ML20031D813
| ML20031D813 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 09/07/1976 |
| 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 8110140241 | |
| Download: ML20031D813 (4) | |
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SUBJECT:
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. Plant Name:' g e fy.,.j.-.,.,.,..., Creens County Nuclea-r Power Plan.t y
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Responsible Branch and Project Manager: ; j'.,., T J. J. Curry
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Systems Safety Branch Involved:'
Core Performance Branch Description ~of Review: ' --
Reques.ted Co=pletion Date:
Second Round Questions.and Staff Positions Septecher 3, 1976
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Review Status:
. Request for Additional Information
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The Reactor Puels Section has prepared the attached second roun'd questions on t.
the Creene County PMR.
The Physics Section has no'second round questions for the Greene County PSAR.
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Denwood F. Ross, Assistant Director for Reaqtor Safety
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GREENE COUNTY STAFF POSITIONS 231.29 The response to lat-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 loads.
Please describe the extent to which these design loads have been confirmed experimentally.'
231 30 The fesponse to 1st-round question 231.2 is incomplete with (4.2.1.1) respect to the discussien of fretting, wear, and deflection.
Please cite the current design limits for these phenomena, outline the on-going or planned R&D progra=s 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 31 The response to 1st-round question 231.5 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 5.
231 32 The response to 1st-round question 231 7 requires amplification (4.2.1.2) regarding (1) the " conservative models" said to be used for rod differential growth and grid pressure drep and (2) the out-of-reactor flow tests and measurements which reportedly confirm the calculatiens that show that grid position is well-=aintained throug.:out 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.
231.33 The response to 1st-round question 231.8 does not provide the (4.2.1.2) requested information on dimensions, spring constants, and experifental 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 postulated flux, temperature, and stress conditions, as asserted in the 1st-round response.
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231 34 Th2 r;cpon:o to tst-round qu stion 231 9 dosa not provida tha (4.2.1 3) esqu:sted dacign b22:a for Zircoloy-4 irrtdiation 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 bases.
The response to 1st-round question 231.10 requires 'larifica-231.35 c
(4.2.1 3) tion because of an apparent confusion of terminology.
The resp,onse appears to treat cladding strain and fuel rod de-flection as if they were synonymous.
An intent of 1st-r.ound question 241.10, 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 36 The response to 1st-round question 231.13 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 1st-round question 231.13 231 37 The response to 1st-round question 231.14 requires amplifica-(4.2.1 3) tion regarding the procedure for reducing the recommended power startup rate in the 0-20% power range.
Please quantify this recommended reduction in power startup rate and provide ex-perimental quantitative verification of the effect of reduction in power startup rate on defect propagation, o
231.38 The response to 1st-round question 231.20 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 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 cn ridging, cumulative damage and stress corrosion cracking.
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1 231 39 The response to 1st-round question 231.24 indicated that in (4.2 3 2) experiments where irradiat'ed Al 0 -B C was exposed to high-23 4 temperature high-pressure water, the B C reacted with the 4
water to form H B0. Thus, if the poison rod cladding were 3 3 perforated, the H B03 would be leached into the coolant.
3 Please discuss the potential safety implications of the re-activity insertion resulting frcm the loss of.B-10 from the burnable poison rods by this mechanism. What would be the resultant increase in fuel centerline and cladding tempera-ture if all the B C were removed from (a) one rod and (b) all 4
the poison rods early in life? Provide rate equations for the hydrolysis of B C and rate of loss from perforated-rods, 4
and calculate these rates at (a) reactor coolant temperature and (b) local poison pellet temperature.
231.40 The response to 1sc-round question 231.17 indicated that a (None) value of 85% of the critical buckling load is used for accident conditions.
Please provide the bases, i.e. justification, for the choice of 0.85 for accident analysis. Note that fuel assembly buckling is under review as a generic item as part of the review effort on the mechanical loading effects of the seismic and LOCA analysis. Please provide justification for the fact the fatigue-creep interaction is not considered in the analysis of the fuel under normal and upset conditions, as asserted in the response to question 231 17.
231.41 The treatment of the seismic and LOCA analyses, including the (None) responses-to 1st-round questions 231.21 and 231.26, is inade-quate. 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 Greene County Mark C assembly cannot be completed until the requested report has been received and reviewed.
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