ML18025C003

Analysis of Vessel Wall Neutron Dosimeter from Browns Ferry Unit 1 Pressure Vessel

ML18025C003
Person / Time
Site:	Browns Ferry
Issue date:	08/31/1978
From:	Lindholm U, Norris E SOUTHWEST RESEARCH INSTITUTE
To:
Shared Package
ML18025C005	List: ML18025C003 ML20090H438 ML20090H445 ML20090H451
References
NUDOCS 8310280149
Download: ML18025C003 (50)

Text

)pp E. B. Norris C

jynI"". " 'ANALYSIS OF THE VESSEL WALL NEUTRON DOSIMETER FROM BRO'WNS FERRY UNIT 1

PRESSURE VESSEL tiy I

sing I,

c'.g

'I '

FINAL REPORT SwRI Project 02-4884-001 I

~i

%'I s fw~<<

~ dW

~If "v7

,":pp" glr.>>

Vd,;J jfg~

I ~

to Tennessee Valley Authority SOS Edney Building Chattanooga, Tennessee 37402 Augusf 1978

': ~P,

~ M

]C,

. Jjp SOUTHWEST RESEARCH lNSTITUTE SAN ANTONIO CORPUS CHRISTI HOUSTON

0

SOUTHWEST RESEARCH INSTITUTE Post Office Drawer 28510, 6220 Culebra Road San Antonio, Texas 78284 ANALYSIS OF THE VESSEL WAIL NEUTRON DOSIMETER FROM BROWNS FERRY UNIT 1

PRESSURE VESSEL by E. B. Norris FIWi'AL REPORT SN RI Project 02-4884-001 to Tennessee Valley Authority 505 Edney Building Chattanooga, Tennessee 37402 August 1978 Approved:

U. S. Lindholm, Director Department of Materials Sciences

ABSTRACT The vessel wall neutron dosimeter capsule from Browns Perry Unit 1 has been analyzed.

The results indicate that the peak value of fast neu-tron flux incident on the reacto" vessel wall is 1.24 x 10 cm sec E

> 1 MeV.

Although this results in a lifetime neutron fluence of 1.56 x 10 cm

, nearly four times that predicted in the FSAR, it is less than the design limit of 1.0 x 10 cm for 40 years of operation.

Based on a conservative estimate of the neutron embrittlement re-sponse of-the core beltline materials, the increase in the reference nil du'ctility temperature may exceed 100 F by the end of the design life of the Browns Ferry Unit 1 vessel.

The bases for selecting.a capsule removal schedule in accordance with Appendix H of 10CFR50 are discussed.

TA3LE OP. CONTENTS

'Erie me~'.c3. -;i. Z~

eall I'w., ca""i: =-'- j I.

SUKQBY OP RESULTS AND CONCLUSIONS II.

INTRODUCTION III.

EVALUATION OP VESSEL WALL NEUTRON DOS~TER CAPSULE V

~

IV.

CALCULATION OF NEUTRON FLUX DENSITY AND FLUENCE V.

DISCUSSION VI.

REFERENCES t

i w

~

J'

~

~Pa e

~ ~

~

~

2 18 23 lt

'llV.

I h'4 Aiq/

L 4

1

~ ~

I.

SUMMARY

OF RESULTS AND CONCLUSIONS The results of the analysis of the Browns Ferry Unit 1 vessel wall dosimeter indicate that the peak fast neutron flux (E

> 1 MeV) at full power during core cycle 1 was 1.24 x 109 cm 2 sec 1.

As a result, a 40-year design life fast neutron fluence of 1.56 x 10 cm is predicted nearly four times the calculated design life fluence given in the Pinal Safety Analysis Report (PSAR), but considerably less than the FSAR design limit of 1.0 x 10 cm Utilizing the radiation damage trend curve in the PSAR; the increase in the minimum reactor pressurization temperature over the design life is projected to be approximately 50 F.

However, it is possible that variations in the chemistries, particu-larly the copper content, of the Browns Ferry Unit 1 pressure vessel belt-line materials may result in sensitivities to neutron radiation emorittle-ment different from the response curve given in the FSAR.

Using the 0.3%%d Cu RTNDT ad)ustment curve in Regulatory Guide 1.99 '

, the total shift fib*

might reach 105 F at the vessel wall I.D. and 88

'F at the vessel wall 1/4t oy the end of the 40-year design life of Browns Ferry Urit 1 pressure vessel.

