ML20099A474

Pressure Vessel Fluence Analysis for 177 Fuel Assembly Reactors. Prepared for 177 B&W Owners' Group Technical Subcommittee for Reactor Vessel Matl Properties

ML20099A474
Person / Time
Site:	Midland
Issue date:	06/30/1978
From:	Whitmarsh C BABCOCK & WILCOX CO.
To:
Shared Package
ML20099A472	List: ML20099A469 ML20099A474
References
BAW-1485, NUDOCS 7902150409
Download: ML20099A474 (68)

Text

_ _ _ . .

1 I

Wp.

• Vv( g 32.*-1 35

...e *. )* :

t ,0, ,

P RESS'.'RI *."eS!E:. F.*.T.iCI KIA:.~t S IS TCR 177-TA 1EAC*CRS 37 C. L. *.hiaa rsh I

a Prepared for 177 B&W owners Group Technical Sub-Committee for Reactor vessel Material Propertiec 1

i o

l l -

t i

SA3CDCI & ".i'.a I Power Gacerati n Oroup Nuclear Power C sers :n Divis*

SW Lync'iburg. Virgi=is 2*.!O5 i Babcock & Wilcox l

I i

~.-

--~ m -.. - - - ,e.-. --- .._--.~,,.m..---- . - - , , _m_-e- - , - - ..-

I I

CD *ITICAM ON ~^

, )

' This report has been reviewed for tact:sical content and accuracy.

/0 f )rr. w .t r ,

$ClM" '&.lJ/, .'_ ,,<-AJ

.g..

>~

. 3. Wi=ne r Date '

Safety Analysis t

'Jeld cesposi: ion a 4 location data have been reviewed for accuracy.

d

11. S. Palme L s. RLx x Cate '

u, me

!achnical Staf*

i i

e 4

. :: . Sabcock & Wilcox t

a-

i

)

3atcoc's & Vilc:s

?cver Genera:::s ;roup ,

Nuclear ?cver Jeteration Mvisi n l

• ynchburg.

. Wrgista Report 2M-;a$

Juse li'!

Pressure vessel Flue =ce Analysis tor !!7-rn taaet:rs C. :.. *.hirmar st Kev Vards: Neut; n 71.ence hasaresent. Featr:n 71ser.ce Prediction. Neutron Oosisatry. ?ressure '.*es-sel Surveillance, ieutron Tra: sport. ? es-surized Vater Isactsr. 01screte-Crdisates Analvsts, %rvet::asco Cassules A35*'A7 A secessary part of a surveillance ptsgra= required to monitor pressure vessel physical properties is an acalytical setsod of predicting f ast flam distribu-tion in the pressure vessel. In addi: ion. :dere is a need :o predic: fleecce (and tterefore fluz) over :ne vessel's design life. vnich necessitates flux extrapolation over zul:1ple fuel cycles. As analytical model was javelcped based on two-dimensional transport theorv. The ref e.ence nodel is in planar geometry with appropriate correction factars for capsula perturbation effects and axial power distribution. As importast feature of 5e ocal is :he lac

• a.

sico of a point-by-paint relative power distribution which has been averaged over the irradiation time pericd. Istegral f *ux comparisocs vita fostseter data f rom fiver operating reactors indicated agreement of f ait flux (E > 1

.MeV) within 133 hased on fissien reactions in 5# p and W. U Cncertainty limits for calculated fast flux were estiasted to be :23: at :he inside surface of the pressure vessel and :3C* for that value extrapolated to 3: U"?Y . 34 sed on this analytical procedure. a generic design curve is presented f or predic:-

ing the pressure vessel fluence expected is a 177-FA reactor. over its 32 ET?Y design life.

. ;, . BabCCCk & WdCox

-s* -

p te -

n - -

4 i

l CCVTYN*S-Page

1. LNTE D;CTION . ........................ .. 1-1

2. AXALTTAL MCD LL . . . . . . . . . . . . . . . . . . . * * . . . . A*1 2.1. Issortant variables ........... . . . . . .... -L 2.2. 12.tial Generic Model Improvement . ... . . .. .. ...  :-2 2.3. Cal:ulaticnal Model ............ .. . . .... 2-3 2.3.;. Configuration .......... . . .. . ....  :-3 2.3.2. Relative Power *,4asity Distributica .... .... -*

2.3.3. Seltline Region . . . .. ... . . . .. . ....  :-o 2.J.4 Capsule Model ...................  :-7 2.3.5. Fluence Extrapolation .. .... . . . . .... . -4 2.6 Future Laprovements ............ . .. . .... 2-9

3. EIP EL*: MENT 4. 3AS IS . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.1. Desiastees . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.2. Dosisater Activation . . . . . . . . .. . . . .. . .... 3-2 3.2.1. Measurement Procedure . . . .. . . . .. . .... 3-2

3. 2. 2. Analytical Procedure . . .... . . . .. . .... 3-2 3.2.3. CosTertson of Deca . . . . . . . . . . . . . . . . . 3-3

4. ANALYSIS CT ::A*.A f.3CIK*.1I.MTY . . . . . . . . . . . . . . . . . . . 4-1 4.1. Surveillance Capsula Analysis . ..... . ..... ... a-L 4.2. Generic T3us Jaca ..... ... .... . . .. .... .-2

5. REFULTS ............................. 5-1 5.1. Generic Desigm 71eence . . . . ..... .. .... .... 5-L 5.2. Soecific teactor riuence . . . . . . .... . . . . .... 5-5.J. Lead T.ctors . . . . . . . . . . . . . . . . . . . . . . . . 5-3 o

6. RIT7'F4f7T ............................ 6* l APPCfD1XIS A. Time Aversging of Relative Power Distributic .s . . .G . . A-L

3. Asial Power Distribution Correctice Tactor . . . . . . . . 3-1 C. Ef f ect1* e Energy Range for Oosiaster leactions . . . . .. C-L D. Weigneed Capture Cross Sections and T:ssion Yields . ... 3-1 E. Equivalence of Activity and Fluz lacios .. .. . .... E-L F. :esign Fluence for Beltline Resign . . . . . . . . . . . . F-1 DIST113CTICN . .......................... G-1

-v- Babcock & Wilcox 1

l 1

?

litt of "JS*.es Page Inble 2-1. Ref e rence Calculation Model . . . . . . . . . . . . . . . . ... -11 2-2. P3 ,Fg Order of ScatterLag Carrection to Reference M del

-il Calculatica . ..........................

-3. Emergy structare at Cross Section Set ....... .... .. -13

2-=. AaLai Carrection Factor to Xef ere=ce M.odel Calculatiot .. ...  ;*-

2-5. Surveillance Capsule Flum Portarbation Factors ... .. ....

-6. Oata f or Pressure vessel fluence Fredictica is 177-TA teactors .  :-15

1. Oostaeter Reactiens . ............ ...... ... 3-o 3-2. Cceperisca of Relative Flux Spectra . .............. 3-7 3-3. Normalisation Tactors for Calculated Flax to Surveillance Capsules ............................ 3-5 3 -= . Anc tila ry Oos L=e t e r 04 t4 . . . .. .... .... .. ..... 3-8 3-5. Activity and Tium Comparison 5etween Beactors . ....... .. 3-10

-1. **ncertainties Related to Oostaeter Analyses. :

... .... .. -J 4-2. Cncertainty Limits Associated With Generic Pressure Vessel Surveillance Flus Analysis for 177-TA Reactors ......... 4-=

5-1. Maximum Pressure 7essel T1uence Frediction f or 177-TA Reactors . 5-=

5-2. Laad Tactors for Tast T1ux in a 177-TA Reactor ....... .. 5-5 5-1. etermisation of a typical Aztal Correction Tactor ... .... 3-3 C-1. Ef f ective Energy Range for Dostacter Reactions ......... C-3 D-1. TissLon Spectrum Weighted Capture Cross Sections for Cosimeter Materiala . . . . . . ................. D-3 3-2. T i s s io n Y i e ld . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3 T-1. Oesign T1uence Tactors for Celtline Region *delds in 177-TA Reactors ............................ T-3 List of Titures Figure Page 2-1. Plan 71ew Through Reactor Core Midplane ............ 2-16 2-2. Cylisdrical Model of Upper Raactor internala . . . . . . . . . . 2-17

? 1. Geometric Model for Calculation of Capsula Terturbation Ef f ect on Fluz . . . ......................  :-1S 2-4 Turve111ance Capsule Arrangement . ............... 2-19 3-1. Surveillance Capsule Dostseters ................ 3-11 .

5-1. ".. ceric Design Tiuence for 177-TA Reactors Sased en Location of Maxirun Exposure on inside Surface of Pressure vessel . . . . 5-6 T- . Tase T1ux Attentuation Through Pressure vessel '4411 ...... T-3

. vt . Babcock & Wilcox f

1. I:r:10:;CCTICN ne reactor vessel surveillance program (RVS?) was established to monitor the combined ef fects of irradiation and temperature on the nachanical properties of the beltline region anterials in a reactor. D e AVS? becana a requirement v in the early 1960's when it van detersised that sapificant differences ac-curred in ene neuersa embrittlement sensitivity between various steels and weld:aents used is reactor construction. Dese regulations are pus 11 shad in Appendiase C and I of 10 CTE $0.1.2 of major importance is the adjustment of the refereni.e oil ductility temperature (17,g ) as a function of fast neutron fluence (E > 1 MeV). This inforumtion will affect operating limitations with respect to normal heatup and cooldown procedures for the reactor vessel. Dese pressure-temperature limits are part of the Technical, Specifications imposed on the operating License required for each reactor.

De initial program, unich was desiped for Oconee. class reacttes, cor.risted of four holder tubes located is tse coolant inlet region between the thermal shield and the pressure vessel.3 Zach holder tube contained two capsules filled with specimens of pressure vessel anterials similar to those used is vessel fabrication. Also centained is each capsule were dostseters for sonitoring flux exposure. During first-cycle operation, sechanical probless with holder tubes were encountered in several reactors, which necessitated the removal of survgillance capsules free all reactors is the original program. De af fected plants are Arkansas **uelear he. Unit 1: Thade Mile 1= land L elear Sta don i Unit 1. Rancho Seco Nuelear Generating Station Unit 1. and Oconee Nuclear Station Units 1. 2. and 3. nose specimens, and others, were then added to the AVS? (with redesigned holder tubes) in th ee reactors scheduled for later startup - Three .411e Island Nucisar Station Unit 2. Cristal' Aiver Unit 3 Nuclear Oenerating Plant. and Davis-3 esse Nuclear Power Station 7ait 1. 3e-cause reactors were now operating wi'hout surte111ance capsules, an analytical deterstr.acion of fluence was required 1.e.'. fluence actunulation by the re-actor vessel in ene react.: had to be correlated vien ene fluence of :est

.g Babcock & Wilcox e . .6- .O

_. . , , - ~ . . .,

specimens irradiated in another reactor. In addi:ian, precic: ton of futu a fluence excosure is necessary for ecssoaent design and for es:an hsatsg .1:n-drawal senedules for surveillance capsules.

Prior to ths RV:;P a generic fluence value was calculated using the analytical procedure availaole at that time and published in 3X4-1010CA.* 3ecause of un-c:rtaintias is react r operating conditicas and the lack cf applicable experi-mental data to benchmark the analytical procedure, considerable conservaties was incorporated into this curve. *. ten surveillance capsules were removed iros five of the six operating 17*-FA reactors. fluence data becane ava114314 for comparison. B is provided an :psortunity to reduce tenserracism in :he design curva by better defining reactor cperating conditi:ns and refining t3e analytical .*1us calculations on walen fluence is based. Bis repor: Jescribes that effort.

By their very n4ture, generic design curves will be conservative for sost re-actors because they represent the acet adverse condi:1ons expected to occur in a class of reactors. Also. 'the fluence curre refers to the spatial loca-tion in the pressure vessel at vni:n the integrated flux (over vessel life) will be a nazimum. Bis carve is based on a fluence calc'ulation using cycle 1 reactor conditions that has bees extrapolated to equilibrium cycle reac:or conditions. D us, the two primary concerns are (1) the ability :s :alculate fluence for a known set of conditions and (2) the abili:y :o extrapolate : hat fluence to conditions existing in an equilibrium cycle. An additional problem is the predictabili:y of reactse creditions associated with an equilibrium cycle (defined as the reattor cycle that will predosinate over :ne vessel life-time of 32 effee:1ve full power years - ET7T). An evaluation of these uncer.

tainties wi11 he tacluded in this repor .

1-2 .

Babcock & Wilcoa

-- --- .- . . . . .. ..  ;.. x - -

l.

i

2. MA1.C1".M. .CC.

2.1.

• nor e a n t **

.. a1* a b l e s

~he analytical datorstsstion -af f ast 11.a C

• i .Me.* L : .a t r eac .e s : .s r e ac t a r pressure vessel can be considered sg :vo Sasi: 3r:J'esat 81) ca. ala*ica af

.e fast fl.s tha; esca;es : e :are and 'is attes.4 ':S s f .*.a t f;.x f rM

• r.

, e

• ore to the ressure vessel. Determisettan of flaente acds a: addit.3nal t 3e plicatis - the fluz anest be time-aversen over the irradia*ien pertad tc, pro-vide a value that, wnen sui:1 plied by ti.ne, will produce f hence.

