ML20031D817
| ML20031D817 | |
| Person / Time | |
|---|---|
| Site: | Green County  |
| Issue date: | 03/31/1976 |
| From: | Meyer R Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20031D800 | List: |
| References | |
| FOIA-81-236 NUDOCS 8110140250 | |
| Download: ML20031D817 (7) | |
Text
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0' W3 1 197g R. C. DeYoung,' As'sistant Director for Light Water Reactors DPM TERU:
P. Cheek, Chief, Core Performance Branch, DSS GREENE COUNTT - FIRST ROUND QUESTIO.5S Plant Name:
Docket Number:
Creene' County Nuclear Power Plant 50-549 Licensing Stage:
'PSAR Milestone Nunber:
05-24 Responsible Branch LVR-5 and Project Manat;er J. J. Curry Systens Safety. Bruh Involved:
Description of Review:
Core Perfornance Branch Initial Questions Requested Coc$1etion Date:
March 22, ir 6 Review Status:
Request for odditional Information The Eeactor Fuels section has prepared the attached first round questions on the Creene County PSAR.
l R. O. Meyer, Section L;oder Core Perfornance Branch Division of Systens fafety
Enclosure:
Office of Nucient Reactor Regulation First Round Qeustions cc:
S. Hanauer F. Schroeder Distribution:
R. Boyd Docket File D. Vassallo NRR Reading File i
J. Curry CPB Reading File D. F. Ross R. O. Meyer P. Check W. licDonald i
H. Tokar r
CFB: DSS CPB: DSS CPB: DSS HTokar:gm ROMeyer PCheck or
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- 8110140250 810806
ig3ROWN81-236 PDR
230-1 230.'O CORE PERFORMANCE 231.0 Fuels 231.1 The fuel handling and shipping design loads are provided.
State (4. 2.1.1.1) -
the bases for these design loads.
Describe the' extent that these design leads have been confirmed experimentally. State whether the relationship between these design loads and stress-strain limits is such that no design limits are exceeded during handling and shipping.
231.2 Section 4.2.1. (Design Bases) of Regulatory Guide 1.70 recommends (4. 2.1.1. 2) consideration of the following in the design bases of the fuel:
(1) the physical properties of the cladding and the effects of design temperature and irradiation on the properties, (2) stress-strain limits, (3) the effects of fuel swelling, (4) variations of melting point and fuel conductivity with burnup, and (5) the requirements for surveillance and testing of irradiated,
fuel rods.
Provide more discussion in the Fuel Rod Design Bases Section of the PSAR to describe how the above items are considered.
231.3 Provide the numerical values used for the Zircaloy cladding yield (4. 2.1.1. 2) strength and ultimate tensile strength mentioned in conjunction with the stress intensity limits.
In addition, state the cladding thermo-mechanical history and associated temperature and fast neutron flux and fluence for which the stress limits apply.
231.4 Provide a list of the conservative estimates made in the fatigue (4. 2.1.1. 2) calculations.
231.5 The relationship between compressive load on the clad and reactor ' *
(4. 2.1.1. 2) coolant temperatures is discussed in Section 4.2.1.3.2 in regard to hydride precipitation. Describe how the specified coolant
' temperature limits and associated clad loadings are used in design basis calculations of clad stresses.
Describe independent checks made at the completion of fuel loading to verify the loc ~ation and orientation of the fuel in the core.
t 231.6 Provide a drawing that shows the details of the spacer grid at the (4.2.1.2) instrument tube location. Discuss how the spacer sleeves restrict the movement of the spacer grids.
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e 231.7 Provida tha dcfisctisn dzsign spscifications (including dincnziens (4.2.1.2.2) and spring constants) and experimental observations for the upper and lower plenum springs. Describe any evidence of permanent deflection due to fuel rod and fuel assembly handling.
State whether gradual deflection of the lower spring would be expected as a function of irradiation and state the maximum deflection, both expected and possible. Discuss the QC procedures which assure that the proper type of spring is in the lower ~ plenum.
231.8 Provide the design bases for Zircaloy-4 irradiation growth and (4. 2.1. 3. 2 )
supply supporting data or references.
231.9 List fuel rod deflections and cladding strain limits and provide (4. 2.1. 3. 2) justification for their adequacy.
Demonstrate how these criteria are satisfied during steady state, transient and accident analyses..
Your responses should provide a comprehensive network for an overview of your design basis in this area.
Include in the discussion a summary of the safety analysis in the form of a stress report for each component under specified. loadings.
Discuss how the different loading categories are conbined to satisfy the design limit for each component of the fuel assembly.
With respect to fuel rod and assembly behavior, discuss the design loading limit for each component of the fuel assembly.
231.10 Provide tables of numerical values (or equations) of material (4. 2.1. 3. 2) properties of both cladding and fuel pellets as functions of te=peratures and irradiation.
The following properties should be included:
(1) Modulus of elasticity (2) Poisson's ratio (3) Thermal expansion coefficient (4) Yield stress (5) Ultimate stress (6) Uniform ultimate strain (7) Thermal conductivity (8)
Specific heat
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231.11 Describe procedures used for sizing the fuel rod plenum, including (4. 2.1. 3. 2) any computer codes used and the fission gas release rate assumed.
State whether this volune is adequate for accidents and transients in which the fuel might reach a temperature in excess of the design temperature. Also, descr!be how creep effects and dimensional g
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230-3
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231.'11 (4.2.1.3.2) stability are accounted for in designing the fuel rod plenum.
