ML20090H451

Analysis of Vessel Wall Neutron Dosimeter from Browns Ferry Unit 3 Pressure Vessel

ML20090H451
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Site:	Browns Ferry
Issue date:	09/30/1979
From:	Lindholm U, Norris E SOUTHWEST RESEARCH INSTITUTE
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i j ANALYSIS OF THE VESSEL WALL NEUTRON j DOSIMETER FROM BROWNS FERRY UNIT 3 i PRESSURE VESSEL by E. B. Norris l

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FINAL REPORT j SwRI Project 02-4884-003

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to Tennessee Valley Adthority 505 Edney Building.

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.t September 1979 l

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__ e SOUTHWEST RESEARCH INSTITUTE l) e SAN ANTONIO HOUSTON

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SOUTHWEST RESEARCH INSTITUTE Post Of fice Drawer 28510, 6220 Culebro Road Son Antonio, Texas.78284 ANALYSIS OF THE VESSEL WALL NEUTRON DOSIMETER FROM BROWNS FERRY UNIT 3 PRESSURE VESSEL by E. B. Norris FINAL llEPollT Swlli Project 02-111114-0 0 3

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to Tennessee Valley Authority 505 Educy Iluilding-Chattanooga, Tennessee 37402 September 1979 Approved-U. S. Lindholm, Director '

Department of Materials Sciences

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ABSTRACT /  ?

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The vessel wall neutron dosimeter capsule from' Browns > Ferri Unic 3'

, .i has been analyzed. The results indicate that the peak value df f ast ceu-tron flux incident on the reactor vessel vall is 1.04 x 109 cd-2.sec *l, '

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E > 1 MeV. Although this results in a lifetime neutron fluence of 1.31 x Y

1018 cm-2, about 3-1/2 times that predicted in the FSAR, it is 1 ss than J

the design limit of 1.0 x 1019 cm

-2 for 40 years of operatioy./,-

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, ,i Based on a conservative estimate of the neutron embrettelement re- , -

sponse of the core beltline materials, the increase in ne rererence nil ductility te=perature may exceed 100 F by the end of the ed' sign. life of' , ' ,

.I the Browns Ferry Unit 3 vessel. Thebasesforselectingacapsulere=o51 '

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schedule in accordance with Appendix H of 10CFR50 are d'iscuss5d. <

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SUMMARY

OF RESUI.TS AND CONCLUSIONS i # 1-II. INTRODUCTION- .l.. i

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EVALUATION OF VESSEL WALL KEUTRON DOSIMETER CAPSULE.

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CALCULATIOp 0F 'JEUTRON-FLUX DENSITY AND FLUENCE '

9 V. ,. . DISCUSSION '

18 VI. . ,; REFERENCES ' .

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SUMMARY

OF RESULTS AND CONCLUSIONS 1

The results of the analysis of the Browns Ferry Unit 3 vessel wall dosimeter indicate that the peak fast neutron flux (E > 1 MeV) at full power during core cycle 1 was 1.04 x 109 cm-2.sec-1 As a result, a 40-year design life fast neutron fluence of 1.31 x 1018 cm-2 is predicted, about 3-1/2 times the calculated design life fluence given in the Final 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 3 pressure vessel belt-line materials may result in sensitivities to neutron radiation embrittle-ment differeat 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 101 F at the vessel wall I.D. and 83 F at the vessel wall 1/4t by the end of the 40-year design life of Browns Ferry Unit 3 pressure vessel.

The capsule removal schedule necessary to meet the require =ents 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 3 pressure vessel beltline material was higher than 17 F, the first specimen material surveillance capsule should be removed after seven full power years of operation.

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

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2 II. INTRODUCTION The Browns Ferry Nuclear Plant operated by the Tennessee Valley Authority (TVA), consists of three 1065 Mwe (3293 Mwt) 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, veld metal and heat-affected zone), a vessel wall do-s1=eter capsule, and three surveillance capsule baskets containing Charpy.V-I 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. (3) Be-i cause of the low level of fast neutron flux density at the vessel wall predicted by design calculations, the first surveillance capsules containing mechanical test specimens are not scheduled for removal until four years of operation have accrued. However, the vessel vall dosimeter capsules are scheduled for re= oval at the first refuelling to provide a check on the design flux calculations.

l This report describes the results obtained from the testing and analysis of the contents of the vessel wall neutron dos 1=eter capsule from Unit 3.

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EVALUATION OF VESSEL WALL NEUTRON DOSIMETER CAFSULE -

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'i The vessel wall neutron dosimeter capsule was removed from the Browns

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4 Ferry Nuclear Plant Unit 3 vessel during a refuelling outage which began on

'I' September 8,1978, at the end of core cycle 1. This capsule, shown in Fig-t I 'tre 1. contained three each pure copper and pure iron dosimeter wi.res. The

l. - nuclear reactions of interest for these wires are:

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63Cu(n,a)60Co '

1 54Fe(n,p)S4,g 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 or.ened in one of the hot cells at the SwRI Radiation Laboratory 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 as either iron or copper.

The dosimeter vires were prepared for analysis by weighing on a precision laboratory balance. The number of target atoms per mg, N , was computed for o

each wire as follows:

No= " x 10-3 (1) where: N =

6.02 x 1023 nuclei per gm atom; e =

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 j.

i NaI(Th) scintillation detector and an NDC 2200 multichannel analyzer. The experimental efficiency, Eff(E), of the system was determined on the day of i

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. counting for each photopeak of interest, 842 kev for 54 Mn and 1173 kev for

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( Standards. The counting system and techniques have been previously checked against two other laboratories, see Table I.

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

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p , Total counts under photooeak of enerty E less " background ** y T(!). (2)

Eft (E) t w F exp - Att ,

T(E)u where: T =

counting time, see;

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weight of wire, mg;

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peak-to-total ratio;

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disintegration rate, day-1;

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elapsed time between TOR and counting l date, days.

T(E)s = intrinsic efficiency factor for the standard source counting geometry; T(Z)u = 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 fo'r each dosimeter wire are summarized in Table II. The last column in Table II lists the saturated activities, A , of s the dosimeter wires computed for the full power level of 3293 Mwe as follows:

A(TOR) =

E (1- exp - AT m)(CXP - Atm) i 4 m=1

• t where: m =

operation period; Tm =

equivalent operating time at selected power level for the mth period, days;

=

tm elapsed ti=e from the end of the mth period to TOR, days.

The' values of T m and to were determined by dividing the Unit 3 plant opera-

._ tion into 20 operating periods, as su=marized in Table III.

