ML19344E499
| ML19344E499 | |
| Person / Time | |
|---|---|
| Issue date: | 08/08/1980 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | Anderson T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 8009020128 | |
| Download: ML19344E499 (4) | |
Text
'
/
'o UNITED STATES 8 Y '),
NUCLEAR REGULATORY COMMISSION g.'.
s WASHINGTON, w. C. 20555 t
o
'g
,o AUG 8 580 Westinghouse Electrical Corporation ATTN: Mr. T. M. Anderson, Manager Nuclear Safety Department P. O. Box 355 Pittsburgh, Pennsylvania 15230
Dear Mr. Anderson:
Subject:
Request Number 2 for Additional Information on WCAP-9500 We are currently reviewing Westinghouse Electric Corporation report WCAP-9500 entitled " Reference Core Report 17 x 17 Optimized Fuel Assembly".
The initial review of Section 4.4 reveals the need for the additional inform-ation indicated in the enclosure.
This information is necessary to complete the review - its expeditious sub-mittal will therefore be to the Westinghouse advantage. Please advise us as soon as possible of your planned submittal date to permit us, in turn, to develop a review schedule.
Sincerely,
/m,c7"2,!sG2 [
+y
, [ ' Standardization and Special James R. Miller, Chief j
/
Projects Branch Division of Licensing t
Enclosure:
As stated cc: Mr. Alex Ball Pastinghouse Electrical Corp.
Nuclear Safety Department P. O. Box 355 Pittsburgh, Pennsylvania 15230
&OO9020 \\ N
n.
'231-1 Question Set 1 Section 4.4 WCAP-9500, " Reference Core Report 17x17 Optimized Fuel Assembly"_
231.0 CORE PERFORMANCE BRANCH 231.1 Section 4.4.2.1 stated that "... the grid design has changed from the (4.4.2.1) standard inconel R-grid design to a Zircalcy design." Besides the different material used for the six grids in the high flux region (Zircaloy), is there any other difference between the new grid and the R-grid or the L-grid? Provide more information (in addition to that in
~
Section 4.2.2.2.4) on the difference between the new and the old grids including values of the grid pressure loss coefficients. Also provide the OFA inlet and exit loss coefficients.
231.2 Provide a comparison of the average fuel centerline temperature for (4.4.1.2) the 0FA vs Byron and Braidwood units.
231.3 Provide a discussion of the changes in the values of the sensitivity introduced by using the WRB-1 cori' elation instead of the (4.4.2.9) factors St W-3 correlation for DNB heat flux.
231.4 Provide a list of all the parameters and rcnge of values treated in a (4.4.2.9) statistical manner using the Improved Thermal Design Procedure for the DNBR limit of the OFA.
231.5 Provide discussions of the method used in applying the statistical (4.4) method to the minimum DNBRs calculated for the nominal and design transient conditions as shown in Table 4.4-1.
4 231.6 Explain the following differences
~
(4.4.1.3) non-UHI Plants _
UHI plants From Section 4.4.1.3 143.5 X.942 143.3 X.925 Flowrate for effective
= 135.2
- 132.6 6 lb fuel rod ccoling,10 m/hr From Table 4.4-1, 173.3 134.7 heat transfer, 106 lbm/ r h
?
f"
~
~.,
231-2 231.7 Describe the basis and magnitude of DNBR rod bow penalty to be (4.4.2.2) applied to the 0FA.
Was the transient analysis code uncertainty (Tical method?+ 1% of LOFTRAN) 231.8 included in the DNBR analyses with the statis (4.4.2.9.5) 231.9 Provide a discussion of the applicability of the statistical (4.4.4.5.4) method described in WCAP-8567 to part-loop operation.
231.10 For both UHI and non-UHI plants, provide bases for the difference (4.4.2.9.6) in the thermal design flow and best-estimate loop flow (2.5% for non-UHI plants and 4% for UHI plants).
231.11 Fractions of the thermal design flow that is allotted as bypass (4.4.2.9.6) flow are inconsistent in Sections 4.4.1.3 and 4.4.2.9.6.
Provide (4.4.1.3) the bases for the assumptions and make it consistent.
231.12 Quantify the "... excellant heat transfer..." and "... the film (4.4.2.9.1) temperature drop..." in the cladding temperature calculation un-certainty described in Section 4.4.2.9.1.
231.13 Quantify the "... conservatively high values of the nuclear peak-(4.4.2.9.5) ing factors..." used in the THINC-IV analysis for DNBRs.
instability, provide discussions on (6AP/6G) external For Ledinegg/6G) internal >o.Is a designed feature or results of 231.14 0 and (6AP (4.4.4.6) tests applicable to conditions I and II events operational ranges?
1 231.15 for dynamic stability, justify the statement "an open channel con-(4.4.4.6) figuration is more stable than the closed channel analysis imder the same boundary conditions." Explain the basis for extrapolating from the previous tests (Reference 76, 77 from Section 4.4) to the current 0FA design.
Provide Reference 72 from Section 4.4.
231.16 For the distributions shown in Figs. 4.4-4 to 4.4-6. indicate the 4
axial step size, amount of axial cross flow (assembly-to-assembly),
inlet flow distribution (velocity of mass flux), pressure drop (along the elements and in the grid spacer), mixing coefficient used, and axial void fraction.
,,,.o_a a
r
-we T'
231-3 231.17 How accurate are the results of the 1/7th scale model pressure drops when compared to actual operating plants (p. 4.4-7, para.
2)? How significant is the pressure drop across the upper inter-nals in the vessel (with respect to total core losses) and how accurate are theoretical predictions in this region?
231.18 Since no hot channel allowances are included in the design for quadrant power tilts, what is the worst case (e.g., a dropped or misaligned RCCA) hot channel factor developed and the re-sulting increase in coolant temperature or decrease in DNBR (p. 4.4-23, para. 2)?
231.19 Verify that the Plant A and Plant B maximum bypass flows are interchanged (p. 4.4-4, para. 4).
231.20 How is w' determined (p. 4.4-9, eq. 4.4-3)? How does TDC vary with spacing distance (16, 20, 26, and 32 in.)? What is TDC for natural turbulence? How sensitive are transverse coolant
,I temperature differences to TDC? How does boiling during transients affect TDC and, thereby, transverse coolant temperature differences?
How is the large difference (0.038 to 0.059) in TDCs physically ex-plained (p. 4.4-10, para. 3)?
231.21 If core hydraulic loads are twice fuel assembly weight during pump overspeed transients that produce 20". flow over design (p. 4.4-15, para.1), how great are the hydraulic loads at design flow? What load are the hold down springs designed to withstand?
The DNBR values for thimble and typical cells for plant safety analyses 231.22 are taken at 1.82 and 1.85 (Plant A) and 1.47 and 1.49 (Plant B),
respectively (p. 4.4-3, para.1, white and blue).
Explain this signi-ficant difference and provide the process by which these DNBR values were obtained.
Is Reference 15 the proper one to quote here (p. 4.4-8, para. 3)?
231.23 1
s
- - ~. - ~ -
Y