The capsule removal schedule necessary to meet the requirements 'of 10CFR50( ), Appendix H, depends on the value of the ad)usted RTNDT at the end of the design life of the reactor vessel.

If the initial RTNDT of the Browns Perry Unit 1 pressure'essel beltline material was higher than 12 P, the first specimen material surveillance capsule should be removed after six full power years of operation.

• Superscrip numbers refer to the list of references at the end of the text.

.:i ','I.4'NTRODUCTION:.'is

.",'.i,'he Browns Zerr'y Ãuclear.plant',op'crated.by the Tennessee Valley Authority (TVA), consists of three 1065 Rue (3293 Hvt) Boiling Mater Reactor (BMR) units built by General Electric Company..(GE)

GE provided each unit with a pressure.

vessel steel surveillance program which consists of baseline Charpy V-notch specimens (base metal, weld metal and:heat-affected zone), baseline tensile specimens (base metal, weld metal and heat-affected zone),

a vessel wall do-I simeter capsule,":-'and three surveillance capsule baskets containing Charpy V-notch and. tensile specimens.

The latter two items were installed in the three Browns Ferry vessels prior to startup.

The surveillance program is described in detail in NEDO-10115.

Be-t3 3*

cause of'he low level of fast neutron flux density at the vessel wall predicted by design'.calculations, the'irst ioWeillance capsules..containing mechanic'al test specimens arot'*scheduled"for'removal'"until'-'four years of operation have accru'ed;"

However, the"vessel wall dosimeter"capsules'are scheduled for removal at the first refuelling'to provide a-check on the design flux.calculations.

This report describes the results obtained from the testing and analysis of the contents of the vessel -wall neutron dosimeter capsule from Unit l.

III.

EVALUATION OF. VESSEL MALL NEUTRON DOS~TER CAPSULE The vessel wall neutron dosimeter capsule was zemoved from the 3rowns Ferry Nuclear Plant Unit 1 vessel during a refuelling outage which began on September 13, 1977, at the end of core cycle 1.

This capsule, snown in Fig-ure 1, contained three eac'h pure copper and pure izon dosimeter wires.

The nuclear reactions of interest for these wires are:

Cu(n,u)

Co 54Fe(n,p)54~

The capsule was shipped to the Southwest Reseazch Institute (SwRI) laboratories the week of October 24, 1977, in a cask supplied by SwRI.

The capsule was opened in one of the hot cells at the SwRI Radiation Laboratory witn a hand hacksaw.

This could be done because of the low level of activity ll exhibited by the capsule.

The contents were examined and visually identified as either iron or copper.

The dosimeter wires.were prepared for analysis by weighing on a precision laboratory balance.

The number of target atoms per mg, No, was computed for each wire as follows:

N

~

x103 N

c o

A where:

N

~

6.02 x 10 3 nuclei per gm atom; c

weight fraction of detector isotope in detector specimen; A

~

atomic weight of detector

element, gm.

The absolute activities of the dosimefer wires were measured with a NaI(Th) scintillation detector and an NDC 2200 multichannel analyzer.

The experimental efficiency, Eff(E), of the system was determined on the day of

II

~

I I'

I II CE}jlhkiQl}CC}h}C

~J I

Ill00 IllI~

~%111V RwllVIIR86861 V811

~6 2c}9R8(210.

~. ~

~r

/$88 7ISSS'UTRON OOSINETER

~8111m 616 SCRIRC8((R P

lg

~ ~tl tc

~ ~

PSO-yP Iog sEEIIOIE Igz 0

IQ il'l

'61 CI

1 I

JJ 6 ~

SCF}G/-d8/6'5D3%5

/vsres!

I. 8SFORE I}SSEMLIIIPP2 PSO-YPI02.

Est/A(ISr RIR 70 REDUCE scc 4'ors /l/s c5tps(I(F wrcRnw( rRcsscRRE

.2 PLACES 70OIIECV TORR. ORvKP/LL

/Itr// <C'LOIIIO O'RRO-'Z(Iuw 2 PLACES ArRTifa5PIIr RIC FA'ESSES ANDSEA(. /RIEL0 IIO(E.

