ML19326D827

Requests Response to Encl Questions Re Fuel Design & Physics Calculations to Complete Review of Basic Safety Rept Supporting Cycle 4 Reload.Requests That Addl Info Be Provided by 800630 to Meet Review Schedule

ML19326D827
Person / Time
Site:	Millstone
Issue date:	06/20/1980
From:	Clark R Office of Nuclear Reactor Regulation
To:	Counsil W NORTHEAST NUCLEAR ENERGY CO.
References
TAC-11348, TAC-11561, TAC-12505, TAC-42846, TAC-43380, NUDOCS 8007030362
Download: ML19326D827 (5)

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION r.

ay WASHINGTO N, D. C. 20555

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June 20,1980 Docket No. 50-336 Mr. W. G. Counsil, Vice President

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Nuclear Engineeri.ng & Operations

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Northeast Nuclear Energy Company P. O. Box 270 Hartford, Connecticut 06101

Dear Mr. Counsil:

In the process of reviewing the Basic Safety Report supporting the Cycle 4 I

Reload of Millstone, Unit No. 2, we find that additional information as delaited in the enclosure is needed to complete our review.

These questions relate to fuel design and physics calculations. Other questions may arise from reviews under way.by the Thermal-Hydraulic Sections of che Core Per-formance Branch and by-the Reactor Systems Branch.

In order to meet the agreed upon schedule for this review, please provide the additional information, previously telecopied to Mr. R. Kacich on June 16,1980, by June 30, 1980.

Sincerely, r

lim Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing

Enclosure:

Request for Additional Information cc w/ enclosure: See next page s

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l Northeast Nuclear Energy Ccmpany Mr. John Shedlosky cc:

William H. Cuddy, Esquire Resident Inspect,or/ Millstone Day, Berry & Howard c/o U.S. NRC Counselors at Law P. O. Drawer KK One Constitution Plaza Niantic, CT 06357 Hartford, Connecticut 06103 Mr. Charles B. Brinkman Anthony Z. Roisman Manager - Washing, ton Nuclear i

Natural Resources Defense Council Operations 917 15th Street, N.W.

C-E Power Systems Washington, D.C.

20005 Combustion Engineering, Inc.

4853 Cordell Ave., Suite A-1 Mr. Lawrence Bettencourt. First Selectman Bethesda, Maryland 20014~

Town of Waterford Hall of Records - 200 Boston Post Road Waterford, Connecticut 06385 Northeast Nuclear Energy Company ATTN:

Superintendent Millstone Plant Connecticut Energy Agency Post Office Box 128 ATTN: Assistant Director, Research Waterford, Connecticut 06385 and Policy Development Department of Planning and Ener97 Director, Technical Assessment Policy Division 20 Grand Street Office of Radiaticn Programs Hartford, Connecticut 06106 (AW-459)

U. S.* Environmental Protection Agency Crystal Mall #2 Arlington, Virginia 20460 U. S. Environmental Protection Agency Region I Office ATTN: EIS COORDINATOR John F. Kennedy Federal Building j

Boston, Massachusetts 02203 Waterford Public Library Rope Ferry Road, Route 156 Waterford, Connecticut 06385 Northeast Utilities Service Company ATTN: Mr. James R. Himmelwright Nuclear Engineering and Operations P. O. Box 270 Hartford, C.nnecticut 06101 o

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ENCLOSURE Questions on Westinghouse Basic Sa'fety Report for Millstone,' Unit'2 l

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Coolant pressure drop calculations for the Westinghouse fuel assembly design indicates a matching of the overall pressure drop with that for the original Combustion Engineering Millstone 2 fuel assembly design. However, at each axial elevation the pressure drops do not match up between the Westinghouse and Combustion Engineering fuel designs. The largest variation in pressure drops for the two designs occurt at the lower nozzle where the Westinghouse design has a higher pressure-loss coef ficient. This variation in pressure drop will result in an inlet flow maldistribution with less direct flow through the Westinghouse bottom nozzle. The BSR should provide justification as to why the resulting cross flow downstream of the bottom nozzle will not produce an unacceptable degree of fretting wear at sites where spacer grid springs and dimples contact fuel rods.

2.

The Westinghouse fuel assembly design has 4 holddown springs, while the original Combustion Engineering design has 5 springs.

Discuss the differences in the static and dynamic response of each fuel assembly design. The raised pad on the center of the top nozzle orifice place prevents the Westinghouse holddown springs from being compressed solid.