For a given re'.stive pcver density (U 3) distr *.tution a=4 core co.:figura :co, fast flax that escapes tse core 1. direc:17 prnortional :s reactor pcver le.e1.

A1.c. for e given reactor core configurattoa. fuel 1:a41:3. saa power level, tse flax that escapes :Se core is a f usctics of L'D distri Jt1CS represented in three dimens1% s. 23 distributtes. is tars is a fust ion of fuel enri h-sent. lumped bur.able poiaan (13P). .:.t centrol rod locatisns. is S&J reae: cts a soeee core design La utilized whereta aca t31rd of the fuel i.ssestlies are replaced by fresh assemblies at the end of each f ael cycle. Thus, enrichsest Jiatribution (and therefore RFC d!Atribucian) variae f*ce one cycle to the sezt due to fuel managensat procedures. is additico.1.37 and fissionaale satorial 1

burnup and control too movement cause the UD to very during a fwel cycle.

Consequently, a detailed description of reactor operata=4 .cada:Lons 1.s required is onder to es*.culate as RFD representative of as irradiacian period. L*D is utilized la the particle transprt codes 30T and M1!.s as a fixed source (114-stos density) input which has been moras113ed to reactor power lesel. Because of this depender.co on re.ctor operating condi: tons. U3 distributica (and there-fore flua escaping the core) can very from one reactor to ar.other. ' owever.

d dif f ersaces are espected to be saml1 because of st= time operatiot.41 :eneigt.r.s.

7er a given es14tive fluz sp. :r.se. attenuation of f1wi between the core and

.he pressure vessel to pri=arily a far.c: ion af the c:ctir. ration of reactor internal cosponenta, their marecials of ::sstruction and tse c:olant densi:7

.g Babcock & Wilcox

Sisce relative flus spectrae at the core adge as deter 31 sea oy the : ore con-figuration and fuel type, all U?-TA reac: cts asve essentially :he sase spec.

a azitfag the react:r ::re. is addition, al 177-tA reactors save ne same sozinal configuration of internal cesponents and materials and operate with nearly the same coola:.: t es:p e ra ture s . D'as. flux attenuatics from core edge to pressure v<.ssel should be about the same for all 177-TA plants. Small differences any result free the possibility of cross mixing of fluxes is the M place (plan view enrough the core). ni. results frca U3 variation is the saisu:hal (9) direction (periphera11f around tne core) as well as in the radial (R) and ariat C) 41rse t ic.ts.

0*uviously, an at: urate calculation of tais attenuation is Jifficul: :castier-Log :ne approxisations required to s.sulate real geometr/ in a cc= pater coce.

Mcvever. if the attenuation calculation for a typi:a1 U D distribution has been benchmarked acd isolated f res etter ef f ec:s. the problen of calculating pressure vessel fluence wcult be primarily t'.at of calculating fast sluz es=

caping from the re..ctor core.

2.2. !ste tal t'e .c rie Wei !sro r eveme it A gensric desigs curve for fast fluence at the reactor pressure vessel, pub-11shed in 34W-101CCA*. was based on 1sformation and sethods availaole in early 1975. Considerable conserracian was intentionally included is this -

model becausa cf uncerr.aisties is reactor operating conditions ar.a the absence of any comparative experimental data. ne calculaticeal model consisted of coe-dimensicnal geometrv. which war then corrected to the peak axial and azi-matsal flux location using separately caleclatet factors. Se trassport code ANISE 1 , wtaich solves taa Beltraann transport code using the method of discrete ordinatee, wee used to calculate a maltigroup radial flux diatributiou frra the core througa che pressure vessel. Se model inc.luded a radial DD distribution for sa estimatad equilibrissa cycle, coolant water at 600F, and a 107. saf ety

• tactor.

Specific itema of concern were the assumptions that rsdial and arizachal ef-f acts were independent, the 600F coolant tesserature was too higs, and the estimations of UDs. De t 'allabt.lity of dosisecry data f rom cycle i surve11-1.nce copoules 'ed to the use of a :vo-dimensional.1-+ gec=etr-* x. del usir.g more realistic coolant tesperatures a=4 RfDs. A detailed desc?iption of this model is included is section 1.3. Bis apdated xdel per=itted a reduction is :he built-in c onservatise.

. Sabcock& Wdcox l

l l

/

^ - _ _ m M. A

O' y yi 4 W, x N

,~

7 Althougn not directly related to the generic destas :urre, ano:ser calculacian is required to defise ene bel:lise region. This requi:es a :vo-discsional model in !L-Z 3ecentry. The initial model" :entained 33sc of the censerva:is=s and deficiencies that existed in the generic :urve calculation.

2 .1. Ca te s!atinal edel 2.3.1. Configuration ,

~

A major ef fort in analytical codeling is the s1JEJ141'!on of actual three-di. Den- *

~

sional (3-D) shapes by using one- (l'-D) and two-d1 s'iscal (*-3) code geene- 7 tries. hiti-dizensional treat =snt and geometric detail are severely lisi:ed by available coeputer storage. In addition, material preparties enou12 se representative of average conditions during the irradia:Lon period. Although inlet coolant temperature Joss increase ac pcwer levels below full power ,

(*.2.5'T per IC: power change), the total time at a fractional pcver level role '

e ative to the length of an irradiation period is usually small, ari this ef fect can be ignored. Thus, approximations and trade-offs are required and are 4,s N

- 3 lected in such a way as to minizias any effect on fias; results.

The ANIS 35 and DOF partic.le transport codes were assa ior 1-3 and 2- cartu-14tions. Bothcodesusethediscreteordinatesmetaodsfsolutlenofthe Soltraana transport equation asd have multi-group and asymetric scattering ,

t

~

~

capability. The reference calculational model is an 1-4 gecescric repesien-tacion of a plan view through the reactor core taidplane and, includes care, core liner, coolanc, core barrel, thermal shield, od pressure vessel. Ose-eighth core syumetry is used based en structural confipration and 173 data (Figure 2-1 and Taole 2-1). Code parms-ters were limi;Q :o F:. order of seat-caring, Sg quadrature, and 2 ~eargy groups in order to include reasonable geo .

setric detail. Flus generation in the core wee ' re> resented by WU fissioc. '

diatribution, which the code dagivee from tne input 173 and a normalization Reflecte[ upper

~

factor to adjust flux level to the desired power density.

and lower boundary conditions are used to represent core and structural sym-metry. The right boundary (outside surface of pressure vessel) requires ad albedo boundary conditics to simulate flux leakage across char boundary 44' i.

to adequately descrtbe the perturt=ation caused by the :enerete prizat-r shield. -

. . n Energy-dependent aimedos were calculated free a t ,-0 model using :he A:fM N e de in cylindrical geometry vi:h ashr axis dimensions and 223 distributun, and includi:.s the pressure vessel cavity and primary i:ccrete snield regisos.

2-3 Babcock & Wilco a

e 4 j h

• s

+ w .w or ,4.+ N o-g we

• v * #f'**' "*******8 -h--8***I*'*' -

h % 9- p 4

f h 4

For those reference cal:ala:i:ns : hat did not require accurste fluxes aest

he edge of :he pressure vessel. the riant albedo could be replaced :y a va:-

uus cenditica that tended :o si=plify :al:ula:icnal effort.

The P: order of scattering does =ot adequately describe :he p:edosisately for-ward scattering of neutrons caserved is deep ;enetration of steel and water sedia. This necessi:a:ed cal:alation of a P /P: corre :icn fac:or using a 3

P3 . Sg. 22-group.

• 3 cylindrical ::ccel is ANISN vi:n major axis dimensione a:d UO. laterence model correction factors were deter =ised by calcu1J:ing

.he racia of fas flax (I

• 1.MeV) f t:a ene 1-D ::ndel to the 2-0 .odel. Sese valaes were cocaired as a fuse:1:n of racia; loca: ion a1:=g the u*:r axis (Table -2). Since :nese factors are primarily dependent en geo=etr*/ and u-terials be: vees the core and the caosule locacice. :he values should be appli-cable. :o all 177-TA reac: ors, and no sigsificant dependence on azimuthal (f) location is expected.

The Sg symetric quadra:ure has generally produced accurate results in dis-

- crate ordica:es colutions for stailar problems. No significant enange is fluzee was attained with the use of Sg quadrature. No further verifLeation of quadratr_re accuracy was made. .

• he 22-r,roup neutron zicroscopic cress sections represent the neutron ;ortion cf a raieldisg-ortee.ced. coupled, aS-group set. OLC-23/Cl.SK.? 21:5 :his se:

the faat neutron energy ra=ge (E > 1 Me7) .is represented by the first il groups plus a por. tion of group 12 (Table 2-3). . Macroscopic cross sections.

required ter transpor: analyses. were thes obtained with the sizing code S

• N . Noaisal composi:1cas were used for the structural metals. Coolast ccupositions were detersized frra the average boron concentration over a fuel cycle and the bull temperature of the region. The core region was a hc=cge-neous mixture of fuel fuel pins. structure, and coolanc. .

2.3.2. Pela:1se ? m r Sensity 31stributien Tor a gives power level the neutron flux escaping :he reac:or core boundar7 is -

direc:17 related to the 1P3 distributica is the core, e.g. as RP3 values are increasee near the core periphery. :he flux 1 creases proportionately. AP3 distributices is taree di w sicas must be c:nsidered. Ouring a :ael :ycle, fael 'urnup and control rod movemen: :e d to cause the H 3 tosary; be:veen fuel cycles the nevement of fuel asse.blies :auses variaticns. For the

.; Babcock & Wilcor e

V

- - _ _~ - . . . _ . _ _ , - _

v is surveillance cdel 1: :a : nven:e:: io 03:413 a ::. e-tverageo U D so :..at as average flus tan be cal:ulated 23 represes: this :ausiest behastor over as irradi. tion per;oc. ~5.a . t heace cas he cb:aised fr:s :he produe: of the average flux and the ti=4 interva'. ovier which 1: applies.

~ Pin-by-pis PJDs are generated is ri:1cality and fuel anage=ent analyses vnica have been demonstrated :o be accurate predic:1ons of actual Uts in

. operating rasc:mes.5 hus. RP3 values at 1.G-es intervals (pis pi::n) are sbtained for Mark 3 fuel assens11es. These data are utill:ed in :ne sur reil-lasce program as input to transport c^446 Ehsre UD velsen ate s.46eru.ily ccaverted ta fission decsi:1es and as s'ach, represent a flus Sour:e in the 1 i

calculaci:n model. Secause :ne rean free path (stp) f or fast neutrons is :ne l core reti:e is 10 :s 33 cs, only a.utzens born is :he sucer :.c reve of assey blies significantly cuntribute ta f twg escaping *ke core. Thus. arti:214: 4:=

tail is required near :he ccre periphery.

UD data are generated at t=out 3 to 12 points is time during a fuel cycle so snat the transient behavior cza be approximated with static calculati:ns. For transport model input, these data are tise-averaged to produce a si gle RPD that will represent the en:1re irradiation period of inte:est. The ave. raging method assumes 11 sear 173 variation with timet except whe.4 control rods are moved, and then a step enangs occurs. An example of this =ethod la presented in Appendix A. Conversion of R.PD data frem the IT spacial sesh used is cri -

icality analysse to that used is the surveillance ref erence model (K.J requires the use of the subsidiary code SCE::ZI.S.

The analy:1:21 procedure for a reference adel calculation is to use ene XY (or 26) RFD data (ti: e-avarraged) for a specific reactor irradiatioe period for which dosimetry data are available and then adjust the results (sec:Lon 2.3.5) to e*;uilibrium eycle cordt:f ons to ebcats generic flaence results.

. RFD data in IT geometry are available for all reactor fuel cycles for which the cycle time and enrichment distributica have been prederersined. 3ecause of the fuel shuffle patters generally followed - one third of :he fuel asse:- _

y blies are replaced by fresa fuel af:ar eaca fuel cycle - the third and suc-ceeding fuel cycles should be similar and are referred to as equilibr um cycles.

51sce the reference model calculati:n produces fluxes is enly the X and Y disensians, an axial correc:icn factor f.T ) is re*;uired :s adjust these

.3 Babcock & Wilcox A

L -----_.__.____.:___-- . . - , , ,. _ -.

? l i

values :o the axial location C disension) of interest. The n RPD data are all evaluated at an average axial iccation (q/4): = 1.0. S e axi.:1 R?D dia-tribution is pri=arily a function of control rod movemen: and is censidered

' to bn independent of the IT (or RS) data. An average axial :orrectica f ac:sr is evaluated for the axial location of interest that represents the axial .

shape of flux at the core edge (A;pendix 3).

• hat shape is assumed to apply throughnut the 19 calculation model, i.e., flux values at .11 RS loca:1:ns use the same avtal correction fae:or. Calcula:ed axial fac::rs for each re-actor analyzed to date are listed in Table 2-4, t

"Aen 1-3 model calculations (ANISN) are >erforned, :he input RPD distributien is obtained frt** the corresponding IT or (RS) data by using the values alcus .

a =ajor axis only. Ous,, A:!!N.zicula:cd flux a should cer*espend to these along a major axfs in a DOT R9,model.

