ML20039G407

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Forwards First Round Questions by Reactor Fuels Section Re Chapters 1,4 & 15 of B-SAR-241
ML20039G407
Person / Time
Site: 05000481
Issue date: 09/19/1974
From: Stello V
US ATOMIC ENERGY COMMISSION (AEC)
To: Moore V
US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML111090060 List: ... further results
References
FOIA-80-515, FOIA-80-555 NUDOCS 8201180173
Download: ML20039G407 (6)
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UNITED STATES

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,1 ATO!.UC ENERGY COMMISSION 4

WASHINGTON, D C.

20545

'9j'yjf I I n 6' Docket l'o. 50-481 E& 19 d l

V. A. ibore, Assistant Director. for Light Water Reactors Group 2. L I

IfilTIAL QUESTI0ilS FOR B-SAR-241 (Chapters 1, 4, and 15)

Plant flame:

B-SAR-241 Licensing Stage:

Q1 Docket fic.:

50-481 Responsible Branch LUR 2-3 and Project :'anager:

D. K. Davis Technical Review Branch Involved:

Core Performance Branch Requested Co pletion Date:

September 13, 1974 Description of Review:

Initial Questions i

l Enclosed are first round questions of the Reactor Fuels Section of the Core Performance Branch relati.ng to Chapters 1, 4, and 15 of B-SAR-241.

VI,ctor Stello, Jr., Assistant Di, rector t.

for Reactor Safety Directorate of Licensing

(

Enclosure:

Questions

,_1 cc:

S. Hanauer F. Schroeder A. Giambusso' W.14cDonald A. Schwencer D. Davis D. Ross L. Rubenstei,n

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P. Check R. Bottimore D. Basdekas' L. Kopp' L. Chandler S, Varga i

D. ilouston E. Lei,ns 4

8201180173 810403 PDR FOIA MADDEN 80-515 PDR E

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241-1

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I iI CORE PERF0PJGNCE BRANCH L

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REACTOR FUELS SECTION 241.0 Provide n

The Mark C fuel assembly is a new fuel assembly

.p 241.1 and the subsequent Post Irradiation Examination.

(1.5.2) vide a consistent'value for the following items:

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2 (124.7 in Table 4.3-2) 241 2 Total 00 Fuel Rod length (153.88 in Table 4.2-1)

(4.1.1)

What The fuel handling and shipping design loads are provided.

To what extent have are the bases for these design loads?

241.3 Is the these design loads been confirmed expericentally?

(4. 2.1.1.1 )

relationship between these design loads and stress-strain limits such that no design limits are exceeded during handling and shipping?

The Standard Format in Section 4.2.1.1 (Design Bases) requires a 241.4 consideration of the following:

(1) the physical properties of the cladding and the effects of (4.2.1.1.2) design temperature and irradiation on the properties; 1

stress-stain limits;

(

the effects of fuel swelling; variations of melting point and fuel conductivity with burnup (I(the requirements for surveillance and testing of irradiated fuel rods.

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Provide more discussion in the Fuel Rod Design Bases to indicate how the above items are considered.

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Provide the numerical values used for the Zircaloy cladding's -

+

yield strength and ultimate tensile _ strength mentioned in 241.5 O ?:f' conjunction with the stress intensity -limits.

In addition,u.,

(4.2.1.1.2)

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state the cladding thermo mechanical history

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stress limits apply.'

Provide a list of the conservative estimates made in the fatig i

241.6 (4.2.1.1.2) calculations.

The relationship between' compressive loa 241.7 (4.2.1.1.2) in regard to hydride precipitation.

design bases?

Describe independent check made at completion of fuel loading -

to verify the location and orientation of the fuel in the core.

241.8 (4.2.1.2) r b

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Provide a drawing of the Mark C Fuel Assembly for t

'1 241.9 4.2-1.

grid i

no (4.2.1.2)

Provide a drawing that shews the details of the spacerDis f'

h pacer grids, at the instrument tube location.the spacer sleeves res f.

