ML19220B914
| ML19220B914 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/03/1976 |
| From: | Norian P Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904270653 | |
| Download: ML19220B914 (2) | |
Text
.
a LG 3 1371 Docket No.tb)
Richard C. DeYoung, Jr., Acting Assistant Director for L'.as, DPM THRU: Zoltan R. Rosztoczy, Chief, Analysis Branch, DSS REQUESTED ADDITIONAL INF0PF.ATION - THREE MILE ISLAND, UNIT 2 Plant Name: Three Mile Island, Unit 2 Docket No.: 50-320 Licensing Stage: OL Milestone No.: 24.0 P,esponsible Branch & Project Manager: LWR-2:
H. Silver Description of Review: Supplemental Questions Requested Completion Date: N/S Review Status:
Incocolete As requested by the Containment Systems Branch, the Analysis Branch is reviewing the containeent mass and enercy release rodel for tne postulated rain steam line break in Three Mile Island, Unit 2.
We require the enclosed additional infonnation in order to cecolete our rev1ew.
Original Signed B7:
PaulE.Nor123 Paul E. Norian, Section Leader Analysis Branch Division of Systems Safety
Enclosure:
Questions cc: S. Hanauer R. Heineman D. Ross Z. Rosztoczy K. Kniel T. Novak F. Eltawila 790427066.3 F. Odar G. Lainas H. Silver J. Xudrick
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AUG 3 1976 Docket No. 50-320 Richard C. DeYoung, Jr., Acting Assistant Director for LWRs, DFM THRU: Zoltan R. Ros:toczy, Chief, Analysts Branch, DSS g.gp REQUESTED ADDITIONAL INFORMATION - THREE MILE ISLAND, UNIT 2 Plant Name: Three Mile Island, Unit 2 Dociet No.: 50-320 Licensing Stage: OL Milestone No.: 24.0 Responsible Branch & Project Manager: LWR-2; H, Silver Description of Review: Supplemental Questions Requested Completion Date: N/S Review Status: Incomplete As requested by the Containment Syscems Branch, the Analysis Branch is reviewing the containment mass and energy release mcdel for the postulated main steam l':e break in Three Mile Island, Unit 2.
We require the enclosee aeditional infomation in order to complete our review.
l w
Paul E. Norian, Section Leader Analysis Branch Division of Syste s Safety
Enclosure:
Questions cc: S. Hanauer R. Heineman D. Ross Z. Rosztoczy K. Kniel T. Novak F. Eltawila F. Odar G. Lainas H. Silver J. Kudrick 84-033
/
222-1 222.0 SYSTEMS ANALYSIS SECTION, ANALYSIS BRANCH 222.1 The following questions concern calculation of mais and energy (15.1.15) release to the containment from a double ended main steam line break as described in Section 15.1.15. These calculations were performed using the FLASH-2 code, a.
Discuss in detail all modifications to the FLASH-2 code.
Provide and justify all equations and assumptions in the inodifications to the code.
b.
Provide a diagram showing the primary loop and steam generator simulatton in the FLASH-2 code, Provide a table of the mass ana associated enthalpy as a c.
function of location for the water in the feedwater system and condensate system between the single feedwater isclation valve and the next safety grade valve that will close in the event of a steam line break, d.
The mass and energy release data presented in Tables 15.1.15-2B and 42.7-1 through 42.7-3 include entrained liquid which enters the containment from the broken steam line. Our axperience is that liquid entrainment will provide a lower containment pressure than the case tf no liquid entrainment is assumed. Provide and justify, including compartsons with appropriate experimental data, the steam separation model used in the FLASH-2 computer program. We believe the assumption of zero liquid entrainment to be most conservative.
e.
Provide and justify the heat transfer coefficients and heat transfer areas (UA) used in the FLASH-2 code for heat flow betwaen the two steam generators and the primary system. We believe that a conservative assumption would be to hold the (UA) at the initial value throughout the transient for both steam generators.
f.
Using assumptions of liquid entrainment for which you have provided appropriate justification, provide the results of a steam line break spectrten analysis which identifies the break size producing the highest containment pressure.
Include mass and energy release data for the most severe break size. This break size is riormally the larnst break for which zero liquid entrainment is calculated.
84~034