ML20090H445

Analysis of Vessel Wall Neutron Dosimeter from Browns Ferry Unit 2 Pressure Vessel

ML20090H445
Person / Time
Site:	Browns Ferry
Issue date:	09/30/1979
From:	Lindholm U, Norris E SOUTHWEST RESEARCH INSTITUTE
To:
Shared Package
ML18025C005	List: ML20090H438 ML20090H445 ML20090H451
References
NUDOCS 8310280150
Download: ML20090H445 (26)

Text

,,

... ,1 ,-

. .(

, , lV ,

lk ,

f '

J / <

en ,-

1. ANALYSIS OF THE VESSEL WALL NEUTRON c

~

2 DOSIMETER FROM BROWNS F RRY UNIT- 2 i 1

PRESSURE VESSEL - i' -

t A

", \.;,

' '[ 6y , t

,3 J, E. B. Norris .~ ,,

j/

of

• a 1 4 ' '

,s t , r t.

I

'l ,

l }

i j

/ 1 ,,

Il.a FINAL REPORT I

, y i- .-

! SwRI Project 02 4884-002 -

r, M

. < 'c.

i J #

'j . <-

.,; to Tennessee Valley Authoritj 505 Edney Building j Chattanooga, Tennessee 37402 -

i.

~,

j .

.Cl I

y]j

Septe mbe r 1979 -

J):

) r

'!j

.'.]

/I s ,

' /

}j 1 SOUTHWEST RESEARCH INSTITUTE

[

h_h ~:_w

,,, SAN ANTONIO ,' HOUSTON g ,

.i 8310280150 831019 <

l PDR ADOCK 05000259 '

M P PDR .

i

.. - _ _ _ _ . _ _ . _ ., , _ _ - . _ _ _ _ _ - . . _ . ._,___,___.,...__._.-__,_.s . . . . , _ _ _ - _ . _ _ _ , _ _ . . _ , . - _ , , _ _ , . ,

SOUTHWEST RESEARCH I N S T I T U T E"

, Post Office Drawer 28510, 6220 Culebro Road o~ '

San Antor.io, Texas 78284

, ' f: - ,

ANALYSIS OFGHElVsSSEL WALL NEUTRON DOSIMETER. FROM'BNOWNS FERRY UNIT 2

< i PRESSURE VESSEL , ,

t

/ 4 [

i

1. I (

i, \?

.1 by p B. Norris y

FINAL REPORT Switi Project 02-Illil1-002

• h I a

, _ r. .i ~

• / / '

, to .

Tennessed %lley Authority -

<505 Edne, y, Iluildin.-

Chatt'a nooga, Tennessee 37402 e '-

'c'

.- r, '

4

d f

/

September 1979

\

p

.f a

f, - \ _!

- ')* ,

'i Approved

,L -

j s'

U.

S. Lindholm, Director r ,/  ;

^

, Department o',I, Materials Sciences j 'h ' '

f- .(,

- al.

<r *  ? * '

,.._m, ',

~

v

\ l .

4 E ' t ABSTRACT

l I

The vessel wall neutron dosimeter capsule from Browns Ferry _ Unit 2' i  ;$.

i- 'has been analyzed. The results' indicate that the-peak value of fast neu-r tron flux incident on the reactor vessel wall is 1.'06' x 109 cm-2 3,e-1, a

' E'> 1 MeV. Although this results in a lifetime neutron fluence of 1.34 x 1018 cm-2, about 3-1/2' times that predicted in the FSAR, it is less than the design limit of '1.0 x,1019 cm

-2 for 40 years of operation. -

Based on a conservative estimate of the neutron embr ttlement re- -

sponse of the core beltline' materials, the increase in the reference nil ductility temperature may exceed 100 F by the end of the design life of

.;i

, the Browns' Ferry Unit 2. vessel'. The bases for selecting a capsule removal ~

, sch~edule in accordarce with Appendix H of 10CFR50 are discussed.

e

,.T.

t 11 i

MM m e - , .. , , , . - . . - , - , - , ,- ,- - - , , - - , ,- , . - -, , - , - - .