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

, Identification Isotooe Half-Life SwRI .0ther 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) 60Co 1913 d 6.03 x 107 5.83 x 107 (a)

,; Bot (Co) 60Co 1913 d 6.23 x 107 5.93 x 107 (*)

l 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) i R9 54Mn 312 d 1.30 x 104 1.32 x 104(a)

R13 54Mn 312 d t 1.24 x 104 1.29 x 104 (a)

R16 54Mn 312 d 1.21 x 104 1.23 x 104 (^)

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7 TABLE II RESULTS OF ACTIVATION ANALYSES OF DOSIMETER WIRES EXPOSED Di BROWNS FERRY UNIT 3 VESSEL FROM 9/1/76 THROUGH 9/8/78 Weight Count Rate A(TOR) (a) i Isotope Foil (mg) As(

(dom) (dps /me) (dos /mg) 60Co Cu-1 460.4 1.048 x 105 3.792 22.20 60Co Cu-2 431.5 1.011 x 105

. 3.732 21.85

, 60Co Cu-3 465.2 1.056 x 105 3.782 22.14 -

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Average 22.06 i

j S4Mn Fe-1 133.3 5.287 x 105 66.11 108.1 i

j 54 Mn Fe-2 136.0 5.401 x 105 66.19 108.2 i'

54Mn Fe-3 136.1 5.504 x 105 67.41 110.2 Average 108.8 (a) Specific activity at ti=e of reactor shutdown, 9/8/78 . Disinte rates are subj ect to a 3% (1 SA) measure =ent uncertainty.(4,5) gration j (b) Saturated activity at the 3293 Mwt power level.

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. TABLE III OPERATIONS SU1DtARY - BROWNS FERRY NUCLEAR PLANT, UNIT 3 operating Period- Dates Reacto- EquivalentI Decay Time Operating . Shutdown- Power Operating (z) Start Stos Dave _ Days in Days B MDeh) Davs (T.,) (e-)

1 09-01-76 09-28-76 28 - 10,228 09-29-76 10-01-76 -

3.106. 710 3 - -

2 10-02-76 10-18-76 -

17 - 15,328 10-19-76 4.655 690

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- 35,824 10.879 668 11-10-76 11-11-76 -

2 i 4 11-12-76 12-16-76 35 - 85,977 12-17-76 26.109 631

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5 12-20-76 04-25-77 -

127 - 343,906 04-26-77 04-29-77 104.435 501 4 - -

6 04-30-77 05-11-77 -

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28,754 8.732 486 11-77 05-11-77 -

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101.898 30.944 448 2 -

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37,558 07-07-77 07-09-77 -

11.405 429 3 - -

9 07-10-77 07-30-77 21 -

35,640 07-31-77 07-31-77 -

10.823 405 1 - -

10 08-01-77 09-02-77 -

33 -

99,649 30.261 09-03-77 09-04-77 -

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371 11 09-05-77 10-14-77 40 -

115,314 10-15-77 10-17-77 -

35.018 329 I- 3 - -

12 10-18-77 11-23-77 37 -

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33.384 299 3 - -

13 11-27-77 12-10-77 14 -

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10.134 272 1 ~- -

14 12-12-77 01-27-78 47 01-28-7S 138 683(b) 42.101 224 01-29-78 -

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15 01-30-78 03-15-78 45 -

134,033 l- 03-16-78 03-20-78 -

40.702 177

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20 08-24-78 09-08-78 16 - 35,860(b) 10.890 0 Totals 1,750,115 534.502(*

(a) At 3293 We (b) Estimated l' ' (c) 534.502 days = 4.6181 x 107 seconds i

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't IV.

CALCULATION OF NEUTRON FLUX DENSITY AND FLUENCE The energy-dependent neutron flux density,'$ (cm-2.sec-1), the spec-trum-averaged activation cross-section, 3 (en ), and the saturated activity As , of each dosimeter wire are related as follows:

I As

&"NU*o (4) r In the early days of nuclear pressure vessel surveillance activity, the value of E 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 assu=ption was useful. However, as methods of analysis were improved, the use of calculated neutron spectra has increased and is now per=itted 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 3 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 set R-Z calculation was made for a vertical plane through the axis of the core and the location of the vessel wall dosi=eter. A one-i eighth seg=ent, shown in Figure 2, was taken to be representative of the R-0 geometry because of the sy==etry involved. The boundaries of the core, core shroud, jet pu=ps, and vessel wall were described in R-0 coordinates. The

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FIGURE 2.

ONE-EIGHTH SEGMENT FOR FLUX CALCULATIONS AND LOCATION OF SURVEILLANCE CAPSULES i

g.

53-m m = -

11 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 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 vall neutron dosimeter capsule location. In addi-tion, the R-0 calculation provides infor=ation 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 =axicum n'eutron flux incident on the vessel (I.D. lead factors) can be derived.

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

10.0 I o(E) $(E) dE 3 (E > 1 MeV) =

f*fo .

(5)

E $(E) dE 1.0 h g;.

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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, 6(E) Ope (E) (barns) OCu(E) (barns) 8.18 - 10.0 .0352 .581 .0380 6.36 - 8.18 .0882 .577 .0144 4.96 - 6.36 .1285 .491 .0023 4.06 - 4.96 .1031 .354 .00025  ;

3.01 - 4.06 .1167 .185 .00010 2.35 - 3.01 .1503 .078 .00006 1.83 - 2.35 .1278 .023 .00002 1.11 - 1.83 .2176 .0014 .00002 1.00 - 1.11 .0325 -

.00001 8

Fe

= .207 (a) 8 = .00296(a)

(a) Spectrum-averaged cross sections are subject to a 15% (1 5%)

uncertainty.(6) l l

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Substituting the value of k., into Equation (4) along with the average value of A s for the iron dosimeters (see Table II), the fast neutron flux (E > 1 i

MeV) at the vessel wall dosimeter location-is calculated to8 be 8.38 x cm'2 Similarly, the fast neutron flux determined from the copper dosimeters

! is 1.14 x 109 cm-2 i The discrepancy between these two values is largely a re-(

sult of uncertainties in ctrrrent evaluated energy-dependent cross sections. (4,5) f According to ASTM Recommended Practice E 482(6) errors as large as 7% (1 S%)

1 in the determination of disintegration rates and t15% (1 S%) in spectrum- -

i weighted group-averaged cross sections can be encountered, which results in a

{ combined error of t 16.5% (1 S%) for the calculation of neutron flux from the input data.

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

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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 dosim-i eter capsule was placed at the azimuthal position of maximum fast neutron flux.

i Therefore, it is concluded that the calculated flux derived from the analysis of the dosimeter wires is a direct measure of the maxi =um fast flux inc I on the pressure vessel opposite the vertical core centerline.

i 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 5.6%

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 the peak fast I

I flux incidenc on the pressure vessel I.D., for the Browns Ferry Unit 3 sur-

} ve111ance capsules is' calculated to be 0.95.

t 3ased on the results of the DOT 3.5 calculations and the dosi=etry re-g sults, 2 the peak fast flux incident on the Browns Ferry Unit 3 vessel during I

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en -2.see-1, E > 1 MeV.

Therefore, the neutron fluence per effective full power year (EFPY) is 3.28 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 1.31 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 neutrons 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 3/4t 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 vall shown by the dashed curve in Figure 4, the predicted flux and fluence values obtained l

at 1/4t and 3/4t for the 6-5/16-in. Browns Ferry Unit 3 pressure vessel are s"-~'rized in Table V.

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} FIGURE 4.

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

l NORMALIZED TO THE VESSEL WALL DOSIMETER RESULT i

l

,f a _ , , _ _ . . . . . . - . _ . . _ . .

