startup Group for.use as a test acceptance crit'eria.

It was later discovered that leak rate testing was never performed to confirm that the volume' of Yeedwater remaining in the lines would last for 30-days.

During the investigation of this potential problem, the feedwater flashing

analysis was reviewed. Three important deficiencies were found in the ~

calculations. First, the calculation considered only the volume of piping from the RPV nozzles out to the inboard isolation valves, not the outboard valves. This is a nonconservative assumption. Second, the sensible heat of j the pipe walls and the valves was neglected. More seal water would be lost

! by boiling while cooling the hot metal. Third, flashing was assumed to occur -

l at the free surface of the water in the pipe. Due to the large v'olume of steam relative to water, any* flashing that occurred in portions of the pipe upstream of the free surface would entrain some of the liquid in the '

downstream piping and carry it out into the RPV.

Our evaluation of these three deficiencies caused us to conclude that there is insufficient water in the feedwater lines immediately following rapid depressurization of the RPV to cover the seating surface .of the isolation valves for 30 days, regardless of the leak tightness of the valves.

EXTENT: -

FSAR Table 6.3-15 lists all potential bypass paths for secondary ,

I containment. Several other penetrations use water seals to prevent bypass leakage: Reactor Water Clean-up (RWCU) from Reactor Vessel, Suppression Pool, Purification Line, Liquid Radwaste Collection, and Reactor Building Closed Cooling Water (RBCCW). The justification of water seals for these penetrations was reviewed to determine whether the feedwater problem bore any l

implications in these cases.

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PLA-1687 Pegs 2 of 2 RWCU and the Suppression Pool drain draw their seal water'from relatively '

inexhaustible sources and therefore were not considered further. Liquid

. Radwaste and RBCCW, however, are. sealed by. water.in piping at a higher _

2 - ~ elevation than their respective isolation valves. For these two  ;

penetrations, a maximum allowable leak rate was calculated such that the j inventory remains for 30-days post-LOCA. This allowable value was compared j with expected leak rates for the valves used and in each case, the expected

. leak rate is much lower than the allowable. A program of testing and 1

maintenance will verify that this is'tru'e throughout the life of the plant.

The major difference between.the seal water analysis for these lines and the feedwater lines is flashing. In all the lines where we now take credit for seal water, the water is relatively cold throughout the postulated accident. ,

SAFETY IMPLICATIONS

l The analysis of flashing in the feedwater lines following rapid ,_

-depressurization of the reactor pressure vessel has been shown to be ,

invalid. This determination has no bearing oa any system other than feedwater. The implication of this determination is that the design of the feedwater system is insufficient to perform as described in the FSAR.

Since the feedwater isolation valves ~will not remain water sealed post-LOCA, the safety implication depends on how well the valves perform when tested with air. The' measure of success of an air test depends upon-the amount of , ~

bypass air leakage assumed fu the analysis of the radio, logical consequences of a large break LOCA. This analysis, presented in Chapter 15 of the FSAR, used 5.0 scf per hour of bypass leakage. The Technical Specifications allotted 1.2 sef per hour of-leakage to the Main Steam Drain lines which are the only other viable bypass ' paths (per FSAR Section 6.2.3.2.3). Therefore, as long as the feedwater lines leak less than or equal to 3.8 sef per hour, the accident analysis is still valid. Though the feedwater design does not support a water seal, there is no adverse safety effect if the demonstrible leak rate is less than 3.8 sef per hour.

, However, the 3.8 sef per hour leakage . criterion was not specified in the

. Technical Specifications; it was not deemed necessary because the water seal should hhve prevented the leakage entirely. The current leakage criteria for .

the feedwater valves is based only on the requirement that all Type C tested (per 10CFR50 Appendix J) penetrations leak less than 0.6 La. During the life of the plant, the valves may have been allowed to leak so much that off-site doses would have been excessive. This could constitute an unsafe condition

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If a LOCA occurred while the valves were in this state.

Since this deficiency could have adversely affected the safety of operations of the plant, and it represents'a significant deficiency in final design, we

conclude that it is reportable as defined in 10CFR50.55(e)(1)(ii).

CORRECTIVE ACTION:

Alternative plans for corrective action are currently being assessed.. Finalized corrective action will be specified in the final report.

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