-~

j, t

TABLE OF CONTENTS

- f.:!.81 l~

I.- '

SUMMARY

OF RESULTS AND CONCLUSIONS ~l INTRODUCTION

• II.: -2

! III. EVALUATION OF VESSEL WALL NEUTRON DOSIMETER CAPSULE 3

~ !.

.j. IV. CALCULATION OF NEUTRON FLUX DENSITY AND FLUENCE 9 i V. DISCUSSION 18

.; VI. REFERENCES 23 e

T 1

e

, 111 i

i

?

,h M amumen . - - - ___ . . . ~ . - - _ . .

I, ., .,- , , - . - . ~. - - - - - . - , . - - - - - - - - - , - - - . - - - - . - - - - - - - - - . . - - . - - - -

l I.

SUMMARY

OF RESULTS AND CONCLUSIONS The results of the analysis of the Browns Ferry Unit 2 vessel wall dosimeter indicate that the peak fast neutron flux (E > 1 MeV) at full power during core cycle 1 was 1.06 x 109 cm-2.sec-1 As a result, a 40-j_ year design life fast neutron fluence of 1.34 x 1018 cm-2 is predicted,

[ about 3-1/2 times the calculated design life fluence given in the Final l

] Safety Analysis Report (FSAR), but considerably less than the FSAR design ~

limit _of 1.0 x 1019 cm-2 Utilizing the radiation damage trend curve in

'the FSAR, 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 2 pressure vessel belt-line materials may result _in sensitivities to neutron radiation embrittle-ment different from the response curve given in_the FSAR. Using the 0.3%

Cu RTNDT_ adjustment curve in Regulatory Guide 1.99(1)* the total shift might reach 102 F at the vessel wall I.D. and 84 F at the vessel vall 1/4t by the end of the 40-year design life of Browns Ferry Unit 2 pressure vessel.

-The capsule removal schedule necessary to meet the requirements of 10CFR50(2), Appendix H, depends on the value of the adjusted RTNDT at the

[ end of the design life of the reactor vessel.

If the initial RTNDT Of the Browns Ferry Unit 2 pressure vessel beltline material was higher than 16 F, I

the first specimen material surveillance capsule should be removed after six full power years of operation.

Superscript numbers refer to the list of references at the end of the text.

e f .

M m vm .- -.

2 II. INTRODUCTION-The Browns Ferry Nuclear Plant operated by the Tennessee Valley Authority (TVA), consists of three 1065 Mwe (3293 Mut)-Boiling Water Reactor (BWR) 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-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.

I The surveillance program is described in detail in NEDO-10115. (3) Be-I cause of the low level: of fast neutron flux density at the vessel wall predicted f by design calculations, the first surveillance capsules containing mechanical i

j test specimens .are not scheduled for removal un_til four years of operation have i

ll accrued' . However, the vessel wall dosimeter capsules are scheduled for removal

!- ~

i .at the first refuelling to provide a check on the design flux calculations.

I

! ,This report describes the results obtained from the testing and analysis of the contents of the vessel wall neutron dosimeter capsule from Unit 2.

1 p

i

.i:

I i

i 6

I t

O M _

.i 3

/

5 III. EVALUATION OF VESSEL WALL NEUTRON DOSIMETER CAPSULE f

The vessel wall' neutron dosimeter capsule was removed from the Browns Ferry Nuclear Plant Unit 2 vessel during a refuelling outage which began on March 18, 1978, at the end of core cycle 1. This capsule, shown in Fig-ure 1, contained three each pure copper and pure iron dosimeter wires. The nuclear reactions of interest for these wires are:

63Cu(n,a)60co 54Fe(n p)54Mn ll The capsule was shipped to the Southwest Research Institute (SwRI) laboratories the week of December 4,1978, in a cask supplied by SwRI. The

. capsule was opened in one of the hot cells at the SwRI Radiation Laboratory 6

with a hand hacksaw. This could be done because of the low level of activity exhibited by the capsule. The contents were' examined and visually identified i

I 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:

No =

x 10-3 (1)

.- where: N = 6.02 x 1023 nuclei per go atom;

~

c =

weight fraction of detector isotope in detector specimen; A = atomic weight of detector element, gm.