2 AAss sprcrRO ErcR LsRK

~Pso.rr/

rk'Sr Cod/PLEAD ASSE8IOLY

~

4CV gwurcd RLLOIIASLE'68RHROE /0-c t

C+IV/R

8. /fELDIIIgPER 209AWZSO I, 6 L6 CR t'

(

~

I- ~

(:

6 6

~ ~

6 1 J

~

I Q

C I

I I

'I.

I

~ I Clr~

cM.Z It.l 56%

~1 VIIR

~IvI86I

.d8.4/j 08 I(6 AEIALsrAA'P u/rII

~/1 X.OIIIIGuCunt'Cr/,RS I 6588165$ 865RRRR 6

SER. Ro, rIIRN. SY IIIRAVF. ~,,

/~88naS

~let SRI(I JV ft C

~'

~

II 6

f, ~

'vt LI t

C ~

I I

I ~

~I EIGURE l.:VESSEL'ALL

~ 6 NEUTRON DOSIMETER

0

counting for each photopeak of interest, 842 KeV for 54Hn and 1173 KeV for Co, with Mn and Co isotopic standards traceable to the U.S. Bureau of Standards.

'The counting system and techniques have been previously checked aga'nst two other laboratories, see Table I.

The specific activity (dps/mg) of 'each dosimeter wire at time of reac-tor shutdown, A(TOR), was computed as follows:

~(TOE)

Toca'- couoca under obocooeak o-'ger E 1eea "back :ouo8"

.< ~T(-)

A(TOE)

EEE(E)

~ t

~ v

~

P

~

exp - Xc)

T(E)

(2) where:

w P

tl counting time, sec; weight of wire, mg; peak-to-total ratio; disintegration rate, day-1; elapsed time between TOR and counting

date, days.

intrinsic efficiency factor for the standard source counting geometry; intrinsic efficiency factor for the unknown source counting geometry.

In this program, T(E)s/T(E)u was equal to unity because the standard and unknown sources were counted using the same geometry.

The weights, counting rates, and specific activities determined for each dosimeter wire are summarized in Table II.

The last column in Table II lists the saturated activities, As, of the dosimeter wires computed for the full power level of 3293 Mwt as follows:

E (1-exp-XT )(exp-Xr~)

A(TOR) where:

operation period; equivalent operating time at selected power level for the mth period, days; elapsed time from the end of the mth period to TOR, days.

The values of Tm and tm were determined by dividing the Unit 1 plant opera-tion into 19 operating periods, as summarized in Table III.

a nv. 'e'ach

>>In t!!peF u "'ABL'E.g e:>!!II 'ita!'

RESULTS OF LVXERLABORATORY GAMMA-COUNTING PROGRAM u/

Sample Identification

~ I Assumed Activit at TOR (d s/m )

Isotone 'alf-Life SwRI Other Top (Co-Cd) -'.i!. 60Co 1913 d

2.69 x 107 "

2.68 Bot (Co-Cd)

Co

.1913 d'.67 x 107 2.48 x 107 (')

107 (a)

Top (Co)

Bot (Co) 60Co 1913 d

6.03 x 107 5.83 Co 1913'

'6.23 x 107 5.93 107 ( )

x'107 (')

2056 2062 R9 R13

~.l,

'5.57 x 106 5.20 x 106 ( )

'60Co "" " 1913 d

, " "312 d'..". '" 1.30 x 104 a

312 d 1.24 x 104 104 (a) 1.29 x 104 ( )

54Mn

. 54Mn 54M 104 (')

"," '";312 d'.',1.21 x.104,,

1.23

.<..4>>, ee I

u e

Co 1913 d

. 1.41 x 10 1.37 x 107 (

":" ~;"al Ourn:I. !"!.sf cO ":L.'f u'. "s."<<u'I! I ':"~.

~

f.-

(a)

BNRL (b)

WNES V

~

! ~

~ f!L e

0 0

TABLE II RESULTS OF ACTIVATION ANALYSES OF DOSIMETER WIRES EXPOSED IN BROHNS FEB'NIT 1 VESSEL FROM 10/1/73 THROUGH 9/13/77

~Zsoeo e

OCo 60Co 60Co Foi'u-1 CU-2 Cu-3 Weight

~(n )

470.9 496.0 475.5 Count Rate (dom) 1.044 x 105 1.115 x 105 1.031 x 105 A(TOR) (a)o

~des/m

3. 697 3.745 3.614 A (b)

~(dDs/m )

24.33 24.65 23.78 Average 24.25 54Mn 54Mn 54Mn Fe-1 Fe-2 Fe-3 158.3 158.7 158.5 5.790 x 105 5.669 x 105 5.231 x 105 60.96 59.53 55.00 132.9 129.8 119.9 Average 127.5 (a) Specific activity at time of reactor

shutdown, 9/13/77.

