ML20040A480
| ML20040A480 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/31/1974 |
| From: | Stello V US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Moore V US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201210167 | |
| Download: ML20040A480 (6) | |
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AWMC UNIM Y CO.V. MISSION
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OCT 31 O, I
Docket ios.: 50-500 l
and 50-601 i
V. A.1;oore, Assistant Director for Light llater Reactors, Group 2, L 1-l
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j Il41TIAL QUESTI0ils FOR DAVIS BESSE Uf1ITS 2 & 3 (Chapters 1, 4, and 15) s d
Plant 1.'ame:
Davis Desse Units 2 & 3 S
Licensing Stage:
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Docket !!o.:
50-500 & 50-501 1
Responsible Branch LtlR 2-2 and Project !!anager:
R. Benedict Technical Review Branch Involved:
Core Performance Branch i
Requested Co.11pletion Date:
October 11, 1974 I
Description of Review:
Initial Questions Enclosed are first round questions of the Reactor Fuels Section of the Core Performance Branch relating to Chapters 1, 4, and 15 of the Davis Besse 2 & 3 PSAR.
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Vict'or Stello, Jr., Assistant Director i
for Reactor Safety i
Directorate of Licensing r
Enclosure:
Questions i
cc:
S. llanauer F. Schroeder i
A. Giambusso W. IicDonald R. Benedict I
D. Ross j
L. Rubenstein i
S. Kim l
S. Varga D. Ilouston i
E. Leins i
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MA DDEN80 515
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241-1 i
241.0 REACTOR FULLS SI.CTION - CORE PEP, FORMA!!CE BRANCH 241.1 The Mark C fuel assembly is a new fuel assembly design.
Provide (1.b) a detailed sc.hedule of the planned l' ark C irradiation program and the subsequent Post Irradiation Examination.
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241.2 The fuel handling and shipping design loads are provided. What (4.2.1.1.1) are the bases for these design loads?
To what extent have these design loads been confinned experimentally?
Is the relationship between these design loads and stress-strain limits such that no design limits are exceeded during handling and shipping?
241.3 The Standard fonaat in Section 4.2.1.1 (Design Bases) requires a 4
( 4. 2.1.1. 2) consideration of the following:
i (1) the physical properties of the cladding and the effects of design temperature and irradiation on the properties; (2) stress-strain limits;
( )) the effects of fuel swelling; (4 variations of melting point and fuel conductivity with burnup; (5) the requirements for surveillance and testing of irradiated fuel rods.
Provide more discussion in the Fuel Rod Design Bases to indicate how the above items are considered.
241.4 Provide the numerical values used for the Zircaloy cladding's (4. 2.1.1. 2) yield strength and ultimate tensile strength mentioned in j
conjunction with the stress intensity limits.
In addition, state the cladding thermo-mechanical history and associated temperature l
and fast neutron flux (and fluence) for which the stress limits apply.
241.5 Provide a list of the conservative estimates made in the fatigue
( 4. 2.1.1.2) calculations.
241.6 The relationship betwcon 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.
If the'se conditions are design bases, describe how applied.
241.7 Describe independent check made at completion of fuel loading to (4.2.1.2) verify the location and orientation of the fuel in the core.
l 241.8 Provide a drawing that shows the details of the spacer grid at the (4.2.1.2) instrument tube location.
Discuss the manner in which the spacer sleeves restrict the movement of the spacer grids.
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241.9 Provide the deflectiun design specificatior.s (includino dimensions (4.2.1.2.2) and spring constants) and experimental observations for the upper and louer plenum springs.
Is there evidence of peraanent deflection l
l due to fuel rod and fuel assembly handling? Wuuld gradual deflection I
of the louer spring be expected as a function of irradiation and what is the maxinuu deflection, both expected and possible?
Discuss the QC procedures which assure that the proper type of spring is in the louer plenum.
241.10 Provide a detailed description of the insulating spacers including (4.2.1.2.2) the che.nical composition and hydrogenous icpurity content.
Briefly describe the QA program associated with the manufacturing or procurement of this insulator.
241.11 Provide the design bases for Zr-4 irradiation growth and supply (4.2.1.3.2) supporting data or references.
241.12 Describe the power history used to calculate fission gas release.
(4.2.1.3.2) 241.13 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 analysis.
If possible the answers should comprehensively provide a network for an overview of your design basis in this area.
The discussion should include a sumary of the safety analysis in terms of stress report for each component and its loadings.
Discuss how the different loading categories are combined to satisfy the design limit for cach component of the fuel assembly.
Uith respect to fuel rod and assembly behavior, discuss the design limits for the accidents such as LOCA and seismic events.
241.14 Provide tables of numerical values (or equations) of a3terial (4.2.1.3.2) properties of both cladding and fuel pellets as functions of temperature and irradiation. The following properties should be included:
1.11odulus 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 241.15 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, is this volume adequate for accidents and transients in which the fuel night reach a temperature in excess of the design temperature?
241-3 l
Also, describe nw creep effects and dimensional stability are accounted for in designing the fuel rod plenual.