Does this pad limit the axial distance that the Westjnghouse fuel assemblies can grow relative to that of the original Combustion Engineering fuel assemblies? Will the spacer grids of the two fuel assembly. designs always line up? What is the safety signifi-cance if grid-to-grid alignment cannot be assured (i.e., will there be neutronic anomalies, will assembly peripheral fuel rods be punctured)?

3.

The BSR states that cladding flattening is precluded during the projected exposure of the fuel.

Provide,the minimum time to collapse as calculated with the COLLAP code. What is the design maximum value of fuel assembly burnup?

4, What is the calculated minimum shoulder gap which allows for differential growth between fuel rods and the fuel assembly? Provide the two Zircaloy growth correlations used in this calculation and describe or provide the' data base from which these correlations were determined.

How were the growth correlations combined with (a) fabrication tolerances, (b) differential thermal strains of the fuel assembly and reactor in-ternals, and (c) elastic compression and creep of the guide thimble tubes? For steady-state operation, at what axially-averaged assembly burnup will interference result in rod bow?

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5.

The NRC staff has not commenced the review of the Westinghpuse generic topical report WCAP-8691, Revision 1, " Fuel Rod Bowing Evaluation,"

which is referenced in the BSR.

Specifically, the BSR uses a formula from WCAP-8691 that projects anticipated rod bow magnitudes due solely to geometrical changes in the fuel rod thickness and diameter and spacer grid span length. This formula has been somewhat controversial-and has not been accepted by the staff. Therefore, we will require that the degree of rod bowing in the Westinghouse reload fuel be calculated with the existing approved method, which is~ relatively more conservative.

In spite of this additional conservatism, however, we do not calculate a need for. a DNBR penalty until an assembly burnup of 36,300 MWD /tU is attained at which exposure the 50% gap closure value is reached.

We require that Westinohouse confirm our calcula-

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tions and verify that no other changes in fuel design variables (i.e.,

grid spring preload, degree of cladding cold work, etc.) are significant to the rod bowing extrapolation for the Millstone, Unit 2 reload fuel.

6.

The Combustion Engineering supplied fuel for Millstone, Unit 2 was designed according to a specific set of Specified Acceptable Fuel Design Limits (SAFDLs).

Please list all of the Westinghouse SAFDLs for the Millstone reload fuel and provide the bases for omissions or additions to the original Combustion Engineering set of SAFDLs.

7.

Some of the accident analyses described in the BSR were performed with the computer codes FACTRAN (WCAP-7908, "FACTRAN, A Fortran IV Code for Thermal Transients in a U0p Fuel Rod) and LOFTRAN (WCAP-7907, "LOFTRAN Code Description").

Our review of these topicals has pro-gressed to the point that there is reasonable assurance that the con-clusicns based on these analyses will not be appreciably altered by completion of the analytical review, and therefore that there will be no effect on the decision to issue a license amendment.

If the final approval of these topical reports indicates that any revisions to the analyses are required, Millstone Unit 2 will be required to implement the results of such changes.

8.

Please either reference or provide a thorough description of the Westinghouse Computer Analysis Code (WECAN), which was used to perform the stress analyses of fuel assembly components.

9.

Comparisons of power peaking in fuel pins adjacent to CEA water holes using TURTLE (diffusion theory) and KEN 0 (Monte Carlo) have shown an ur.derprediction by diffusion theory, as expected.

Please provide additional information, such as comparisons between KENO calculations and experimental measurements of water hole power peaking, to justify the KEN 0 calculational uncertainty used.

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The fuel rod support grid for Cycle 4 will be Inconel-718 rather than Zircaloy-4 as used in Cycle 3.

What are the effects of this material change on power distributions and other physics. parameters?

11.

Power. distributions calculated by TURTLE appear to underpredict the power in the peripheral assemblies while overpredicting the power in the center assemblies.

In view of the large errors in predicting CEA worth near the core periphery due to these power distribution inaccur-

- acies', justify the use of TURTLE without some type of baffle correction scheme.

12. The CEA drop analysis was performed without automatic rod control (CEA motion inhibit) or turbine load reduction.

Is thi.s.the operati.onal plan for Cycle 47 13.

The parameters used in the analysis of the CEA ejection accident appear to be representative of Westinghouse cores and differ from the previous Millstone, Unit 2..fueLsupplier in several areas such as ejected rod worths, ejection time,(del.ayed neutron, fraction, feedback reactivity weighting, and power peaking F).

Please provide a comparison between the-Cycle 3

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q and Cycle 4 values of these rod ejection initial assumptions and discuss the reasons for and effects of any differences.

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