Disintegration rates are sub]ect to a ~3X (1 SX) measurement uncertainty.(4~5)

,(b) Saturated activity at the 3293 Mwt power level.

0 I!

ll 0

I

TABLE IXI a.~OPERATIONS

SUMMARY

BROWNS PERRY NUCLEAR PLANT, UNIT 1 rz>>c>>> I-sf~ i

.L cgt'I

);,">;wf

~

> ('t >'>'n;>

~ )

lt i

Operacing Period g a"t 10-01-73 10-09-73 2

10-12 73 11-10-73 3

11-13-73 12-12-73 4

12-2C-73 01-19-74 5

01-23-74 02 13-74 6

~

02-18-74 03-01-74 03-08 74 04-04-74 8

04-12 74 05-08-74 9

05-26-74 06-22-74 06-29-74 09-20 74 10-04-74 11 19-74 11 23-74

~

02-03-75 02<9-75 02-27-75 03-06-75 03-23-75 09-14 76 10>>14-76 10-16-76 11-18-76 11-19<<76 01-05-77 01-10-77 04-2"-77 19 04 23-77

(

7 12 13 14 15 16 17 18 Reactor Equivalent (a)

Pouer,.

Operating

~ÃIDth) ~>T )

res Ope 10&8-73 10-11-73 11-0&73

~

11-13-73 12-11-73 12-23-73 01-18-74 01-22-74 02-12-74 02-17-74 02<<28-74 03-07-74 0&3-74 04 11 74 05-07-74 05 25-74.'6-21 74 06-2S-74 09 19-74 10-03-74

'1-sg-74 11-22 74 02-02-75 02-08-75 02-26-75 03 05-75 03-22-75 09-13-76 10 13-76 10-15-76 11-17-76 11-18-76 01 OC-77 01<<09-77 04-21-77 04 22-77 09-13-77 racing Shutdoun Davs Dave 0.30 5.60 9.35 9.05

11. 49 6.48 15.45 12.32 17.54 66.97 35.98 62".35 12.35 13.53 4.32 16.33 33.42 S0.14 113.50 526.47(b) 8 29 28 26

'>]

11 27 26 27 83 46 72 18 17 30 33 47 102 1CC 977 3

18,450 30,793 12 29>789 4

37,S43 5

21,343

~

50,891 8

40,578 57>746 7

2"0 5"5 14 118, 491 4

205,316 40,664 7

44>'554>

541 14 >215 2

53,790 1

110,042 5

263>890 1

>>3 769

'otals 1;733,666 7

. 18 Decay Tine in Days a-1436 1C04 1372 1334 1309

~

1293 1259 1225 1180

'1090 1030 954 930 906 335 300 252 145 0

(a) At 3293 Hut (b) 526.47 days

>> 4.5487 x 107 ssconds

Ct 0

EV.

CALCULATION OF NEUTRON FLUX DENSITY AND FLUENCE The energy-dependent neutron flux density, j (cm sec 1), the spec-2 trum-averaged activation cross-section, a

(cm ), and the saturated activity As, of each dosimeter wire are related as follows:

~g A

No (4)

In the early days of nuclear pressure vessel surveillance activity, the value of a was based on the assumption of a fission spectrum energy distribu-tion for the neutron flux at the surveillance capsule location. It was recog-nized that this assumption was probably in error, but since correlations be-tween neutron exposure and vessel steel mechanical properties were empirical, the fission spectrum assumption was useful.

However, as methods of analysis were impro'ved, the use of calculated neutron spectra has increased and is now permitted by NRC Regulatory Guide 1.99( ) for application to reactor pressure vessel wall locations.

The neutzon flux energy and spatial distribution were calculated for the Browns Perry Unit 1 pressure vessel with the DOT 3.5 two-dimensional discrete ordinates transport

code, a 22-group neutron cross section library, a Fl ex-pansion of the scattering matrix and an Sg order of angular quadrature.