2.3.3. 3eltline Keeion' Closely associated with the generic curEe calculation is the definitica of the beltline region in a reactor. For this purpose a 2-D R* calculaiional model ,

is useded to obtain the required isofluence lines. A:i=uthal dependence (9 dimension) is el'*mted by correcting calculated fluxes to the 9 locatica with maxim m flux. The methodology for the calcula:1onal model is essentially the same as that described is section 2.3.1 except for geo=etry. Radial bound-aries, which represent cylindrical surfaces in RZ geometry, relate to disen-sions along a sajor axis. The cavity and -the primary concrete shield regions external ta the pressure ver,mel are included to account for axial flux s: ream-ing. Because of computer storage limitations, two calculational models are necessary upper half of the core plus rela:ed internal structure and lower -

half of the core plus structure. Figure 2-2 is a schematic of an upper care medel; a st=flar model is used for the lower reactor region. Input RPD dis-tribution is for=ulated from the product of the 1-D radial R?D (major axis) and an estimated eouilibrium cycle aw*mt PEDM. Both RPD fiscributicas a.e time. averaged over an equilibriu: .. fuel cycle. .

CMculated flux along the midplane of the L*. model is normalized to the RS reference model af:er allowance for the reittive axial pcver at that 1: cati:n.

This flux distribution is then c:nverted :o fluence (nul:1plicacica by :ise) and plotted as lisiting isofluence values for various c:nsentrations of ;nos-phorum and copper.*

4

-6 Babcocit & Wilcox j '

A ,

1 NL y

+ s j

)

I 2.3... Caegule Model The presence of a surveillance capsule in a reacter per: urbs the flux measured

'oy the des *zecers. Therefore, ccuparison of experi= ental to calcula:ed flux requires that this effec: be determined. Since the capsu.e was not incluted is the 13 reference zedel because of computer storage li=1:2tions, it was neces-sar7 :o obtain a correc: 1on factor that could be applied to reference nadel data. Thus, a capsule perturbation factor. JCZ), was calculated to represent the relative flux change (as a func: ion of energy) that would escar ?! ;te capa sule were included in the reference model.

A ;-3 ca~ culattenal mouel (?;3, ;h-group) was f or=ulaced in XT aec=e:ry :o explicitly represent the survel' lance capsule (Figure 2-3). A1:nougn exact geenetry could not be represented in the XY ccord*nate system, boundaries were f allowed reasonably well, and aross sectional areas of capsule =aterials were equated to real areas. A so.ur:e was input into a 1-co-thick ::re barrel re-gian that reproduced thes sane relative flux energy spec: rum at :he ou:er edge of the thermal shield region that existed in a 29 referen:e nocel calculation.

This procedurs was nece:ssi:sted by code limitations (no provision for angular

.'12x input). Inclustos of a separation distance fres the ther:41 snield outer surface to the core 54rrel was required to gene: ate the approcriate angular ,

flux frem the isotropic source input. Several iterations were required to crea:e the correct spec: rum; :he oatput spectrun from one calculat*cn was used to adjust the source input to a suc:eeding calculation. ~he capsule

ross section Ln Figure :-) is representative of the Charpy specimen regica (Figure -4). Any flux diff erences that might be caused by using a tensile spec:.aen region were considered small.

Two calculations were made - one with the Figure -3 cenfiguration and ene with water substica:ed for the capsule materials. The ratio of the average flux in the specimen region from the vo calculations was defined as J(E).

Ta:: ors vern talculated for energy ranges consis:ent with dosi=e:er reacti:ns and reported fluxes of interest (Table :-5). A f actor greater than one indi-ertes that the capsule nacer*als do not remove (by devnsca::eri.73 and absorp-tien) as sany neutrons as an equivalent water regten. The sv=retric 1:ca:::=

of dosisaters at the outer edge of the specinen region led to :he use of afeci-sen region average !!ux as :na :al:alation :rt:erien. *o verify :nis seies:.:n.

I

Babcock & WHeox I

l t

s fac: rs were also calcula:cd using the average flux is tse Al filler region anc resul:s were essencial17 the same.

Ccaparison of cosimeter seasurements to calculated. fluxes requires that a ref-erence location be selected. The dosi=e:ers were s p erically located vi:h respect to :he center of the capsula (plan view). Thus, an average of the feur fosimeter measurements should correspond to the average flux is the speci-nes regica. Point (or sesh interval) values are =cre conveniently obtaimed

'ftra ecde calculations. Data from the emptule mdel calculation andicated that :he flux value located at the center of tha capsule was tha sa a as the srecimes regicn average value. Therefore, all c:=parisons of =easured to cal-c . lated data will be based ca values calculat:1 at :ha loca:1:n of the :;sule ce ter.

2.3.5. Fluence Extraeolation The prediction of fluence for future fuel cycles vi:h ksown power distributions (fuel distributiens and burnups) is of primary concern. In a III-TA reactor the transport of fast neutron flux from the care edge to the pressure vessel is solely dependent on relative spectru:n at the core edge, structural ccepo-ser.t geeseery, and materials of construction. Since all 177-TA plants are quite similar in these respects, the ratio of fast flux at :he pressure vessel (37,) to fast fluz that escapes the core (s c,) should be constant. Thus, the followiss procedure is proposed for predicting pressure vessel flux (and c:: .e-fore fluence) prior to performing a complete analysis.

Spatially averagej PDQ fast flux (E > 1.85 e7) in the core liner is a quasti:7 readily available from criticality calculations, which are included in the fuel

=nagement analyses required for all operating plants. These data are censid-ered to be proportional to the fast flux which escapes the core and ulti=ately reaches the pressure vessel. Values from each time step are weighted in order .

to obtain a time averaged flux that can bs used te calcalace fluence. These

' values were observed to increase up to 201 from beginning- to end-of-cycle depending on control rod sanipulation. Pressure vessel time-averaged flux cas then be predicted by

'eeIAI =4 pv(3) e pv(A) = a,,(5)

.a Babcock & Wilcox i

r 1

where A and 3 represent fluence periods which are generally defined for a specific reactor and a specific fuel cycle.

Since our interest is in fluence (or fluz) at the mard == location, as:mut::41 flux variations are important. However, implicit in chia procedure (use of spatially averaged flux) is the assumption that significant variaticas is azinuthal flux peaking do not occur between fuel cycles. To date, sufficient experience has not been obtained to evaluate this assumption al:houga indica-ticus are that it is cor:ect.

Available fium data are listed in Table 2-6 for 177-TA reactor plants vnica are presently operating witnout surveillance capsules. Concarison of ;re-di:ted-to-nor M ia.J .alues f Ga a sur reillance capsule analysis indi:ates good agreement. However, capsule data are sparse and, in general, represent similar conditions (cycle 1 condi:icus) A more rigorous comparison using cycle 2 and 3 results is censidertd necessary to verify this procedure for extra.sla*.ing fluzes from one reac: r cycle to another. However. :he proce-dure ' toes have a good theoretical basis and comparisons to reference .acdel calculations for MO-1. cycles 3 and 6 show good agreement.

For ec3parative purposes, pressure vessel maz1=ua flux (cycle averaged) has been calculated for all reactor cycles for which P00 analyses are availaole (Table 2-6). Note the similarity between reac: ors for assumed equilibrius cycles (cycle 3 and later). Fluxas for Sr 3 are somewnat higher prt=arily be-cause of its higher power level (2772 ISit). MO-1 (cycles 4 5. and 61 1.s a apecial case because Arkansas Power E Light has recently sade a decision to u e an 18-month L3P fuel cycle. Othe; utilities are contemplating such a move but none is def taite as yet.

_2_. a . yiture I:rprovemem

. The analysis procedure described hereis was developed duries the course of analyzing the initial group of out reillance capsules. Since analytical zodel development is a continuing process, a nuncer of areas where technAques ceuld be improved have been noted.

Of :he items unfer consideration na;or emphasis should ha given :o the use of the code :XTI 17. ah improved version of ::C". Improved data storage, re-star: capabilities. ::nvargence, and variabi*ity of quadrature and dinensions should perzi: P3 scattering detail and explicit capsule geometry :o se 29 Babcock & Wile ,x i l

, l w - , - , - , - +4

incorporated into the reference model. This would eliminate the need for those correction factors csed is the present sethod.

Other improvements, judged to have lesser individual significance are:

1. Use of component dime:sions that exist at reactor operating corditiens as opposed to " col 6** cimensions.

2. Calculation of axial flux shape that exists at the capsule and pressure vessel locations as c7 posed to us1=g the relative power distribution in peripheral fual assemblies.

3. Determination of the " lux distribution within a capsuis to better initne specific dosimeter and/or specimen exposure.

4 Correctica of dosiseter measurements f or attenuation of fl.ax through dest-aeter tune and wire holder wall thicknesses.

5. Rafinement of spectral shape use.1 for flux veighting of dosisater capture cross t.ecetans, l

_tg Babcock & Wilcox 4

I

=

4

I I

I

• able 2-1. laferecce Cale21stion Fedel l Outer radial dimension along Cosconent Material maior axis, es ,

Core Homogenized sizture of UO2 fuel. Or-4 cladding. 163.79 H 0 coolant, and Incenal structure Core liner SS 304 163.497 3ypses ceolant  !!:0 (100 ;; 5) 3 3do'F and 2:00 pai . 9' 3 Core barrel SS 304 134.15 Inlet coolant (a) H 1 0 (500 pps 5) a 555'F and 2000 psi '36*69 Thermal shield SS 304 191.77 Inlet coolant (a) H 2 O (500 pps 5) 3 ,'

555'F and 2000, psi 6*694 Pressure vessel (b) A308 Class 2 steel 238.444 (4) Full power condition.

(b)55 304 vessel cladding is included as an integral part of the pressure vessel, minimum thickness = 0.J17 cm.

-i 5

-11 Babcock & '#11cc 4 ,

s

. - . , - . _ _ . - , 7

g _ 4E d- 4A4 4-A - -

s .

i,.

i ,.

Table 2-2. P3 /?! Ceder of Scatterist Correction to Reference Model Cal:ulati:?.s Pressure

' Energy range, Capsule vessel MeV locacica vall

>0.1 1.22 1.29 30.5 1.2J -

- 31.0 1.23 l'.29

>2.3 1.23 -

52.5 til) -

e f

T 7

, e

.;; Sabcock & Wilcox i .- .

1 4

0

"\

_ , . .__. - _ _ ., . . _ - - ,;.,,,.,_ , ,~r,.,.~,-- _ . . , . . . . . ~ . . . . .

Table'2-1. fr.erty Stricrure of Cross fecet:n fet l Upper Euergy Average (#

ene r gy , width.. Latharry energy.

Croup MeV MeV vidth MeV 1 15 2.8 3.20 13.5 2 12.2 2.2 3.00 .'

• 1.1 I 3 10 1.82 0.20 9.05 i

i 4 S.18 1.32 0,25 7.19 5 6.36 1.40 0.25 5.60 5 4.96 0.90 0.20 4.45 7 4.06 1.05 0.30 3.50 8 3.01 0.55 G.20 ,, 2.73 9 2.46 0.11 0.05 2.42 10 2.35 0.52 0.25 2.C8 11 ,

1.83 0.72 0.50 1.42 1 1. 11 0.56 0.70 0.78 i 13 0.55 0.439 1.60 0.25 14 0.111 0.107 3.50 *1.93(-2) i 13 3.35(-3)IDI 2.77(-3) 1.75 1.41(-3) 16 5.83(-4) 4.8:(-4) 1.75 2. 44 (-4) 17 1.01(-4) 7.2(-5) 1. 23 5.a5(-5) 18 2.7(-3) 1.83(-5) 1.00 1.76(-5) 19 1.07(-5) 7.64(-6) 1.25 5.74(-6) 20 3.06(-6) 1.94(-6) 1.00 1.96(-6) 21.c l' 12(-6)

. 7. 06 (. .') 1.00 6.83(-7) 72 4.14(-7) 4.14(-7) -- --

(a)3ased on 1/E flus variation over the energy group.

(b)saad as 3.35 = 10*3 2-13 Babec<k & '#ilcox f

t f

4 Table 2-4 Axial Correction Factor to ReIerence Model calculati.-n Axial correction f setoc Pressure Capsule vessel Reector_ Cvele location # vallib)

Ceonee 1 1 1.1(# 1.1(#'

1.1I ' 1.1(#I Ccones 2 1 1.14 1.17 cconec 3 1 1.17 1.19 3I-1 1 1.1(c) 1. J (0' ANO-1 1 1.11 1.13 (a) Averaged over capsule length for that portf ors of cycle while'cagoule was in reactor.

.hmtamon value.

("' Estimated.

+

Table 2-5. Surveillance Capsule Flux Pcrturbation 7 acto:s Energy range,I MeV '4 Dostmeter raection J(E)

>0.1 - 1.90 123

i. >0.5 2373p(3,g)137Cs 1.81 8M

• 1. 0 23SU(n.f)l37Ca 1.71 / #7

>2.3 . 58r.(a,p)5aco 1.46 / 84 .

32.5 5**Fe(n.p)563, g,;$ y, ? f

'USes* Appendix C sor deter ination of ef fective energy range for dosimeter reactions.

