ML20040A425
| ML20040A425 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 04/20/1976 |
| From: | Check P Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201210062 | |
| Download: ML20040A425 (11) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D. C. 20655 f
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4 skre R. C. DeYoung, Assistant, Director for Light Water Reactors, DPM FIRST ROUND QUESTIQUS FOR EFAR-205 Plant Name:
BSAR-205 Docket Number:
STN-561 Licensing Stage:
PDA Milestone Number 05-23 Responsible Branch LWR-1 and Project Manager:
T. Cox Systems Safety Branch Involved:
Core Performance Branch Requested Completion Date:
April 26, 1976 Review
Description:
First Round Questions Review Status:
Awaiting Information
]
Enclosed are first round questions on the PSAR for' the standard plant BSAR-205 from the Physics Section of the Core Performance
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t Paul S. Check, Chief Core Performance Branch Division of Systems' Safety
Enclosure:
Questions cc:
S. Hanauer F. Schroeder R. Boyd J. Stolz
[T.Cox D. Ross W. Mcdonald W. Brooks M. Dunenfeld 8201210062 810403 PDR FOIA MADDEN 80-515 PDR nn
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230.0 CORE PERFORMANCE M dd
....i Reactor Fuels dygfiday/
231.0 231.1 The fuel handling and shipping design loads'are provided.
(4.2.1.1)
State the bases for these. design loads. Describe the extentdu 6"bA6i thee these design loads 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 Specify yield and ultimate stress values as a function of (4.2.1.1) temperature and irradiation. Also, specify design limits with regard to fretting wear and deflection.
231.3 Provide the numerical values used for the Zircaloy cladding (4.2.1.1) yield strength and ultimate tensile strength mentioned in f
conjunction with the stress intensity limits.
In addition, state the cladding thermo-nechanical history,amd associated
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temperature and fast neutron flux and fluence for which the j
stress limits apply.
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'1. 4 The relationship between cladding compressive load. coolant 4
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,.2.1.1) temperature, and zirconium hydride precipitation is dis-
-f cussed in Section 4.2.1.3.2.
In Section 4.2.1.1.2, describe k,
how the specified coolant temperature limits and associated cladding loadings are used in design basis calculations of cladding stresses.
231.5 Provide a drawing that shows the details of the spacer (4.2.1.2) grid at the instrument tube location. Discuss how the spacer ~
sleeves restrict the movement of the spacer grid.
L 231.6 Provide the deflection design values (including nominal (4.2.1.2) dimensions and spring constants) and experimental observa-tions 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 i
lower spring would be expected as a function of irradiation and state the mav4=um deflection, both expected and possible.
Discuss the QC procedures that assure that the proper type i
of spring is in the lower plenum.
231.7 Provide the design bases for Zircaloy-4 irradiation growth (4.2.1.3) and supply supporting data or references.
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230-2 231.8 List fuel rod deflections and cladding strain limits and (4.2.1.3) provide 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 the 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 loadings categories are combined to satisfy the design limit for each component of the fuel assembly.
Mith rcspumi iv fmmi.vd eud-essc;bly Lchai-iora Discuss the design': din,lixi{foreachcomponentofthefuelassembly.
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& M is.4 (.<al w 231.9 Provide tables of numerical val 6es (or equations) of material (4.2.1.3) properties of the cladding 6nd fuel pellets where specified)as functions of temperature and irradiation.
The folloving prop-erties should be included:
(1) Modulus of elasticity (2) Poisson's ratio (3) Thermal expansion coefficient (cladding and fuel pellets)
(4) Yield stress (5) Ulti= ate stress (6) Uniform ulti= ate strain (7) Thermal conductivity (cladding and fuel pellets)
(8)
Specific heat (cladding and fuel pellets) 231.10 Describe procedures used for sizing the fuel rod plenum, in-(4. 2.1. 3) cluding any computer codes used and the fission gas release rate assumed.
State whether this volume is adequate for accidents and transients in which the fuel might reach a temperature in excess of the design temperature.