An R-8 calculation was made for a horizontal plane perpendicular to the vertical axis of the coze, and 'an R-Z calcul'ation was made for a vertical plane through the axis of the core and the location of the vessel wall dosimeter.

A one-eighth segment, shown in Figure 2, was taker. to be representative of the R-8 geometry because of the symmetry involved.

The boundaries of the core, core

shroud, 5et pumps, and vessel wall were described in R>>8 coordinates.

The

0

)I

LOCATION Wall AZIMUTH 30'EMOVAL PERIOD YEARS 10 "Wall" "'"120'

'" '12 '""

'il~

~ 4 Wall

~ l ~ Wall.-. "i 300':..30'2

~ I Capsule <

Capsule Dosimeter Z70 V

Q +l

~ l I L ~

"~ i rill,I 45' Capsule I

~

iWJ i IVC'I l

ir ~ > ~

~

~

Core Shroud pWn 'rrzM

~.r )r(.'Jeo, the i~l't

~,,

~ ~.-..

c. r;cj: 30 4 ~.

w4 ~ l i

Jet

~ Pumps I

~ ~

~

Vessel Wall 04

~

~

FIGURE 2 ~

ONE-EIGHTH'EGMENT FOR FLUX CALCULATIONS AND LOCATION OF SURVEILLANCE CAPSULES

0 l'1 0

- 0

core was subdivided into two regions, an inner region with one-sixth of the control rods inserted and having a 0.4368 void fraction, and an outer region with all control rods withdrawn and having a 0.4543 void fraction.

The core materials within each region were homogenized over their respec-tive areas.

Stainless steel was assumed to be 18K Cr, 8X Ni, and 744 Fe, and the pressure vessel was assumed to be 98Z Fe.

The coolant outside of 0

the core was assumed to have no voids.

An average power distribution in the core was derived from data sheets supplied by TVA.

The same assumptions were used in modeling the R-Z geometry.

Both of these calculations provide 'nformation on the neutron energy spectrum at the vessel wall neutron dosimeter capsule location.

In addi-tion, the R-8 calculation provides information on the radial and azimuthal variation in neutron flux, and the R-Z calculation predicts the radial and vertical distribution. of the neutron flux.

By combining these factors, the relationship between neutron flux at the surveillance capsule locations and that at the point of maximum neutron flux incident on the vessel (I.D. lead f'actors) can be derived.

The neutron spectrum at the vessel wall dosimeter location, as deter-mined with the R-6 model, and the group-averaged cross-sections for the dosimeter reactions of interest are given in Table IV.

The spectrum-averaged cross sections, o, were determined from the relationship:

10.0 E

o(E) y(E) dE 1.11 cr (E > 1 MeV)

~

0 0 Z

$ (E) dE 1.0

0 0

c~vt lilac 'aiP ce,I cnI 0: ~ -.

~

ejgLE Iv. 'f.fsf'.

I.i

~ a~'.f li'

...ILI~-JJ..

~ ~ i: I DETiZHINATION OF REACTION CROSS SECTION FOR IRON DOSIMETERS If

~

54Fe(n,p)54N Cross Section, Energy-Range:";. Normalized'Neutron'-

8. 18 10. 0

.0337 l

6.36 - 8.18

. 0844

4. 96 6. 36

. 1230 f

4.06 4.96

. 0985 f ~

'581

.577

'I

.491

.354 3.01 4.06

.1117

.185 I

2. 35 3. 01

. 1439

.078

~ ( "u

)I(

O'I:

~ g 4'a y If~u"' tII:f'~I1'"'~r( i 1.83 - 2.35

.1224

.023

~

I 1.11 1.83

.2084

.0014

~

1.00 - 1.11

.0740

~

f ( ~, ~,

~, ~.4 I ~

(

'A ffg \\

@fan Q(

~t on, fII

( f \\ b

~

.198('Fe f.f ~ op%bi f >et<a('

' Ij.on ft i'zs at f.fnc su+vel '

'Qc 63Cu(n u)60Co Cross 'Section, o

(E) (barns)

. 0380

. 0144

~ ~ f

. 0023

.00025

.00010

.00006 "f~ ff f

.00002

.00002

'I

.00001 a

~.00283(a)

I-n ~g~.