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

FOR BROWNS FERRY UNIT 3 PRESSURE VESSEL WALL I

I p Vessel Wall Relative Fast Fast Neutron Flux Fast _ Neutron Fluence (cm-2) -

, g Location Neutron Flux Density (cm-2.sec-1) 1.463 EFPY(b> 4 EFPY 40 EFPY I

I.D. Surface 1.00 1.04 x 109 4.80 x 1016 1.31 x 1017 1.31 x 1018 1/4t 0.67 7.0 x 108 3.2 x 1016 8.8 x 1016 8.8 x 1017 3/4L 0.24 2.5 x 108 1.2 x 1016 3,1 x 1016 3.1 x 1017 I

ff'

?

9 H

(a) E > 1 MeV. Calculated flux and fluence values subject to a i16.5% uncertainty.(6) -

(b) End of core cycle 1 J.

.e b .

18 r

V. DISCUSSION The predicted value of the. peak neutron' fluence (E > 1 MeV) for the Browns-Ferry Unit 3 pressure vessel after 40 EFPY of operation is given in ihe Final Safety Analysis Report (FSAR) as 3.8 x 10 17 cm-2-(E > 1 Mev). I The analysis of the vessel wall dosimeter capsule projects that the peak

. neutron fluence will. be 1.31.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 cm-2, .

. A similar trend has been noted in several other BWR plants with which SwRI i 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 projected peak fast fluence factor of 3.28 x 1016 cm-2 per EFPY can be employed to predict the change in the l- reference nil ductility temperature (RTNDT) as a function of reactor power generation.

i 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 (g > 1 1 MeV). The proposed relationship between fast neutron fluence and the change in the RTNDT of the Browns Ferry Unit 3 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 j the end'of core. cycle 1, and (2) an additional abscissa relating neutron

I .

19 t

9 Reactor Operation, EFFY 1 4 10 40 L

2- 200 I I I i N

's E

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NEUTRON FLUENCE (>1 MeV) (p:), nit i

i i

l I

FIGURE 5.

VESSEL MATERIAL NEUTRON DBRITTLDENT CURVE FROM BROWS FERRY UNIT 3 FSAR i

l hik M F h ._ Mur5EEiG2EERLNummean:arz* i: :. Sm=a

20 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 3 pressure vessel will begin to increase after an exposure 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 af ter 40 EFPY is ,

less than 50 F above the baseline (unirradiated) value.

L

,The neutron embrittlement sensitivity curve from the FSAR (Figure 5) corresponds closely with the RT ET adjustment curve of Regulatory Guide 1~.99(2) for 0.15% Cu and 0.012% P, see Figure 6.

However, in a recent re- -

1 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 3 vessel at the end of design life would be 101 F at the I.D. surface and 83 -F at the 1/4e 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 E 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 RTET 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 #pr.:omimately 50 F, or at one-fourth service life, whichever is earlier.

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Based on an end-of-design life increase in RTET of 83 F determined from the 0.3% Cu response curve in Figure 6 at the 1/4t fluence obtained from the vessel vall 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 RTynT RTNDT Capsule Removal 5 (17 F) s 100 F 10 EFPY

> (17 F) > 100 F 7 EFPT Since the results of the analysis of the vessel vall dosimeters indi-cate that the fast neutron flux is higher than that predicted in the FSAR, the current pressure-te=perature limits for operation and testing should be reviewed to determine if they are consistent with the projected adjusted values of RTET between the first refuelling and the time of removal of the first surveillance capsule. If not, revised pressure-temperature limits should be established in accordance with Seccion III, Appendix G, of the ASME Boiler and Pressure Vessel Code.(8) l o

e 0

l

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%m ,_ . . . . . . .. . . . . .

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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 July 1969. Property Surveillance of GE BWR Vessels," NEDC-10115, 4.

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

5.

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

.6.-

Juilaf E 482-76, " Standard Recommended Practice for Neutron Dosimetry

-for Reactor dards, Part Pressure

45. Vessel Surveillance," Annual Book of ASTM Stan-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 Docket Nos. J. E. G111 eland, TVA, to A. Schwencer, NRC, regarding 50-259, 50-260, and 50-296, dated August 23, 1977.

I i

i i

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g, 5 l ATTACHMENT 5 i

4 A.

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^*

~ - -...

, e i

BROWNS FERRY NUCLEAR PLANT FABRICATION AND TESTING OF REACTOR VESSEL MATERIAL O

I 1

l TENNESSEE VALLEY AUTHORITY i

DIVISION OF CONSTRUCTION SINGLETON MATERIALS ENGINEERING LABORATORY o

Knoxville, Tennessee I

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FABRICATION AND TESTING OF REACTOR-VESSEL MATERIAL 1

4

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^ t T. F. Ziegler, Chior, l' chanical Dranch,,1720 CST 2-C-(2) '-

Frank Van !!cter, Chior, Construction Scrvices Branch, 500 SPT-K c ,

August 11, 1983 r, .

'lr >; .

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DP.0' itis FEDRY !!UCLEA!! PLA!!T - FABRICATIO" A!!D TESTEIG. OF REACT 0!! VESCEL

's '.

!!ATERIAL l -

'f. f,cy p

j.

/ -_ i

)

As requested in L29 830210 993, a Charpy V-n61ch (CVI!) testing program was _

conducted to determine transition to'. 1perature curves for' Browns Forry

^

Nuclear Plant units 1, 2, and 3 reactor vessel core beltline materials ,

(base metal, wold cetal, and !!AZ). A complote listing of the. data in the t form of impact energy, mils lateral expansion,, and perceht shhar as a function of impact temperatures are ?,1ven in the tablet of attach =ent 2. '

- Also included are transition curves' for each caterial. *

.t Asucmaryofthesedataexpressedin,termsbrtransition,tenperatured.at30 '

ft-lbs, 50 f t-lbs, and 35 mils lateral 6xpansion and upper' shelt enArgios are given in the table of attachment 1. All transition tcroeraturas listed are Graphic interpolated values based on the data given in attachmint 2.

Tho upper shelf energies are averages of impact values for those camp 1'co which exhibited 99-percent or greater choar in their fracture surfaces. u

, ,; s i]

Attachnent 3 cives plots of porcent snear as a function of inpact tecperature for cach naterin1. <

[ 9 CVN specicens for unit 1 natorial were fabric'ated at SME from material ,

shipped fron Drowns Ferry Power Stores. Cu:!delinesprovidedthropshGE Specification 21A1111. attachtent D, "!!aterial Tests and Test Specicehs,"l' f ,'

wore followed in detornining the location, orientation, and notch direction- f of each specimen. All samplos were tosted in accordance with AST!f FJ3

• i /

Impact test machine information involving physical dimensions, wind' age ud I - /

friction measurements, cooling bath and tceperature control, and latest ( / 9 ,

calibration records is given in attachment 4. j

,3 '. >r s

j

/ l Usinal eigned by i 7

Fran,ig Van Meter ,.

