ML20133C267
| ML20133C267 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 01/03/1997 |
| From: | Huffman W NRC (Affiliation Not Assigned) |
| To: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 9701070098 | |
| Download: ML20133C267 (5) | |
Text
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WASHINGTON, D.C. 20656-0001 January 3, 1997 Mr. Nicholas J. Liparulo, Manager Nuclear Safety and Regulatory Analysis Nuclear and Advanced Technology Division Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, Pennsylvania 15230
SUBJECT:
COMMENTS ON THE REVISED AP600 PRHR HEAT EXCHANGER FINAL TEST REPORT
Dear Mr. Liparulo:
Westinghouse letter NSD-NRC-96-4868, dated November 5, 1996, submitted revision 2 to WCAP-12980, AP600 Passive Residual Heat Removal (PRHR) Heat Exchanger (HX)
Test Final Report. The revision addresses previous Nuclear Regulatory Commission (NRC meeting, summary) dated Februarystaff comments and concerns documented in a J 21, 1996.
The NRC staff has reviewed the revised PRHR HX test report WCAP and has additional questions and comments which are included as an enclosure to this letter.
The staff has also begun to review the draft safety evaluation report open item (DSER) response to open item 21.3.3-1 on the applicability of the straight tube PRHR HX test data to the "C"-tube configuration.
This information was submitted by Westinghouse letter NSD-NRC-96-4660 dated March 6, 1996. The staff had previously informed Westinghouse that it would not review the open item response until it had received and reviewed the revised PRHR HX test report.
Based on a preliminary assessment, the Westinghouse response to open item 21.3.3-1 does not appear to address the staff's major concern as to the performance margin available in the design if critical heat flux (CHF) in the horizontal tube bundle were exceeded.
In numerous discussions with Westinghouse on this issue, the staff has suggested that a sensitivity study be provided demonstrating that even if a substantial portion of the PRHR Heat Exchanger (HX) was deactivated due to CHF, it could still perform its design basis function. During the December 18 and 19, 1996 ACRS, thermal-hydraulic subcommittee meeting on the AP600 test program, Westinghouse implied that such an analysis had been done but had not been submitted to the NRC for review.
If Westinghouse continues to use arguments that have their basis in single tube or single wire data that are of questionable utility when analyzing a bundle of tubes, the staff remains doubtful that the PRHR data can be proven applicable to the C-tube design, especially on the horizontal section.
The staff notes that when the response to open item 21.3.3-1 was prepared, Westinghouse was in the process of revising the PRHR test program report, s
including reanalysis of all of the data.
The staff must conclude that any N(0, insights from the additional analysis or revisions to the heat transfer V
correlations that may have resulted from the reanalysis has no bearing on the open item response.
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NRC FILE CENTER COPY 9701070098 970103 PDR ADOCK 05200003 A
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Mr. Nicholas J. Liparulo January 3, 1997 l
To. resolve this is*ue. Westinghouse should consider providing to the staff an i
appropriate set of sensitivity studies,~using a conservative estimate of the heat transfer from the active portion of the HX andsvarying the deactivated portion (due to CHF) of the HX.over a" range of both ' forced and natural convection conditions. Acceptable results of such a sensitivity study, along with insights from ROSA and.a commitment. (in Chapter 14-of, thetSSAR) to perform testing. in-plant (during power. ascension) to1 verify, the performance capability of the PRHR HX (or, alternatively, a Tier I commitment by Westinghouse to test a prototype heat exchanger before the. firstcplant,.is. built)- would enable the staff to complete its safety evaluation of the PRHR test.-program.
D Many of the enclosed questions and comments'on 'the'PRHR final test report do not merit the issuance of formal RAls. To expedite'the review process, Westinghousa is requested to provide brief written responses to each item.in the enclosed material which can then be used to conduct detailed discussions during a subsequent telecon or meeting. The staff requests that Westinghouse provide the written responses to the encicsed items in advance of any further discussions (except requests for clarifications by Westinghouse).
The staff expects that the enclosed questions and comments will be included in the open item tracking system so-that the status and disposition'of these items can'be tracked.