( ( ~.---'... ne I,'rn;I g". -:

. If. ~,

~ I I t" I(

'I <<:<!;.,

f1

~

q

(

fg P

~ fI I

'I f 2

~ ~ I

)

vf II' fff[

Is

~

f f

I (a)

Spectrum-averaged cross sections are subject to a 15/ (1 SX) uncer tainty. (6) 1 1 g

Sf'

0 0

13 Substituting the value of a into Equation (4) along with the average value Fe of As for the iron dosimeters (see Table II), the fast neutron flux (E > 1 MeV) at the vessel wall dosimeter location is calculated to be 1.03 x 109

'm".

Similarly, the fast neutron flux determined from the copper dosimeters is 1.31 x 109 cm The disczepancy between these two values is largely a re-sult of uncertainties in current evaluated energy-dependent cross sections..(4 5 According to ASTM Recommended Practice E 482( ), errors as large as

+7Z (1 SZ) in the determination of disintegration rates and i15Z (1 SZ) in spectrum-weighted group-averaged cross sections can be -encountered, which results in a combined error of +16.5Z (1 SZ) for the calculation of neutron flux from the input data.

It therefore appears reasonable to average the results obtained from the two dosimeters.

The azimuthal variation in fast neutron flux, as calculated with the R-8 I

model and. shown in Figure 3(a), indicates that the vessel wall neutron dosim-eter capsule was placed at the azimuthal position of maximum fast neutron flux.

Therefore, it is concluded that the calculated flux dezived from the analysis of the dosimeter wires is 'a direct measure of the maximum fast flux incident on the pressure vessel opposite the v'ertical core centerline.

However, the axial flux distribution, as calculated with the R-Z model and shown in Figure 3(b), indicates that the peak fast neutron 'flux is 6Z-

~

'igher at a position 20 cm below the capsule location.

Therefore, the lead

'actor, the ratio of the fast flux at the capsule location to the peak fast flux incident on the pressure vessel I.D., for the Browns Ferry Unit 1 sur-veillance capsules is calculated to be 0.94.

Based on the results of the DOT 3.5 calculations and the dosimetry re-

sults, the peak fast flux incident on the Browns Ferry Unit 1 vessel during

Oo

~

~

~

~

~

~

s

~

I I

I I

I I

L'I

0 e

15 the first core cycle is calculated to be 1.24 x 10 cm 'sec, E > 1 MeV.

Therefore, the neutron fluence per effective full.power year (EPPY) is 3.91 x 10 cm E > 1 MeV.

Assuming 100X availability over the 40-year design life of the plant, the design life neutron fluence received by the vessel is predicted to be 1.56 x 10 g cm E > 1 MeV.

The neutron flux is moderated as it moves from the core and penetrates the pressure vessel wall.

The radial dependence, of the fast neutron flux obtained from the DOT 3.5 analyses is shown as the solid curve in Pigure 4.

The dashed curve through the pressure vessel wall represents a conservative estimate of the fast flux attenuation by steel which is acceptable to the NRC. <>>

Since the pressure-temperature limits for reactor operation and test-ing are based on requirements of the ASME Boiler and Pressure Vessel Code~@,

I the fluence at the 1/4t and 3/4t positions within the pressure vessel wall are of specific interest.

Utilizing the conservative estimate of the attenu-ation'of fast neutron flux by a pressure vessel wall shown by the dashed curve in Figure 4, the predicted flux and fluence values obtained at 1/4t and 3/4t for the 6-5/16-in. Browns Ferry Unit 1 pressure vessel are summa-rized in Table V.

0 0

16 1013

( ~ I,

~

~

1012

'I

~

=

~

I I r.

1011

~I O.

~!

Iu N."-.

1010 I

' i

~ I Cl O.

, ~.r, gr!

~!

0QV ClCZs'J

~

~I CO I ~

Ol nfl Cl

~\\

I Cl

-. ~a)

(((0

$4 109 gee..~i r.

~

r

~ I I

lw ice::est.

]gz q~ ~i~ cp

I r

~

- I

~

~

I 'P

~ =,!

10'07 220 240 260 280

Rad1us, cm 300 320 340 FIGURE 4.

CALCULATED NEUTRON FLUX BETWEEN CORE AND PRESSURE VESSEL I.D.

