ML19275H405

Analysis Branch Input to First Round Questions

ML19275H405
Person / Time
Site:	Midland
Issue date:	03/13/1978
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML111090060	List: ML19208A315 ML19208A325 ML19250B014 ML19250B445 ML19250B480 ML19250B482 ML19254C975 ML19254G119 ML19254G128 ML19255K396 ML19255K397 ML19257E274 ML19257F047 ML19257F050 ML19260B823 ML19260E912 ML19260E913 ML19260E914 ML19261C469 ML19262E467 ML19262F230 ML19263D836 ML19263F605 ML19263G068 ML19268C086 ML19268C087 ML19269F130 ML19271A085 ML19271A172 ML19273B051 ML19273D308 ML19274A409 ML19274E085 ML19275B428 ML19275B444 ML19275C327 ML19275C328 ML19275G904 ML19275G918 ML19275H308 ML19275H317 ML19275H339 ML19275H341 ML19275H375 ML19275H401 ML19275H405 ML19275H493 ML19275H655 ML19275H660 ML19276D594 ... further results
References
FOIA-80-515, FOIA-80-555 NUDOCS 8006240759
Download: ML19275H405 (3)

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222-1 222.0 SYSTEMS ANALYSIS SECTION, ANALYSIS BRANCH 222.1 Provide a discussion of the calculational methods used in the evaluation of the steamline break accident including all the codes used.

222.2 Provide a detailed flow diagram for the primary and secondary systems identifying all the components considered in the steam-line break accident analysis.

222.3 Describe how the initial and transient power distributions were calculated.

Provide the initial and transient power distributions used in these analysis.

222.4 Describe in detail how the thermal-hydraulic effects were evaluated including calculations of DNBR.

222.5 Provide transient axial and radial power distributions for each case analyzed.

Describe how these peaking factors were considered in the thermal-hydraulic calculations.

How were these peaking factors calculated?

222.6 Provide all the time dependent reactivity feedbacks during the steamline break accident (provide for all the cases analyzed).

222.7 Provide nuclear and thermal-hydraulic analyses for the first 15 seconds of the steamline break accident for both BOL and E0L conditions from full power.

222.8 For the high pressure safety injection system and the flow of borated water from core reflood tanks, describe the flow path into the core, the method for evaluating the time for these fluids to reach the center of the core and the method for determining the resultant reactivity feedback.

222.9 Describe in detail how the coolant flow reduction in the hot channel is evaluated following the steamline break accident.

Discuss the potential for coolant flow blockage due to fuel swelling.

222.10 Describe in detail how the time dependent pressure drop in the fuel channel was calculated following a steamline break accident.

222.11 Provide a plot of the core coolant density and average fuel temperature following the steamline break accident for the time period from zero to 15 seconds.

222.12 Describe how the peaking factors in the hot channel were determined.

222.13 Provide a discussion of the calculational methods used in the evaluation of feedline break accident including all the codes used.

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w 222-2 222.14 Provide a detailed flow diagram for primary and secondary systems identifying all the components considered by this feedline break analysis.

222.15 Provide and justify the worst single failurc considered for the feedline break accice..t.

222.16 Pros.

a spectrum of feedline break accidents and identify the worst case.

222.17 Provide the discharge rates out of the feedwater line including the mass inventory in the steam generators during the feedline break accident.

222.18 Provide the time dependent primary and secondary pressures during the feedline break accident.