- If you have any questions regarding this matter, you may contact me at (301) 415-1141.
Sincerely, original signed by:
William C. Huffman, Project Manager Standardization Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No.52-003
Enclosure:
As stated cc w/ enclosure: See next page DISTRIBUTION:
Docket. File PDST R/F PUBLIC TMartin DMatthews
.TRQuay TKenyon BHuffman JSebrosky DJackson JMoore, 0-15 B18 WDean, 0-17 G21 ACRS (11)
TCollins, 0-8 E23 Alevin, 0-8 E23 RLandry, 0-8 E23 GHolahan, 0-8 E2 FEltawila, T-10 E46 DOCUMENT NAME:
A:
PRHR-COM.INL
- See previous concurrence OFFICE PM:DPST:DRPM l SC:SRXB:DSSA l
D:PDST:DRPM
.I NAME WHuffeastT W Alevin
- TRQuay O @
01/g /97 Uh DATE 01 @ 7 ~
01/ /97 To ecosive e ospy of shie document,indiosee in the box: "C" = Copy without enclosures "E" = Copy with enclosures 'N' =
No copy J
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Mr. Nicholas J. Liparulo Docket No.52-003 Westinghouse Electric Corporation AP600 cc: Mr. Nicholas J. Liparulo, Manager Mr. Frank A. Ross Nuclear Safety and Regulatory Analysis U.S. Department of Energy, NE-42 Nuclear and Advanced Technology Division Office of LWR Safety and Technology Westinghouse Electric Corporation 19901 Germantown Road P.O.-Box 355 Germantown, MD 20874 Pittsburgh, PA 15230 Mr. Ronald Simard, Director Mr. B. A. McIntyre Advanced Reactor Program Advanced Plant safety & Licensing Nuclear Energy Institute Westinghouse Electric Corporation 1776 Eye Street, N.W.
Energy Systems Business Unit Suite 300 Box 355 Washington, DC 20006-3706 Pittsburgh, PA 15230 Ms. Lynn Connor Mr. John C. Butler Doc-Search Associates Advanced Plant Safety & Licensing Post Office Box 34
-Westinghouse Electric Corporation Cabin John, MD 20818 Energy Systems Business Unit Box 355 Mr. James E. Quinn, Projects Manager Pittsburgh, PA 15230 LMR and SBWR Programs GE Nuclear Energy Mr. M. D. Beaumont 175 Curtner Avenue, M/C 165 Nuclear and Advanced Technology Division San Jose, CA 95125 Westinghouse Electric Corporation One Montrose Metro Mr. Robert H. Buchholz 11921 Rockville Pike GE Nuclear Energy Suite 350 175 Curtner Avenue, MC-781 Rockville, MD 20852 San Jose, CA 95125 Mr. Sterling Franks Barton Z. Cowan, Esq.
U.S. Department of Energy Eckert Seamans Cherin & Mellott NE-50 600 Grant Street 42nd Floor 19901 Germantown Road Pittsburgh, PA 15219 Germantown, MD 20874 Mr. Ed Rodwell, Manager Mr. S. M. Modro PWR Design Certification Nuclear Systems Analysis Technologies Electric Power Research Institute Lockheed Idaho Technologies Company 3412 Hillview Avenue Post Office Box 1625 Pa,o Alto, CA 94303 Idaho Falls, ID B3415 Mr. Charles Thompson, Nuclear Engineer AP600 Certification NE-50 19901 Germantown Road Germantown, MD 20874
_ _ _ _ _ _.. _ _ _ _ ~
9 l..
1 Staff Coments on PRHR Report The revised report represents a significant improvement over the original version i
submitted to the NRC.
In general, the data appear to have been analyzed i
appropriately; there are, however, some inconsistencies and errors that need to be j
corrected.
The outstanding question of applicability to the current "C"-tube HX design is not addressed to any significant degree in the report. Additional comments i
include:
Technical i
i 1.
It would be helpful for the reader to know which orifice plate was used for each j
test, to better understand how flow data are computed.
