ML20054J800

Forwards Request for Addl Info Re Technical Bases for Core Disruptive Accident Energetics Analyses

ML20054J800
Person / Time
Site:	Clinch River
Issue date:	06/21/1982
From:	Check P Office of Nuclear Reactor Regulation
To:	Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT
References
NUDOCS 8206290580
Download: ML20054J800 (5)

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<s.r of Distribution June 21, 1982 I nncimt file CRBR Reading NRC PDR Docket No.: 50-537 Local PDR NSIC CRBR Staff Mr. John R. Longenecker Licensing and Environmental Coordination Clinch River Breeder Reactor Plant U. S. Department of Energy, NE-561 Washington, D.C.

20545

SUBJECT:

CLINCil RIVER BREEDER REACTOR PLANT, REQUEST FOR ADDITIONAL INFORMATION The staff requires additional infonnation which is described in the enclosure.

The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.

If you desire any discussion or clarification of the information requested, please contact R. M. Stark, Project Manager (301) 492-9732.

Sincerely, c.e swe.c4 ht twi :

Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation

Enclosure:

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Dr. Cadet H. Hand, Jr., Director Barbara A. Finamore Bodega Marine Laboratory S. Jacob Scherr University of California Ellyn R. Weiss P. O. Box 247 Dr. Thomas B. Cochran Bodega Bay, California 94923 Natural Resources Defense Council, Inc.

Daniel Swanson 1725 I Street, N.W.

Office of the Executive Suite 600 Legal Director Washington, D.C.

20006 U. S. Nuclear Regulatory Comission Eldon V. C. Greenberg Washington, D.C.

20555 Tuttle & Taylor 1901 L Street, N.W.

William B. Hubbard, Esq.

Suite 805 Assistant Attorney General Washington, D.C.

20036 State of Tennessee Office of the Attorney General L. Ribb 450 James Robertson Parkway LNR Associates Nashville, TN 37219 Nuclear Power Safety Consultants 8605 Grimsby Court William E. Lantrip, Esq.

Potomac, MD 20854 City Attorney Municipal Building P. O. Box 1 Oak Ridge, TN 37830 George L. Edgar, Esq.

Morgan, Lewis & Bockius 1800 M Street, N.W.

Washington, D.C.

20036 Herbert S. Sanger, Jr., Esq.

General Counsel Tennessee Valley Authority Knoxville, TN 37902 Scott Stuckey, Chief Docketing and Service Section Office of the Secretary U. S. Nuclear Regulatory Commission Washington, D.C.

20555 Raymond L. Copeland Project Management Corp.

P. O. Box U Oak Ridge, Tennessee 37830

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION CS760.178 The staff and its consultants have held five meetings with your representatives for discussions of the technical bases for the core disruptive accident energetics analyses presented in GEFR-00523.* These discussions have focused on questions asked by the NRC staff and its consultards at the January 27, 1982 meeting.

Those questions were:

A.

Initiation Phase 1.

Can TOP accidents become prompt-critical in such a way that internal fuel motion in lower power channels is the key factor in the energetics determination?

Is such an event possible only for midplane failures with low sweepout? How is the degree of sweepout determined?

What is the effect of intrasubassembly incoherence on sweepout?

2.

An LOF-d-TOP might still occur if the sodium void worth is 50-60 percent higher and internal fuel motion in TOP type channels can occur.

What are the reactivity un-certainties of sodium void, Doppler, axial expansion and lead channel fuel motion? How do you interpret the significance of these uncertainties?

3.

What is the potential for autocatalysis due to plenum fission gas acting on the fuel column to force axial compaction as disruption occurs in the initiating phase of the LOF?

B.

Meltout Phase 4.

To what extent can steel blockages form throughout the l

core to prevent fuel removal through normal axial blanket flow channels during the early phase of the LOF? What is the location and character of the steel blockages in these channels?

5.

What is the basis for maintaining continuous subcriticality in the high heat loss environment of early meltout phase?

What are the fuel losses (quantified) taking into account uncertainties in removal path geometries, driving pressures and freezing mechanisms?

l The meetings were held on January 27, 1982 in Bethesda, Md.; February 18, 1982 in Bethesda, Md.; March 10, 1982 in Sunnyvale, CA; March 25, 1982 in Bethesda, Md. and April 27, 1982 at Argonne National Laboratory.

l

. C.

Boiled Up Pool Phase 6.

What degree of subcriticality is required to prevent pool recriticality from thermal and fluid dynamics upset conditions? What is your position on the potential for small recriticalities to amplify? What is the justi-fication for your position?

7.

In assessing benign termination from the boiled-up pool justify the fuel removal mechanisms and rates.

In parti-cular assess the potential for upper pool sodium entry via rapid condensation of steel vapor pressure.

D.

Expansion Phase 8.

What is your estimate of the force required to produce a mechanically induced relief path via upper internals structures displacement?

To facilitate our ongoing review of CDA energetics please provide, in a format suitable for reference, documentation of the relevant infonnation submitted in these meetings together with any appro-priate additional information you may have that applies to these guestions.

CS760.179 Dimensioned design layout drawings are required by the staff and its consultants to provide an accurate basis for the geometries used in analysis of hypothetical core disruptive accident energetics.

Please provide design layout drawings, including dimensions, materials and weldments, for all the components and structures in the reactor vessel including:

a) fuel, blanket, control and removab'le radial shield assemblies; b) the upper internal structure and all its components; c) the core support plate and all its components; d) the reactor vessel inlet plenum, including the inlet piping and core support cone; e) the core barrel, the core former rings, the fixed radial shielding, the horizontal baffle, the fuel transfer and storage assembly (FT&SA) the bypass flow modules, the reactor vessel thermal liner, and any component or structure connected to the core barrel; and

I 3-f) the reactor closure head, with its 3 rotating plugs, including the reactor vessel walls, the reactor vessel thermal liner, the gas entrainment suppressor plate, the thermal and radio-logical shielding plates, the upper internal structure (UIS)

Jacking mechanisms, the liquid level monitor plugs, the vessel l

flange, the riser assemblies (inner and outer for all plugs),

the riser dip seals, the riser elastomer seals, any component 1

cr structure connected to the reactor closure head, and all penetrations thereof.

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