ML19344E080

Forwards Nonproprietary Response to Questions on Cycle 4 Basic Safety Rept

ML19344E080
Person / Time
Site:	Millstone
Issue date:	08/14/1980
From:	Counsil W, Fee W NORTHEAST NUCLEAR ENERGY CO.
To:	Clark R Office of Nuclear Reactor Regulation
References
A01085, A1085, TAC-11348, TAC-11561, TAC-12505, TAC-42846, NUDOCS 8008260596
Download: ML19344E080 (19)

Text

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NOITITIEAST trrII. FRIES 8

., MA fo*RE CONNECTICUT 06101 3 "7'l ;;?",y'." F,T.i~

g ...; . , , , ;g . .;~7 August 14, 1980 Docket No. 50-336 A01085 Director of Nuclear Reactor Regulation Attn: Mr. Robert A. Clark, Chief Operating Reactors Branch #3 U. S. Nuclear Regulatory Commission Washington, D.C. 20555

References:

(1) R. A. Clark letter to W. G. Counsil dated June 20, 1980.

(2) W. G. Counsil letter to R. A. Clark dated July 7, 1980.

(3) W. G. Counsil letter to R. A. Clark dated July 22, 1980.

(4) R. A. Wiesemann letter to 11. R. Denton dated February 29, 1980.

(5) R. A. Wiesemann letter to D. G. Eisenhut dated July 27, 1976.

Gentlemen:

Millstone Nuclear Power Station, Unit No. 2 Response to Questions on Cycle 4 Basic Safety Report In Reference (1), the NRC Staff requested Northeast Nucleer Energy Company (NNECO) to provide additional information regarding fuel designInand physics calculations References (2) and (3),

to continue review of ttie Basic Safety Report (BSR).

NNECO responded to questions identified in Reference (1), and indicated that this information was non-proprietary.

It has subsequently been determined that the information docketed in References (2) and (3) is proprietary to Westinghouse Electric Corporation and is respectfully requested to be withdrawan from the Public Document Room.

The non-proprietary version of the responses is provided as Attachment 1.

Due to the proprietary nature of portions of the material contained in References (2) and (3), NNECO requests that it be withheld from public disclosure in accordance with the provisions of 10CFR2.790 and that this material be safeguarded. The reasons for the classification of this l

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material as proprietary are delineated in the affidavits previously docketed by References (4) and (5).

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY s # Y$

W. G. C6dnsil Senior Vice President Attachment

. , DJ By: W. F. Fee Executive Vice President

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DOCKET NO. 50-336 ATTACilMENT 1 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 RESPONSES TO NRC QUESTIONS ON BASIC SAFETY REPORT NON-PROPRIETARY AUGUST, 1980

Responses to NRC Questions on Westinghouse Basic Safety Recort for Millstone Unit 2

1. Please submit page 5-14 which is missing from the BSR.

RESPONSE: Please find enclosed with this submittal copies of the missing page.

2. 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 occurs at the lower nozzle where the Westinghouse design has a higher pressure-loss coefficien t. 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 botton nozzle will not produce an unacceptable degree of fretting wear at sites where spacer grid springs and dimples contact fuel rods.

RESPONSE: As stated in Section 3.3 and shown in Table 3.1 of the BSR, the Westinghouse fuel assembly design is hydraulically compatible with the Combustion Engineering Millstone 2 fuel assembly design. The geometric similarity of all the assembly corponents and subsequent testing have assured that the overall assembly pressure drops as well as the pressure drops at each axial elevation are matched for the two designs.

The lower nozzle loss coefficient mismatch which was a concern at the Feb. 14,1977 meeting (Reference 1) was resolved by a re-design of the lower nozzle and subsequent testing to determine the effects of rods lifted and not lifted. This nozzle re-design and test results were presented at the January 26, 1979 meeting (Reference 2) and showed that the W lower nozzle was hydraulically compatible with the CE lower nozzle.

Thus there will be no inlet flow maldistribution between the two designs.

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F References M. M. Mendonca, NRC. Memorandum to R. L. Baer, NRC,

Subject:

(1) ,

" Westinghouse Design and Testing Program for Millstone 2 Reload Nutrber 1 Meeting - February 14,1977", April 19,1977.

(2) W. G. Counsil, Northeast Nuclear Energy Company, letter to R. Reid, NRC, " Millstone Nuclear Power Station, Unit No. 2 Cycle 4 Reload Regarding January 26, 1979 Meeting", March 21,1979.