NOR%0 IZED'O THE VESSEL WALL DOSIMETER RESULT

0

TABLE V CALCULATED PEAK NEUTRON FLUX AND FLUENCE FOR DROWNS FERRY UNIT 1 PRESSURE VESSEL WALL Vessel Wall Location I.D. Surface 1/4t 3/4t Relative Fast Neutron Flux 1.00 0.67

0. 24 Fast Neutron Flux Densit cm-2-sec-1 1.24 x 10~

8.3 x 108 3.0 x108'0 EFPY 1.56 x 1018 4

EFPY 1.56 x 1017 1.442 EFPY 5.64 x 1016 x lpl6 1.4 x 1016 l.p x lpl7 1.0 x 1018 3.7 x 1017 37 x]016 Fast Neutron Fluence (cm

)

(a)

E > 1 MeV.

Calculated flux and fluence values sub)ect to a 16.5X uncertainty.

+

~

(6)

(b) End of core cycle l.

0

18 t

t t ~

I~i The predicted I

, i Browns Ferry Unit 1

'.V.

DISCUSSION Jt't M

I"

~ at value of the peak neutron fluence (E > 1 MeV) for the

't pressure vessel after 40 EFPY of operation is given in the Final Safety Analysis Report (FSAR) as 3.8 x 10 cm (E > 1 MeV).

i

~

'he analysis of the vessel wall dos/meter capsule projects that the peak

~I='I neutron fluence vill be 1.56 x 1018 cm

, nearly four times the predicted t

r,.

~

~

value, but considerably less than the PSAR design limit of 1.0 x 10 cm".

A simiiar:trend

'has been noted in several other BWR plants with which SwRI I

has been associateP.

Foi..example, the neutron dosimetry analyses performed on the-first capsules removed from the Elk River, LaCrosse, Millstone Point I

tt,

~

~

1, and:Pilgrim r'eactors indicated that the fast neutron flux densities were VN higher than the design values by factors ranging from 2 to 6.

y~l l

The estimat'ion ef a 40-year neutron fluence from less than two years I

'N l C'f operation is--a'.large extrapolation and willbe sjxbject to revision at the

~&

time ok-the next,caps'ule removal, currently scheduled after four years of I

operation In the meantime,

however, the projected,:peak.fas fluence factor

..I of 3.9X-xt10 cm-2 per EFPY can be employed to predict the change in the reference,nil ductility temperature (RTNDT) as a function of reactor power

~

I generation.

I!

The threshold vilue.of neutron fluence for the 550 F embri.ttlement of I

ferritic steels is generally taken to lie between 101'nd 1018 cm 2 (E >

1 MeV).

The proposed relationship between fast neutron fluence and the change in the RTNDg of the Browns Perry Unit 1 reactor'essel, as given in

7) I I

~ (.

Pi,gure 3.6-2 of the FSAR, is reproduced in Figure 5,.'

Added to this figure are (1) an arrov indicating the fast neutron fluence on the vessel I.D. at the end of core cycle 1, and (2) an additional abscissa relating neutron

19 QD QJ CC W

Ct W

I o

I z

Ctl Cl C7 200 150 100 50 Reactor Operation, EFPY 1

4 10 40 0

18 1017 1078 10)S NEUTRON FLUKNCS t>1 McV) tps) ~ net FIGURE 5.

VESSEL MATERIAL NEUTRON EMBRITTLEMENT CURVE FROM BROWNS FERRY UNIT 1 FSAR

e

20 fluence and effective full power years as determined from the vessel wall surveillance capsule.

Figure 5 indicates that the RTNDT of the Browns Ferry Unit 1 pressure vessel willbegin to increase after an exposureof 1.35 x 1017 cm 2, E > 1 MeV.

The I.D. surface would reach this fluence in about four EFPY, but it would require over f've EFPY of operation to.

reach this fluence at the 1/4t location in the pressure vessel wall.

Also, the predicted shift in RT~T at the I.D. surface after 40 EFPY is less than 50 F above the baseline (unirradiated) value..

The neutron embrittlement sensitivity curve from the FSAR (Pigure 5) corresponds closely with the RTNDT adjustment curve of Regulatory Guide 1.99( ) for 0.15%

Cu and 0.012% P, see Figure 6.

However, in a recent re-sponse to the NRC(9),

TVA submitted information from GE indicating that the copper contents of plates might be as high as 0.2% and those of welds might be as high as 0.3%.

