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t MEMORANDUM FOR:
T. H. Novak, Chief, Reactor Systems Branch, DSS f
THRU:
P. S. Check, Chief Core Performance Branch, DSS b
FROM:
R. O. Meyer, Leader, Reactor Fuels Section, CP3, DSS
SUBJECT:
FUEL DAMAGE ACCEPTANCE CRITERIA FOR B-SAR-205 MAIN LINE STEAM BREAK ANALYSIS f
This memorandum is in response to the March 17, 1977 memorandum from G. Mazetis to P. Check on the title subject. Recapitulating the main points of Mazetis' memorandum, the B-SAR-205 steam line break analysis shows DNBR <1.3 for a small percentage of rods (<1%).
Babcock and j
Wilcox states in the PSAR that even though the DNBR drops below 1.30 for the hot spot of the core during the steam line break accident, no
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fuel cladding failures are expected. This conclusion is based on the contention that there is insignificant zirconium-water reaction in terms l
of the amount of wall thickness reduction that occurs as a result of the reaction; i.e., 0.00035 inch out of a total wall thickness of 0.0235 inch.
I Mazetis also points out that although fuel damage during a main steam line break is acceptable, providing that the radiological doses do not exceed 10 CFR 100 limits, B&W is using the analysis to bound several excessive heat removal events of moderate frequency (SRP Sections 15.1.1 to 15.1.4).
Lastly, Mazetis requests that CPB confirm B&W's contention
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that no fuel perforations occur for this " temperature excursion."
i Accordingly, we have examined B-SAR-205 Section 15.1.14.2.4.1 "Results of f
28-Inch Steam Line Break Case," Section 15.1.14.2.4.2 "Results of 42-Inch l
Steam Line Cases," and appropriate sections of the SRP. In brief, B&W's
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contention that no claiding failures will occur is, for the present, j
unacceptable for two main reasons.
The Zr-H O reaction is only one of several potential fuel (1) 2 failure mechanisms of concern in steam line break events.
Others include pellet / cladding interactions (PCI), creep-collapse, ballooning, adverse hydride precipitation, etc.,
none of which are addressed in the steam line break analyses in the PSAR. Note that the PSAR plots of peak cladding temperature (PCT) versus time for the 42-inch line rupture f
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Contact:
M. Tokar, NRR, 27603 8201210416 810403 PDR FOIA MADDEN 80-515 PDR
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T. H. Novak -
(with offsite power) case and 28-inch rupture (with loss of offsite power) case (Figs. 15.1.14-3 and 15.1.34-4b, respectively) show that while the maxi-mum temperature (N1170 and ll60*F) is reached ir.
4 and 12 seconds, respectively, the temperatures then drop off rather gradually; at 10 seconds (the maximum time shown for the analysis of the 42-inch break) the PCT is still N1100*F, and at 24 seconds, (the cut-off for the analysis of the 28-inch break) the PCT is s1150*F. Thus, we have no information on the temperature-time predictions beyond a few seconds for these analyses, nor do we have any analyses of the potential effects of these predicted thermal responses on the mechanical performance of the affected rods.
(2) Acceptance Criterion 2b of SRP Section 15.1.5 " Spectrum of Steam System Piping Failures..." indicates that "if the DNBR falls below these values," (e.g., 1.30 or 1.32, as appropriate), " fuel damage (rod perforation) must be 1
assumed unless it can be shown, based on an acceptable
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fuel damage model, that' fuel failure has not occurred."
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Legalistically, therefore, from the standpoint of literal interpretation of the wording of this acceptance criterion, B&W has made a legitimate attempt to meet the criterion by proposing a fuel damage model.in place of the one for minimum DNBR. We have had a preliminary indication from F. Schroeder, however, that even if we believed that B&W's position on this was a,cceptable, i
our evaluation would have to undergo review by the
" Regulatory Requirements Review" committee. This would, at the very least, have an undesirable effec.t on the B-SAR-205 review schedule.
In summary, we find that B&W's contention that no fuel failures will occur for the steam line break, because of insufficient Zr-H O reaction, to be 2
unacceptable. There are other potential fuel failure mechanisms, which i
B&W has not addressed. The PCT-versus-time curves must be provided for longer times, and the potential for PCI, creep collapse, ballooning, etc.,
must be analyzed before the DNBR <l.3 failure criterion could be super-ceded by another model.
B&W should be informed that we are philosophically receptive to the idea of " improved" fuel damage criteria, but that we must retain the current criteria (e.g. DNBR <l.3, PCI failure, etc.) until we
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T s
T. H. Novak.
are provided a technical basis, via analyses, provided in reports and supported by technical data, for changing these criteria. The March 11, 1977 Ross letter on fuel damage criteria should again be brought to B&W's attention. This letter invited B&W's participation in our efforts to study the mechanisms of fuel failure, to identify the probable failure mechanism for each event analyzed in Chapter 15, and to establish more realistic fuel failure criteria. The main steam line break analysis is one exemplar case of direct application and benefit from this study.
jib 6fr o R. O. Meyer, Leader Reactor Fuels Section Core Performance Branch T
Division of Systems Safety cc':
R. Lobel F. Coffman Fuels Section
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G. Mazetis k
D. Ross T. Cox S. Newberry d
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