3. The Westinghouse fuel assedly design has 4 holddown springs, while the original Co2ustion Engineering design has 5 spr'ngs. Discuss the dif-ferences in the static and dynamic response of each fuel assently design.

The raised pad on the center of the top nozzle orifice plate prevents the Does this pad Westinghouse holddown springs from being compmssed solid.

limit the axial distance that the Westinghouse fuel assentlies can grow relative to that of the original Corbustion Engineering fuel assemblies?

Will the spacer grids of the two fuel assently designs always line up?

What is the safety significance if grid-to-grid alignment cannot be assured (i.e;, will there be neutronic anomalies, will assembly peripheral fuel mds be punctured)?

RESPONSE

The W fuel assenbly holddown springs are designed to pmvent lift off of the fuel assembly from the bottom core plate during normal operation.

This design requirement is consistent with the CE fuel assembly design of the holddown spring as discussed in section 3.3.1.4 of the Millstone Unit 2 FSAR. Since both the CE and W fuel assembly holddown loads are similar, the fuel assembly spring rates am judged to be similar, thus the static and dynamic response of each fuel assembly should not differ significantly.

The raised pad at the center of the W top nozzle orifice plate will not interfere with the free growth of the fuel assetbly. It is located such that it contacts the holddown flower prior to the springs being compressed solid and prevents the nozzle extensions from topping out in the blind holes in the upper core plate. This promotes uniform loading of the guide tubes during any hypothetical accident which could cause the fuel assembly to lift.

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l Since the raised pad does not prevent fuel assembly growth, there will always be grid to grid overlap between W and CE fuel assentlies even when considering irradiation growth.

4. The BSR states that cladding fla'ttening 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? ,

RESPONSE : The clad flattening time is predicted to be2 50000 EFPH for (I) Region

, region 6 fuel using the current Westir.ghouse evaluation model.

6 fuel, comprised of 6-1 and 6-2, has a projected residence time through 3 cycles of ~27000 EFPH.

5. What is the calculatcd minimum shoulder gap which allows for differential growth between fuel rods and the fuel assently? Provide the two Zircaloy growth correlations used in this calculation and describc or provide the data base from which these correlations wem determined. How were the growth correlations combined with (a) fabrication tolerances, (b) dif-ferential themal strains of the fuel assembly and reactor internals, For and,(c) elastic compression and creep of the guide thimble tubes?

steady-state operation, at what axially-averaged assembly burnup will interference result in rod bow?

RESPONSE : To insure that no axial interference between the fuel rod and 5e assembly nozzles can occur, the W Millstone Unit 2 fuel rod was de-signed so that the minimum roo emperature clearance value would be equal to or greatgr than ,

_ percent times the fuel rod length _

Extensive operating experience with other fuel asserblies using the same materials and having equivalent fuel duty has shown that this design value for clearance is conservative with respect to temperature and irradiation induced length changes of the fuel rod and fuel assembly.

Therefore, under normal steady-state operation, there will be no inter-ference of fuel rods since the growth allowance will preclude rod bow.

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6. The NRC staff has not connen.ced the review of the Westinghouse 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 peralty until an assembly burnup of 36,300 MWD /MTV is 6?tained at which exposure the 50% gap closure value is reached. We require that Westin,ghouse confirm our calculations and verify that no o'J1er 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.

RESPONSE: The degree of fuel rod bowing in the Westinghouse reload fuel has been recalculated with the existing, more conservative approved method:

Ly (M.S.II)/Ig(M.S.II)

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= 0.59 l

15x15 /I 15x15 where L = span length I = cross-sectional moment of inertia The average burnup at which a gap closure of 50% is attained is 32,000 MWD /MTU. By the time the fuel attains a burnup of 32,000 MWD /MTU, it is not capable of achieving limiting peaking factors due to the decrease in fissionable isotopes and the buildup of fission product inventory.

This physical burndown effect is greater than the rod bowing effects which would be calculated at those burnups. Therefore the effect of rod bow need not be considered in the analysis of the Millstone II core.

The fuel design variables were selected according to standard Westinghouse design practice. Therefore, the rod bowinq extracolation method is valid.

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7. 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 basis for omissions or.

additions to the original Combustion Engineering set of SAFDLs.

RESPONSE: There are only two Specified Acceptable Fuel Design Limits (SAFDL's) covered by the Reactor Protection System (RPS), and these are discussed in Chapter 6'of the Basic Safety Report (BSR). These SAFDL's are:

1) The peak linear heat rate must be below that which would cause incipient fuel . centerline melting. The :nelting point limit is con-servatively tak'en as 4700'F to bound the effects of fuel burnup and uncertainties in the melting point.