The absolute activities of the dosimeter 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

=

e1

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

l

. o 5

counting for'esch photopeak of interest, 842 kev for 54Mn and 1173 kev for 60C o, with 54Mn and 60 Co isotopic standards traceable to the U.S. Bureau 'of Standards. The counting system and techniques have been previously checked against two other laboratories, see Table I.

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

A(Toa) T tal e unts under ph t ceak of energy E less "backtround" g T(E), (g)

Eff(E) r v P exp - Act T(E)u where: T = counting time, sec; w = weight of wire, mg; P = peak-to-total ratio; A = disintegration rate, day-1; c1

= elapsed time between TOR and counting date, days.

T(E)s = intrinsic efficiency factor for the standard source counting geometry; T(E)u = intrinsic efficiency factor for the unknown source counting geometry.

In this program, T(E)s/T(E)u was equal to uni,ty 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:

A(f0R)

= (1 - exp - ATm )(eXP - Atm) s m=1 where: m = operation period; Tu = equivalent operating time at selected power level for the mth period, days; ts = elapsed time from the end of the mth period to TOR, days.

The values of Tm and tm were determined by dividin'g the Unit 2 plant opera-tlan into 23 operating periods, as se=marized in Table III.

4 M *  ;-m

I 6

)

TABLE I RESULTS OF-INTERLABORATORY GAMMA-COUNTING PROGRAM Sample Assumed Activity at TOR (dps/mg)

Identification Isotope Half-Life SwRI Other Top'(Co-cd)- 60Co 1913 d 2.69 x 107 2.68 x 107 (a)

Bot (Co-Cd) 60 Co 1913 d 2.67 x 107 2.48 x 107 (*)

Top (Co) 60 Co 1913 d 6.03 x 107 5.83 x 10 7 (*)

Bot (Co) 60Co 1913 d 6.23 x 107 5.93 x 10 7 (a) 2056 60Co 1913 d 1.41 x 107 1.37 x 107 (b) 2062 60Co 1913 d 5.57'x 106 5.20 x 106 (b)

R9. 54Mn 312 d 1.30 x 104 1,32 x 104 (a)

R13 54Mn 312 d 1.24 x 104 1,29 x 104 (a)

R16 54Mn 312 d 1.21 x 10 1.23 x 104 (a)

(a) BNWL (b) WES e

r i +

7 TABLE II RESULTS OF ACTIVATION ANALYSES OF DOSIMETER WIRES EXPOSED LN BROWNS FERRY UNIT 2 VESSEL FROM 8/3/74 THROUGH 3/18/78 Weight Count Rate A(TOR) (a) As Isotope Foil (mg) (dpm) (dps/mg) (dos /mg) 60Co Cu-1 456.7 8.764 x 10 4 3.198 21.89 60Co Cu-2 459.8 8.636 x 104 3.130 21.42 60Co Cu-3 467.8 8.568 x 104 3.053 20.89 Average 21.40 54Mn Fe-1 137.4 4.791 x 105 58.11 118.4 54Mn Fe-2 135.3 4.773 x 105 58.80 119.8 54Mn Fe-3 135.0 4.653 x 105 57.44 117.0 Average 113.4 (a) Specific activity at time of reactor shutdown, 3/18/78. Disinte rates are subject to a 3% (1 S%) measurement uncertainty.(4,5) gration (b) Saturated activity at the 3293 Mwt power level.

t________ ~ =

L.