4 l
2.
Uncertainties in actual data are appropriate and useful. However, given the importance of the information derh ed from the raw data, estimates of the t
l uncertainties in derived quantities, such as flow rates, heat transfer rates, and j
heat transfer coefficients should be provided.
I 3.
The discussion in Section 8 is somewhat misleading and rather difficult to
)
j follow, in some cases, due to errors in the figures (see below). The data do i
seem to support a conclusion that the tank is well-mixed radially (e.g. Fig. 8-25). Vertical mixing, however, is not as efficient, and the discussion on p. 8-5 seems to focus on long-term behavior (19500 sec). Comparing the top and bottom i
tank elevations overlooks, for instance, the behavior shown in Table 8-8, before the top of the tank saturates. Here (column "D"), while there is only 42*F i
gradient between the 23.917 and 1.917 ft elevation (the reference in the text to the 4.917 elevation as the lowest is not clear), there is almost 95'F between the 1.917 and tube-top (19.917 ft) elevations.
i 4.
The staff understands that the report is limited, for the most part, to a discussion of the test program without any specific consideration of applicability to the current AP600 PRHR HX design (which has been addressed by Westinghouse in RAI responses). Nevertheless, some of the discussion in Section 8 is also misleading in this regard.
For instance, in discussing the plume tests, Section 8-E (p. 8-3) states, "...there should be no interactions between rows of HX tubes exist (sic) when the distance between the rows is 2 feet or more." While this may have been a pertinent observation for the 1989 HX design, it has no relevance whatsoever to the current design.
Similarly, when discussing
" zones of influence" of the plume, the report indicates that there is a minimum of tube-to-tube interaction. Since the tube spacing (center-to-center) u 1.5" within rows of the current HX design and 3" between rows, and the plume measurements show an influence, in some cases, more than 1.5" from the tube centerline, it seems reasonable to believe that there will, in fact, be some degree of tube-to-tube interaction in the current design. Note, too, that the measurements presented in Figs. 8-7 and 8-8 represent plumes between the tank wall and a tube; no data are shown that would correspond to a plume (or interacting plumes) between tubes.
Enclosure
j 2
i Editorial 1
5.
Figure 9-18 does not appear to be a plot of PRHR boiling data as stated, but rather a copy of Fig. 9-3, the single-phase (inside tube) heat transfer j
data.
I 6.
Figure 8-24 is mislabeled and there are other data problems in Section 8.
Looking at Table 8-7 and comparing the data to Fig. 8-24 (column "D") at about 5000 seconds, location "G" is not the hottest, as plotted in the figure; "C"
1 (19.917 ft) is hottest. Also. in column "J" of the tables, the 4.917' elevation is markedly colder than the 1.917' elevation (see Tables 8-8 and 8-9, for instance); it appears that the thermocouple may have a significant bias, which should be noted in the figures or in the text. Even if the curve is relabeled,
)
though, there appears to be a general discrepancy between the trbles and the curves.
)
1 i
7.
Section 4.4.1, p. 4-4, first sentence--the tube length should be 18.16 " feet,*
not " inches."
8.
Section 9: Why is Eq. 9-10 different from the "PRHR Data" equation shown in Fig.
j 9-3?
l 9.
In Fig. 9-1, what is the distinction (if any) between the " triangle," " square,"
and " diamond" data? There is no legend to distinguish between them.
4 10.
In Fig. 9-3, the " dashed line" in the legend for the Dittus-Boelter and Petukhow-Popov correlations loth identical; they also cannot be differentiated in the figure itself.
i 11.
Figures 9-21 and 9-22 do not appear to be labeled properly; the relevance of the curves cannot be determined.
12.
Page 9-7, eqn 9-20, liquid thermal diffusivity isn't used in the equation and j
should not be listed in the nomenclature.
j 13.
Page 9-11, eqn 9-25, right hand side, denominator under tne radical, rho sub r should be rho sub v.
j 14.
Page 9-14, second to last paragraph, " Figures 9-22 and 9-23" should be 9-21 and j
9-22, both times it appears.
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