I

.y Babcock & Wilcox l ~ l A

i

,3 Tabla *-6. Oata for Pressure 7essel Flue:ce Predi::i u

• n 177-FA teacters 7 n at peessure vessel surface.

Spatial and ti,e 3, fg , ; y,yj 3

averased flu 7,,g (I > 1.85 e7; in 7ahs from :tornalized talas 88F8 *T' )

zeactor -evele 8:e 'e: #8C1 fres :ansule a-alrsts ccoces 1 1 A.37(10) 2 7.05(13)Id) 2.03(10) 2.08(10) 4 5.81(13) 1.68(10) -

Cecnee 2 1 4.3 2(13 ) t al'. 1.37(13) 2 5.60(13) 1.6;(13) -

3 6.13(13) 1.*7(10) -

4 6.13(13) 1.77(10) --

Oconee 3 1 4.73(131 1.36(10) 1.39(10) 2 5.68 (13) 1.64(10) -

3 6.11(13) 1.79(10) -

DfI-1 1 4.'82(13) 1.39(10) 1. U(10) 2 5.48(13) 1.58(10) -

3 5.60(13) 1.61(10) -

4 5.93(13) . 1.71(10) -

A30 I 3 6.10(13) 1.79(10) . 74(10)Id 4 (L37) 3.96(13) 1.14(10) --

5 (L3P) 4. 45 (C) 1.23(10) -

6 (13P) 4.37(13) 1.26(10) 1. 3 2 (*.0) (*

• SMCD 2 6.83(13) 1.99(10) -

3 6.56(13) 1.89(10) -

I*I Cses Oconee 2. cycia 1. as reference valsa.

fertial cycle prior to removal of surveillance capsula.

(c4 Dire.it calculation using ref erence analytical odel and ap sule ..t..: sal-Laatiou factor from cycle 1 caalyses.

(d)1and as 7.05 m 1013

-13 84ococw & Wikox 1

t u

db d

9#

1. .a.

. as 4,'gio *t < %s* >c.> g5 aC 1'

ts**,4* 2

't .#

1

/ f 4 9".t. co4.f.i 4 t **S ,,

u3)

-c e

, ,f wo e *

g# . g to 3" C" C

,,s** O I C.

w IC 4 .'! . s'- 'A ' I l=

a3 -

1. L .!-

50 ,G l I i l l 5. St I g 2 1 2 3%

3a '

d4 d = I=

- , - i g o 4

3

---y ----

i 1__ .

ac 4 i .

4 a e Ia

>; I I -

I -

• l-ew I I

. 3 i 1 a .= .

-- L- y -& _._I.4 I

1 o l I'7 7 I  : 5 l l!

12

" C l , 1 l= --m---+-- dg.

I 1

" I I l ~=

4 l l l ==

i 1

A

-.+.

• 4 l I

I I i l

---+--

i 1

o

/ '_-

> C C ll.16 Babcock & Wilcox I

I

- - . f .. . . . .

4 r

l 1

I m

us- .nn +

e,. , ,1em u su I

• i m

• 5e ,

1 O i '

1-u s

~2

.. g tseest einseesJ v

1-z.a3 e 3 1**T**3 *atst a-I .2 1

- n e e. c eu.:au; D 3

.u, ..e G 1**T**) esensei t2%~

~

a Ql *@

22e l

fw . Asuti esos  :

7"> m

~

1-w .a.

e 3-

~

2

= - a g

• s g . .= ,

p ~m *

~

}

1*

e e 3

=

e

=

3 a

e

=

s.

=

1

]

w 4

7

~ .

3 "5

]e r =. . . -

3 w a.

.e. - -

e Is .m. n 3

3 - .e .e e

e - ,e z i

I - e e a. .

. e g ' ' - m 4

. 2 a-a n m J J

= 1 w . .

3 1 5 e a e .e.

e e .e., Il e a m. a m e. .e

'estt2*se.o e2es N '

u 'runas 7****t;*s O s

2.p Babcock & Wilcos h

4

I

a. - .nn. 3 i

• I 3 1 3 i *;

e. .

3 I i

. I

• l i g  !

I C E  !

I

.5 E d Q -

0i I ( ) esenns peto I o 3t O I . *

"; % 1

~

9 A

3e

~3 0

i [ ',

l 2 I .

at I

• b <

, t a i a

a i

I j $'a 3 & -

.i h I C  %

4 I .

E i I

PISTU WC .

I j santoo3 ** tut

. unpr.not possatP4 o

.y Babcock & Wilcos

^ - -_ a i

i  :

-b

i l

1 l

• 1

)

'e u .

, ,eJ .i Ue,

• %. :d

• e r. <,

, o :- .

g. t >mt )-

.a

.+E

<1 m -

l

• 9< +s ll $hk D i J.

t olo ,

1. $ {

l ^J:f<

,: . !- .~ k. . :

%d,

.r .. .3, l

, ':p:w:;w' e '

s

-g g . ^:

'.: _: ve/ ',.

=  ;

g g

• s;. ~ :3 2

1>.

• W *l#

j  %,

1.t w

g. 1...rg.! 4

< g e

4 ,

a g

• :s

- s P 3.m .

s  :.Li

.2 e L ,h ;J T

A; : .e i e .4 *o vi 2

s :h ;e

• I- $ r U l .

w 1 L. ._.e!I.1

} '

iC,, / I 3 l .

?.Ct. . 'b

~

- =

+1gp-l-t a 's.-

2: ,

. -m -

4 .ijN d T t el Miff e.,t,d.

'u f.l  !

tl

.g. 1 . g II GN' a.'e s

'. ,3V -

b

..s.. , . ,

j  : ,

e >

, ..w .

I E

"O*p

'M _. . _ . .

2 13 Babcack & Wilco j l

l l

l l

b l

~-

t BLANK FRAME

~

FOR i

PROPER PAGINATION P

4 4

t

. - , . - . - . _ . ..,_..,,.____..,n-. _--,,.,.,.___,c-,. , , _ , . _ , _,,.. -,.,,,, ,,-.,._.-.._.n..,__,,..,. . , ..

3. EXPI2DtIX;A!. 3 ASIS 3.1. Sosimeters Dosimetry for long tera f ast fluence seasurements taquises 66c th. i.4e=&

sacerial have a sigsi:.icant ca;ture cross section over :he appropriate acer n tan t a. *hia is necessary so t'.at suff*:ien reactions oc:ur vi:h neutre".3 's .

the energy range a: interest. In asat: ton, a reactAon procuct wita a lang half-life is desirable, preferably on the order of the irradiation period or greater. This insures that the dosAnecer activity is a record of reactions that have occurred over the entire irradiation period. If short-lived prod-ucts were utilized, reactions tha: occur early in the irradiauon period voi.1d decaytothepointthatthemeasuredactivitywouldrepresentoniytheflux that existed during the latter part of the ti.ie period. Cbviously, for a cen-stant flux irradiasees, pooduct half-life woul1 not affect the analye,is; but that condition rarely exists in co=mercial reactor' operation.

For pressure vessel surreillance, dosinater reactions of interest are liated in Table 3-1. However, c Lly the fission reactions in # p N and 133'J satisfy both energy and longevity requirements. The 5'31 and I'Fe dosimeters are use-ful for supporting data with appropriate ad;ustment for energy range and :i=e-averaged fluz. *he HCo dosimetert are used to esti= ate :stal and :.iersal fluz.

The target materials are %0 mil diameter wires of high purity materials whi:n are contained individually is holder tubes of Al or Cd-Ag alloy. These holder t:.bes are asaeabled end-to-e:.a is a dasimeter tune (yigurw 3-1). tou: identi-

, cal dosimeter tubes are contained in each surveillance capsule (see'tica 2.3..).

3ecause of their synastrical location relative to the specimens, average data were considered representative of the interiot region of the capsole. Also.

the elevation differences (two dosimeter tubes in the upper section and two is the icver section) were considered insignificant because of the relatirely flat salal flux shape at :he capsula elevation. As part of the integrated sutreil-Isace progras, capsules and dosimeters are being redesigned to provide addi-ticaal target anterials :or 5et:er spectral analysis and redundancy.

..e 3.t Babcock & Wilcox J ..

}

. t 1.2. 'Sost=eter Activation 1.2.1. Measure:ent Procedure Af ter a prescribed irradiation period. capsules were re=oved f rom-4 reactor and transported to the Lynchburg Research Center (LAC) for disassembly. Oo-siseter wires were mually removed and t.:en counced with a Ce-Li detector connected to a multi-channel analyzer. arget :*.aterial vei, tats were cbtained from wire weights and associated chemi:as and isotopic =cepositions. Mea-sured activities were reported as LC1/3 of tarde: =ateria*.

Because wire dosimeters are flax accumulators, to distinction is =ade with respect to the- ea.er:7 of tae ittat ient flut *r to !N rate-af 9t ?ry of em$ure.

Thus, a signifJcant a=alytical effor: is required to define spectral shape of the flux acc its magnitude as a function of tise.

3.2.2. Analf t teal Procedura To provide a basis for comparison of sessured-to-analytical data, calculated flux was converted to activity. This is accomplished vi:n the following equa-tion.

r .

- yre

! -ag(7- 2)

?

o = o. , y1o.).oE o (Z"<E> y r, p - e-1.t

- 1 "h*'*

D = dosiaster activity. 3C1/3 of carget material n.

f = fission yield of target as:erial n (* 1 for non-fission materials).

N = Avogadro's number.

M = acotic weight of target asterial n. grams, c(E) = stoup averaged Meroscopic activation cross section for material n. b,a an/ctos.

(E) = E(Ele (1) = asutron flux obtained f rom reference model r.siculation with appropriate correction fac:sts, n/cm 2-s.

K(E) = ::.:stned fie:s correc:Lon factor = Fj/?;

• F, a J(E).

s(E) = neutron flux fres reference model calculation, n/es:-s.

, F = fraction of reaccor fall power during jth time intervs1.

(*Jacayconstanttor tactope 1. s*I.

c) = ::me interval for irradia*.ica period j. s.

3-: Babcock & Wilcom

c T=:Je4* denaar :. 4 ft:2 reac::: s art.p :: a h::; .T. for capsula resoval. s.

) = time interval frca rea::or startup to end of jth irradiation period, s.

For each reac:or the flux integral. [g :(E)s(Z). was calculated in the refer-encemodelas[.3(I)HE) 4 and then nul:1 plied by the appropriate ccerectica fae: ors as follows:

K(E) [ c(Elt(I) = [g o(E)K(1)f(Z)=[g oL&)ott).

• he ac,1vation cross see:1on :(E), were *otained by flux we:gnting ENDF/31**

sata over a fission spectrus f or ne required group structs.re t see Appenoix 31.

Although initial weignting vaa based on a fission spectrum in *ieu of kaculedge of the actual spectrus seen*' by the dosimeters, a later c:=parisen of calcu-la:ed spectra was nade (Table 3-2). The nec effect on dosi:eter activity of cross section over the various spectra, a*: hough not determined quantitatively.

was considered sasil based on a similar unreported ccuparisen. Of course.

there is no guarantee that ene calculated spectra are correct even though the calculation techniques are considered theoretically correct. Experinental spectrus determination was not feasible because of the prev::usly noted lack of long-lived doeineter reactions available.

, [ -i gt -%gC-?)

The power integral. ; T 1-e e

• is a measure of Laocope decay prior to reactor shutdown. (No al'levance was sede for subsequent decay be-cause sessured data vere reported for zero ti=e at:er shutdown). Fractional power traces. ootained frcs w1ine 1*strumentatico, were input CJ an aux-iliary computer code for each isotope of interest. Thus a power integral was generated for each isotope over each reactor history. 1 p11:1: in thia p*ssedure is the assumption that relative power density does not vary with power level.

3.2.3. Comoartson of Sata in arder to benchmarx calculated f *.uxes and to verify the analyti:a! procedure used to determine flux distribution and spectra. neasured costmeter activi:ies were compared :o calculated activities. 1splicit in this prosecure is :he as-sunction :nat seasuree-to-calculated ac: vity ratios are ecuivalent :s !;ux ratios. This can be snewn to se ::ae if :ne :alculated flu spe::rus :s 33 Babcock & Wilcos f

identical to the actual spectrum seen by the desi=eter *:see Appendix E). On this basis the surseillance capsule dosimetry data were correlated. t~hese data have previously been published for esca capsule analysis in references 11 through 16).

Because of their effective energy range and long half-life product, tne 237.;p (n. f) l 37Cs and 238g(n,g}137Cs reacticas were coesicerad to best repre-sest fast flus to the surteillance capsules averaged cver the irradiation -

paricd. ha corrasponding, normalizacica fa::::: (-*

• su-ai ' a-"1 M " *d d t n ratio) are listed in Table 3-3 for each reactor. The data fall within a nar-row band of 3.99 :o 1.17, which suggests that :alculated flux .ay be sceevnac lov. Ilovever. it should be notad that all are within the uncertainty h nd (see section 4) although a bias does occur. Considering ene many variables in the analytical procedure, the comparison is quite g cd and is considered to be a verification of the analytical procedure for calculating fluxes in

- the capsule. By inference, calculated fluxes in the pressure vessel should have about the same accuracy. For conservatisa, a normalization factor of 1.1 was selected to apply to all calculated fluxes.

