ML20040B504
| ML20040B504 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 01/06/1982 |
| From: | Mazur D WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | Faulkenberry R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| References | |
| 10CFR-050.55E, 10CFR-50.55E, GO-1-82-03, GO-1-82-3, NUDOCS 8201260125 | |
| Download: ML20040B504 (5) | |
Text
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Docket Nos. 50 l+60/513 50 55(e) Report C97ED H:::
Washington Public Power Supply System 72-5000 P.O. Box 968 3000GeorgeWashingtonWay Richland, Washington 99352 g9)}!
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January 6,1982 I
G0-1-82-03 IEEM -
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Nuclear Regulatory Commission i
Region V l
1450 Maria Lane g $LL Suite 210 g'v'
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j Walnut Creek, California 94596 1
Attention: Mr. R. H. Faulkenberry j
t' Chief, Reactor Construction Projects Brancj l'g y,, 7 1
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Subject:
NUCLEAR PROJECTS 1 & 4
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f' DOCKET NOS. 50-460 and 50-513 x
P0TENTIALLY REPORTABLE CONDITION
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RCS ATTACHED PIPING
References:
- 1) Telecon TJ Houchins, Supply System to DF Kirsch, i
Region V Nuclear Regulatory Commission, dated November 13,1980
- 2) G01-1-80-380, DW Mazur to RH Engelken, Directcr dated December 12, 1980
- 3) G01-1-81-198, DW Mazur to RH Faulkenberry, dated July 2, 1981
- 4) G01-1-81-331, DW Mazur to RH Faulkenberry, dated October 5, 1981 In reference (1) the Supply System informed your office of a potentially reportable deficiency under 10CFR50.55(e) and references (2), (3) and (4) i were interim reports on the subject condition.
In reference (3) the Supply System committed to providing quarterly updates on the status of the subject deficiency.
In keeping with j
that commitment, Attachment A includes the same information previously i
provided and Attachment B has been updated.
It is still not possible to l
provide a final completion date for the subject design analysis; there-fore, the Supply System will continue to provide your office with interim status reports on a quarterly basis.
If you have any questions or desire further information, please advise.
i Very truly yours, i
i
. W.
azur i
Program Director WNP-1/4 cc:
C. R. Bryant, Bonneville Power Administration 399 j
V. Stello, Director of Inspection and Enforcement - NRC A. Toth, Resident NRC inspector iET)
Engineering Files 1/4 847
.5 FDCC 899 8201260125 820106 f
PDR ADOCK 05000460 S
PDR g,y
r Attachment A l
WNP-l/4.
Dccket tios. 50-460 and 50-513 Reportable Condition 10CFR50.55(e)
(Potential') Reactor Coolant System Attached Piping Second Interim Report
_ BACKGROUND In G0-1 331 dated Oct. 5,1981, the Supply System provided a report on this subject to the f4RC.
This was an interim report on the planned i
reanalysis being performed on piping attached to the RCS. Commitments were also made to provide the f1RC with quarterly updates on the status of the reanalysis. The purpose of this letter is in response to that comitment and appraises the NRC of our on-going program and the results to date.
DESCRIPTION OF DEFICIENCY The original B&W ECCS analysis assumed that the piping connected to i
the RCS ioop in which the LOCA pipe break is assumed to occur remains intact during the event.
B&W had not analyzed the cold leg (for a i.
second break), pressurizer spray and surge line, or the in-core piping to determine if these assumptions were correct.
In addition, B&W recom-
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mended that the RCS attachment piping analyses under the AE's responsibi-l lity be reviewed to confirm that the basic analytical techniques used are conservative and that the piping is adequately designed to maintain function and integrity under LOCA conditions.
Since the original report, B&W has performed additional analyses and determined for the reactor coolant pump suction and discharge piping that potential breaks at the pump nozzles will not result in subsequent breaks in the reactor coolant piping adjacent to the pumps. Likewise,.
the pressurizer surge-line analysis has been completed and the results demonstrate that this line (attachment to hot leg piping) is not of The remainder of the piping analyzed by B&W (spray line and concern.
in-core piping) is still undergoing evaluation.
The A/E for WNP-1/4, United Engineers and Constructors (UE&C), performed a review of all owners supplied RCS attachment piping and confirmed that a Loss of Coolant Accident Anchor Displacement Analysis (LAD) i had been performed. Further investigations raised questions as to whether the Static Displacement Method (SDM), which was utilized in the analysis, is conservative.