2) The DNB themal limits must not be exceeded.

The SAFDL's as defined by Westinghouse am equivalent, or mom conser-vative, when' compared to the CE SAFDL's as given in Chapter 1 of CENPD-199:

a) The reactor fuel shall not experience centerline melt.

b) The departure from nucleate boiling ratio (W-3 DNBR) shall have a minimum allowable limit of 1.3.

Further discussion of fuel design criteria may be found in Chapter 2 of the BSR and Chapter 3 of the FSAR. ,

8. Some of the accident anlayses described in the BSR were perforr.ed with the compcter codes FACTRAN (WCAP-7908, "FACTRM, A Fortran IV Code for Themal Transients in a U0 Fuel 2 Rod") and LOFTRAN (WCAP-7907, "LOFTRM Code Description"). Our review'of these topicals has progmssed to the point that them is reasonable assurance that the conclusions based on these analyses will not be appmciably altered by completion of the analytical mytew, and therefore that there will be no effect on the decision to issue a license amendment. If the final approval of these topical reports ing p.. .u . , . . , , . , ,- . . . . , . . . . - .. .. . ,

will be required to li.pleaent the resuits of suoi changes.

r RESP 0f 4SE : We do not expect that any revised analyses will be required as a result of your review of FACTRAN or LOFTRAN.

9. Please either reference or provide a thorough description of the Westing-house Computer Analyses Code (WECAN), which was used to perform the stress analyses of fuel assembly components.

RESPONSE: A description and benchmark problem solutions have been sub-mitted to the NRC via Reference 1 below'.

Ref.1 - WCAP-8929, " Benchmark Problem Solutions Employed for Verification of the WECAN Computer program", April 1977.

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

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

RESPONSE: The total water hole peaking factor bias to be used in INCA can be calculated from (TURTLE-KEN 0) plus (KENO-experiment) differences which is equivalent to the bias (TURTLE-experiment). By inference, the difference between TURTLE and water hole experiments was calculated using INCA msults. For purposes of licensing TURTLE for use with large water hole lattices, the INCA comparisons described below justify a water hole peaking factor bias of 2.8% to be used in INCA for neasured peaking factors. .

The total water hole peaking factor bias (TURTLE-experiment) was cal-

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culated from a comparison of INCA and TURTLE values of the ratio, hot rod to assembly average power, in cycles i, 2 and 3. This comparison provides the water hole bias because:

I l a) The rctio, hot rod to asserrbly average power, from INCA is the same as the hot pin relative pcwer predicted by the design code (PCQ) used for INCA input in Cycles 1, 2 and 3.

b) The hot pin power always occurs next to a water hole.

c)

The water hole peaking factor bias used in INCA for Cycle 3 is 4.8%(I) which results from extensive comparisons between PDQ and water hole experiments.

d) The total water hole peaking factor bias to be used in INCA with TURTLE input is bounded by correcting the Cycle 3 bias of 4.8% by the bias between TURTLE and PDQ hot pin powers.

Comparisons betweeri TURTLE and INCA peak to average rod power for six INCA maps in cycles 1 and 2 are given in the BSR and results for two Cycle 3 maps are shown in Figures 2 - 5. The results for all 3 cycles are given in Figure 1 and indicate a bias of 2% between TURTLE and PDQ.

The total water hole peaking factor bias to be used in INCA with TURTLE is thus 4.8 - 2.0 = 2.8%.

11. 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 char.ge on power distributions and other physics parameters?

~ RESPONSE: The effects of the nuclear and thermal expansion properties of Inconel-718 grids in Westinghouse - supplied fuel assemblies and Zircaloy-4 grids in CE supplied fuel assemblies were considered in the evaluation of physics parameters (e.g., reactivity coefficients) for the Millstone 2, Cycle 4 care.

(1) CEN-88(N)-NP, Increased Water Hole Peaking in Operating Reactors (Millstone 2), March 30, 1978.

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12. Power distributions calculated by TURTLE appear to underpredict the power in the peripheral asserrblies 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 inaccuracies, justify the use of' TURTLE witnout some type of baffle correction scheme.

