ML20090H732
| ML20090H732 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/19/1984 |
| From: | Tessier J ARGONNE NATIONAL LABORATORY |
| To: | Guttmann J Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20090B375 | List: |
| References | |
| CON-FIN-A-2311, FOIA-91-106 NUDOCS 8404270010 | |
| Download: ML20090H732 (3) | |
Text
ARCONNE NATIONAltABORATORY 9X)OScur Cu.tvu,Mcpu.t!ros 604H Ug h n 317/97/.
3338 April 19.1984 Mr. Jack Guttmann Reactor Systems Branch Division of Systems Integration Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. _20555
Subject:
Completion of Ginna Steam Generator Tube Rupture RETRAN Calculations
- FIN A2311
References:
1.
Letter, T. Y. C. Wei to J. Guttmann, " FIN A2311 Task !!:
Ginna Steam Generator Tube Rupture Calculations " March 6, 1984 2.
INPO Draf t Report, " Thermal-Hydraulic Analysis of Ginna Steam Generator Tube Rupture Event, September,1983.
Dear Mr. Guttmann:
l Reference 1 reported on the completion of subtasks !!.B.1 (Secondary System Behavior) and II.B.2 (Stuck Open PORV) for the subject FIN. We have i
now completed the remainder of the required RETRAN calculations using input supplied by INPO and-the Modo 3 versions of the computer code.
Subtask !!.B instructed us to re-calculate the Ginna tube rupture event using HETRAN02 at ANL and the input decks developed cooperatively by RG&E and INPO. This analysis was done to verify that the INPO model, when run at ANL reproduces their results, reported in Reference 2.
The ANL calculations were i
made with the latest released version of the code, viz. RETRAN02/ Modo 3, i
whereas the INPO results.were obtained using a pre-release version, Modo 3A.
This required that changes to the input decks be made in order that it run on Modo 3; these changes were identified in the documentation accompanying the code distribution package. More important, from the standpoint of schedule i
and budget, coding deficiencies or errors caused significant delays and at-tendant losses in man and computer time. These dif ficulties were overcome, L
of ten through discussions with El personnel who were most helpful in resolving
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i i-M,r. Jack Guttmann 2
April 19,1984 the problems. In two instances changes to the FORTRAN source decks were suggested and made. Thus, replicating the INP0 calculations was not as simple and straightforward a task as originally thought and the models and/or solu-tion techniques are not absolutely identical. It is to be expected, there-fore, that there be some diffc ences in calculated results and such is the However, observations made to date indicate the results are comparable case.
to well within the range of uncertainties inherent in such calculations. Our current evaluation is that INP0's calculations and conclusions derived there-from are confirmed insofar as is possible with a state-of-the-art computer code as RETRAN. We are presently continuing to examine the results, and in particular, are preparing comparative results in graphical form to illustrate the level of agreement in the two sets of calculations. Code errors dis-covered have.been formally reported to EPRI, El and EPSC for general reso-lution as required by their procedures, except for one problem wherein the request for submittal of input decks has been referred to RGaE for approval.
In subtask II.B.3 (Failure to Terminate HPI), we were requested to deter-f mine the consequences in the primary and secondary systems if HPI were not j
terminated as in the actual event. The operator secured the HPI pumps at one hour. and twelve minutes af ter tube rupture in the Ginna event which inter-rupted the ongoing cooldown of the RCS and dramatically reduced the discharge of radioactive water to the environment; some release continued because the charging pumps remained-on and their flow exceeded the letdown rate. INPO estimated that the faulted steam generator (SGB) safety relief valve (SRV) did not completely close until-three hours and two minutes af ter the tube rup-1 ture.
For the purposes of this subtask, INPO's modeling of operator control.
of the intact steam generator (feed and bleed) was unaltered; the only changes in the model were to inhibit tripping the HPI and to maintain a constant flow area for SGS SRV equal to that assumed when llPI was terminated in the actual event. The calculation was continued for eight minutes beyond actual HPI termination, and the response trends in the primary and secondary systems are l
as anticipated. Primary system (RCS) pressure remains above that of SGB by approximately 300 psi.
This maintains the tube rupture flow into SGB and attendant release through its SRV at a rate of nearly 600 gpm which approxi-mately equals the sum of safety. injection and charging flow rates.
Sustained injection of this relatively cold water into the RCS. causes a moderate rate of cooldown to continue. For example, examination of the calculated fluid temp-erature in the reactor vessel downcomer gives an estimated rate of approxi-mately -70'f/ hour during the end period of the calculation; the results also P
indicate a slow redaction in cooldown rate as anticip:ted. Based upon a limiting rate of -100*F/ hour, these results show that a certain thermal margin still exists even for continued operation of the HPI system.
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.Mr. Jack Guttmann 3
April 19, 1984 The fore calculations. going summaaizes the salient results of the indicated RETRAN We are >ce;aring a final report that will provide details on all RETRAN calculattua> made for this task.
Sincerely yours.
s M
J.
essier Light Water Reactor Systems Analysis Reactor Ana ysis an y Division T
s T. Y. C.
'e k Light Water Reactor 5,yf ems Analysis Reactor Analysis and Safety Division TYCW:JHT:Jkb R. Mattson, Director, Division of Systems Integration, NRC/NRR cc:
B. Sheron, Chief, Reactor Systems Branch, NRC/NRR R. W. Houston, Assistant Director for Reactor Safety, NRC/NRR S. M. Boyd, NRC/NRR J. Carter, NRC/NRR B. L. Grenier, NRC/NRR W.'Jensen, NRC/NRR N. Lauben NRC/NRR L. W. Deitrich, RAS R. Avery, RAS LWR Systems Analysis Section RAS Files: 8M627, A15 I
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