The original SDM analysis utilized, as input, static values over applicable LOCA breaks which contributed to the anchor displacement at the terminal end of the pipe run. liistori-cally, this approach has been consistent with standard industry practice and was believed to be conservative.
However an analysis of two lines as described below has confirmed that the SDM was not conservative 1
and additional analyses and modifications are required.
1 A dynamic analyses was performed in an attempt to demonstrate that the original SDM is conservative. Using worst case time i
history displacements selected by B&W, a time history LAD analysis was performed for a line in the Decay Heat Removal System (DHR-4-1-14) j and Main Steam System (MSS 3-1-28).
The results were as follows:
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The combined pipe stresses (including LOCA and other loadings) did not increase significantly and remained within the allowable 1
limits.
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i 2.
Pipe support loads 11 creased significantly, in some cases up to 30 to 50 percent.
Two supports (DHR) were determined j
to have loadings which exceed the design allowables.
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3.
The component nozzle loads were also detarmined to increase
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significantly. The DHR nozzle loads exceeded B&W allowable l
limits, whereas, the MSS nozzle loads were within B&W allowable j
limits. B&W is currently evaluating the increased DHR nozzle loads for acceptability.
Although the design allowables are xceeded for the pipe supports, this loading only occurs under faulted conditions. The loadings
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calculated do not exceed faulted allowable loads for the supports, however, it will result in deflections of those supports. Since 1
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the original piping analysis was performed without exceeding design conditions for supports and hence, no deflection, any ne.v analysis l
which assumes support deflection will result in higher values j
for pipe stress.
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Attachment B lists the current status for all systems which must be reanalyzed to determine if modifications are required.
4 SAFETY IMPLICATIONS j
Although the two pipe supports identified above would not fail under j
these revised loading conditions, the loading placed on the piping j
as a consequence of the support deflection could result in unacceptably high stresses which would require further modifications. It has not j
yet been determined that this is the case.
i 4
Since the Dynamic Analysis performed has shown a significant increase j
in load over the previous Static Displacement Method and the acceptability of the piping in the proximity of the two, supports which have been determined to be overstressed (design allowables) cannot be assured
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without further analysis, this has been reported under paragraph (iv) j!
of 10CFR50.55(e) since extensive analysis is required. Specific scfety implications involving each deficiency which is found, cannot be determined until the analyses are completed.
CORRECTIVE ACTI0ft PLAfiflED As a result of analyses performed to date, it will be necessary to review /reperform analyses for all attachment piping to the RCS to determine whetaer design conditions are exceeded following an accident. For j
those cases where design allcwable loadings are exceeded, it will be necessary to perform additional evaluations to determine whether hardware modifications are required, i
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CURREfli STATUS ReanalyseshasbegunontheFe$dwaterSystem,AuxiliaryFeedwaterSysten, Core Flood System, and the Decay Heat Removal System, and should be completed before the next quarterly report. The reanalyses of the other designated systems have been held up pending resolution of the steam generator overfill study.
It is still not possible, at this time, to provide a final completion date. The Supply System will continue to provide interim reports to the fiRC on a quarterly basis.
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ATTACHMEllT B LOCA ANCHOR DISPLACEMEffT REAtlALYSIS PROGRESS REPORT No. 2 12/29/81 1
System Tot al
% of Work Remarks Caapl eted MSS 30 Reanalysis of three lines is being held pending results of Steam Generator Overfill S t udy. For the one line completed, pipe stresses, nozzle loads and support loads increased; however, are within acceptable l imi ts.
FWS 75 Reanalysis of all of these lines is in pro-TWA gress. Results will be available for next report.
DHR Pipe stresses and nozzle loads increased; Drop Line 100 however, are wi thin acceptable limi ts. Due to increased loads resulting fran the dynamic reanalysis, f our (4) supports out of nine (9) found to be unacceptable and must be modified.
CFS &
75 Analysis of all of these lines is in progress.
Results will be available f or next report, f1]S (HPI) 0 Time history data is required frcm B & W.
Reanalysis has not started.
I nstr unent 0
Time history data has been received from B & W.
Lines (10)
Reanalysis has not started.
1
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