Utilizing,the 0.3%

Cu response curve in Figure 6, the predicted shift in RTNDT of the Browns Perry Unit 1 vessel at the end of esiga;Q.ife.would be 110 P at the I.D. surface and 88 F at the 1/4t wall

--~.'position'.

Since the capsule lead factors are near unity, one-fourth of the

'=; end-',.of-life fluence should be reached in 'approximately 10 EFPY.

m~~~~SectMn-II.3 of Appendix H of 10CPR50(

) describes three cases which j

..=',"-.:-SYvern. the surveillance specimen capsule removal schedule.

The first case, k

which applies when the adjusted RTNDT of the reactor vessel steel will not

.~ceed;400" F at the end of the design life, requires that a specimen cap-sule be removed at one>>fourth of the design life.

The second and third

cases, which apply when the adjusted RTNDT of the reactor vessel steel ex-ceeds 100 P at the end of the design life, requires, that the first specimen capsule be removed when the predicted ad)ustment of the reference temperature is approximately 50 P, or at, one-fourth service life, whichever is earlier.

(ao 1000 (5 Cu 0.08) + 5000 P 1ii l/2 P 0.008)) (I/'fO

~ 300 t

200 I

)f

~ 100 5

~

so o

0.3 0.2

=.0.15%

'O.OI

~ Lu s)ppF Lttht>

u W

OW C

P Ml 0.08 000 2X10'0 4

0 8

10>>

4 8

8 10 2

4 0

0 10

~

2 FLUENCE, n/cm (E) 1MeV)

Figure 1

predicterf Art)ustment ol nelerence Tempereture, "h", es e Function ol Fluence eml Copper Content.

For Copper errrf Phosphorus Conlents Other Then Those Pfotterf, Use the Errpressfon for "h"Gben on the Flguro.

(Note:

Das)ted lines represent GE recommended extrapolations to 20'F for BWR operation.)

I FIGURE 6.

REFERENCE TEMPERATURE ADJUSTMENT CURVES FROM REGULATORY GUIDE 1.99~ )

22 Based'n an end-of-de'sign life increase in RT~T of 88 P determined from the 0.3X Cu response curve in Figure 6 at the 1/4t fluence obtained from the vessel wall dosimeter, the capsule removal schedule necessary to meet the requirements of 10CPR50, Appendix H, are as follows:

Initial

~RT ~

End of Design Life RT,

~

Time of 'First Capsule Removal

~ (12 P)

> (12 P) s 100 P

> 100 P

10 EPPY 6 EFPY Singe the results of the analysis of the vessel wall dosimeters indi-cate that the fast neutron flux is higher than that predicted in the

PSAR, the current pressure-temperature limits for operation and testing should be reviewed to determine if they are consistent with the projected adjusted values of RTNDT between the first refuelling and the time of removal of the first surveillance capsule.'f not, revised pressure-temperature limits should be established in accordance with Section III, Appendix G, of the ASME Boiler and Pressure Vessel Code.(8)

0

23 VI.

REFERENCES 1.

Regulatory Guide 3..99, Revision 1, Office of Standards Development, U.S. Nuclear Regulatory Commission, April 1977.

. 2.

Title 10, Code of Federal Regulations, Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements."

3.

"Mechanical Property Surveillance of GE BUR Vessels,"

NED0-10115, July 1969.

4.

ASTM E 523-76, "Standard Method for Measuring Past-Neutron Plux Density by Radioactivation of Copper, " Annual Book of ASTM Standards, Part 45.

5.

ASTM E 263-77, "Standard Method for Determining Past-Neutron Plux by Radioactivation of Iron," Annual Book of ASTM Standards, Part 45.

6.

ASTM E 482>>76, "Standard Recommended Practice for Neutron Dosimetry for Reactor Pressure Vessel Surveillance," Annual Book of ASTM Stan-dards Part 45.

7.

Telecon, E. B. Norris to Ken Hogue (NRC Staff), January 19, 1977.

8.

ASME Boiler and Pressure Vessel

Code,Section III, Appends G,

"Protection Against Non-ductile Pailure."

9.

Letter from J. E. Gigleland, TVA, to A. Schwencer, NRC, regarding Docket Nos. 50-259, 50-260, and 50-296, dated August 23, 1977.

l'r I

I