241.10 (4.2.1.2) l 4.2 Provide the details of the spacer sleeves in Tab e g

241.11 (including

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(4.2.1.2)

Provide the deflection design specificationsd L

l observations i

241.12 l assembly handling?

for the upper and lower plenum springs.

(4.2.1.2.2) permanent deflection due to fuel rod and fue ted as i

Would gradual deflection of the lower spring be expec i

deflection, a function of irradiation and t. hat is the max mumDiscuss the in the lower plenum.

both expected and possible? assure that the proper typ h and supply Provide the design bases for Zr 4 irradiation growt 241.13 supporting data or references.

t during (4.2.1.3.2)

Discuss the behavior of the bottom fuel rod supporf the fuel cyclic thermal axial expansion and contraction o 241.14 d provide (4.2.1.3.2)

List fuel rod deflections and cladding strain limits anDem d accident analysis.

justification for their adequacy.

241.15 are satisfied during steady state, transient ide a (4.2.1.3.2) i in this. area.

r network for an overview of your design bas s t analysis

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The discussion should include a suanary of the safe y loadings.,

in terms of stress report for each component and i,ts I

are[ombined,to' fuel, assembly ( ' ' ;

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it loadi.ig categorie h

. Disc'us'sihow 'the ilifferasatisfy the design limit for eac

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With respect ot fuel rod and assembly

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i i events.

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) of paterial' ',.

~l Provide tables of numerical values (or equations functions of..

F properties of both cladding and fuel pellets asThe followi f

241.16

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temperature and irradiation..

(4.2.1.3.2) 1 be included:

f

+

11odulus of elasticity r

1.

Poisson's-ratio Thermal expansion coefficient 2.

3.

z 4.

Yield stress Ultimate stress S.

Uniform ultimate strain l

l 6.

Thermal conductivity l

7.

l 8.

Specific heat i

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241.17 Describe procedo. A used for sizing the fuel rod plenum, (4.2.1.3.2) including any computer codes used and the fission gas release rate assumed.

Is this volume adequate for accidents and transients in which the fuel might reach a temperature in 2

excess of the design tempe~rature? Also, describe how creep effects and dinensional stability are accounted for in designing the fuel rod plenum.

In a transient, is it possible for the cladding terperature to become so high that clad swelling -

will occur due to internal pressure?

L' hat will the end of life internal pressure be for these fuel rods, (both average burnup and peak burnup)?

L' hat temperatures are assigned to each of the follcuing 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 241.18 The following items should be addressed in the flow-induced (4.2.1.3.2) vibration program:

(1 natural frequency limitation of the fuel assembly, I

(2 natural frequency relative to primary system frequency, and (3 stiffness limitations on the spacer grid assembly and individual grid spring.

241.19 Discuss the corrective actions to be taken to avoid water.

(4.2.1.3.2) logging rupture.

241.20 Discuss all procedures used during fabrication to assure that,-

(4.2.1.3.2) no axial gaps are introduced during the' loading of the fuel s'

rods, such as weighing, counting pellets, fluoroscopic _

examination, etc.

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241.21 Internal and external surfaces of the Zircaloy tubing are J,

(4.2.1.3.2) cleaned with dry cott'on swabs and~ acetone ~ saturated cloth J ';' 4 respectively.

Discuss. the effect of residual lint 'on the internal surfaces upon the fuel performance, specify acceptable- -

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limits, analyzed results and ultimate chemical disposition.

241.22 Give safety factors applied in the fatigue design,' creep rupture. ~

(4.2.1.3.2) fatigue creep interaction and instability (buckling) analysis..

for the 17x17 fuel assembly..

241.23 Provide steady state, transient and accident response of (4.2.1.3.2) guide tube including dimensional sta.bility.

241.24 Evaluate the effects of fuel rod bowing together with spacer.

(4.2.1.3.2) grid response including time dependent behavior due to creep.

