Text

In-Vessel Neutron Spectral Analysis
	ML20147D670
Person / Time
Site:	Browns Ferry, Cooper
Issue date:	09/08/1980
From:	Armijo J, Burns L, Sabol W; GENERAL ELECTRIC CO.
To:	;
Shared Package
ML20147D629	List: ML20147D634; ML20147D670; NLS8800063, Forwards Responses to NRC 880113 Questions on Proposed Change 48 & Revised Tech Spec Pages 155 & 156 Showing Increase of 2 F in Nonbeltine Portions of Curves of Figures 3.6.1.a & 3.6.1.b.GE Rept NEDO-24793 Also Encl;
References
	NEDO-24793, NUDOCS 8803040087
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{{#Wiki_filter:.. NEDO-24793 80NED0ll Class I August 1980 DRF Bll-23 BROWNS FERRY UNIT 3 IN-VESSEL NEUTRON SPECTRAL ANALYSIS G. C. Martin /N .rkA d a Reviewed: M b Reviewed: L. S.' Burns, Jr. W. W. Sabol. Manager Chemical and Radiological De ign Analytical Chemistry mb 9!l' /0 Approved: i .-l J.' 3. A3bijDanadr # Cor Materials Technology

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Prepared for Tennessee Valley Authority Contract No. 67C60-91750 Change No. 61 NUCLE AR ENGINEERING DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNI A 9512$ GENER AL h ELECTRIC 8803040087 880222 PDR ADOCK 05000298 P DCD

r-F DISCLAIMER OF RESPONSIBILITY This document was prepared by or for the General Electric Company. Neither the ~ General Electnc Company not any of the contnbutors to this document: A. Mak9s any warranty or representation, express or implied, with respect to the accuracy completeness, or usefulness of theinformation containedin this docu-ment. or that the use of any information discbsed in this document may not ,a infretge privately owned nghts; or B. Assumes any responsibstity for Isability or damage of any kard which may result from the use of anyinformation disclosed in this document. 1 11

.NEDO-24793 ~ CONTENTS' i ABSTRACT ix 1. INTRODUCTION 1 2. hUMMARY 3 3. EXPERIMENTAL 5 >l 4. RESULTS 19 .j ' '5. DISCUSSION 33 'm ~ ~ 6. REFERENCES 51-APPENDICES A. DISINTEGRATION RATES AND REACTION RATES, BF3 - CYCLE 1 53 l~ B.. LOCALIZED POWER HISTORY CALCULATIONS BY SCIENCE APPLICATIONS, INC. 63

i-C.

PHOT 0 FRACTION AND ASSEMBLY PERTURBATION CALCULATIONS BY SCIENCE APPLICATIONS, INC. 71 DISTRIBUTION 73 1 l I: l l iit/iv

NEDO-24793 TABLES Table Title Pg 1 Documents for Neutron Fluence Assemblies 6 2 Neutron Dosimeter Weights 7 3 Cadmium 0xide-Copper Thermal Shields 11 4 Dosimeter Nuclear Parameters 17 5 Neutron Spectral ~ Analysis Positions 18 6 Browns Ferry Unit 3 In-Vessel Neutron Flux Densities at 1098 MWe 20 7 Browns Ferry Unit 3 In-Vessel Neutron Fluences (August 19, 1976 - September 9, 1978) 22 8 Measured Flux Gradients in Capsules Located Between Shroud and Vessel 28 9 Determined Thermal Flux Densities 29 10 Determined Fast Cross Sections 31 11 Fissien Dosimeter Reaction Rate Ratio Comparisons: Reactor Power Versus Localized Power - BF3 Cycle 1 34 12 BF3 Power History (August 19, 1976 - September 9, 1978) 36 13 Adjustments to Reaction Rates for All Dosimeter 1 Radionuclides Using Localized Power Histories - BF3 Cycle 1 37 14 Effects of Photofission, Assembly Perturbation, and Localized Power History on Measured Fast Neutron Reaction Rates - BF3 Cycle 1 40 15 Significant Adjustments Made to Measured Fast-Neutron Reaction Rates 43 16 Integral Fast Flux Density Results as a Function of Photofission, Assembly Perturbation, and Localized Power History Effects - BF3 Cycle 1 45 7 17 Estimated 2a Uncertainties in Flux Densities and Fluences (see Tables 6, 7, and 9) 49 18 Estimated 2a Uncertainties in Fast Cross Sections in Table 10 50 v/vi