8 TABLE III OPERATIONS

SUMMARY

.- BROWNS FERRY NUCLEAR PLANT, UNIT 2 Operating deactor Equivalent

• Period g,,,, ,

'Cecay Time Operating Shutdown Power Operating in Days (s) Start Stop Days Oays (Mcth) Days (%) (e-)

08-03-74 08-13-74 11 -

635 0.193 1313 08-14-74 08-24-74 -

11 - - -

,2 . 08-25-75 08-28-74 4 -

423 0.128 1298 08-29-74 08-30-74 -

2 - - -

3 08-31 09-03-74 9 -

5.919 1.797 1257 09-09-74' 09-10-74 -

2 - - -

4:- 11-74 09-19-74 9 - 8,674 2.634 1276 09-20 09-20-74 -

1 .

5 21-74 10-07-74 17 -

26.031 7.905 1258 10-08-74. 10-25-74 -

13 - - -

6 10-26-74 11-01-74 7 - 9,031 2.742 11-02-74 1233 11-02-74 -

1 - - -

7 11-03-74 11-24-74 22 - 33,797 11.732 1210 11-25-74 11-27-74 -

3 .- - -

8 11-28-74 12-03-74 6 9.757 2.963 1201 12-04-74 12-07-74 -

4 - - -

9 12-08-74 01-11-75 35 -

84,603 25.692 '

1162 01-12-75 01-12-75 -

1 - - -

10 01-13-75 01-19-75 7 -

10,602 3.219 1154 01-20-75 01-23-75 -

4 - - -

11 01-24-75 02-11-75 19 - 38,926 11.821 1131 02-12-75 02-15-75 -

3 - - -

12 02-16-75 03-22-75 36 -

109,756 33.330 1092 03-23-75 08-31-76 -

528 - - -

13 09-01-76 10-22-76 52 -

67,279.1 20.431 $12 10-23-76 10-23-76 -

1 - - -

14 10-24-76 12-10-76 -48 0 91,682.3 27.842 463 12-11-76 12-12-76 -

2 -- - -

15 12-13-76 01-24 43 -

125,854 38.219 413 01-25'77 01-25-77 -

'l - - -

16 01-26-77 03-10 44 -

122,653 37.247 373 03-11-77 03-11-77 -

1 - - -

17 03-12 04-16-77 36 -

95,936 29.133 336 04-17-77 04-21-77 -

5 - - -

18 04-22-77 05-02-77 11- -

23,'53 7.092 320 05-03-77 05-03-77 -

1 - - -

19 05-04-77 03-11-77 -100 - 277,033.4 84.128 219 08-12-77 08-14-77 -

3 - - -

^20 08-15-77 10-18-77 65 - 183,554.9 $3.741 151 10-19-77 12-14-17 '- 57 - - -

21 12-15-77 02-05-78 53 -

133.319.6 42.004 41 02-06-78 02-06-73 -

1 - - -

22 02-07-78 02-13-73 7 -

11.107.1 3.373 33 02-14-78 02-14-78 -

1 - - -

23 02-15-73 03-18-78 32 -

37.765.2 26.652 0 Tacals 1,567,691.6 476.068 (a) At 3293 Mwe (b) f476.068 days = 4.1132 x 107 seconds 4 -

%~ M a.e'**-*--> '~

9 IV. CALCULATION OF NEUTRON FLUX DENSITY AND FLUENCE W e energy-dependent peutron flux dersity,'$ (cm-2 3,e-1) , the spec-trum-averaged activation crons-section, 3 (cm ), and the saturated activity As , of each dosimeter wire. are related as follows:

$=ND.

o (4)

In the early days of nuclear pressure vessel surveillance activity, the value of 3 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 improved, the use of calculated neutron spectra has increased and is now

-permitted by NRC Regulatory Guide 1.99(2) for application to reactor pressure vessel wall locations.

