ML20040A061
| ML20040A061 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 09/07/1976 |
| From: | Cox T Office of Nuclear Reactor Regulation |
| To: | Happell J BABCOCK & WILCOX CO. |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201200284 | |
| Download: ML20040A061 (5) | |
Text
{{#Wiki_filter:- -.4 U.S. NUCLEAP, REGULATORY COMMISSI0f1 WASHIllGTON, D. C. 20555 301-492-7000 Telecopier---------------------492-7617 ( Automati c) '~ Ve ri f i ca ti on flumber---------------------492-7 3 71 Date Telecopier Number Verification Number No. of Pages (If automatic) (Excluding ^ " * * * * '.'-2d 3EY 7773 9/7 7/. I aanuai fo1~_Wf57// ~ coversheet) 4 T0: Name State, City & Company Phone Yi#fc/ Alt 4,ShkU60%,bkY N FROM: riame Organiza ion / Location Phone Mail Stop /OW PAv/ % /38 A7243 /27 / OVC-U2-;i^MH b DV6PN /. b N*I[h.1;;';T: AUO1V 8201200284 810403 PDR FOIA MADDEN 80-515 PDR
.. a 7EF k hnee2L 80f-374-7773 /ff H ,2,36,O CORE MAWW pgy al3 /. o fea.che-fia), ~ m 231 30 The response to 1st-round question 231.1 is incomplete because (4.2.1.1) it does not provide the requested information on experimental confirmation of the fuel handling and shipping design loads. Please describe the extent to which these design loads have been confirmed experimentally. 231 31 The response to lat-round question 231.1 is incomplete with (4.2.1.1) respect to the discussion of fretting, wear, and deflection. Please cite the current design limits for these phenomena, outline the on-going or planned R&D programs which should yield confirmatory information on the specific design limits, and present fall-back positions. Discuss how deflection is accounted for in the summation of stresses in the fuel assembly (as suggested in the response to question 231.2). 231.32 The response to 1st-round question 231.8 lacks detail. Please (4.2.1.2) describe how the specified coolant temperature limits and associated cladding loading are used in the fuel rod fatigue analysis. Show by means of an example how the coolant tem-perature limits and associated cladding loading are used to " identify the conservative conditions for input to the stress analysis," as asserted in the response to question 231.4. 231.33 The response to 1st-round question 231.5 requires amplification-(4.2.1.2) regarding (1) the " conservative models" said to be used for rod differential growth and grid pressure drop and (2) the out-of-reactor flow tests and measurements which reportedly confirm the calculations that show that grid position is well-maintained throughout life. Please show in greater detail how these calculations and experiments provide support for the conclusion that the frictional force on the fuel rods is sufficient to maintain grid position throughout life. 231.34 The response to 1st-round question 231.6 does not provide the (4.2.1.2) requested information on dimensions, spring constants, and experimental obaervations of the upper and lower plenum springs. Please provide the requested information and, in addition, show quantitatively that the resistance to creep and relaxation of age-hardened A-286 alloy is sufficient to withstand the worst postuated flux, temperature, and stress conditions, as asserted in the 1st-round response. m 1
i N, 231.35 The response to lat-round question 241.7 does not provide the (4.2.1.3) requested design bases for Zircaloy-4 irradiation growth. Design "boses" are not synonymous with " values," as appears to be implied by the response. Please provide the design bases as requested, and briefly outline the data which sup-port these bases. 231.36 The response to 1st-round question 241.8 requires clarifica-(4.2.1.3) tion because of an apparent confusion of terminology. The response appears to treat cladding strain and fuel rod de-flection as if they were synonymous. An intent of 1st-round question 241.8, however, was to establish the displacement limit of B&W fuel rods from a rod bowing viewpoint. Such a displacement limitation, when used in fuel design, should reflect a DNB correlation and power peaking factor calcula-tion. Provide the as-manufactured displacement limitation as well as the one imposed during operation. Discuss how one confirms that these limitations are not exceeded. 