Frank Van tieter PVG:ASY hr ,

Attachments cc (Attachments): (x' )

U.11. Childres, SHE-K  ! ,

E. Fonda Harwell,1410 CST 2-C (2) '"

J. P. Layno~150'0' CST 2-C - - 6 MEDS, USD63 C-K ,

Principally prepared by P. V. Guthrio, extension 2771.t

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l i Attachment 1 BROWNSFERRYNUbLEARPLANT CVN TESTING OF REACTOR PRESSURE VESSEL MATERTAT.

SUMMARY

OF DATA PROM TRANSITION CURVES f- .

i.

i  ! Temperature

• at Temperature

• at Temperatur,ee at Upper 30 ft-lb 50 ft-lb ..

_ Materia 1 7 Tanact Enerry - 3:i mils Lateral shelt "F Tmnant Enernv_ Ernansion EneravH "F "F ft-lba

!3 Unit 1 BM <

<-20 -16 ga HT C2868 -19., 149.5

/

/ ,l Unit 1 BM <-20 +10 0 7

HT C2884 124.5 Unit 1 HA2 <-20 +20 HT C2884 0 97 8 i

Unit 1 +10 '

+24 MS Weld Metal ,

140***

\

' Unit 2 <-40 '

-15 -24 #

Base Metal e

'142.2

'g Unit 2 <-40 HAZ

<-40 < -4 0 I 130.5

\-

', Unit 2 -13 , 0 -7 Weld Metal 117.5 Unit 3 *-40 '

Base Metal -37 <-4o 154 3 c .

- Unit 3

-24 -2 HAZ ,

-15 125.8 Unit 3 9..

Weld Metal

-30 -23 s' I' -28 i 131.5

, . 1.; -

' Interpolated from data of attachment 2. #

• • Average value of impact energy for samples with 99% shear.

eveExtrapolated value of 5 shear versus impact energy. '

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e BROWNS FERRY NUCLEAR PLANT REACTOR PRESSURE VESSEL MATERIAL CHARPY V-NOTCH TESTING Sample iest Impact

' Lateral Location Specimen ge F Energy Exoansion Shear ft-lb mils  %

Base BF16-11 -20 52.5 Metal 42.0 29

, BF16-12 -20 31.0 27 0 Unit I 17 41.8 Average of 2

. HT C2868 I

. A-BF16-9 0 67 5 56 5 38 BF16-10 0 74.5 60.0 44 f 71.0 Average of 2 BF16-5 +20 68.0 52.5 33 BF16-6 +20 91.5 68.5 52

' 79.8 Average of 2 BF16-1 +60 117 0 71.0 74 BF16-2 +60 99 5 68.5 60 108.2- Average of 2 BF16-3 +100 126.5 67 0 80 BF16-4 +100 128.0 89 0 96 127 2 Average of 2 BF16-7 +140 156.5 93 0 99 BF16-8 +140 142.5 89.5 99 149.5 Average of 2 Tested by b

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RPV MATERIAL -

CVN TRANSITION CURVE-UNIT 1~

SASE METAL CNT C2868) 160- '

150 -

140-4 130-

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160 180 4

TEMP. F 6 g' i .

3 BROWNS FERRY NUCLEAR PLANT REACTOR PRESSURE VESSEL MATERIAL CHARPY V-NOTCH TESTIN(1 Sample .

Test Impact Lateral Location Specimen gF Enerev ft-lb Exoansion Shear mils 5 Base BF17-11 -20 35.0 27 5 17 Metal BF17-12 -20 35.0 27.5 17

~

Unit I 35.0 Average of 2 HT C2884 BF17-9 0 53.0 40.5 27 BF17-10 0 65.0 51.0 38

, BF17-13 0 42.5 35.5 33 BF17-14 0 26.0 22 5 27

~

46.6 Average of 4 BF17-5 +20 54.0' 42.5 42

. BF17-6 +20 53 0 41.0 42 53 5 Average of 2 BF17-1 +60 68.5~ 53 5 58 BF17-2 +60 74.0 57.0 50 BF17-15 +60 92.5 69 0 67 78.3 Average of 3

'1 BF17-3 +100 117 0 77 5 84 BF17-4 +100 103 0 69 0 52 110.0 Average of 2 f-1 BF17-7 +140 113 0 77 5 94 BF17-8 +140 124.5 63 0 99 118.8 Average of 2 i

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• • • BROWNS FERRY NUCLEAR PLANT >>>

RPV MATERIAL -

CVN TRANSITION CURVE ~

UNIT 1 BASE METAL CMT C2884) 160-150 -

140-ISO-1 g 120-E J 110-I

, P 1OO-h.

$ 90-1 O f

h 80 -

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TEMP. F t

1 BROWNS FEHRY NUCLEAR PLANT REACTOR PRESSURE VESSEL MATERIAL CHARPY V-NOTCH TESTING Sample Test Impact Lateral

}'3

Location Soecimen gF Energy Exoansion Shgar_

ft-lb mils  % ,

HAZ BF18-11 -20 46.0 42.5 25 Unit I BF18-12 -20 22.0 18.5 21 HT C2884 BF18-15 -20 42.0 32.0 33

} 36.7 Average of 3 BF18-9 0 50.0 41.0 27 l BF18-10 0 35.5 30.0 27 J BF18-13 0 46.0 37 5 33 43 8 Average of 3 BF18-5 +20 46.0 36.0 36 BF18-6 +20 36.0 32.0 85

,, BF18-16 >20 70.0 50.0 50

• 1 50.7 Average of 3 BF18-1 +60 69 0' 56.5 68 BF18-2 +60 50.5 44.0 59 e 59.8 Avera6e of 2 a BF18-3 +100 73 0 61.0 69 BF18-4 +100 99 5 65.5 99 86.2 Average of 2 BF18-7 +140 96.0 71.5 99 BF18-8 +140 98.0 67 0 99 97.0 Average of 2 e,

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• • • BROWNS FERRY NUCLEAR PLANT >>>

RPV MATERIAL -

CVN TRANSITION CURVE UNIT 1 HAZCHT C2884) 160-150 -

140-ISO-g 120-D

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-100 60 -40 -20 0 20 40 SO 80 100 120 140 ISO 180 TEMP. F*

D-i BROWNS FERRY NUCLEAR PLANT REACTOR PRESSURE VESSEL MATERIAL CHARPY V-NOTCH TESTING Sample Test Impact Lateral Location Soecimen ge F Energy ft-lb Exoansion Shear mils  %

' i Weld' -BF19-11 -20 8.5 75 6 Hetal BF19-12 -20 11.0 10.0 6 Unit I 9.8 Average of 2 BF19-9 0 23.0 22.0 12 BF19-10 0 75 8.0 6 15.2 Average of 2 BF19-5 +20 55.5 48.5 29 .

BF19-6 +20 42 5 38.0 29

.] 49.0 Average of 2

~

, BF19-1 +60 59.5 53.0 44 BF19-2 +60 69 5 58.5 52 64.5 Average of 2 BF19-3 . +100 78.5 67 0 60 BF19-4 +100 83 5 68.0 55 81.0 Average of 2 BF19-7 +140 92.0 67 0 65 BF19-8 +140 94.0 70.5 84 93 0 Average of.2 -

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• • • -BROWNS FERRY NUCLEAR PLANT >>>

RPV MATERIAL UNIT - CVN TRANSITION CURVE 1

WELD METAL 9

4 160 -

160 -

140 -

ISO -

, g 120-D 4 110-I t-ti. 1OO-i $ 90-O

[ 80 -

z W 70 -- i 1-o 60 -

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! H 40 -

1 30-

20 -

1 1n-O . . . . . . . . . .