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1 241.25 241-4

('"4 (4.2.1. 3. 2) The that isanalytical calculatio

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A detailedused description, ~ c'omplete descrito describe the de n of fuel clad mechanic l application,of equatiassumptions, ma themati s gn bases

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ption is a

intera neededshould be given.ction used 4

wit in th (1)h data. e quations,ons a flow hart, all em,iincluding e

or cal equa,tions fuel swelling driInclude the a

c sample ' calculation sequence effects assumption of and ap rical con of.

((3)) radial and/o"bambooing"used), ven cla comparisons tants, 2

of the

. he fuel and cladding.ifferentialcladding du t

r axial d Explain how ng thermal expansionpellet end effe would effecta transient in uhich f Shouldthis is taken int the fuel-cladding of 241.26 mechanicaluel rod pouer the amount of mechaccount in the o

(4.2.1. 3. 3)Discuss fu anical model interactionwas increased particular,el assembly seismi interaction bedescribed aboveand of the fuel rod a design. limit?

a

@l.2.1. 3 should be given. from themethod of obtaining dc model l.27 simple eta Discuss in detail springiled stressmethod.

4)

In irradiated fuel rod mass beam and deformation the mode examination plan is A surveillance, inspecti response Cladding Including s.

An

.28 exampl STM-E-453, "e i

Examina tionof one post irradiatitesting l-on and

?.1. 4. 2)

Identify all weldthe Determination type of liechanical Propertiof Fuel Ele on non-destructiveand icportance fored joints in the metallographic assembly and'categori weld safe inspectioindicating what coexaminationtesting,ty.

es. "

De e.g. scribe both destructive ze by

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localized corrosion and dime,nsional nstitutes n

7 2.5)

Give the followinof fuel ~ rods part of an and\\

acceptable resultinspections,, '

tem at various burnups:of.the QA Progiam?

1. perature g

properties ~ '

Is x-ray 2

swelling A1 0 -B melting pointthermal expansion e

C as, function

~e 3

- 2 3 4 a

4 specific heatthermal conductivi of S.

6 ty compatibility with Zi l

How is helium rcaloy A1 0 -8 C and steamrelease

'L 2 3 4 or hot water ifaccounted for?

Is there

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cladding perforates? reaction between

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Discuss operating exper j

1-including reactor names,ience of BMI PtiRs with burnable poison V " b';

Itow does this raterial behave as the poisloading dat parameters.

burned?

t What models does B&W of these poison rods?.

on is use to predict the behavior'

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Can swelling of fuel rods du' ring a post l t d

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~ O with control rod guide tubes to the point of hi d1.0CA~ interfere.

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uae rod movement?

creep and thermal dimensional changesDescribe the bases for th

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n ering control. -

interaction and fuel rod bowing., spacer grid fuel rodu ng l'

L l-burnup and peak burnup 1 Supply calculations of the following

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l densification and burnup:7x17 fuel rod as a function ofparameters for an average 1.

hot pellet diameter (nominal values should be used):

gap conductance 2.

3 hot gap 4

fuel centerline temperature 5

internal gas pressurefuel volumetric average temperature 9.

gap thent.al conductance i

cladding inside diameter temperaturejump distances f s

g or total jpecify a reference or supply i

j sputer program used for these calculationsa complete description of the j

terials properties and models used including all

. umed for these calculations and explain why, it is typicalSup what can be expected in reactor.

1 ape assumed for this e.

calculation.

Supply the axial power

!crgy obtained from this calculation is used aExplain how the s

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2 2 fuel rod heat up calculation in the EC

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s input to.

CS analysis..

g tvide a detailed engineeri

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ident.

In particular, justification should be' ing failure a'nalysiEM the ~

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lowing:

A dropped fuel assably could not strike g ven for the '

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assembly in the storage rack (provide drawimore thkn~ one' fuel b

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of the drop with no fuel rod failure (provideThe fue ng).-

iThe worst fuel handling accident that b,

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calculation spent fuel p(provide d t ilool is the dropping of a fuel assem 001 floor could occur in the ).

m eight is assumed).

ed calculations' and state what drop ea I

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