Ancillary dosimetry data are listed in Table 3-4 which generally support the observations above. !!owever, both the I*Te(n.p)5**n and I3Ni(n.p)3ICo re-actions were shown to have normalization factors consisten.ly less enan 1.0 (overprediction of flus by calculational procedure). Although no explanation presently asists for this phenonenon, it is suspected that the calculated flus spectrum is overpredicted in the energy range greater than de MeV. If such a condition does exist. any effect on 23"Np(n.f) and '38;(n.f) reactions would be imch less. $2: and 7: respectively, because relatively more of those reactions occur at lower energy. Any effect on integrated fast flux (or flu-

! ence) would be similarly diminished because of the relatively easil fraction of fluz with energy greater thaa %4 Me?. Consequently. the importance of .

these data are discounted with respect to normalizatica of calculated fast flux. However, the above discussion does illustrate the danger of directly applying doeinstry data to other energy ranges. Should the specimen evalua-tion procedure be altered in the future to include an energy dependent danage analysis, spectral shape of the flas will beccee more i_portant and such sus-pected discrepancies shou,1d_ ,be recenciled. In addi:1:n. :he imoortance of

' oconee 1 data was discounted because of undefined probless that o::urres in the seasurement of cycle 1 dosimeters.

3., Sabcock & Wilcom 1

Jane versit:staen frer :se precia:: n af flaa s a:1 art:y :e .een ;**-TA rea ::rs (sectisa 2.1) for similar irradiation periods (fuel :yclast is :staines v :.=.

. partsg activity sessurecents with the pcwor integral divides Jut. *?is rem 1ves

' tne ef fects of reac:or operating hiatory and length of the irradiation ;eried free the activity. Thus, a siguificant variation is measured activity was re-duced :o an also c constant value Cable 3-5) vnich Laat:a:es :.a: tae fas:

fluz was essentially the same is esca reactor. De variatian is pcVer ante-dral values is indicative of different operating hist: ries and irractatt:n size s. Calculated fast fluz was also essentially :onstant setween reac ars.

Analytical models differed only in the input pcves diatributtans vei:h were ai ilar for cycle 1 La esca teactor.

4 I.

33 8abcock & Wilcos es wa.o. -g

-g mn- .v .s

l 7:5:e ? *. Test eter Sea:tt:-s o.s Target Half. Ef f ective ene r,;y "'

safertal

• tsaetten produer life (b) rinte. Me7 2 37.gp id) a,f 137Ca *) 30.03 y *0.5 '

(

13tg(fl n.f 13?C s

• 30.03 y =1.1 88N1 III o,p tag, pg, j g ..,)

1*TeIhI n.p 8*Ma 312.6 4 82.5 IA HCo n,y 88Co 3..to y totaA UoNI a., 4 04 5.;ta y * . 3 e '.*

'All asterials escept I*Te and one of the.IICJ sre contaiaed is C4-A4 holder tubes to eliminate thermal neutres fl a.

See refores:e 17.

I"I 3ee Append 1.2 C.

'1.44 weight ". Np,100 taetopic.

• Although all f Lseien products are produced III*a 6 is of prisery importance because of its hash yield and tons half-life.

III l0.34 wetast : ," 99.27: 1sotopic.

I8I l00 weisat : 31, 67.77: 1sotopic.

(h)100 wetshe : Te, 5.82: 1sotopic.

III .36 O weisst : Co 100: 1setopic.

i 1

'l 1

36 Sabcock & Wilcot l I

2 I

.y 1

\ ,

Tale 3 Ce-eartoon af telative F'.y Icectes fpectra Normalised ~'

es E

• I wev f leser eseray la M:0 at Capsule

j. CrsJe bound. "e4* F!setsn escoule locaeten eec.t e r ** *a ll  ?'a 1

j. I 12.2 3.0002 3.0014 0.0011 3.001s 0.0313 L" 2 13 3.0013 0.0064 0.0054 0.0064 0.0053 I

3 a.11 3.scs: 0.018 0.013 0.013 0.014 l

[ '

j 4 6.14 0.021 0.010 0.341 0.047 0.033 5 s.1e 0 ;:1 3.;9: 0.074 0.38' 1.3e0 [

6 a . P+ s 4.32 0.174 0.06) *). 36 ) 3.3 7

? 3.31 0.139 0.118 - 0.098 0.091 0.37$

8 2.46 0.13: 0.122 0.109 0.106 0.092 9 2.35 3.334 0.039 0.037 0.03a 0.331 13 1.63 0.178 0.132 0.162 0.133 0.1J) 11 1.11 0.323 0.218 C.334 0.3 0 0.J8:

12'*I 1.0 0.064 0.046 0.060 0.066 3.399 l r I*I Fattial group.

l i

f i

py SONOCh & wilco 2 l

j*'

. . . , , .4 . - . .

I I &

L__ __________________f1_________________

f.

Taole 3-3. 3ormalazatton Factars fer 041:ulate:

F'.su in f arreill4 ce Ca?sules bos'. aster activity.

• C '* a Dosisater -- Narzalization(al Neactor Cvele ees:tten Messared W ag, factor l . Ocosee 1 1 33*Np(s.f;;3*C. (b) 3.19 --

21sgg,,g3:A? . (g) o,,4 l cconee 1 1. 2175p(n.fli3*cs 9.): S.50 1.36 i 21sgg3,ggtstce 1.44 1.72 1.13 l Oconee 2 1 23'Jp(s.f) 3*Cs 6.73 ,1.31 1.13 ,

138C (e.fil37Cs 1.21 1.22 . 0. 79 l

Oconee 3 1 2 3 ?3 ,(, 3 11?C4 5.42 ..o) 1.17 21ag(,,g}11?C4 1.05 0.94 1.1 TMI-1 1 21ty,3,;)t37Cs 7.23 4.40 1.11 23eU(s.fil37Cs 1.41 1.:^. 1.13

! ANo-1' 1 1375p(n.fil370s 3.20 4.39 1.i?

l 23 eggs,g}t17Cs 0.97 0.93 1.04 Normalization factst = esasured activity /calculeted activitp.

I I Measured values were Laconsistent with sta114e data as1, the:Jfore.

I were coseidered to be tacorrect.

I

\

d 33 Brecock & Wilcos 4

k

. s t

I

T b

(

' ii :35::

l, . fl=

.5 n,J : : u_

i

1

.. 2 i Si -

91

• t -

3.l Jl .

b**

't '.1 . S 2 " 3 ::

! j{ %21 r.e' 'd44: 4 tl S 1 g

+ ~. '

3 .

j~ 1 ) t

~ I, # <

8 ". O 2 g 4'

1 3J. 'S I J .: J

(

~jl.' "si 4*t

' kl s t l

I Ji (" d aI *1 3l I

1. J l t3: sz$s

$ =385 4 4 i d 4 i

^!s; 6

Ila/jI '

l i t i t$232 ( .

t +

3 44444s i <

gg

s.

!/

= '.= - - - t f.

i.

t n -

l ,

i 1 111..6 - '  !

ddi1 53 1

l 4 t

i

• M Sabceck & Wilces 4

h*

E

(, <

(t- .e

,( + 1

. - r+ A=.pr. :

' u. )

(

I a

Tatie 3-1. Aestittv ses ?!ws caeartsen 3etween Sea:tsrs t"

g Calculates "3 Measured ,, flu ts acttstty. Power 1

• FI. capsiale.

9/es 2 -e' tatetral - '. C1/t Reactor M .C1/1 oconee 2 1 6. M 0.3:64  !!0 2. ad e,10 s oconee 3 1 a..! 3.021e :33 .4Ji10) 31-1 1 T.23 0.0 93 247 2.6*(10)

A.%= 1 ~1 S.20 0.0 04 253 2.4a(131 ,

. w-

'4) Average of f aut destanters f or 3*Np(n.fis!.04.

(b)E a F1 = esasured activity' power Latogral.

f1 .*I I * (3.1 e it.-;s

• F1 = flas 1stegral = sgtis(t).

g (c) Average value for 1 > 1 Mef.

1 i \

3.g 3 Satcock & Wilcos .

( i W,

n. k L k

n _ - - _ _ _ _ _ _ . - - _ _ - _ - - _ _ _ _ _ _ . _ .__-_ __

. 1 J

r

.e; c. s.: .

.y-4 -

t 1

5 '3 I

!!.If!!'...i.....i.

..;... .i.

1 Ir jd i/

s Le l . 1y: i 7 ,U1 I h'i r U.,;J  :

p.,.!!! 1 1:s 1.f:j :Isl,

m. .

a t 3p H

re slJ.!!,;s::

,1 '

s :,' L.q '

w - t.! .i ,I r.li:j p -

4,1 _

,a ,4i:upa:.au u . .1.1. i

. 2 t

3ie,,-,.d 4

s;::4,11:;, , , 0 4i,,311l::q::,-4:44,4 3 ,I I 81,8'118'!'s8!'if8'!'t 0 ?; l i li -i H l:i: 18:l:PH+1 i 4

3 g 4 in l

. . 11 1  ! is i Ii!4 4 3

t a!D}"i ih .e 9 A

. ,/

• l

=

! *dj!,,I E

83 g E

- . h,4 .1

'j F-7 -

. l

..n n.

Ilil  !

NiA ka l

!d2 a L. ,

E *;'  ?

= g p'l- li. . Y.r '

E s?'.' . /@ .

5

%,e  ! . .

ll '

n l ej:

p -

i .,

I .

Y  ! ll ,1 I >

I  !

'  !;l!

a I/ O a J' Lg l

1

-- i 7

1 1

4  !

1 3.g. Bat. cock & W$ cor l

l l

j I

)

l

\

t

.,,,,,32 a - _ - - - - - --- - - - - - - - - - -

l l

l

_. .W

BLANK FRAME t 4

FOR i

PROPER PAGINATION i

I i

I

. g ..

[ 4*

1: -

m)

Y t ,

i 4

4 ANALYSIS OT OATA 'OICIETAl';TT

. - t .:

Flux values determined in the surveillance program are sub,'ect to un:ertain-ties celated to dostanter activation measurements, calculated conversion of

. activity to flux, and prediction methods for extending data to future reactor operation. A recant study eas perforned at 35W =a define uccer:ain:v lic.1:a e

fm neasured and calculateJ flux dataI8 and those results are su=arized in this secticr.

4.1. Survet*1ance Caosuf Analysis '

Ranges of data uncertainties for the various phases of this analysis are list-ed in Table 4-1. Limit values for seasured ac:1vities were calculated in ref-erence 18 and include components for dosimeter wire weighing, compost:1on of target material in the wire, ga m a counting statistics, instrumentation cali-

. beation, and possible misalignment of the capsula holder tub'e in a reactor.

Major contributors to the overall uncertainty 11mi:4 were calibration anc po-sitioning errors. Detector calibration was :M and the capsule position was assuaset to be withis =0.25 inches of its desig ated location. Limi: values for the 2173p(n.f) reaction emceeded other reaction data due to greater uncer-tainty of the composition of 237 Np in the desiaster wire.

Evaluation of uncertainty limits related :o calculated activities includea considerationofs.chcomponentsasapprow*mationsisthetranspr[theo.7 andel, microsecpic cautrca cross sections, material compositions, cc.ree-dimen-sional relative power distributions, capsula model geometric approximations, i 1 fission yields,- aversting of neutron cross sections over energy groups for dosimeter reactions, isotope decay constants, and seasurement of reac:or power history. 3ecause of their interdependent and sometines couplex effects on activity, limits for nest of these components were necessarily estimated based ,

on experience and engineering judgment.

1 Conversion of measured ac:ivity to flax requires the ase of neu:rna cr:ss sec-  !

i tions for dosiseter reactions, fisalon yields, and kn wledge of reac:or power

.t Babcock & Wilcor I

. l L <

t

.s y i

?

.. ; W . , . _ _ ,, + --. -a, -. . , _ .

0 I

l history duri:.g the irradiatica period. I:plicit in this 4:41ysis is the as-sumption that : elative pcwer distributi:n is independent of ;cwer level. In-clusior. of limits foc these cc=ponents :suses the uncer:sinty limits cf the measured activities to increase (Table 4-1). For scos reaciicas an extrapo-lation is required to extend measured flux to the fast flux energy range (>l MeV). ?cssible errors in spectral shape were esti= aced to have a :1C: effec:

on total fast flux. This additional uncertainty tends to decrease confidence in 5"Fe(n.p) and Sani (c.p) reaction measurements for fast flux calculations relative to the two fission dosimetars. '

4.2. Generie Flux Data Since calculated activities for the ini:141 group of surveillance capsules

. (Tables 3-3 and 3-4) were essentially all within uncertainty limits for sea-suced activities, i: vould seem reasenable to conclude that the analytical procedure has been verified. (A 1.1 factor was applied to calculated fluxes to account for an apparent negativa bias and to insure conservative results).