-100 60 20 0 20 40 60 80 100 120 140 160 180 TEMP. F

T.

BROWNS FERRY NUCLEAR PLANT REACTOR PRESSURE VESSEL MATERIAL CHARPY V-NOTCH TESTING Sample Test Impact Lateral Location Soecimen pF Enerstv ft-lb Exoansion Shear mils  %

Base E6Y -40 36.5 30.0 17 i

I Metal E6D -40 34.0 29.5 17 Unit 2 35.2 Average of 2 ESA -20 44.5 37 5 23 E65 -20 42.0 36.5 27

} 43 2 Average of 2 E6J 0 89 0 62.0 49

E7P 0 68.0 56.0 29

-1 78.5 Average of 2 E6C +60 103 5 76.5 67 E6A +60 116.0 81.5 69 109.B ,

Average of 2

[ E6M +100 123 5 71.5 85

• E6B +100 147.0 95.5 99 135.2 Average of 2 E54 +140 135.0 70.5 99 E7U +140 144.5 79 0 _99 139 8 Average of 2 Tested by W-

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• • • BROWNS FERRY NUCLEAR-PLANT www

.RPV MATERIAL -

CVN. TRANSITION CURVE

~ UNIT 2 BASE METAL 160-160-140-4 130-n) 120-ID

-j 110-F IL 100-

$ S0-O j @ se- -

~

z Id 70 - s 1 F 4

0 60-4 .

, h SO-

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-!00 -80 -60 -40 -20 0 20 4O 6O 80 100 120 140 160 180 TEMP. F ,

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l y

BROWNS FERRY NUCLEAR PLANT

. REACTOR PRESSURE VESSEL MATERIAL

- CHARPY V-NOTCH TESTING Sample Test Impact Lateral -

Location Soecimen gF Energy ft-lb Ernansion mils Shear 5

HAZ _EEE -40 72.5- 54.0 38

.'. Unit 2 EDP -40 63.5 50.5 29 68.0 Average of 2 1

EDJ -20 '85.0 66.0 41

, EJ6 -20 91.5 67 0 41 88.2 Average of 2 EJ7 0 109 0 75.0 57 EEH 0 90.0 67 0 57

- 99.5 Average of 2

- EDA +60 138.0 69.0 80' EEK +60 120.5 82.0 80 129 2 Average of 2 EJM +100 105.5 83 0 95' EE2 +100 134.5 97 5 99 120.0 Average of 2 1

. EEU +140 109 0 90.0 29 EE3 +140 148.0 100.0 99

- 128.5 Average of 2 Reviewed by M d. /

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9 -

a N

w

]

]

iksiii W td Lesd 64 taal danist intif - dead imid - ined initi daad W disid W &ltif -M M

• • • BROWNS FERRY NUCLEAR PLANT ***

RPV MATERIAL -

CVN TRANSITION CURVE UNIT 2 HAZ 160-150- '

1AO-130- A g 120-(D I

3 110-t-

It 100-

[ 90-O h 80-z .

Id 70 -

l-O 60- ,

4 k 50-H 40 -

30-2L,-

1C-O- . . . . . . . . . . . . .

-100 -80 -60 -40 -20 0 20 40 60 80 100 120 140 ISO 180 TEMP. F s

BROWNS FERRY NUCLEAR PLANT REACTOR PRESSURE VESSEL MATERIAL CHARPY V-NOTCH TESTING' Sample Test Impact _ Lateral 3

Location Soecimen pF Enerstv ft-lb Exoansion mils Shear

- '] Weld' EBP -40 55 5.0 0

.J Hetal EC6 -40 75 75 6 Unit 2 6.5 Average of 2 ECA -20 29.5 27 0 6 ECB -20 13 5 14.0 6 21 5

}. Average of 2 ECC 0 66.0 55.5 23 -

EA6 0 40.0 35.5 12 s

], 53 0 Average of 2

, EA7 +60 74.0 62.5 49 EBY +60 80 5 68.0 49 77.2, Average of 2 EBM +100 100.5 80.5 72 EB3 +100 95.5 75.5 74 98.5 Average of 2 EAB +140 113 5 89 0 99 ECL +140 121.5 92.5 99

, 117.5 Average of 2 q

~

sevie..e ,,

c a e. W2.

e

? .

M 3

b

. . - - , . , - _ , . . . , , . . - , ,v ., , -

. .. . _- - U L 4l BROWNS-FERRY NUCLEAR PLANT ***

RPV MATERIAL'- CVN TRANSITION CURVE UNIT 2 WELD' METAL 160 - '

150 -

140 -

130-

! g3 120-

! O j 110 -

l-

h. 100-

[ 90-O y 80-Z h1 73-F-

U 60 -

4 e se - -

H 40 -

30-20 -

10-O9 . . . . . . . . . .. . . . .

-100 SO -40 -2G O 20 4e 60 se jee jag 34e jeg jeg TEMP. F .

.} -

3 BROWNS FERRY NUCLEAR PLANT REACTOR PRESSURE VESSEL MATERTAL CHARPY V-NOTCH TESTIN0 Sample Test Impact Lateral i

Location Soecimen gF Energy ft-lb Exoansion Shear mils  %

Base 66D -40 43 0 36.5 17

} Metal-Unit 3 666 -40 44.5 43 8 so.5 Average of 2 23 65C -20 74.0 58.5 44 665 -20 83 0 66.5 -38 78.5 Average of 2 664 0 109 5 88.0 61 -

656 0 101.5 78.0 52 105.5 Average of 2

, 64Y +60 134.5 90.0 8O 66J +60 155.5 88.5 99 145.0 Average of 2 66E +100 149.5 72.0 99

_' 65E +100 155.0 75 0 99 152.2 Average of 2 i

65P +140 154.5 75.0 99 66L +140 157 0 66.5 99

, 155.8 Average of 2 1

i Reviewed by o N. I

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l .

l 1 -

l i

l l . , . .

• • • BROWNS. FERRY NUCLEAR PLANT.*wn

, RPV. MATERIAL CVN TRANSITION CURVE UNIT'3 BASEMETAL 160-a 150-140-1 ISO-

. g 120-(D

! 110-I t-

.; h. 1OO-

/ GO-0

@ 80-

• Z hl 70 -

e o 60-1 < '

k 50 -

l H 40 -

) 30-20 -

i i 1 ;j -

i O --

i i

-100 -80 -60 ~40 -20 0 20 40 60 80 100 120 140 160 180 1 TEMP, F '

1 t

i

1 BROWNS FERRY NUCLEAR PLANT REACTOR PRESSURE VESSEL MATERIAL CHARPY V-NOTCH TESTING Sample Test Impact Lateral j Location Soecimen ge Energy Exoansion Elm J. F ft-lb mils 5 HAZ 6BM -40 16.0 15.0 6 Unit 3 6D2 -40 15.5 13 0 11 15.8 Average of 2 3 6CC -20 39.5 36.5 27 6DJ -20 27 5 24.5 26 33 5 Average of 2 6CY 0 29 5 29.0 23 3 6BU 0 76.5 66.5 44 J 53.0 Average of 2 6D6 +60 90.0 73.0 63

} 6C1 +60 105.5 97.8 79 5 Average of 2 68 6BP +100 119.5 81.0 81 1 6CD +100 116.0 92.5 99 117 8 Average of 2 6CA +140 122.5 91.0 99 6D7 +140 139 0 97 5 99 130.8 Average of 2 1 r. ted ,, A l~ f d~ 4 k.