Also, de-are accounted scribe how creep effects and dimensional, stability'en3 for in designing the fuel rod plenum. /In a transi State
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whether it_is possible-for~the~ cladding temperature to become sohighfthatcladdingswellingwouldoccurdue"tointernal pressure.
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-cladding annulus (c) fuel pellet end dishes (d) fuel pellet open porosity l
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internal surfaces upon the fuel performance, and specify acceptable limits, analyzed results and ultinate chemical disposition.
,N (DEd2Dhive safety factors applied in *he f::.tigue. design,_creepN 231.15 q
(NONE)
' rupture, f atigue creep interaction and instability (buc,k. K g
(ling) analysis for the fuel assemblie.s
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(NONE) the guide tubes in terms of stress, strain, dimensional stability, and deflection.
231.17 Evaluate the effects of fuel rod bowing together with spacer (4.2.1.3) grid response.
Include time dependent behavior due to creep in your evaluation.
231.18 Discuss the analytical calculation of fuel / cladding mechan-(4.2.1.3) ical interaction that is used to describe the design basis.
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 used in the equations, a sample calcu-lation and a comparison with data.
Include 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 ex-g
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pansion of the fuel and cladding.
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231,11 Add're. ' 'N ing it' ems'iOpedloMdGEeLvib7atip (NONE) jp@ ' mf (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 individual grid spring.
231.12 Describe the corrective actions indicated in the last 4MBN84L sentence of the paragraph titled, " Potential for Water (fflel.33)
Logging Rupture."
231.13 Discuss all procedures used during fabrication to assure (NONE) that no axial gaps are introduced during the loading of the fuel rods, such as weighing, counting pellets, and fluoroscopic examination.
231.14 Internal and external surf aces of the Zircaloy tubing are (NONE) cleaned with dry cotton swabs and acetone saturated cloth, respectively. Discuss the effect of residual lint on the internal surfaces upon the fuel performance, and specify acceptable 1Dnits, analyzed results and ultimate chemical
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231.15 (DELL g'Give safety f actors applied in the.f atigue_ design,. creeps q
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[ rupture, fatigue creep interaction and instability (buck.
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231.16 Provide steady state, transient and accident responses for (NONE) the guide tubes in terms or stress, strain, dimensional stability, and deflection.
231.17 Evaluate the effects of fuel rod beving together with spacer (4.2.1.3) grid response.
Include time dependent behavior due to creep in your evaluation.
Discuss the analytical calculation of fuel / cladding mechan-231.18 (4.2.1.3) ical interaction that is used to describe the design basis.
A detailed, co=plete description is needed, including a general description, assumptions, mathematical equations, all sequence of application of equations or a flow chart, empirical constants used in the equations, a sample calcu-lation and a comparison with data.
Include 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
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(3) radial and/or axial differential thermal ex-pansion of the fuel and cladding.
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231.18 Explain how a transient in which fuel rod power was in-(4. 2.1. 3) creased would affect the fuel / cladding mechanical interac-tion and how this is taken into account in the model de-scribed above.
Discuss operating design limits (transient conditions) or total lifetime limits based on calculated fuel cladding mechanical interaction effects.
Discuss the fuel assembly seismic model and analysis metho'd.
231.19 (4.2.1.3)
In particular, describe a method of obtaining detailed stress and deformation of the fuel rod from the. simple spring-mass beam mode response.
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231.20 Identify all welded joints in the assembly and categorize (NONE) them by type and importance for safety. Describe both de-
-3 structive and non-destnfhtive weld testing, e.g., localized corrosion, metallographic examination and dimensional in-spections, and indicate what constitutes an acceptable result.
State whether x-ray or equivalent inspection of fuel rods is part of the QC program.
If not, explain why.
231.21 Give the following properties of A10 -B C as a function 23 4 (4.2.3.2) of temperature at various burnups:
(1) swelling (2) thermal expansion (3) melting point (4) thermal conductivity (5) specific heat (6) compatibility with Zircaloy Describe how helium release is accounted for.
D2 scribe any reaction between A1 0 -B C and steam or hot water if 33 4 the cladding perforates.
