ML20031D814

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Forwards Reactor Fuels Section & Reactor Physics Section Second Round Questions & Portions on BSAR-205.Util Must Discuss How Deflection Is Accounted for in Summation of Stresses in Fuel Assembly
ML20031D814
Person / Time
Site: 05000561
Issue date: 09/01/1976
From: Ross D
Office of Nuclear Reactor Regulation
To: Vassallo D
Office of Nuclear Reactor Regulation
Shared Package
ML20031D800 List:
References
FOIA-81-236 NUDOCS 8110140245
Download: ML20031D814 (5)
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?, 9, D.L.SECOND ROUND. QUESTIONS FOR ESAR-205...,,. '. '.

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f; Second Round' Questions and Staff' Positions

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, Request for Additional Information Review Status:"

The Reactor Fuels Section~and the Reactor Physics Section of the Core

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staff positions on BSAR-205.

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B-SAR-205 STAFF POSITIONS 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 js 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 231.4 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.

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The response to 1st-round question 231.5 requires amplification

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( 4. 2. 's. 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 observations 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.

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231.35 Thi r&cp:ns3 to 1Et-round qu;stion 241.7 do2a n:t provida tha

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(4.2.1.3) r:questsd dssign b:cca for Zircaloy-4 irrrdiction gr:wth.

Design " bases" 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 base,s.

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.

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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-=anufactured 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 1st-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 informction, please outline how the results of specific portions of the Mark C

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fuel asseEbly development program will be used to provide the inforcation requested in 1st-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 0-20% 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 15 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 mechanica.. 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.

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231.40 Th2 r:rpons2 to 1ct-r und quzstion 231.21 indicatsd th:t in (4.2.3.2) experiments where irradiated A1 0 -8 C was enposed to high-23 4 f

te:perature high-pressure water, the B C reacted sith 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

re=oved from (a) one rod and (b) all the poison rods early in life.

Provide rate equations for the hydrolysis of B C 4

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 lat-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 bowing 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 d,ata;

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o (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 assembl}

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.

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232.26 The response to Question 232.17 is iuadequate. Please (15.1.2) identify the 205 FA plant for which the analysis was perforn.ed.

232.27 Ihe response to question 232.11 (as presented in the, response to Question 2.2.71) implies thst 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!

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' EVALUATION OF CRYSTA!! RIVER, UNIT 3 REPLACIMEh"r FUEL ASSDOLY 3A33 '

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analyses of the subject replacement fuel assenbly.

'A copy of tbn evaluation is enclosed and is provided as input to SER Section 4.2.1,

" Fuel, Mechanien1 DesiSn."

Ralph 0. Meyer, Section Leader Reactor Fuels Section Core Performance Branch

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Evaluati*n of Crystrl Riv r, Unit 3 R;piccimtnt Fu,1 Asntmbly 3A33 Backgrcund*

Damage incurred to Crystal River, Unit 3 fuel assembly 3A33 during handling necessitated replacement with a new assembly. The replacement

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3A33, assembly is made up of 40 fuel rods removed from the original assembly and 168 replacement fuel rods.

On March 26, 1976, Babcock and Wilcox submitted a report (Reference 1), which described the replacement fuel assembly along with analyses performed by B&W on the potential for cladding creep collapse, fuel / clad interaction, fuel densification and fuel swelling.

In response to requests by the NRC staff for additional information, this report was supplemented by a subsequent submittal on May 4, 1976 (Reference 2).

Description of Replace =ent Fuel Rods The only sigsificant differences between the 168 replacement fuel rods in replacement assembly 3A33 and the remaining fuel rods in' the first cycle loading of Crystal River, Unit 3 relate to the internal spacers and the fuel pellets.

Spring spacers and Zitesloy tubular spacers replaced the corrugated tube spacers and zirconia ceramic spacers in the original rods.

The newer types of spacers are r epresentative of those currently in operation in several B&W reactors and thus are not of concern. The fuel pellets, however, differ slightly in enrichment, density, and active length, as shown in Table I.

Of particular interest to the Regulatory staff were the potential effects of the low-density, 90.9% theoretical density (TD), pellets on thermal and mechanical performance.

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Summary cf RiguletrJry Ev71untien Analyses of the thermal-hydraulic performance of the replacement fuel rods in replacement assembly 3A33 were performed by B&W and reperted in the March 26, 1976, letter (Reference 1).

A comparison of the results of these analyses to analyses of the 40 original fuel rods in assembly 3A33 and the fuel rods in the remaining 176 fuel assemblies is shown in Table II.