Reviewed by kM d 3 <

3 1

1 1

• • • BROWNS. FERRY NUCLEAR PLANT- ***

RPV MATERIAL -

CVN. TRANSITION CURVE UNIT 3 HAZ 160-150-1 140.-

j 130-g) 120-(D d 110-I E

h. 100-i

$ 90-0

h 80-i z hl 70-i f

i o 60 - ,

4 .

) h GO-

! H

, 40-

30-4 20-O , , , , , , , , , , , , , , , , , , , ,

-100 60 -40 -20 0 20 40 60 80 100 120 140 160 180 TEMP. F' .

I 4

.l; .

'I I BROWNS FERRY NUCLEAR PLANT REACTOR PRESSURE VESSEL MATERIAL CHARPY Y-NOTCH TESTING Sample Test Impact Lateral Location Specimen gF Enersey -

ft-lb Ernansion h mils  %

Weld .672 -40 17 0 19 0 6

- Metal 6AL -40 5.5 5.5 0

' Unit 3 11.2 Average of 2 67U -20 60.5 52.5 25 67M -20 53 5 47 5 19 57.0- Average of 2 6AY 0 58.0 52.0 25- -

67K 0 72.0 60 5 31 65.0 Average of 2 6AK +60 92.0 71.5 55

-676 +60 97 0 74.0 63 94.5 ,

Average of 2 674 +100 .88.5 64.5 60 6 %' + '. 00 131.5 63.0 99

• . in. u_a ' 2

~

-1. E. -il , 1i3.5 E3 5 ,

Ei 67L +140 131.5 73 0 99

.. 125.5 Average of 2 Tested by M Reviewed by d , .l O

m M

1 d

I

N' M ^ M N N U M $ M W Odad W & 6 M M & ikiild &

• • • BROWNS FERRY NUCLEAR PLANT ***

RPV MATERIAL -

CVN TRANSITION CURVE UNIT 3 WELD METAL 160'-

150-140-130-g) 120-(D j 110-e

h. 100-N GO-O h 80 -

Z '

W 70-F l

0 60- -

4

$ GO-H 40-30-2r -

13-O . . . . . . . , . .

-100 60 -40 -20 0 20 40 60 80 100 120 140 160 180 TFMP. F*

I

.f J.

5

~

I.

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A J

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I ATTACHMENT 3 4

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9. 6 M 0. 2 B 5 0 1

+ 5 .

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= - = ^

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6 '

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RE EH TC O TA .O NMM 1 A' B LV PP1

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AFI EON E L U R CG U UNR .0 T NIO 6 A TF R YS '

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0 9 8 7 8 S 4 3 2 I 1

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W W best M idW Ihli M M M M Ydid 6aud fend W IEdhd M l mind inmid N 4

BROWNS FERRY NUCLEAR PLANT CHARPY V-NOTCH TESTING OF RPV MATERIALS SHEAR VS TEMPERATURE FOR UNIT 1 BM CHE C2884) 100- 0 90-80-70-0' Id 60- o I

. D

{

Id 50-

0

$ 40-0.

t 30-2 ~

Y=A+BMX A=29.4379 IO~ B=0.52SS3 R^2-0.993 O . . . . . . . . . . .

-40 -20 0 20 40 SO 80 100 120 140 160 TEMPERATURE, "F ,

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

.- - w -Q e w J J " " U i

~ .

. BROWNS FERRY NUCLEAR PLANT CHARPY V-NOTCH TESTING OF RPV MATERIALS SHEAR VS TEMPERATURE FOR UNIT 1 HAZ < H t. C2884) 100- -

O i

90-4 80 -

70 -

0 Id 60 -

I 0)

F SO-Z .

Id O

h 40-

! Q.

4

• u

! 30-o .

$ 20-Y=A+SMX .

1C- A=33.0275 i i B=0.50413 i

O R^2-0.S88 3 ~40 -20 0 20 .4 0 60 80 100 120 140 16O.

TEMPERATURE, ' F' ,

' s e

U- U U' u' LJ 'Ld LJ LJ La 'W W LJ W W La Ld W km BROWNS-FERRY-NUCLEAR PLANT ~

CH RPY V-NOTCH TESTING OF RPV MATERIALS SHEAR VS TEMPERATURE FOR UNIT 1 WELD METAL 1OO-90-80-o 70-i 4 td 60-I o 0) i

$ SO-Id 0 0 '

h 40-a .

30- ,

20-Y=A+S*X i

~

A=15.4456 o B=O.43842 R*2-0.970 O . . . . . . . . .

~40 -20 O 20 40 60 80 1OO 120 140 16O l TEMPERATURE, F ,

.._ _. ,___; u BROWNS FERRY' NUCLEAR' PLANT CHARPY V-NOTCH TESTING OF RPV MATERIA'LS i

SHEAR VS TEMPERATURE FOR' UNIT 2 HAZ

-100 -

O C-90 -

80-o 70-2 R

4 hl 60-d

' I n o F 50-Z l' hl.

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^

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! 30- -

- 2,0 - ,__ sr 4

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1O-

~.

' A=44.!206

' '~

' 3 0,54S35-Oi .

R^2-0.SS1,

- 4 0 ;- ' -20 / - . . .

0 20 40 GO 80 , -

100 120 '

j40 160 ,

TEMPERATURE, F .,

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. BROWNS FERRY NUCLEAR, PLANT CHARPY V-NOTCH TESTING'OF RPV MATERIAf.S ' -

SHEAR VS TEMPERATURE FOR UNIT 2 WELD METAL f

100- ,

i:

90- ,

- r 80 -

• ~ r // *W 70- cs ,.

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i -40 -20 0 20 40 60 80 100 120 140 160 i

TEMPERATURE,- 'F l'

4. .

~

BROWNS FERRY NUCLEAR PLANT

CHARPY V-NOTCH TESTING OF-RPV MATERIALS s

SHEAR VS TEMPERATURE FOR UNIT 3 BASE METAL 4

100 -

^

U U 90 -

80 -

i' 70 -

i E

4

,' W 60-

I i U)

L y 50-

W O

j h 40- 0 O. .

i 1

30- -

l i i 5

20- o i Y=A+BMX

' ~

A=38.St42.

1 t

S=0.8557I i O . ..

R^2=0.997  ;

{ -40 -20 0 20 40 SO 80 100 l 120 140 160 i TEMPERATURE, F ,

1 4 .