RESPONSE: As noted in the BSR, CEA-3 control bank worth on the periphery of the core was underpredicted relative to measurement in Cycles 1, 2 and 3.- This results from a slight underprediction of the power in the peripheral assemblies. The BOL, HZP control worths were calculated again with a baffle correction of [ ]a,c and am shown in Table 1 along with the original results. As seen, the. CEA-3 agreement is improved by the baffle correction. The total control worth, 2-7, remains virtually unchanged as expected. The baffle correction will be used for Cycle 4 design. O F Table 1 Effect of Baffle Correction on Cycle 3 'CEA-3 Worth % Difference

• CEA Worth, %Ap Measured No Corr. Baffle Corr.

7 0.64 0. 31 -1.56 6 0.25 -16.80 -12.00 5 . 0.17 -24.70 -13.53 4 0.88 - 8.64 .10. 80 3 0.67 7.61 3.23 2 , 1.15 0.70 -0.35 Center (7-1) 0.03 -27.90 -21.67 Sequential Worth, '-7 3.76 - 3.21 -3.72

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13. The CEA drcp analysis was performed without automatic rod control (CEA motion inhibit) or turbine load reduction. Is this the operational plan for Cycle 4.

RESP 0tlSE: CEA motion inhibit and turbirie load reduction tend to mitigate the consequences of a CEA drop. Therefore, no credit was taken for these functior.s in the CEA drop analysis. Turbine load reduction will not be operational; but the CEA motion inhibits will be available during Cycle 4.

14. The parameters used in the analysis of the CEA ejection accident appear to be representative of Westinghouse cores and differ from the previous Millstone, l' nit 2 fuel supplier in several areas such as ejected rod worths, ejection time, delayed neutron fraction, feedback reactivity weighting, and power peaking (Fq ). Please provide a comparison between the Cycle 3 and Cycle 4 values of these rod ejectica initial assumptions and discuss the reasons for and effects of any diffentnces.

RESP 0flSE: Comparison of CEA ejection accident parameters are given in Table 1 for Cycles 3 and 4. In Cycle 3, the ejected rod worth is larger than the Cycle 4 value at HFP; at HZP, the values are the same. The difference at HFP is probably due to the assumption used for control bank insertion prior to ejection. As explained in Ref.1, the method used by Westinghouse in Cycle 4 is to assume that the rod is ejected from control bank CEA-7 [ ],(a,c) The delayed neutron fraction used in Cycles 3 and 4 are the same. The Fg after ejection at HFP is slightly higher in Cycle 3 than in Cycle 4. Differences in radial and axial power distribution due to the burnup characteristics and location af the fuel in Cycle 4 account for the di f ference.

1. WCAP-9272, March 1978, " Westinghouse Reload Safety Evaluation Methodology" e

At HZP, the Cycle 4 gF af ter ejection is larger than the Cycle 3 value. Again, sore of the difference can be accounted for by the change in fuel burnup and location between Cycles 3 and 4. Another important contribution is the axial shape assumed as a pre-condition. The Westing-house methodology is to [ ]a,c which accounts for the large value of Fq at HZP. The CEA ejection time was assumed to be 0.1 seconds in Cycle 4. This value has no impact on the results compared to the ejection time of 0.05 seconds used in Cycle 3. The feedback reactivity weighting used in Cycle 4 was applied only to the Doppler feedback, and conservatively accounts for the increased feedback due to the highly peaked power distribution following the CEA ejection. In addition, this weighting factor was applied to a conser-vative prediction of the zero to full power normal operation Doppler power defect of only 0.84% ak. This methodology is described more fully in WCAP-7588 Rev.1-A and in reference safety analysis reports. Although the Cycle 3 analysis employed a spatial Doppler feedback weighting factor, the value used was not _ reported for that cycle. However, values of 1.24 to 1.34 were reported for the full power cases and 1.94 to 2.52 for the zero power cases analyzed for Cycle 1 and reported in the Millstone 2 FSAR. These values are very close to the values used in the Cycle 4 analysis. i a c Table 1 , Comparison of CEA Ejection Accident Parameters P _HlP, HFP_ Cycle 3(I) }[ Cycle 4(2) Cycle 3(I) )[ Cycle 4(3) Parameter 0.29 0.17 . 0.65 0.65 Ejected rod worth, ak 0.47 0.47 0.47 0.47 Del Ayed r,eutron fraction 5.83 5.70 14.5 18.8 F after ejection q .05 .1 .05 . 1 Ejection time 1 .30 (4) 2.50 Feedback reactivity (4) weighting

1. Letter, Counsil to Reid, Millstone Unit No. 2 Power Uprating, Feb. 12, 1979

2. WCAP-9660, BSR, February,1980 l

3. RSE, Millstone Unit 2 Cycle 4, May,1980

4. Value not reported .

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