The results of the analyses show that, except for the engineer-ing hot channel factor, all thermal-hydraulic performance parameters for the replacement fuel rods in fuel assembly 3A33 are not more restrictive than for the fuel rods in the limiting fuel assembly in the remaining 176 fuel assemblies. The effects of the higher engineering hot channel factor for fuel assembly 3A33 were not addressed under " Fuel, Mechanical Design," but vare, treated under " Nuclear."

Cladding creep collapse analyses were performed by B&W in accordance with material properties and design procedures set forth in Topical Report B&W-10084P-A, entitled " Program to Determine In-Reactor Performance of B&W Fuels" (Reference 3).

The evaluation was completed using the NRC -

approved CROV creep ovalization analysis code described in Section 3 of

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the cited report.

In addition other conservations were introduced, as described in reference 2.

Results of the analyses indicated a collapse time > 14,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, compared to the required 10,320 hours0.0037 days <br />0.0889 hours <br />5.291005e-4 weeks <br />1.2176e-4 months <br /> associated with the single cycle burn of asse.nbly 3A33.

Pellet / cladding mechanical interaction (PCMI) and fuel swelling effects were addressed by B&W in their cladding strain analysis.

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Of th2 pallst dsncitics uccd in assembly 3A33, ch3 90.9% T.D. p211sts represent the limiting PCMI case at the peak pellet burnup seen by the assembly. Accordingly, cladding strain analyses were performed on the 90.9% T.D. fuel corresponding to the worst-case specification dimensions and the as-built, 2e dimensions. The analyses were performed in accordance, with material data and design models set forth in Section 3 of Topical Report B&W-10054, Revision 2, entitled " Fuel Densification Report" (Reference 4).

This represented the same approach as used in the Crystal River SAR except that additional conservations were introduced for the 3A3; analyses as listed in refereces 2, p. 2.

The results of the analyses indicated that the total circumferential strain resulting from PCMI for the worst-case.-speciication analysis and the as-built dimensions analysis were 0.80% and 0.48%, respectively, as compared to the EiW " design" value of 1.42%.

In summary, the mechanical design and tnermal analysis aspects of the 168 replacement fuel rods in Crystal River, Unit 3 replacement fuel assembly 3A33 have been analyzed by B&W, using NRC-approved codes and methods

'h and in accordance with material data and design models approved by NRC.

Evaluations of the potential for cladding creep collapse, pellet / cladding mechanical interaction, fuel densification and fuel swelling were made, dC The results of these analyses have shown that the 3A33 fuel rods are within,

acceptable design limits for first cycle operation of Crys; <1 River, Unit 3.

The information provided on the results of the analyses of the 3A33 replace-

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ment fuel assehbly provides an acceptable basia for demonstrating their adequacy.

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u Riforcnce')

1.. J. T. P.odgers. Asst. V.P., B&W, to D. A. Butler, Chief, LWR Research #4, " Report on Replace =ent Assembly 3A33," March 26, 1976.

2.

J. T. Rodgers to D. A. Butler, " Supplement to Report on Replacement Assembly 3A33," May 4, 1976.

A. F. J. Ick2rt,_e_t,al, " Program to Determine In-Reactor Performance 3.

t of B&W Fueld," B&W-10084P-A, Nov. 1974.

4.

R. A. Turner, " Fuel Densification Report," B&W-10054, Revision 2, May, 1973.

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COMPARISON OF FUEL PARAMETERS NO.

FUEL ASSEMBLY FUEL RODS ENRICHMENT

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3A33 40 1.93 92.5 144 s

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[1.98 95.35 23-1/2 (Upper Zone) 1.94 90.9 95-3/4 (Central Zone) >142-3/4 1.98 95.35 23-1/2 (Lower Zone)

Remaining Assemblies 208 1.93 92.5 144 in Batch 1 e

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FUEL ASSEMBLY IN REHAINING 176 FUEL THERMAL-HYDRAULIC CRITERA 3A33 ASSEMBLIES 1.

Linear Heat Rate Limit Based on D

Central Fuel Helting, KW/Ft.

z For Fuel Density:

a. 95.35% TD 21.46
b. '90.9% TD 19.96 19.7
c. 92.5% TD 2.

Average Linear Heat Rate, KW/Ft.

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Average Fuel Tempeintures (Stored Energy), F

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(1) 95.35% TD 1285 (2) 90.9% TD 1327

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(1) 95.35% TD 2840 (2) 90.9% TD 3066 (3) 92.5% TD' 3110-l 4.

Engineering Hot Channel Factor 1.026 1.014 5.

D::LR Penalty Due to Fuel Densification,%* 1.9 2.9

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