231.22 Discuss operating experience of B&W pressurized water
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(4.2.3.2) reactors with burnable poison rods, including reactor names, loading dates, and' irradiation parameters.
Describe how 10 is burned.
What models does this Al 0 ~B C behaves as B 23 4 B&W use to predict the behavior of these poison rods?
231.23 Discuss the potential swelling of fuel rods during a (4.2.3.3) postulated LOCA with respect to interference with control rod guide tubes to the point of hindering control rod move- -
Describe the bases for your response to this question, ment.
including creep and thermal dimensional changes, spacer grid / fuel rod interaction and fuel rod bowing.
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-Supply calculations of the following parameters for an (4.4.2.8) average burnup and peak burnup 17x17 fuel rod as a function of burnup:
(Nominal values should be used.)
(1) gap conductance (2) hot pellet diameter (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 cladding or total (9) cladding inside diameter temperature Specify a reference or supply a complete description of the computer program used for these calculations, including all materials properties and =odels used.
Supply the power history assumed for these calculations and explain why it is typical of what can be expected in your reactor.
Supply the axial power shape assuced for this calculation.
Explain how the stored energy obtained from this calcula-tion is used as input to the fuel rod heat up calculation in the ECCS analysis.
Provide an engineering failure analysis of the design s1.25 (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, fustify the assumption that damage occurs to only 64 fuel rods due to a drop of the fuel assembly.
231.26 Describe independent checks made at the completion of fuel (NONE) loading to verify the location and orientation of the fuel in the core.
231.27 Describe the effects of blowdown forces on the fuel rods (NONE) during a LOCA. Name and give detailed descriptions of the computer codes used to calculate these forces and ex-plain how the thermal hydraulics calculations are used as input into these codes.
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Viscuss h the different loading categories ay ombin
, f(NONE) to satisfy the design li=it for each component of the fuel 1
d asse=b17 231. 8 A8' List the fuel rod stresses caused by lateral differential (NONE) thermal expansion between the end fitting and the spacer grid.
231.3 :61 Provide a detailed description of the fuel fatigue analysis.
(NONE)
Include a sanple calculation for a 17x17 rod fuel assembly showing the assu=ptions that are cade for fuel rod tolerances, fuel te=peratures and cycling history. Also describe the primary creep rate expression that is used.
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JhCP!IS$1Pff 232.1 APSR maloperation implies that some limits are placed on the l
(4.3.2.2) location of the APSR's.
Plense clarify.
If limits are used in performing desig;n analyses will these limits become subjcets of Technical Specifications?
232.2 Provide local (intra-ec11) pcahing factors for the varicus (4.3.2.2) types of ascemblies used in the core (different enrichments nad burnable poison loadir.gs, e.g. ).
232.3 U;2date the diccuasic:. cf the r.acicar unccrtair.ty f acttw 1
(4.3.2.2) to incit.di the experience gained with operctiig, rcr.ctors.
In particular, include any ccmparit,ons between calculation and c:<periment for transient conditions.
4 232.4 Sustained acimuthal ::enon oscillations are predi-red not (4.3.2.2) to occur. Ilouever, it is not clear that a peakini; factor prol.lca docs not occur for damped c?.Jrcuthal oscillations, such as might eccur after a dropped red.
Pleanc co:c.ent on the behavior of the core in r.uch a case and provisions r
f f or avoiding peaking f actor problems.
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.3 What is the manimu:. peakins, factor increasc caused by (4.3.2.2) mislovling fuel in the core? 1Tnat ie the-reaxiraum increasc _. _
that trould go undetected by the incore udectors?
232.6 Tables 4.3-9 and 4.3-10 ar inconsirtent vith recract to
-i (4. 3. 2.4 )
Dappler and Moderator tes:pe'cature defic. irs at EOL, first cycle.
Picate clarify.
232.7 Discuss the effect of failure of the ICS to keep the (4.3.2.0) temperature in the two loops equal.
- 23. 8 Piovide additional data on the new fuel storage racks to (C.1.4.2) support the corclusion that k will nor excced 0.95.