In a transient, is it possible for the cladding tempet ature to become so high that clad swelling will occur due to internal pressure? What j
will the end of life internal pressure be for these fuel rods, (both average burnup and peak burnup)? What temperatures are assigned to each of the follouing void regions when determining the fuel rod pressure?
- a. fuel rod upper end plenum
- b. fuel-clad annulus
- c. fuel pellet end dishes i
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- d. fuel pellet open porosity l
241.16 The following items should be addressed in the flow-induced t
(4.2.1.3.2) vibration program:
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(1) natural frcquency limitation of the' fuel assembly, (2) natural frequency relative to primary system frequency, and (3) stiffness limitations on the spacer grid assembly and j
individual grid spring.
241.17 Describe the corrective actions indicated at the end of Pottntial f
(4.2.1.3.2) for Water Logging Rupture.
241.18 Discuss all procedures used during fabrication to assure that no
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(4.2.1.3.2) axial gaps are introduced during the loading of the fuel rods, such as weighing, counting pellets, fluoroscopic examination, etc.
i 241.19 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 i
the fuel performance, specify acceptable limits, analyzed results and ultimate chemical d'isposition.
l 241.20 Give safety factors applied in the fatigue design, creep rupture, i
(4.2.1.3.2) fatigue creep interaction and instability (buckling) analysis for l
the 17xl/ fuel assembly.
j 241.21 Provide steady state, transient and accident response of guide (4.2.1.3.2) tube including dimensional stability.
241.22 Evaluate the effects of fuel rod bowing together with spacer grid (4.2.1.3.2) response including time dependent behavior due to creep.
241.23 The analytical calculation of fuel clad mechanical interaction (4.2.1.3.2) that is used to describe the design bases should be given. A detailed, complete description is needed, including a general description, assumptions, mathematical equations, sequence of application of equations or a flow chart, all empirical constants i
used in the equations, a sample calculation and a comparison with da ta.
Include the effects of:
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241-4 (1) fuel suelling driven cit.Cding strain (specify fuel suelling assumption used),
(2) "bambooing" of the cladding due to fuel pellet end effects.
(3) radial and/or axial differential thermal expansion of the fuel and cladding.
Explain how a transient in which fuel rod power was increased would effect the fuel-cladding mechanical interaction and how this is taken into account in the model described above.
241.24 Discuss fuel assembly seismic model and analysis method.
In (4.2.1.3.3) particular, a method of obtaining detailed stress and deformation of the fuel rod from the simple spring-mass beam mode response should be given.
241.25 Discuss in detail the surveillance, inspection and testing of (4.2.1.3.4) irradiated fuel rods. An example of one post irradiation examination plan is ASTM-E-453, " Examination of Fuel Element Cladding Including the Determination of liechanical Properties."
241.26 Identify all welded joints'in the assembly and categorize by type (4.2.1.4.2) and importance for safety. Describe both destructive and non-destructive ucld testing, e.g., localized corrosion, metallographic examination and dimensional inspections, indicating what constitutes an acceptable result.
Is x-ray inspection of fuel rods part of the QA Program?
241.27 Give the follnwing properties of A10 -B C as a function of temperature 23 4 (4.2.3.2.6) at various burnups:
- 1. swelling
- 2. thermal expansion
- 3. melting point
- 4. thermal conductivity
- 5. specific heat
- 6. compatability with Zircaloy is there a reaction between A10 -B C and steam or hot water if the 23 4 cladding perforates?
241.28 Discuss operating experience of B&W PWRs with burnable poison (4.2.3.2.6) including reactor names, loading dates, and irradiation parameters.
Itow does this material behave as the poison is burned? lihat models does B&ll use to predict the behavior of these poison rods?
l 241.29 Can swelling of fuel rods during a postulated LOCA interfere (4.2.3.3.7) with control rod guide tubes to the point of hindering control rod movement? Describe the bases for the conclusion including creep and thermal dimensional changes, spacer grid fuel rod interaction l
and fuel rod bowing.
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241.30 Supply calculations uf the follo.:ing paru :tcrs for an average (4.4.2.U.4) len nup and ped burnup 17x17 fuel rod as e function of densification and buinup:
(nouinal values should be uscd):
- 1. gap conductance
- 2. hot pellet diameter
- 3. hot gap
- 4. fuel centerline temperature
- 5. fuel volumetric average temperature
- 6. internal gas pressure
- 7. gap thermal conductance
- 8. jump distances for fuel pellet and cladding or total
- 9. cladding inside diameter teuperature Specify a reference or supply a complete description of the computer program used for these calculations, including all materials properties and models used.
Supply the power history assumed for these calculations and explain why it is typical of what can be expected in reactor.
Supply the axial power shape assu;aed for this calculation.
Explain how the stored energy obtained from this calculation is used as input to the fuel rod heat up cdiculation in the ECCS analysis.
241.31 Piovide a detailed engineering failure analysis of the Fuel (15.1.23) llandling Accident. Compare the postulated accident for in-core during unloading versus in the spent fuel pit.
Provide details of drop distances, kinetic energy imparted to the fuel assembly, basis for fuel rod failure determination, and the number of fuel assemblics involved in the accident.
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