Consequently, uncertainty limits associated wi:h fluxes derived from s.easured activities should be applicable to directly calculated fluxes. A nominal value of 225: was selected fres Tabla 4-1 for determination of average flux * -

in surveillance capsules. Cacertainty limits associated w1:h spacial extra-polation from capsule to =sximsa location in the pressure vessel and :o account for variation of relative power distribution between fuel cycles cause :he over-a n uncertainty limit to increase to :30% for predic:ed fluence over the pres-sure vessel life (Table 4-2). Power distribution effects are es:f=ated to be nredictable within :20: with a nachod based on comparison of total fast flux escaping the areactor cere (section 2.3.5). It should be emphasized that this 230 value is dependent on the availability of a core cri:1cali:7 analysis for an equilibrium fuel cycle and that this fuel cycle be utilized over most of the vessel life. Also, the uncertainty limits of a consideraole number of the ,

ccaponents included in the overall value have been estinated due to the ab-sence of specific data. The net effect is to dimisish the credibility of the overall uncertainty limits listed in Table 4-2. Ecwever, an at:empt was made to select conservative values (maximum expected variations) for each identifi- '

able conponent.

3ecause operating procedures, fuel enrichments, etc.. vary between 177-TA re-acters. :he ;over distribution fr:3 a specifi: fuel cycle does not correspond 4-2 Babcock & Wilcox

.r

_ _ - r , . , _ , . , . . . . -- + , . . . - , - , - = n y

to that in a ge=eric acalysis. Selection of aeerage cycia ;cwer districu::ans will in:roduce an additional uncer:alnty. In general, hc-ever, a cytia repre-

' senting maxiaua fluence conditions (max:=n fast flux esca:i=g :na rese:or

' core) is used so that any extension of the uscar:ainty limit is on :he nega-tive side (acd usually ignored).

Table 1-1. Oncertainetes Reisted to Sostseter Ansiises, I tesc: ten 2373 ,r3,g). :2sr(3,f) 56y,<3,7) j!yt<3,3)

Measured dosimeter activity 8 :21  : 13 :13 :13 Calcula:ed dosinecer activity (8) :34 :29 :29 :29 Conversion of activity :o flux :11(:21)(b) 211 :11 :11 Flux (based on measured activi:y)  ::t(:30)(b) ,:17 :17 :17 Tast flux '

(E > 1 Mev) 224(:30)I } :17 :26 26 ,

(s)An addi:iscal :onvenant was added to :he ref erence 13 u=certain:v analysis to account for rela:1ve power distribu:1on averaging over the irradiation period.

(b)7alues in parenthesis are based on the yield data used La the ini:Lal sur-veillance capsule analyses. Cpdat ed yield data utli ,r:d sce lower -*21ues (Table 3-2).

O 41 Babcock & Wilcox "

+

=

, m - - ,

s l

!anle !.-2. ~

Cr. certainty Limi:s Associa:ec *.*ith 3acer*: I Pressure vessel Surreillance Tiux Analysis for 17? A Reseters t'acertaine r. y("

Tast flux based ou capsule dosiseter meaoweasents - 21 F2**'n= fast flux in pressure.

vessel region .

23 Pressure vessel fluence pre-dic:1on over 32-year vessel life :30 ("'

(a) values sesociated with fast flux are considered appli -

cable to fluence because reactor operati=g power sea-surements (and their duration) seasurements are rela-tively accurace.

(b) Nominal value frem all dosimeter reacticus .

(c) Based on a reference fuel cycle.

4-4 Babcock & Wilcox

~,,.

b

.P l

)

s

.e

1

,D l A

yi

~

I

5. Fli".~* !S 5.1. Generic testen Fluence A generic desigs curse was const:ac:ed sc that the maxi =us fluence :ha vill eccur is the pressure vessel of any 177-TA reac:=r based :s present 34*4 design ,

parameters could be predic:ed. Calcula:ed flax (bench = ark.c :s capsule fo- ,

simetry) at the pressure vessel it. side sur' ace ( .11* off a sa*cr axis) for Cconee 2, cycle 1, was used as a reference value and :nen ad us:ed :o accou== -

for generic model condi: tens as fo;; ws:

Calculated flux ,

I ' Ratio of. Ratio of cycl (E > 1 Me7) at i

- t generic '  !, core escape . .ax,e-averaged!

aver inside surface of5 alt.ence ~ ' .** "" #

(ge eric) pressure vessel j'{ocon*e#e 2:  ! cycle to Oconee 2. cy-

. for Oconee 2 e i  ; power cle 1 cycle .

Upper uncertaisty ,

lini: of predictedi i Pres'sure vessel

, .,e

=

equilibrius cycle ;. = l.esign a s e nds

~

conditions  ; L ,

"he terms of this equation are based an the following :ocdi:1:ns. A deneric power racing of 1772.'We represents ese zazi=us powar presently considerec -

for 3 W 177-TA reactors. For a given relativ- power distrita-ion, flur is-caping the core should be directly proper:icn41 to power level. Change is relative power distributics is accounted fer by the ratio of fast flux e,s-caping the core. This direct proportion 411:y to pressure vessel fluence is based on the premise : hat all 1**-TA reactors have the same flux attenua :cn

narac: eristics fr:s core edge :o pressure vessel (sec:ica 1.3.5). A review of available core calcula: Loss ;erf:r:ec f:r fuel .anagece'nt analyses inc -

cated that ANO-1, :ycle 2. snould represent :ne equilibrium :ycle expec:ec ::

.. ave.the greatas: fiwe per uni: ;cwer escania; :ne ::re fanc :nerefore great-est pressure vessel fluence per =::: ;cver). Ta ensure ::a: :he generic :urte 1 1

51 Babcoctt & Wilcox 3

9 mee-SpD$ en-e.y6_9 A pp - e sw * - * *

• _y%MS 9 dv4Q-. 4 4*M'E- W*

w

1 .

represents .the extre=e case, the upper uncertainty limit of 10% assign $a to predicted power distributiens is included. Thus,

7 7 9' '6 20 = 101 4 Fluence = (1.39 = 1013) j3,j ,4.62 = loa) G.D n.01 = M

= 2.1 = 100 n/cm2 .

Fluence as a function of operating time is plotted in Figure 5-1, with the inclusion of a c3C: uncertainty limit (Table I.-2). 'A fluence of 2.1 = 10' n/cm2 c30% which corresponds to 32 ETPY of reactor operation, is predicted to occur at the m*m location on the pressure vessel and under the nost adverse conditions anticipated for 177-FA reactor. Possible chaages is fuel

= =gement procedures, such as the use of an 18-nonth .3P fuel cycle, under consideration by several utilities, could result in fluecce reductions of up to 25 .

Additional confidence in these data was obtained from the analytical procedure developed in section 4.2. A direct calculation of pressure vessel fluence based on AE> 1, cycle 3 power distributions and the calculational norm =14 ration factor from cycle 1 capsule data (section 3.2.3) yielded an end-of-life fluence of 2.1 = 1018 n/cm2 .

5.2. Soecific Reactor nuence Fluence data derived from dosimeters in specific reactors have been extrapo-laced by use of a technique described is sections 2.3.5 end 4.2. The last cycle for which core escape fluxes were available was considered to be the equilibrium cycle, whier. was then assumed to be repeated to 32 EITY (pressure vessel design life). These data indicate that a somewhac lower fluence vill be reached at end-of-life than. predicted by the generic curve (Table 5-1).

This difference is accounted for by the higher power level and extreme cycle conditions utilized fai the generic analysis. A 230 uncertainty limit should also apply to specific plant extrapolated data. Note -Jtac the ANO-1 fluence of 1.3 = 1018 n/cm2 is considerably lower than all other systems, primarily be-cause of Arkansas Power & Light's decision to use the 18-month, L3P fuel cycle starting with the fourth cycle. Fluence levels for each pressure vessel veld j considered to be in the beltline , region can be determined frem data listed in Appendix F for oconee 1, 2, and 3, Three Mile Island 1 and 2, Crystal River 3.

Rancho Seco, and ANO-1.

5-2 Babcock & Wilcox  !

i

""m k

i 1

• i l

As surveit here capsule losisatry data secor.e available fres future reactor l

operation, it win be used if necessary to refise the generic curve. lang ters _

u irradiation data vill reduce the required extrapolatica period (out to 32 ~ ?T) and provide a better indication of equilibrium cycle power distributiens, voich should. tend to reduce the present uncertaisty limit. Also, as data fren mul-tiple fuel cycle irradiatica periods are obtained, a better definition of ex-trene cycle conditiona could permit an additional reduction is generic fluence.

5. 3. I.ead Factors A convenient =echod for translating fast flux values presented in this repert (inside surface of pressure vessel) to other locations is by use of a lead factor. Lead factor is defined as the ratio of fast flux at a specific loca-s tion to the maximum fast flux in the pressure vessel. Is this analysis saxi-mum flux occurs at the inside surf ace of the pressure vessel on a radial tra-verse located *.11* off a major axis (19 geometric model). Values for pertinent locations are listed in Table 5-2. These ratios were determined from a radial flux distribution obtained from a one-dimensional radial ANISN calculation.

Although ,the calculational model was based on major axia dimensicas, the lead Jactors sho.uld be applic3Ma re aar sadiat traverse (is this case 11' off axis) because flux distribution near the pressure vessel is primarily a function af attenuation in the radial direction due to symmetry is planar geometry of near-by regions (Tigure 2-1).

Lead factors for different azimuchal locations (points not on the same radial traverse) are scaewnac sensitive to variations that occur in azimuthal flux distribution between reactors; as contrasted to the shape of radial flux dis-cributions which is relatively consistent for 177-FA reactors in regions near the pressure vessel. In general, however, maximum and minimum azimuchal values-were located at about 11' and 26* respectively. he corresponding azi-uth a' displacement factor in' Table 5-2 represents the assumed equilibrium cycle (ANO-1 cycle 3). Since cycle averaged flux and fluence are directly propor-

'tional, these lead factors are also applicable to fluence. hus, any fluence value from Figure 2-5 or Table 5-1 can be converted to a different location by multiplication with a lead factor.

-3 Babcock & Wilcox

~ ~

s m,ea 6

,4; Tablu S-l. Heulmum Psemeure Vessel Fluence Ptsdiction for fli-FA Ncacture Faut fluences ,gcm#

8.imed on Estrapolattun l'uwe r , capsule tu eigullibrium Estratiol ttusi lhica r t a in t y React or _ NW4 , i t s e.lla t ion cy g to 32 ttt'y limit Oconee 1 2568 8.9(t17) 1.8(*18) 1.7(*19) tint '.

tkunes 2 2568 5.3(*17) 1.9(t18) 1. 8 (* 19 ) 3301 ocunee 3 2568 4.2(*17). 1.5(t18) 1.9(t19) 1101 TMI-I 2515 5.7(t17) 1.8(*18) 1. 7 (* 19 ) t ilit AHO-1 2368 4.l(617) 2.6(tl8) 1.3(419) 1101 Generic 2772 -- --

2.1(*19) 1101 Y

8' (a)Eelullibrium cycle ammuned to be the last cycle (r 3) f ur wl.ich f uel manage-ment analyste data (cure esempe flumen) ar's available.

4

)

'i.

K R,

P

. I

=

D

• M 4

amie 3-2 . Laad Factors for Tast Flux is a 177-FA Rese:or

.aca: ten Radial Azi=uthal displace: wnt displacement Lead Radial Azimut hal(" fae:cr factor facer Capsule center 11* 1.3 1.0 1.3 Vessel Laside surface 11* 1.0 1.0 1.C T/a 11* 1/1.8 1.0 1/1.3 3 T/4 11* 1/7.5 1.0 1/7.5 Vessel outer surfaces 11* 1/17.5 1. 0 . 1/17.5 Capsule center .

26* 1.3' 1/1.4 1.3 New capsule centerI "I 11' 4.5 1.0 4.5 New capsule caeterI"I - 25' 4.5 1/1.4 3.2 I*I Location of w

• mm (11*) and =1 == (26*) of azimuthal flux variation based on A30-1, cycle 3 power distribucion.

IDI T = thickness of pressure vessel.

(*I Capsule location in the three host reactors - Thsee?X11e Island 2 Crystal River 3, and Davis-3 esse 1.

b l

5-5 Babcock & Wilcor

's

"' 8 - as ea,a-aw. aqp * = s n,g , , ga p ge giw --

Ceehadw 93 +w.4 k

t Tigure 3-i. Gentri: Castg: The::ce f r ;- -T., .ea:- rs 34 sed on Location of y.axmus Exposure on Isside Sur' ace of Pressure 7essel 2.6 p

2.6 -

/

/

2.4 -

/

/

2.2 -

/

/

j 2.0 -

/

a N

I 1.3 -

/

A J'/

(z 1.6 -

/

s

- 1.4 -

/

e3 1.2 - ,/ f* ~s #

s' 59 f *** p E 8 f .#

C

• 1.0 -

/ 9

M/

< %f' / .

9

T o.8 2

o.s

/

/

/

b W)pf

/ /

/

o.4 -

l s

/ /

p /

0.2 - / p

/

/' ' ' ' ' ' ' ' ' ' e . . , ,

0 la :o 22 24 26 28 30 32 o 2 6 6 3 10 C 16 16 Effective fall Fever Years (JFT) i I

I 5-6 Babcock & Wilcox

- s -

- y 4

6. RITERINCIS I

Code of Federal Regulatiocs, Chapter 10. Part 50. A;pendix C. "Tracture*

Toughness Requirements."

2 Code of Federal Regulaticas. Chapter 10. Part 50, Appendix 3. " Reactor vessel Materials Surveillance Program Requirenents."