At ggf i
least the following infornation shoald be provided:
(1) Calculation method used and :nethods used to I
verify its accuracy.
(2) Assumptions used (fuel enrichment, credit for ntructural materials, etc.).
I (3) l'neertaintics used (a)
Calculational bias (b) Celculational uncertainty (c).hhnnical :ncet.ainty
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(4) No:t.inal value of keft for the tacks Q
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230-2 232.9 Provide assurance that fresh fuel racks will remain sub-(9.1.4.2) critical (keff f 0.98 with uncertaintias included) for
',j accident conditions.
Include in these conditions the flooding of the racks with partini density water.
(If the racks are not suberitical for some densities of water each applicant will have to provide assurance that achieving such densities is incredible).
232.10 Provide additional data on the spent fuel storage racks (9.1.4.2) to support the conclusion that k,ff will not exceed 0.95.
The same information as that sought in Question 232.8 should be provided.
232.11 Justify the use of a symmetric cosine axial power shape (15.1) in generating the shutdown reactivity curve, since bottom-and top-peaked distributions are permitted during normal operations.
In particular discuss the effect of the as-sumed shape of the initial axial distribution on the min-imum DNBR achieved during Loss' of Flow and Increased lient Removal Transients.
~ 232.12 Justify the use of the same shutdown reactivity curve for s'
i.1) both full power and zero power transients.
232.13 Revise the SAR to reflect the fact that the additional (15.1) conservatism claimed for the shutdown reactivity curves due to the assumption of minimum tripped worth does not apply at the end of the first cycle where only a 1.0%
shutdown margin is predicted.
232.14 There are references at several places in Chapter 15 (15.1.1)
(e.g., p.15.1-4b and Table 15.1-3) to Table 15.1-2 which bear no relation to the contents of the table.
Please clarify.
232.15 Discuss the case of withdrawal of a single CRA at zero (15.1.1) power. Does the increase in peaking f actor resulting from this transient cause the heat generation rate to exceed the thermal limits in any part of the core?
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232.16 The sensitivity studies show that the maximum vessel (15.1.2) pressure during the transient increases with more negative values of the moderator temperature and Doppler coefficients.
Discuss the consequences of this transient at EOL when both of these coefficients have their most negative values.
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230-3 232.17 Discuss the consequences of a single CRA withdrawal as a (15.1.2) function of initial power. Assume operation of the control rod groups at the limit of their insertion band and discuss the effect of a single CRA withdrawal on peaking factors.
232.18 Correct the legend on the " Thermal Power" ordinate of (15.1.2)
Figures 15.1.2-la, lb and ld.
232.19 Discuss the effeebtof the azimuthal tilt produced by drop-
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(15.1.3) ping the rod, Q th_e, hot pin resultst particularly for the return to 100% FP in the EOL case.
Is there a trade-off between dropped rod worth and location such that a s= aller rod worth may have greater hot pin consequences?
232.20 Discuss the consequences of this accident on DNBR, par-(15.1.3) ticularly in view of the reduced pressure at BOL.
232.21 What assumption was made with respect to tripped rod worth (15.1.8) in the analysis of the rod ejection accident? Was a rod assumed to be stuck in addition to the postulated ejected rod?
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'12.22 Describe the manner in which the feedback coefficients,
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3.1.18) particularly the Doppler coefficient, are obtained for the point kinetics calculations.
In particular discuss the manner in which "three-dimensional" effects are treated.
232.23 The point kinetics results in Table 15.18-3 are different (15.1.18) from those given in the PSAR for the Greene County Nuclear Power Plant.
Please clarify the discrepancy.
232.24 Justify the use of*the design peaking factor in the point (15.1.18) kinetics calculation rather than some larger peaking factor representative of the situation with the ejected rod removed from the core.
232.25 For the limiting rod ejection accident provide, in addition (15.1.18) to the number of rods experiencing cladding failure, the number of rods in which the fuel reaches the melting temper-ature.
Indicate whether the same rods experience both types of failure.
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