-The neutron flux energy and spatial distribution were calculated for the Browns Ferry Unit 2 pressure vessel with the DOT 3.5 two-dimensional discrete ordinates transport code, a 22-group neutron cross section library, a P t ex-pansion of the scattering matrix and an'S8 order of angular quadrature. An R-0 calculation was made for a horizontal plane perpendicular to the vertical axis of the core, and an R-Z calculation was made for a vertical plane through the axis of the core and the location of the vessel wall dos 1=eter. A one-eighth segment, shown in Figure 2, was taken to be representative of the R-0 geometry because of the symmetry invcived. The boundaries of the core, core shroud, jet pumps, and vessel wall were described in R-0 coordinates. The i

4 w ,

= w

+ --

9 - #-g- + e r - --- --- - -aw-a

PERIOD 10 LOCATION AZIMUTH YEARS Wall 30* 4 Wall 120* 12 Wall 300* 32 Wall 30* 1 Capsule QD Capsule [

r t C\

M, Dosimeter

/ \

n.-

(

+ n.

N.

\ /

s

\ '

/ f Capsule [ 45'

'#0 Core Shroud PUN'rMW 30' I O o p 15

• Jet Pumps

/

/ l } oa Vessel Wall /

l i

i FIGURE 2. ONE-EIGHTH SECMENT FOR FLUX CALCULATIONS AND i LOCATION OF SURVEILLANCE CAPSULES l - . . . - . . . - . -

s-.. ,

11 core was subdivided into two regions, an inner region with one-sixth of the controlcrods 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 18% Cr, 8% Ni, and 74% Fe, and the pressure vessel was assumed to be 98% Fe._ The coolant outside of 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 information on the neutron energy spectrum at the vessel wall neutron dosimeter capsule location. In addi-tion, the R-0 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 factors) can be derived.

( The neutron spectrum at the vessel wall dosimeter location, as deter-I mined with the R-0 model, and the group-averaged cross-sections for the dosimeter reaccions of interest are given in Table IV. The spectrum-averaged cross sections, 3, were determined from the relationship:

10.0 E a(E) $(E) dE

[ 3 (E > 1 MeV) =

1f f .

(5)

E $(E) dE

! 1.0 l

l

. __h.

.n e -

-m. - - - - - , . ,,n,- . - - . - - , - - - - - - , - - - .. -,,-n--.., ,,-----c -- - . - -n,. ,

12 TABLE IV

. DETERMINATION OF REACTION CROSS SECTIONS FOR DOSIMETER WIRES 54Fe(n,p) S4Mn 63Cu(n,a) 60Co Energy Range Normalized Neutron Cross Section, Cross Section, (MeV) Flux, 4(E) 0Fe(E) (barns) - GCu(E) (barns) 8.18 - 10.0 .0351 .581 .0380 6.36 - 8.18 .0879 .577 .0144 4.96 - 6.36 .1283 .491 .0023 4.06 - 4.96 .1029 .354 .00025 3.01 - 4.06 .1167 .185 .00010 2.35 - 3.01 .1505- .078 .00006 1.83 - 2.35 .1280 .023 .00002 1.11 - 1.83 .2179 .0014 .00002 1.00 - 1.11 .0327 -

.00001 g

~

_ op , = . 20 7,(a) Cu

= .00295 (a) e (a) Spectrum-averaged cross sections are subject to a t15% (1 S%)

uncertainty.(6)

I 2

a ._.o.

_ A

[

13 Substituting the value of iry, into Equation (4) along with the average value 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 9.12 x 108

-2 cm Similarly, the fast neutron flux determined from the copper dosimeters  ;

is.l.11 x 109 cm-2 The discrepancy 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(6) , errors as large as- 7% (1 S%)

in the determination of disintegration rates and 15% (1 S%) in spectrum-weighted group-averaged cross sections can be encountered, which results i.n a combined error of 16.5% (l'S%) for the calculation of r_utron 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-0 model and shown in Figure 3(a), indicates that the vessel wall neutron dosin-eter capsule was placed at the azimuthal position of maxi =um fast neutron flux.

Therefore, it is concluded that the calculated flux derived fro = the analysis of the dosimeter wires is a direct measure of the maximum fast flux incident on the pressure vessel opposite the vertical core centerline.