3 Reactor vessel Material Surveillance Progras,* RAW-10006. Rev 1. Babcock &

"Jilcox, Lynchburg, Virginia, May 1970.

H. 5. Palme. C. S. Carter, and C. L. Whitsarsh. Reactor vessel Material i

Surveillance Program - Compliance With 10CT150, Appendix H. for oconee Class Reactors, 3A*J-101C0A, Sabcock & Wilcox Lynchburg virginia. Tebruary 1975.

l .

5 J. J. Sapyta E. M. Nevien and N. M. Hassan, User's Manual for 3&*J's Ver-sion of ANISN. VPr3-T% 121, Babccck & Wilcos Lynchburg Virginia, Oecember 1971.

6 User's Manual for the 30T-IIW Discrete Ordinates Transport Computer Code, WANL-D'E-1982. December 1969.

I CASE Group Coupled Neutron and Canna Ray Cross Sectica Data. RS!C-OLC-23. Radiation Shielding Information Center, Cak Ridge. Tennessee.

'8 E. A. Bassan, et al. , Power Peaking Nuclear Reliability Factors, 3AW-10119P.

3abcoca & Wilcaz. Lynchburg, Virginia, June 197; .

3 U. M. Hervig and N. M. Eassan, SORREL - DOT Input Generation Code, NPCD "1-122, Sabcock & 'Jilcoz Lynchburg, Virginia. November 1977.

I3 W. T. Brunsen to L. 3. 'Jimmer, Memorandum, " N at ?over Shapes cor S ==

Heating Analysis," 12!11.9, Sabcock & 'Jilcox, February 11, 1973.

II A. L. Love. *r., e: al., Analysis of Capsule OCI-T Trom Ous'e ?over company,  !

Oconee, Unit 1 - Reactor 7essel Materials Sur reillance Progras. Revision 1.

EN M e:1. Rev. 1, 34tcock & Wilcox, lynchburg, *firginia. Septa =ber 1975.

6-1 Babcock & Wilcox i

I 1

I

.(

A. L. Love, Jr., et al., A:alysis of Capsule CC**E  : ua ? ver C =;asr Oconee Nuclear Statica, " sit 1 - Reactor Vessel Materials Surveillance Program, 3AV-It.36,*3abcock 5 Wilcox, Lynchburg, 71rgisia, Septencer 1977.

• 1 A. L. Love, Jr., et al. , " Analysis of Capsule CC :-C 7 := ltke Pwer Can-peny Oconee Nuclear Station, Unit 2 - Reactor Vessel Ma:erials Sur re111ance

• Program. yt*J-1437, Fabcock & Wilcox, Lynchourg, 71rginia, May 1977.

I* A. L. Love, Jr., et al., Analysis of Capsule OC*::-A Tr:a 0fes Power Ces-pany Oconee Nuclear Station Unit 3 - Reactor 7es'sel Materi.</ 3rreillance Program, 3AW-1438, 34becek & Wilcox, Lynchburg, Virginia, J 17 19*7

'I A. L. Love, Jr., et 21., Asa17 sis of Capsula Of:-II Trcs Mac : poli:aa Edison Company Three Mile Island Nuclear Statien, Unit 1.- Reactor 7essel 1

Materials Surrefitance Program, BAV-1439, Babcock & *Jilcox, Lynchburg, .

Virgisia, January 1977. t IS A. L. Love, Jr. , et al., " Analysts of. Capsula ANI-E Tr:n Arkansas Power &

Light Company, Arkansas Nuclear one, Unit 1 - Reactor vessel Materials Survet11ance Program,' 3AV-1440, Sabcock'& Wilcox, Lynchburg, Virginia, April 1977. -

t 17 1. C. Helmer and 1. C. Creenwood " Evaluated Decay Sche =e Data." *:uclear Technoloty E (1975), pp 253-273.

18 L. A. Hassler, Reactor vessel Surse111ance Program Cccertaisty Report (to be published).

IS M. E. Meek sad 3. F. Rider, "Ccapilation of Fission Product Yields.**

Nco-12154-1, Vallecitos .%e1 mar Center, General Electric Company, January 26, 1974.

20 W. N. McIlroy and L. 5. Ec7.* g " Fuels and Materials Fast-Raactor 3csimetry Data Development and Testing," Nuclear Technolotv 25 (1975), pp 180-223.

6-2 Babcock & Wilcox t

I 1

r I

)

\. ,

/

I AF7CfDIX A .

Time Averar,ing of Relative Power Distributions i

t A-1 Babcock & Wilcox

'*~**'**"~'"* .. --

f 0

L 0 a

t

- v To provice a single ?J2 :..a: w ~1 represe=t an en: ire irradiacies ;eriod :he m

'l following calculation setsod is used. It is based on the assumption that the values vary linearly vita burnup except when control red moveme:t oc:urs and then a step cha'nge occurs. For example, let q3 453' 9133' S!!;' 9 3 3' #

-+ 953. represent the ' relative power at a specific location at 0, 50, 100, 150 200 and 250 days during a 250-day fuel cycle. Ac 200 dayv~ :strol rods are moved. Thus. ,

T * ))l'g(25q, + 50q.; + 50g gg, + 75q153

• 2'4233

• 229253 I Th13 averaging calculatica is performed hy the code SCR.IL.

t Effec:ively, this forzulation indicates that q 3

applies frea 3 to 25 days. q,[,

75 days. qg 3, fres 75 to 125 days. qis3 from125:o200 days.q[g from 200 fres 25 :o to 225 days, ana q 253 from 225'to 250 days.

A t

A-2 Babcock & NVilcox i

I 4

maou s Axial Power Distribution Correction Factors 5-1 Babcock & Wircox

(

4 i

i For a ;ives fuel asse=aly, esd-of-cycle hurtups as a functics f axial loca-tion are obtained from fuel =anagement analyses. Relative power distribu: ion is obcat=ed by dividing local burnup by :he average burnup for : hat fuel as-sembly. Frem this distributien, peak (or correction) factors are calculated for the surieillance capsule elevatica and for the avfm= axial location.

• (Although, in pri:ciple, fac: ors for any av4=1 iccation could be calculated).

These factors are then weighted to accouse for fuel assembly loca:1on. From Figure 2-1 it can be seen that values from the E. I, and L revs should be representative of the flaz escaptsg the core that reaches the capsule located at Li* off axis.

deighting fac: ors were calculated to account for expecenzial attenca:icn of fast ceutrons hors in is:erior fuel assesolics. ~hus,

'T Attenuation Colu=c distance, en **eiehting

- factor 15 0.0 1.D-li 21.8 e 2 . .,-(:. 37) (:2 3) = 0. 22 13 43.5 e "E25 = e (3 07)(4 3 6) = 0.05 where g = 0.07 cmII is :he macroscopic removal cross section calculated in the core region for neutrons and has been averaged over the energy rasse E > 1 MeV and weighted with a fission spectrum. All neutrons hors in a fuel assembly -are assu:.ned to be located at the outer edge so that the at:enuation distance is a multiple of a fuel assemoly pitch. The rapid reduction of this factor indicates that only the outer two aasamblies significan:17 contribute to fluz escaptng the core.

The =vf=1 power distribution factor. T . can be calculated for each rev as z

follows:

e "

= *

• l'q T q 1 f .' 16 7

1*

j - 3.12 4 l;H15t 9 -81.!'+0.05{i_E;3ll 9 9

, . i i ,

j ve.ere (q/i3 :5) is the relative poser densi:y in asseroly dif. etc.

L 3-2 Sabcock a 'Nilcox I

t i

r

- Values of ;/g should be near :ne sa=e elevati:n f or esca asse o;y. A core average value is :o:ained by (171:4 equal weight to :he va*ues af H. K. 4:4 *-. ,

Admi::edly this 14 an approxt:a 2 pr:cedure but IPD values do not vary signifi-cantly from pis :s pis as shown is :he foll:ving example fres Oconee 3. :ye'e 1 data {!able 3-1).

The use cf axia; fac::rs :o c:rrec: At flux =a;:ulations at isca:icas cu: side the core is predicated on the assespeion :nac axial f *ax snape iu the core is the sa:e as the axial power shape asi :na: :nis shape remains in:ac: curside the core. For these reac:or prooless :nis a; pears :o be a-reasonable assc=p- -

• ti:n because si :.e regular geome:ry f the attenuatten regions (s: rue:ura" componen:s) at the elevati:n of interesc 4: ore heign:). Is prac:1ce. (caleu- s lations is RZ geometry of sisilar reactor models) the axial snape has been observed :s flat:en so=evr.at at i:creasi:g dis:ance frem t'.

. core. This wculd tend :o make flux precic:1ons wi:n :nese axial f actors slightly high. .

t

-Table 3-1.  ::eterminatten of a Typical .o al Correcrien Teeter F

Average over Peak Tuel Weighting surveillance axial assemble factor :sosule length loca:ien E15 1. 3 1.17 1.13

..,4

s. J . ., ., ..,o .

113 0.05 1.13 1.17 3 Row --

1.17 1.13 4

K13 1.0 1.17 1.13 Kia 0.22 1.16 1.1.*

K13 0.05; 1.13 1.17 4

K Row --

1.17 1.13

,,5

-. ,3

.. .. .o

.,4, 3.,, ,

.. .3

. . . ... ,a.

.3 3. 3 e, .

... 3' ,,,

. lov --

1.16 1.*s Care average -- 1. ;7 * *

.3 4

3-3 Babcock & Wilcox e

h ;

4

~

+s

• e

BLANK FRAME I

I FOR 1

PROPER PAGINATION

=w

-,a- .,,- ,- . . , - - - - --,,-,,--,,,-.-n

- - - -- - -- - . , - , - - , , - -en-

f l 4

(

l 1 s r

APFDfDLT C Effective Energy Ra=ge for Dosimeter leacti:u C*1 Babcock & Wilcox l

t

., \

l

- 1

)

In order to properly evaluate the flux data derived frca dosiseter reactions. l the energy ra=ge over vnich those reactions occur must be known. This effec- .

tive energy is defined as that energy above which '.95: cf the reactions occur and is a function of the flux spectrum. Values were calculated using fission

~

spectrJm-averaged 11croscTriitE capture crass sections and the relative flux .

spectrum at the dosimeter location. Reaction rate data vare calculated for the flux spectrum that existed at the capsule location f on the reference model calculation. These data, itemized in Table C-1, indicate effective en-ergy ranges of 20.5 Me7 for 237 Np(n.f), >1.0 Me7 for 138 I3 0(n.f), >2.3 MeV for ,

Ni(n.p), and $2.5 MeV for 5'Te(n.p) reactions. These energies are socewnat' dependent on ene group structure used in the flux spectrum calculations and, therefore, should not be considered absolute values. Eevever, they are con-sistent with other neutron transport calculations perfor=ed in the surveil-la-c.e capsule analysee.

(

4 C-2 Sabcock & Wilcox w

% a

.~w

, 'e ,

~ Table C-1. Effective Energy Range for Doulmeter Reactfonis

"""'S' Energy Lower en.rgy Norme11 eJ $8 H 5

, grog t.ound. HeV flum 23/Np(n.f) 238 H(n,Q l (3g}, Fe(u d g

7.o(-4) 1.74(-3) 3.56(-3) 3.74(-3) 4.22(-3) i 12.2 2 to 2.s5(-3) s.87(-3) 1.69(-2) 2.41(-2) 2.59(-2) s.la s.19(-3) 3.09(-2) 5.54(-2) s.69(-2) 9. 36(-2 )

3 4 6.36 2.24(-2) s.19(-2) 1.53(-1) 2.51(-1) 2.82(-1) 5 4.96 4.11 (-2) 1.50(-1) 2.70(-1) 5.11(-1) 5.5a(-1) 6 4.c6 3.50(-2) 2.07(-1) 3.63(-1) 6.75(-1) 7.19(-1) 3.01 5.25(-2) 3.04(-1) 5.01(-1) s.36(-1) s. 72(-1) -

7

- a 2.46 5.43(-2) 4.02(-1) 6.43(-1) '9.24(-1) 9.46(-1)

I 9 2.35 1.75(-2) 4.34(-1) 6.us(-1) 9.42(-1) --

n 0* 10 1.s3 6.7sy-2) 5.57(-1) s.61(-1) -- --

11 1.11 1.24(-1) 7.68(-1) 9.96(-1) -- --

~

12 5.5(-e)(*) 1.36(-1) 9.4 5 (-1) --

13 1.1(-1) 1.94(-1) -- --

14 3.35(-3) 2.44(-1) --

I*Iseed se 5.5 = 10'I I

5M

(

e e

b r- a - -- - ----- - - - - _

5 h

i BLANK FRAME FOR  !

I PROPER PAGINATION l

l f

I i

9 l

1

~

i 1

l l

4 APPCGII D Weighted Capture Cross Sections and Fission Yields X Babcock & Wilcox m.g-

Micr:se pi: captur- cross sections for the desbeter reactions of interest, and over the attergy group structure used for transport calculations (Table

- ' 2-7).'" 9ers obtained by beighti:rg CCF/317 data over a fission spectrum. Con-sequently E

o c(E) 6(E)dE a= -

s(E)dt

'E.r for each energy group. The resulting data are listed is Table D-1.