In a transient, stat: whether it is possible for the cladding temperature t; become so high that clad swelling will occur due to internal pressurn.
State the end of life internal pressure for these fuel rods, (both average burnup and peak burnup).
State the temperatures assigned to each of the following void regions when determining the fuel rod pressure:
(a) fuel rod upper end plenum (b) fuel-clad annulus (c) fuel pellet end dishes (d) fuel pellet open porosity 231.12 Address the following items in the flow-induced vibration program:
(1) natural frequency limitation of the fuel assembly, (2) natural frequency relative to primary system frequency, and (3) stiffness limitations on the spacer grid assembly and indi-vidual grid spring.
231.13 Describe the corrective actions indicated in the last sentence of the paragraph titled, " Potential for Water Logging Rupture."
231.14' Discuss all procedures used during ' fabrication to assure that no axini gaps are introduced during the loading of the fuel rods, such as weighing, counting pellets, fluoroscopic examination, etc.
231.15 Internal and external surfaces of the Zircaloy tubing are cleaned (4. 2.1. 3. 2) with dry cotton swabs and acetone saturated cloth, respectively.
Discuss the effect of residual lint on the internal surfaces upon the fuel performance, specify acceptable limits, analyzed results and ultimate chemical disposition.
231.16 Give safety factors applied in the fatigue design, creep rupture, fatigue creep interaction and instability (buckling) analysis for the fuel assemblies.
231.17 Provide steady state, transient and accident responses for the
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guide tubes in terms of stress, strain, dimensional stability, and deflection.
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230-4 Evaluate the effects of fu21 rod bowing togsthtr with spresr 231.18 (4.2.1.3.2) grid response.
Include time dependent behavior due to creep in your evaluation.
Discuss the an&lytical calculation of fuel clad mechanical inter-231.19 action that is used to describe the design bases. A detailed, (4. 2.1. 3. 2) complete description is needed, including a general description,
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assumptions, mathematical equations, sequence of application of equations or a flow chart, all empirical constants used in theIn-equations, a sample calculation and a comparison with data.
clude the effects of:
(1) fuel swelling driven cladding strain (specify fuel swelling assumptions used),
(2) "bambooing" of the cladding due to fuel pellet end effects, and (3) radial and/or axial differential thermal expansion of the fuel -nd cladding.
Explain how a transient in which fuel rod power was increased would affect the fuel-cladding mechanical interaction and how this is taken into account in the model described above. Discuss operating design limits (transient conditions) or total lifetime limits based on calculated fuel clad mechanical interaction effects.
Discuss the fuel assembly seismic model and analysis method.
In 231.20 (4.2.1.3.2) particular, describe a method of obtaining detailed stress and deformation of the fuel rod from the simple spring-mass beam mode response.
Discuss in detail the planned surveillance of irradiated fuel 231.21 (4.2.1.3.4) rods.
(See letter, Ross to Suhrke, March 10, 1976.)
231.22 Identify all welded joints in the assembly and categorize by type and importance for safety. Describe both destructive and non-destruative weld testing, e.g., localized corrosion, metallo-graphic exu aination and dimensional inspections, indicating what constitutes an acceptable result.
State whether x-ray or equi-If valent inspection of fuel rods is part of the QC program.
not, explain why.
Give the following properties of Al 0 -3 C as a function of
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231.23 23 4
.(4.2.3.2.5) temperature at various burnups:
(1) swelling (2) thermal expansion l
(3) melting point (4) thermal conductivity
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230-5 231.23 (4.2.3.2.5)
(5) specific heat (6) compatibility with Zircaloy Describe how helium release is accounted for. Describe any reaction between A1 0 -B C and steam or hot water if the cladding 23 4 perforates.
231.24 Discuss operating experience of B&W pressurized water reactors (4.2.3.2.5) with burnable poison rods, including reactor names, loading dates, and 15radiati n parameters.
Describe how this Al 0 4C behaves 23 as B is burned.
What models does B&W use to predict the be-havior of these poison rods?
231.25 Discuss the potential swelling of fuel rods during a postulated (4.2.3.3.7)
LOCA with respect to interference with control rod guide tubes to the point of hindering control rod movement. Describe the bases for your response to this question, including creep and thermal dimensional changes, spacer grid / fuel rod interaction and fuel' rod bowing.
231.26 Supply calculations of the following parameters ~for an average (4. 4. 2. 8. 4) burnup and peak burnup 17x17 fuel rod as a function of densifi-cation and burnup:
(Nomiral values should be used.)
(1) gap conductance (2) hot pellec diameter o
(3) hot gap (4) fuel centerline temperature (5) fuel volumetric average temperature (6) internal gas pressure (7) gas thermal conductance (8) jump distances for fuel pellet and r.ladding or total (9) cladding Anside diameter temperature l
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230-6 Sptcify a rafsrenes er cupply a completa description of tha com-puter program used for'these calculations, including all materials properties and models used.
Supply the power history assyted for these calculations and explain why it is typical of what can be expected in your reactor.
Supply the axial power shape assumed for this calculation.
Explain how the stored energy obtained from this calculation is used as input to the fuel rod heat up calculation in the ECCS analysis.
231.27 Provide a detailed engineering failure analysis of the design (15.1.23.2) basis fuel handling accident.
In particular, provide a detailed mechanistic description and calculation of the accident, including assembly drop height and a justification for the selection of the height and the location of the drop (such as in fuel storage pool or over the reactor). Provide experimental data to support your calculation of fuel rod damage.
Also, justify the assumption that damage occurs to only 64 fuel rods due to a drop of the fuel assembly.
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