231.37 The response to lat-round question 231.11 does not provide the (None) requested information on the currently used' stiffness limita-tions on the spacer grid assembly and individual grid springs. In addition to providing this requested information, please outline how the results of specific portions of the Mark C fuel assembly development program will be used to provide the information requested in lat-round question 231.11. 231.38 The response to 1st-round question 231.12 requires amplifica-(4.2.1.3) tion regarding the procedure for limiting the recommended power startup rate in the 0-205 power range. Please quantify this recommended limit in power startup rate and provide ex-perimental quantitative verification of the effect of reduction in power startup rate on defect propagation. 231.39 The response to 1st-round question 231.18 addresses the 1% strain (4.2.1.3) limit which is based on average cladding strain. The R-2 re-actor power ramp tests, referred to in the response, were, however, performed on low exposure rods which were still ductile and, therefore, only demonstrated the ability of the rods.to withstand pure mechanical loading. Describe any research pro-grams on analytical modeling development currently in progress or planned to evaluate the effects of local cladding strain due to pellet cracking on ridging, cumulative damage, and stress corrosion cracking. /
y j 231.40 The response to 1st-round question 231.21 indicated that in (4.2.3.2) experiments where irradiated Al 0 -B C was exposed to high-23 4 temperature high-pressure water, the B C reacted with the 4 water to form H3B0. Thus, if the poison rod cladding were 3 perforated, the H B03 would be leached into the coolant. 3 Please discuss the potential safety implications of the re-activity insertion resulting from the loss of B-10 from the burnable poison rods by this mechanism. Describe the re-activity anomaly that would result if all the B C were 4 removed from (a) one rod and (b) all the poison rods early in life. Provide rate equations for the hydrolysis of B C4 and rate of loss from perforated rods, and calculate these rates at (a) reactor coolant temperature and.(b) local poison pellet temperature. 231.41 The response to 1st-round question 231.17 on fuel rod bowing (4.2.1.3) refers to examination measurements on the Oconee 1 Mark B (15x15) assemblies which will be used as a basis for pre-dicting bowing in the Mark C (17x17) assemblies. Please discuss how the bcWing data from 15x15 Mark B assemblies will used for 17x17 Mark C bowing predictions; i.e. how will 15x15 Mark B assembly data be related and applied to the 17x17 fuel? Also provide the following information: (a) Status of the 15x15 rod bowing data collection; (b) Schedule and scope of the 15x15 examination program; (c) Manner by which the 15x15 data and analysis will be reported to NRC, and approximate date-for submittal of a topical report; (d) Plans for obtaining 17x17 fuel assembly bowing data; (e) Out-of-pile (if any) mechanical experiments which will provide input to a mechanistic bowing model. 231.42 The treatment of the seismic and LOCA analyses for the Mark C (None) (17x17) fuel assembly is inadequate. An in-depth safety analysis of the seismic and LOCA response of the Mark C (17x17) fuel assembly has been requested (letter, Ross to Schwencer, July 25, 1974) and a commitment to submittal of a topical report in early 1976 (at least one year prior to the filing of the first FSAR incorporating the Mark C fuel assembly) was made by B&W (letter, Malley to Schwencer, September 3, 1974). Our evaluation of the B&W seismic and LOCA analyses for the Mark C assembly cannot be completed until the requested report has been received. fY
'A 4Q ,. 23a. O MeecM V 232.26 The response to Question 232.17 is inadequate. Please (15.1.2) identify the 205 FA plant for which the analysis was performed. 232.27 The response to question 232.11 (as presented in the response to Question 212.71) implies that power shapes with "large" negative offsets were used in the deriva-tion of the power range scram reactivity curve. Please confirm and indicate the range of negative offsets con-sidered. In particular was consideration given to a ~ scram while in the recovery from a load following ~ transient? =9 A}}