However, the' axial flux distribution, as calculated with the R-Z codel

, and shown in Figure 3(b), indicates that the peak fast neutron flux is 5%

higher at a position 20 cm below the capsule location. Therefore, the lead factor, the ratio of the fast flux at the capsule location to tha peak fast flux incident on the pressure vessel I.D., for the Browns Ferry Unit 2 sur-veillance capsules is calculated to be 0.95.

Based on the results of the DOT 3.5 calculations and the dosi=etry re-sults, the peak fast flux incident on the Browns Ferry Unit 2 vessel during

. k, - -

4- . _ . . . _ . _ - _ _ _ . _ _ . _ ,

g yM* -  %

9W co .1. 55 i.o .\ 0

@,3r(ce 6v 09gg top +, o.24x 10 8

@ o.34-0.B9

.BD -m - 0. + 1

] 46

./ '

/ o.17 o'95 o.59

,'o g m o.67 j .

[- / 96 O m - I. o l i 3d - , -o.76 f_ . e 1, o l l _.,- 0.86  : -Lo6

-*- 0.65 -

-o.77 -

--: - 0.78 IS* ~

---

. o.So

-0.75

~ o.3 7

. < 0.69 OA8 0'\ w o.48 444 8

_ 3,g, e a59 x.10 24 Neu. tron Flux , E >1 M.y 03o ,

w ( J .se d

0. 33 ko s o

(a) Azimuthal Distribution at Core (b) Vertical Distribution at 30* Azimuthal Vert 1 cal Centerline Position FICURE 3.

5 CALCULATED DISTRIBUTIONS OF FAST NEUTRON FLUX INCIDENT ON ,TIIE BROWNS FERRY UNIT NO. 2 PRESSURE VESSEL I.D.

l 15 the first core cycle is calculated to be 1.06 x 109 cm-2.sec-1, E > 1 MeV.

Therefore, the neutron fluence per effective full power year (EFPY) is 3.34 x 1016 cm-2, E > 1 MeV. Assuming 100% availability over the 40-year design life of the plant, the design life neutron fluence received by the vessel is predicted to be 1.34 x 1018 cm-2, 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 Figure 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.(7) This conservatism is preferred because as the flux greater than 1 MeV decreases with distance into the pressure vessel wall, the popula-tion of ne itrons in the energy range of 0.1 MeV to 1 MeV increases and there is some degradation caused by these lower-energy neutrons.

Since the pressure-temperature limits for reactor operation and test-ing are based on requirements of the ASME Boiler and Pressure Vessel Code (8) ,

the fluence at the 1/4t and 3/4e positions within the pressure vessel wall are of specific interest. Utilizing the conservative estimate of the at-tenuation 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 6-3/16-in. Browns Ferry Unit 2 pressure vessel are su=marized in Table V.

i i

16 13 10 gasigss#ign issgi44 ji s s i j a i s i j a is i j s e s ij in iij ai si gi a r a j aii g 12 ,_

10

~

=

11 -

h 10 -

-=

~

d

• - Z g

- d -

g -

- s 1

o

• o 8 -

d y &

a U

10 8 e

o 10 V u u u 0

a a e u 0 U C3 U 3

~ _

ko $ 8 $ ~

! - a h w

e u n -

C o

k

$ 109 - -

a = -

z - =

.\

s

\.

8 10 _

~

7 ' ' '

10 '

220 240 260 280 300 320 340 Radius, cs FIGI 4. CALCULATED NEUTRON FLUX BETWEEN CORE AND PRESSURE VESSEL I.D.

ii NOREV.IZED TO THE VESSEL WALL DOSDfETER RESULT

&w w. . _ _ . - --- -

L

I '

i=L - __

. TABLE V

)

CALCULATED PEAK NEUTRON FLUX (a) AND FLUENCE (a)

FOR BROWNS FERRY UNIT 2 PRESSURE VESSEL WALL Vessel Wall Relative Fast Fast Neutron Flux Location Neutron Flux Fast Neutron Fluence (cm-2)

Density (cm-2.sec-1) _1.303 EFPYlb) 4 EFPY 40 EFPY I.D. Surface 1.00 1.06 x 109 4.36 x 1016 1,34 x 1017

, 1.34 x 1018 1/4r 0.67 7.1 x 108 2.9 x 1016 9.0 x 1016 9.0 x 1017 3/4t 0.24 2. 5 x 108 1.0 x 1016 3. 2 x 1016 3.2 x 1017 4

).