+

Fission yields used to convert calculated flux to ac'tivity are itemixed in

• able > 2. Data from ref.rence 19. usea is the capsule analyses, were cansid-ered the best compilation at that ciae. However, updated values fr a reference 20 will be used in future analyses. Note that a significant increase will oc-cur in calculated 137Np fission pr: duct activities. This would improve experi-mental data comparisons presented is section 3.3.

>.2 Babcock & Wilcox

n

- . . .-. . ~ ... .

i e

s

'aole 3-1. Tlasion spectrum 4eigated Capture Cross

~

lectices f:t Oest=eter Matarials "Jeinhted sieroscoste eross sectien barng/ aces Lower energy Cesuo Sound, MeV 2373 ,f3,f3 :19ern,g) If3gr3,7) 5.y,(3,3) ,

1 12.2 2.321 1.073 0.1195 0.425 2 10 2.340 0.981 O.622 0.537 3 8.18' 2.308 0.991 0.659 0.583 4 6.36 2,09 '

O.9165 0.638 0.572 5 4.96 1.54 0.6c0 0.5;0 0. 73 6 a.06 1.533 0.562 0.403 0.325 7 3.01 1.616 0.553 0.264 0.206 8 2.46 1.691 0.550 0.139 0.336 9 2.35 1.695 0.553 0.089 0.3525 10 1.83 1.676 0.535 3.051 0.022 11 1.11 1.593 0.229 0.0128 0.0115 12 5.5(-1)I*I 1.217 0.008 4.8(--) --

13 1.1(-1) . 0.1946 1.3(-4) -- --

la 3.35(-3) 0.0410 -- -- --

"I tand as 5.5 a 10~1 Table 3-2. Fission ?ield

?feld, !

I' ' "'I Fission Droduct Ref. 19 Ref. 20 Ref. 19 Ref. 20 137Cs 3 4 'i'f 6.67 6.28 6.194 ts2r 5. 7 fII 5.55 5.30 5.32s 13 ?tu y 6. !!Ik 5.75 6.35 6.336 3-) Saticock &Wilcox j 1

1

-i 9

l

. _ . . . s.na ,_ ,,~.--~..,~_,.......___.=..._..-2 .-a - -,.-_-. -.; a , _.n.

I I

l r

'l i

BLANK FRAME l

i i

I k

I  !

I h

FOR 1

d

! e 1

i PROPER PAGINAT ON l

1 >

1

}

t l

t i

i I

/

t s

5 N

%dh 9

~ .

AFFDDII E Equivalence of Activity and Flux 24cios I

s A

= * , e Pg. 7e I-1 Babcock & W11cox 4

5~g .'#

I 2

J p

+, ,

I 1

fr:s Activity of a dosimeter used to seasure fluence can be obtaine.

D = K = FI = PI sci /g of target =aterial ig}

g , f(6.023 = 1023}

• M(3.7 = 10*)

~

- FI = fivx integral

= [g a(E)t(E),

PI = power integral

-A t -1,(T-t,)

=.[,F(1-e g 3)e

. 3 (see section 3.2.2 for cars definitions). .

1 -

• is a constant for a given reaction

• D FI seas = PI seas mens ,

U cale Neale " IIcale And,- if the power integral has been calculated correctly, PI = ?!

Ic' It should be noted that for long-lived isoccpss, this ters is relatively is-sensitive to errors is power history and is prisar117 dependent en total ir-radiation time which is generally easy to deter =1se. Ihus.

F D, FI , [9 a(E)e(E)

=-

Dc ,y, = M,,y, - }g C aW For a given spectral distribution I

I, a(s) (s) = se where a = average cross section over the energy range of interest.

4 = fluz integrated over the energy range.

3 If the calculated fluz spectrian is identical to the actual spectt as, then the

- averaged cross sectices will be equal.

E-2 Babcock & Wilcox v

O

,,*'C_,, - - . , _ - , , . _ . - . . , , , - - ,

't *

• P se_* &

' seas -csic -

and i: follo.s that U I seas . scas " ' seas . ' seas D

eale 3 calc * ' calc ' calc e

9 6

/

I 1

I-3 Babcock & Wilcox 0 9%

• h 7 v

-. l l

\

l BLANK FRAME FOR

• =

PROPER PAGINATION -

1 l

, l

( 'l l

l -

l l

~

APPD*DII F Desi;n Fluence for 3eitline Region ".* elds

'l

w. . .

I i

1 7.; 320COC4 % WiIC3X

~ . -

k laters uacion of fluence for a specific location La :he ;ressure veJsel of a reactor requires knowledge of flux distribution is :hree dinensicas, i.e..

.both 19 ano 11 model calculations. Such an effort for equilibrium cycle con-ditions for each reactor would probably be both excessive and impractical.

Not only would the analytical cffer: require considerable sanhours and computer hours, but equilibrium cycle cendi: ions as presently :enceived will pecbably be changed in the future. Therefore, specific reactor veld fluences.have been based on the generic desigs curve (Figure 5-1) which should exceed expected values in nost reactors. If problens result, data for specific reactors can be recalculated with specific reactor power distributio'ns.

In Table F-1, fac: ors are presented :s ccavert the saxirua fluence presented is Figure 5-1 to account for spacial location and power level for beltline region velds is each reactor. A specific fluence can be obtained f res

~

" Fluence Wald fluence = frca =F =F =F =F F 9 A K

, Fig. 5-1, where . = power ., actor p ,

seecific reactor sover in W 2771 F, = azimuchal spatial factor flux se seecific azi=uthal location flux along 11* off axis radius F = =wtat spatial factor A

fluz at seecift: axisi location flux at sartsua ax M location Fg = radial spatial factor (lead factor) fluz at ssocific radial location fluz ac inside surface of pressure vessel

• Relative factors for azimu:hal and axial locations were obtained from analyti-cal models that utilized equilibrium cycle power distributions. Weld fluence .

values calculated from Fir.Lre 3-1 and Table F-1 vill have an estimated uncer-

ainty of =fC , although the conser rative procedures used should bias the re- .

sults on the high side.

F-2 Babcock &Wilcox

t t t rc11!y,Y pcoqes I-2

x: *1e

• T .er-e n ie.-

• • .ct t: 2x: e:': v5 *n ec . ,.

t-eynixI '4?'I VI C1 1** *t t TerTNatlee!

='

t" .'t t" gy p m *sm'i TI vt *nm sHe3 suessnae t" (t.;s 's t** t *1 te; en TTeve senat CO !! *1*tli e

• 0* 29 *S.n T5 TTow' Jea8*

• " C" t" t*1 C 171 *65* d? en TTm md.*.

00129 *f!* O TTeve seese f*1 C*t C Stt *16 e ** tiet elsees C" 6;*t 0*t (e) et**** w taeo tatC r ** ((*)e*1 t1 oc 319 *eet! T5 C tit 't!ct TT ' TWTPatiteet

-* c' of't is *C 30f.1

• t!i* * * *ttow* 38*80 Te=TPR319est

(*t C *t t*1 16*C temos tut) fa 'f1 ns moeda Wesquet t*1 (f-le *1 f *1 26*C (*) en TTm ana8*

T18e* anae!

f ** t" (*1 ft*C 57 &. 83 Tim 8*48.*.

Tiews seem t*T 9t *C C*t te*C C4 m 83 stes etsses (e) etsees wizao 1-3C

(*! (t*Pt*5 t *1 26*O Weegsset

(*t (F -)t

• 5 (*1 (6*C (e) etTim amart so SC *ti # TT*** ana*f P*! C* t c*t (f*O C II; *45 e es TTMS J.48$

T1*ws Jedde t*1 9;*C t*1 (6*C Oc: e e3 stM etsses f (f*C (e) etme utsee g eme C" (t-)**5 C" eme:nes' t *? (t*>f*C f*t 16*O (e) 83 TTous 8eas1 Tie ** 8 east f *! C*t f*1 ts*w C M. es Times ame8t.

Ytees seese C*t 16*c est m et stM etsses C *! 9 *t (g* (e) ots*** set sg g eeeeao t" ((-st g c*t teetsmatteet c" t *; et*C (1*O tem gwO Otti 75

• Tim aonet Te=18P 3195et C *t C *! C *C tt*C (*We*' 4"U tatt T5

• Tine asees f eet oa 219me, ,

t** 6?'t 64*C tt*C (swee useo (Jn v5

• Tim mtweasset suunissar t*; ((-19 g C1 (6*0 (e) en ttm amat ,_

Treue sonst 16*C fft! Tg on troge seeds C*! r*t C *t (tiet16**kt a Ttees ames e3 et 1eeeasset

(*1 9:*C C*1 Ef*C g C :Tl '6:n 75 Ttees esotees.senet E!!! vs et steg steses r*! g!*C t*t 36*C C*1 ed*e (e) eTHM HW  !"M cat g,y((*)=*$

• " F m ew usae, ~ 4 >e* a

• 3 a e re J J J 4

• anns o p t 's

• • ,t ee *Terev *T m =rev ue ..; totie.s 8"*d sa: areg yj _;- u; s pia..

ve-tet e::T:T*E .1=; s:c:wi s ar. i c2 se; *T-2 e cc 1 o 1 -_

I

~

s

1 i I I

i i

.33.4

.r*..

J P. * . * ,

f eet :s: f ac t re e

.. ifs tutar. aa n.natna &. asasi. See.a..

1. #* #4 **

touter ista t :se we 9 e 3.) Seties sessie (a) 3.49 1. 3 S..t.33 a.3 Bessie seit ta SA 1769 .J1 3 3.99 1. 3 3. *4 .3 4 W ene.1 4109. eJ1 JD

~; poor eneta to a 70 3.49 1.3 1.3 a. )

tener neell

meer enen to eel 3.49 1.0 S.se.31 .3 ,

settamme

.poer one11. 418. *cCt 3.89 2.93 1. 3 .3 Lang stes taa t W t. 13C1

' seer enen. 1A Liao .heen seems) J.89 3. se  ;.3 .I

.;aesstes saaa taaene 1me 3.t;es ww.. tas 1.3 .3 9..t.33 ;3 beste seit to W 223 - 1.3 1.0 3.*6 1.J

. pose emeA1 .

eper see.; to 4 126 1.3 .3 1.3  ;.J

meer one d

.eees soeu to faa 1.3 1.3 S.st.33 1.3 ee.'s mam

oser amen. 4 29 theta sement 1.3 3.49 1.3 4. J 1eesteestmes teser emeal. W 'O. '3

:3 1. 3 2.99 1.J *J 1aes tzuseaal 4 29. 2.** 23 4 29. ;203 asa>& autles essage tal 4.93 1. 3 3..(.33 1.J seeste telt to W 1331 0.93 1.J S.?6 1.4 esper ea+C 7 peer seen te W 112 3.91 1.3 1.3 a.d 1 seer seau taver eneal te (a) 4.93 1.J S.at.33 1.J esse amme

pset amen. W 14 Desa samme) 3.93 3.76 1.3 1.J 1emoiresAmal taeor emen. W La (beta sammas 3.93 3.73 E.J 1. 3 toesttensas

'*8 se eeA4 uom:121satsaa esmeere we awalaale.

+ "I tL! 3 repreeness til si tu wead esser'.at -res free the tastae stanseer of sne aceseere easee&. 391 2 . , . a att ei tse en&d essertal ammeeree free tae entstae ehemeter ad one pronewe woment.

Seme St$.6 e LVI .

f d e g,,,g g , , ,, .,3,,,, ,,,,,,, ,g, ,,, ,3, ,,;, ,,,,, g,4 , gg g ,,, g,,,,,, ,,,,,

C'8v .4ee. we ret taewe namer . ,r.e . ve et re.stam. r. stat .crut se veta r:ees.ee. :: .

e.n.e e, .at :masta a in. ,re.e..e .ee .r. af 1 t. r, estees ..e .e eauun e uueiaca. sne rua et rue e:ee .t a. usa as ata.a io cut ::= n tu re.ue,e e..et ta. 4e stamuu aus steero r.t.

t F-.e Babcock & Wilcox T

4

- - , - - - g - - .

l

'r.*

i Figure F-1. - Fast Flux Ac:enuation "hrougis Pressure vessel Wall 1.0 J '

i I l

~

. 1 Clad _

Pressure 7essel i

I

. J I I ..t

' l

'l l 1 i 1

I

-l l I

> I v

c i

- 1

. I g w I I I

BB

= I

- 1

- 1 I

. I I

• l l

I

s.

I

.se l .

I

! i 1 l 0.1 - l 4 I I I

'l 1 s

1 .

I l

l 1

I

. J -

I l .

. - T/4 . T/2 3T/4 I 3 2 a 6 S 10 12 li 16 IS 20 2:

i Thickt. ass, em J

. F-5 Babcock & Wilcox

} emWA  ?'

6 ,l

. . ; , . . = -,

s 4 (, . .

• e-- ./%ax_;. . . ,

v

s. '

aQ% q , $' ; t- 6Y &h40 $ - O ^ ' '

, y g s

, - A .*

- ,Le <$. Ave.= ~

I O L I

l -

_ _ , _ . . - - ,- , . ,, _ , ~. . . _ - . - _ . _ - _ . , . , , _ . . , , .