BROWNS FERRY NUCLEAR PLANT CHARPY V-NOTCH TESTING'OF RPV MATERIALS ,

2 SHEAR VS TEMPERATURE FOR UNIT 3 HAZ 100 -

i G

90 - i 4

se- <

1 70-E

< W 60-I V)

{

w 50- i O

l 40-4 0. i j o 30- . ,

I o

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i Y=A+SwX I O '~

o A=23.2068 s=e.67478

! C i -- 4 0

. R^2=0.SS4

-20 0 20 3

40 60 80 i

100 120 140 160 i

1 TEMPERATURE, 'F .

I

- - - '- - w .-- u u u 'u u sa u y y y y 6

. BROWNS FERRY. NUCLEAR PLANT CHARPY.V-NOTCH TESTING OF RPV MATERIALS SHEAR VS TEMPERATURE FOR UNIT 3 WELD METAL 100 -

90 - o

'80- o 70 - -

E Id 60- O I

' 0) 1 j y sO-

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O '

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30-o 20-Y=A+BwX I

A=t8.0862 IO- S=0.631SS R^2-0.989 G3 . . . . . . . . . .

-40 -20 0 20 40 60 80 tOO 120 140 160 TEMPERATURE, F ,

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DEPARTMENT OF THE ARMY

' ARMY MATERIALS AND MECHANICS RESEARCH CENTER wATERTOWN. M ASSACHUSETTS o2172 DRXMR-STM 18 March 1983 1

Tennessee Valley Authority Singleton P.aterials Engineering Lab ATTN: Mr. Childres Knoxville, TN 37902

Dear Hr. Childres:

- ' A set of Charpy impact test specimens broken on the 264 f t-lb capacity Tinius Olsen machine, Cerial No. 100400 has been received for evaluation along with the completed questionnaire.

The results of the tests indicate the machine to be producing acceptable energy values at both energy levels (see inclosed table) .

This machine satisfies the proof-test requirements of ASTM Standard E-23.

1 ' If this machine is moved or undergoes any major repairs or adjustments, this certification becomes invalid and the machine must be rechecked. Removal of the pendulum, replacement of anvils or adjusting the height of drop are examples of such major repairs or adjustments. It should be noted that if a specimen requires over 80% of the machine capacity to fracture, the machine should be checked to assure that the pendulum is straight, the anvils or

' striker have not been damaged and that all bolts are still tight. This i

certification is valid for one year from the date of the test.

Sincerely yours, 0%5b 1 Incl ROGER M. LAMOTHE Table Chief, Mechanical

, Dehavior & Testing Branch P.G25 ca ROL u m. , ,ma mic Fho l T '--ll ]

F/3 p ..

I I I l -

ca. i

, [, M

l

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i

^

ARMY MATERIALS AND MECHANICS RESEARCH CENTER Watertown, Massachusetts 02172 Date of Test: 1 January 1983 TABLE COMPARISON TESTS ON CHARPY IMPACT !ACllINES Tennessee Valley Authority, Singleton Matls. Eng. Lab.

Facility Knoxville, TN 37902 Make of Machine - Tinius Olsen Serial No. 100400 AFBIRC Varivtion (ft-lb) (ft-Ib) Actual Allowed liigh Energy 73.6 73.9 +0.4  % , +5. 0".

Low Energy 13.5 14.0 +0.5 ft-Ib +1.0 ft-lb DIR Form 105 Rev 1 Apr 81

f' .., ,

DEPARTMENT OF THE ARMY

/ a ARMY MATERIALS AND MECH ANICS RESEARCH CENTER 3

WATERTOWN. M ASSACHUSETTS O2172 C fjj3 6 81

} ROL Tvac l l

DRDIR-MQ l1ji: 3J ul.y is51

j. ,

J l Pvo

.  ; j Tennessee Valley Authority co-stmu.nIwos he l /

Singleton Laboratory tovsn. l l

' ' ~

- Singleton Terminal RECEIVED " '

A'ITN: Mr. Mark Burleson Singleton (Blount County), TN 37777 J.UL311981 SINGLUON YARDS

Dear Mr. Burleson:

Your check for Charpy impact test specimens has been received. The specimens 1 set (s) of 10 specimens are being forwarded under separate cover and 2 copies of a questionnaire are inclosed.

Prior to testing, your machine should be checked to assure co:npliance with Sections 4 and 5 of ASTM Standard E-23. Specimens are to be tested at -40

- in accordance with the testing procedures outlined by ASul (Section 8 of E-23).

An accurate machine will ' produce values within 1.0 f t-Ib or 5.0S (of the q nominal values), whichever is greater. The nominal values for this specimen 4 series are (1) DD10 series - 13.S ft-lb and (2) EE10 series - 73.1 ft-Ib.

Because the cause or causes of erroneous values at one energy Icvel may not g be the cause at the other energy IcVel, calibration or correction curves J should not be used. .

This Center will, if requested, evaluate the results of your tests and return a report of the findings to your facility. If your machine produces values outside the nominal values, the report will suggest changes in machine design, repair or replacement of machine parts, a change in testing techniques, etc.

Facilitics desirous of the evaluation must return the broken specimens and completed questionnaire to the Director, Army Materials and Mechanics Research Center, ATTN: DRDIR-MQ, Watertown, Massachuset,ts 02172. Shipping charges for the return of the broken specimens and completed questionnaire are the responsibility of the customer. Overseas shipments should be shipped parcel post only. We can no longer pickup airfreight shipments. Airfreight ship-

~ments.will be returned by Customs. '

Sincerely yours, 1

$ l. L.

2 Inc1 ci PAUL W. ROLSTON Chief 1 1. Questionnaire

2. Wrapping Instructions Quality Engineering Branch

,4 i D1R Form 1FL-1 Rev 15 Apr 81

QUESTIONNAIRE ON CHARPY IMPACT TEST PERFORtWICE EVALUATION

• 2, -

FOR THE ARMY 11ATERIALS AND MECHAtlICS RESEARCH CEtlTER

... m -

DSA(OCAS) Office Company Tennessee Valley Authority, Singleton Materials Engineering Laboratory

-Address Knoxville, Tennessee 37902 TESTIt!G MACHIllE -

} 1. Manufacturer Tinius Olsen- 2. Serial !!o. 100400

3. 'When was machine manufactured (or originally purchased)? 12/70 ,-

l' 4. Machine capacity 264 ft-lb

.c5. Capacity used in test 264 ft-lb .

. 6. Test

• velocity at impact, feet per'second 16.8 ft/see

. 7. a. Is nachine securelv bolted to concrete foundation? Yes X No -

~

b. Ilhat is the diameter of bolts used in a? 3/4 in. .

1

, c. Are these bolts J or lag bolts? no .

, d. If none of above, explain. Phillips red-head anchors used to i

attach machine to 5-in. floor slab.

?

8. Arc dimensions of striking edge and anvil suppor s in accordance with ASTM E 23-66 as shown below? Yes x No . If not, list accordingly.

, . e

i. .. . .

Strikins Edse .