'l i

(a) E > 1 MeV. Calculated flux and fluence values subject to a 116.5% uncertainty. (6)

(b) End of core cycle 1.

l I

I 1

s

, a

r 18 V. DISCUSSION The predicted value of the peak neutron fluence (E > 1 MeV) for the Browns Ferry Unit 2 pressure vessel after 40 EFPY of operation is given in the Final Safety Analysis Report (FSAR) as 3.8 x 1017 cm-2 (E > 1 MeV).

The analysis of the vessel wall dosimeter capsule projects.that the peak-neutron-fluence will be 1.34 x 1018 cm-2, about 3-1/2 times the predicted value, but considerably less than the FSAR design limit of 1.0 x 1019 cm72, .

A similar trend has been noted in several other BWR plants with which SwRI has been associated. For example, the neutron dosimetry analyses performed on the first capsules removed from the Elk River, Lacrosse, Millstone Point 1, and Pilgrim reactors indicated that the fast neutron flux densities were higher than the design values by factors ranging from 2 to 6.

The estimation of a 40-year neutron fluence from less than two years of operation is a large extrapolation and will,be subject to revision at the time of the next capsule removal, currently scheduled after four years of operation. In the meantime, however, the pro'jected peak fast fluence factor of 3.34 x 1016 cm-2 per EFPY can be employed to predict the change in the reference nil ductility temperature (RTNDT) as a function of reactor power generation.

The threshold value of neutron fluence for the 550 F embrittlement of ferritic steels is generally taken to lie between 1017 and 1018

, cm-2 (E >

1 MeV).- The proposed relationship between fast neutron fluence and the change in the RTNDT of the Browns Ferry Unit 2 reactor vessel, as given in Figure 3.6-2 of the FSAR, is reproduced in Figure 5. Added to this figure are (1) an arrow indicating the fast neutron fluence on the vessel I.D. at

, the end of core cycle 1, and (2) an additional abscissa relating neutron t

Ma m mm. . _

I 19 Reactor Operation, EFPY l 4 10 40

.$ 200 ' I I

E s

E je 150 5

f' 2 100 j-e b

+ l /

- .2o 5-c: 50 dr

o

-[7 l

0- .

y 1018 1017 ctB gg33 g% NEUTRON FLUENCE (>1 MeV) (pt), mt l

l

? .

l FIGURE 5. VESSEL MATERIAL NEUTRON DGRITTLDfE:iT CURVE FROM BROWNS FERRY UNIT 2 FSAR l

l l

l v' A -_ - -,

20 5

. fluence and effective full power years as determined from the vessel wall surveillance capsule. Figure 5 indicates that the RTET of the Browns Ferry Unit 2. pressure vessel will begin to increase af ter an expcsure of 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 about six EFPY of operation to reach this fluence at the 1/4t location in the pressure vessel wall.

Also, the predicted shift in RTET 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 (Figure 5)'

corresponds closely with the RT NDT adjustment curve of Regulatory Guide 1.99(2) 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 RTET of the Browns Ferry Unit 2 vessel at the end of design life would be 102 F at the I.D. surface and 84 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.

Section II.3 of Appendix H of 10CFR50(2) describes three cases which govern the surveillance specimen capsule removal schedule. The first case, which applies when the adjusted RTET of the reactor vessel steel will not exceed 100 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 F at the end of the design life, requires that the first specimen capsule be removed when the predicted adjustment of the reference temperature is approximately 50 F, or at one-fourth service life, whichever is earlier.

I W

-4 3 acmm s.- _ .-e -v + -..-.. _

1 g :_.

