ML19220B050
| ML19220B050 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/10/1977 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| TASK-TF, TASK-TMR NUDOCS 7904250232 | |
| Download: ML19220B050 (3) | |
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/ccketFiles NRR Reading AB Reading NOV 1 b 13II Docket No. 50-320 MEM0PANDUM FOR:
D. B. Vassallo, Assistant Director for LWRs DPM FROM:
D. F. Ross, Jr., Assistant Director for Reactor Safety, DSS
SUBJECT:
DRAFT SAFETY EVALUATION FOR THREE MILE ISLAND, imIT 2 Plant Name: Three Mile Island, Unit 2 Docket No.: 50-320 Licensing Stage: OL Responsible Branch and Project. Manager: LWR-4 ;
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DSS Brar involved: Analysis Branch Descriptior Of RO Mw: Safety Evaluation Review Status: Inc wplete As requested by the Containment Systems Branch, the Analysis Branch has reviewed the applicant's methods for determining mass and energy release to the containment following a main steam line break. We have requested additional docunentation frem the applicant concerning the ElmAM code which is used in these analyses. Although the applicant stated in snswer to question 222.9 that this infomation would be provided, it has not yet been received. We will require this infomation in order to co=plete our review. With this exception, we have concluded that the analytical methods used by the applicant for calculation of mass and energy release to the contairment following a main steam line break are censervative.
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3,y.naas D. F. Ross, Jr., Assistant Director
Enclosure:
for Reactor Safety Request for Additional D. vision of Systems Safety Information cc:
S. Hanauer Z. Rosztoczy R. Tedesco R. Mattson P. Norian G. Lainas S. Varga J. Guttmann J. Shapaker H. Silver W.Jensen F. Eltawila, Centact "Ix"i79 7 9 04 2 5 0 23Z
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UNITED STATES
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NUCLEAR REGULATORY COMMisslON f.. 1 j
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,....J yov i 0 ETI Docket No. 50-320 MEMORANDUM FOR:
D. B. Vassallo, Assistant Director for LWRs, DPM FROM:
D. F. Ross, Jr., Assistant Director for Reactor Safety, DSS
SUBJECT:
DRAFT SAFETY EVALUATION FOR THREE MILE ISLAND, UNIT 2 Plant Name: Three Mile Island, Unit 2 Docket No.-
50-320 Licensing Stage: CL Responsible Branch and Project Manager: LWR-4; H. Silver DSS Branch Involved: Analysis Branch Descriptior of Review: Safety Evaluation Review Status:
Incomplete As requested by the Containment Systems Branch, the Analysis Branch has reviewed the applicant's methods for determining mass and energy release to the containment following a main ste2m line break. We have requested additional documentation from the applicant concerning the EWHAM code which is used in these analyses. Althcugh the applicant stated in answer to question 222.9 that this information would ce provided, it has not yet been received. We will require tnis information in order to complete our revSw. With this exception, we have conclucco that the analytical methods used by the applicant for calculation of mass and energy release to the containment following a main steam line break are conservative.
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-. ' c D. F. Ross, Jr., Assistant Director
Enclosure:
for Reactor Safety Request for Additional Division of Systems Safety Information cc:
S. Hanauer
- 7. Ros2 toc::y R. Tedesco R. Mattson P. Noriin G. Lainas S. Varga J. Guttmani, J. Shapaker H. Silver W. Jensen F. Eltawila Contact NRR:W. Jensen Ext. 27911
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MAIN STEAM LINE BREAK WITHIN CONTAINMENT - BREAK FLC'.I The mass and energy release to the containment following the postulated double-ended ructure of a main steam line was calculated using the CRAFT-1 computer code with modifications to maxicaize heat flow from the primary into the secondary system and into the containment. The heat transfer from the primary system is sufficient to heat the secondary system break flow to suoerheated steam. The steam flow is released to the containment at sonic' velocity using the Moody critical flow correlation. The main steam isolation valves are assumed to fail so that both steam generators discharge comtletely into the containment.
Feedwater flow into the steam generator from the main feedwater lines is calculated usina the RELAP 4 code for the portion of the system between the feedwater pump suction and the steam generators. Between the feedwater pumo suction and the main condenser, the systems are modeled using the EWHAM code. The RELAP-4 code is a two-phase blowdown code and is used for the portion of the system where two-phase flashing occurs. The EWHAM code is a single phase transient code and is used for the portion of systems where the flow remains subcooled. Using these methods, the feedwater flow was calculated to increase to accroximately two times the normal value before closure of the isolation valves.
Following isolation valve closure, the unisolated mass in the feedwater system was added to the steam gener; tor using the RELAP J code. Sufficient energy was transferred frca the primary system to transform the feedwater flow into steam before release into the containment.
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We have reviewed the applicant's methods and assumptions for the 'ortions of the analysis using the CRAFT-1 and RELAP 4 codes and have concluded that the release rates to the containrent are calculated so as to maximize the containment pressure and temperature.
We have requested additional documentation concerning the EWHAM code.
The applicant has indicated that this information will be provided in the near future. Contingent on our review of the additional documentation of the EWHAM code, we have concluded that the analytical methods used by the aoplicant to calculate mass and energy release to the containment following a postulated main steen line break are conservative.
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