~" a Striking Edge Specimen \ '

l ,

Support A

R Support '

8

+

h

- W .-.-

ANVIL SUPPORTS STRIKING EDGE '

AI.LO!JABLE ACTUAL ALLOT?ABLE ACTUAL DIMENSIONS DIMENSIO'IS DIMENSIONS DDfENSIONS A - 80* approx. a - 30* cpprox.

R-0.0391.002" r ,- 0.315 I .01" _ _ _

11 - 1.574" I .002" w - 0.157" approx.

B - 90* 10 min.

RMR FORM 34 13 JAN 69

..s

~

. - g. . .. .,

bA

. *: - , .a Y.' . . . .~ . s' b. ' $ E . :*

-lJ [;w* ' f, y*,* * ,  ; ., h* * . y . . * .s

, *-), ./ 2, #

' '. f . . ' .

-~ , , - . - - p. , ~ , , - , . , , . . - - a-.-,, - -.,--..n--,---m-,--, . , - - , . - - - - , , - . . , . - . - - . , . , , -

. e.

9. Check appropriate pendulum design below. .

A B -

p 4 _. . .

e .

f.

-_y,,._.. * . .> .

's-

. ,i ..u C .X. . . . . - . . ..... .. .. D . .

m * .;-.

,y _.

w .. , , . , . .,

. . u. ,%. , .

- g js-- - i:g;::r. ,

• ...' . , , , - p't - d .,

y _

- . .s -r. .

.s.,.. ;r .

E (other) Please sketch - 'J .

.s.

$? ' **

-.'-. -;. d

' - * ...:... . :li.yd,:;rl :.: . ., ! Wj $ .. . u . .t.- - . .  :.- .

~

.......e +"3 .# ,

.. ,s..

• *.* M P ., . g /. . .,. . . .

g s.a. ... .

. .. . . ..v. . i.. e. .n. . <..ur. . .

,., . ;; g . 2 .,. .. . . . .

o ..

,..;..s...,...

. .. .g...

...........g..g...,....

s: 20 nn

10. If side support is as shown below', list dimension'd O. -

v. .

. .: .a . .r--. -

d

..L......

M

.o

.N/ >

s .s . . . . .

Side */ 3 7, ' ' . .. ..:. - -- . 'd * '

.?

.! :. Support .;, .

'.~' ' . .

t . s . . .. .. '

t. . . - .. 4. 3 : .

. 11.'If side support is not as 'above, sketch apnropriate one (or enclose photograph). ..-,

h .

.. 2 .. _ . . . .

s . .

). . . i

- 4. .

I e

. . . =, .

( . .

.s., ...

. . - . * '~

XHR FORH 34-1 i .;L 13 JAM 69 .

. :;2s.. .

.....?

~

12. If prctectiva shrouds are used to contain broken soccinens, the f ollow-ing statements are to ba completed by checking (/) or filling in the blanks, as appropriate (use sketch below for reference): (1) The shrouds are x are not of hardened steel. The hardness is R,. 47 . (2) The thickness of the shrouds is0.077"(use decimal neasure.a.ients) . (3) L'ith relation to the clearance between the shrouds and the pendulun overhang, the clearance at 'the sides and top (in f ractional me'asurements) is (a) 1/4 in., (b) 3/8 in. , (c) 3/8 in. , and (d) 1/4 in. .

Pendulum

• w shro' uds %

o -> -

l  !

. I e- .a s' \\\\\\ //////// 3'

/

N -

\ .

/*

/2 sN

\ -

.h spacers _

13.

ifith accurately dimensioned standard specimen on machine supports,

. check appropriate sketch below.

'. f~ .:?a*.' . ..

^~ ~ ^

strikins -

B c_

x Edg e ----

i -

Specimen '

Y:3T

' l

.[..

g -

( , -

~ ' support *

• b E! '

- :a

14. It pendulum striking edge is as shown in "c" above and energy indicator l

1 is engaged, uhat is dial reading? 264 ft-lb ~

15. If specimen is then removed and the pendulum is dropped fro- latched position, what is dial reading? 0.0 ft-lb l

' R roRH 34-2 JAN 69 1

..s, .

3 .4 ' ~

- d.

l.

..sl~.'.. .

N * .~ * %4 ' . * - s' - C* Y  % * ~

. .N F

- 4. i

16. , Does pendultim pass through center of anvil supports within 0.010"? ,

Yes x No . If not, how far off-center is it?

I

17. When (1) the energy indicator is set at zero, and (2) the pendulum is raised to its nornal release height, released, and allowed "to cycle 5 times -(to and fro), and (3) the energy indicator *is set at approxi-mately 20 ft-lbs so as to be contacted on the sixth forward swing by the pendulum, what is the value recorded from the energy reale on this I

S sixth forward swing?8.0 f t-lb .

18. When was this machine last certified utilizing the standard specimens .. ig obtained .either..from.DCASR or f rom .the Army Materials and Mechanics 5

'Research Center? Oct 1981 . .... ._.

TESTIflG TECHt!IO.UE ' ,. ' ,i i

l

. .. . . .  ; .+. .

1. Test temperature must be'-40*F I.2*F

-l.:. i. .

(a) Bath medium ethanol

, !n- ,e., .' .

' . .(b) Size 'of bath 9' liters E

.~..

s%

(c) Was bath agitated? yes (d) Are specimensat 'least 1 inch from bottom of container and at least 1 inch below top of liquid? (Applicable only to liould media) Yes x No .

' (c) Method of controlling bath temperature Mercury contact thermometer 5 . .

(f) Temperature tieisuring methad Platinum res!. stance bridge'

l.I .

f

.Yes x no ,

". .(.g)

Was temperature'-measuring t r._.... instrument calibrated? --

(h) If so, how? 'Against TVA Central Lab Standards (i) Time specimens held at temperature. 10 nin (j) Uhat instrument is used to rcr.ove speci:.s :- from coolant? ,

.W Charpy centering tongs

. . q: ., ,, 3 , . .

. . . (k) Is it cooled with,s'pecimen? Yes .X No -

s .

I- (1) Time interval used in removing specimen f rom bath and fracturing

~

5 seconds maximum .

~

. c . ..x.

.g.' .,e XHR FORM 34-3 * . .  ; , ,

,/

-/

f i -

[O , ,

i

2. How is specimen positioned in testing machine to insure that knife edget of pendulum strikes directly opposite to notch?

With Charpy centering tongs

3. In a few words describe the good and/or bad features of the Charpy machine used in these tests.

Easy to operate and relatively maintenance free.

a

.) ,

was all the foregoing information physically witnessed by the Govern-ment representative? Yes No . (this statement to be answered only if the machine in question is being certified for use on an Army contract).

TEST RESULTS ENERGY - FOOT-POIRms Series DD10 Series EE9 Series Spec. No. Value ec. No. Value Spec. No. Value 0294 14.0 '0396 70.0 0824 14.0 k)I67 76.5 0170 13.5 0909 71.5 7 ,

0767 14.5 0589 76.0

.i.

0334 14.0 20004 ,

75.5 q;

-?.4 Average Values 14.0 'J 73.9

- Date of Tests nr,r.m u ,- 1 , lon?

h

.S ignature of Company Representative e AEdA N Title Signature of Government Representative Title

] .

t i XHR FORM 34-4 13 JAW 69

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