I 1 l

U U

• g .

r ,

y _ ,,,,,m,,,,un,,,,,,,,,,,,,,,,,._,_.,.,_,,,,,,,,,,,,,,,,,,,mm,m,,,,m . .

i k l' l 2 D 1/2 l .'

r A = 140 41000 (% Cu - 0.08) + 5000 (% P - 0.000)] lf/10 I - -'-

400

~ ~ ' ' *

? '

brin 0 '

4 j ,300 '

( '  ! -

i

[j:

? i r' '

j.:,

y 7 '

g. .

li r: : l .--

(5 u g - -- - -

y .

g . . .

I! ! ! H 2-r 200 jf q k? -

'l y8 , '

3 p ,

, ~

f:

{  : l .! . j .-

. f. ,y' [

t l {t

. 'y '

E l l '. _ . i ,

M?

,#{..--/, -

l' g 1 l l q l 100 l  ; g; - ,,, - - -

r--- -

JM,ry <

n; ' e '.<' g.

WM '

5 h  :-  !

l 3 7 ,

%, I N'

..l ,a; r i +-

.E t i b - -

l ,,

' , * .s b5 - -

i l h -

h i t i ,

s

" y\

..- j-, -

ll i  :

7g 50 l a

}o ,0.25 l

.s--. ..- 1

- -  :- -e,-

0.3r s .o.30 0.20  :  : 0.15% Cu. .U.10% Cu j -  : ;<~

l l I 'l ' 's i ' ' I! lo

-3 %l' ="i655 LOWEll lMil l'

i I ;l y

' l l'{

.,+ j l '. .

' ~

.I l ",)l! ! -

~

% Cu = 0.00

Qi  % P = 0.000 l -

? I ;

- . i

? ,

1 s l l 1 1 -

l e l 5  :

l II

[,:r -/-

a f

l, NI -

- - h5r :I :

i l ' '

[.4{ / -'fi ! [; : i h li L. , ,I i , 1 i 2X10 D 4 G 0 10 17 2 4 6 0 10 10 2 4 6 0 10 D 2 4 0 F LUENCE, n/cm2 (E > 1MeV)

Figure 1 Pseilicted Atijustment of fielerence Ternpesatuse, **A", es a function of Fluence and Copper Content. .

For Copper and Phosphorus Contents Other Than Those Plotteil, Use the Empression for **A** Given on the figure.

(Note

Dashed lines represent GE recommended extrapolations to 20*F for BWR operation.)

FICURE 6. REFERENCE TEMPERATURE ADJUSTMENT CURVES FROM RECULATORY GUIDE 1.99(1) w S

1

l 22 Based on an end-of-design life increase in RT NDT Of O' * "' ""

from the 0.3% Cu response curve in Figure 6 at the 1/4e fluence obtained from the vessel wall dosimeter, the capsule removal schedule necessary to meet the requirements of 10CFR50, Appendix H, are as follows:

Initial End of Design Life Time of First RTNnT RTNDT Capsule Removal 5 (16 F) s 100 F 10 EFPY

> (16 F) > 100 F 6 EFPY Since the results of the analysis of the vessel wall dosimeters f.ndi-cate that the fast neutron flux is higher than that predicted in the FSAR, the current pressure-temperature limits for operation and testing should be reviewed to determine if they are consistent with the projected adjusted values of RT NDT between the first refuelling and the time of removal of the first surveillance capsule. If not, revised pressare-temperature limits should be established in accordance with Section III, Appendix G, of the ASME Boiler and Pressure Vessel Code.(8)

~m sneammaman _ - - - - . - - . - -

/ 23 VI. REFERENCES

~

1. Regulatory Guide 1.99, 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 BWR Vessels," NEDO-10115, July 1969.

4. ASTM E 523-76,' " Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Copper," Annual Book of ASTM Standards, Part 45.

5. ASTM E 263-77, " Standard Method for Determining Fast-Neutron Flux 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, Appendix G,

" Protection Against Non-ductile Failure."

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

i lL

- me e - z = a_ = ~ .