NEDo-24793 / ILLUSTRATIONS Figure _ Title g l' In-Vessel Neutron Dosimetry' Assembly 8-2_ Typical Location of Neutron Dosimetry Assembly (Side View) 8 3 In-Vessel Assembly Locations in BF3 (Top View) 3 4 BF3 In-Vessel Dosimetry Capsule 12 5 Precise Capsule Locations in Water Cap Between Shroud and Vessel 13 6 Dosimeter Locations Within-Capsules 15 7 ,,,,l -. 3F3 Differential Spectra - 4 cm from Shroud 25 8 BF3 Differential Spectrum - 30 cm from Shroud 26 9 BF3 Differential Spectra - 7 cm from Pressure Vessel 27 10 Power History Comparisons: Localized Power Versus Reactor Power - BF3 Cycle 1 38 't !]g j..I l l vii/v111 l t l t

or.DU-24/93 ABSTRACT 9 In response to vendor and utility needs in reactor material irradiation work, a multiple dosimeter system has been used~to determine neutron spectra, flux densities and fluences, and integral cross sections at several locationc in the water gap between the' core shroud and reactor pressure vessel. Thh irradiation was conducted at the Brovna Ferry Unit 3 nuclear pouer plant. The vork vas performed at the General Eleatric Val 2ecitos Nuclear Center. ACKNOWLEDGMENTS The author vould like to thank G. Simons and W. Hagan of Science Applications, Inc. (SAI) for their Browns Ferry 3 localized power history, photofission, and perturbation calculations, and to A acknouledge H. Till of the Electric Power Research Institute for providing resources to SAI for these valuable calculational phases of this work. The cuthar also vishes to thank the fot2ouing contributors: F. Brandt for ' establishing the dosimetry program, R. Ernst for the design, analysis, and testing of the capsule support frames,. J. Peterson and F. Smith for the preparation of the dosimeters, H. Helmholz for coordinating the removal and return of the capsule assembliec, E. Hoshi for the disassembly of the capsules, and

I.4 L. Xessler and H. Sebastian for the radioactivity analyses.

ix/x

NEDO-24793 1. INTRODUCTION The Nuclear' Regulatory Commission (NRC) in 10CFR50 Appendix H requires a. reactor owner to provide data on measured and calculated neutron fluence in the reactor vessel. In the past the measured data have been provided by iron and copper wires included as flux dosimeters adjacent to a vessel sur-veillance sample. Typically the dosimeters were removed'and evaluated at ,the first refueling of the reactor. In order to evaluate the fast (>l MeV) flux density from these samples, a neutron energy spectrum or spectrum-weighted cross sections are required. This type of measurement lacked the quality required to. determine accurately the vessel flux and to improve the calcula-tional methods. .4 To provide the degree of accuracy desired, an experiment was devised to measure the neutron flux density in a number of locations between the core shroud and the vessel of the Browns Ferry Unit 3 (BF3) nuclear power plant. The experiment was devised to use a number of dosimeters, each dosimeter ( being sensitive to a specific energy range, which allowed a more thorough estimation of~ spectral response. This process improves the quality of measure-The combination of locations to show radial, axial, and angular flux ment.- variations plus an improved spectral fit provides a much stronger base for comparisons between measured and calculated flux densities. This report describes the. experiment and the neutron flux / spectra results obtained-from the analysis of contents of the neutron dosimeter capsules. 4 1/2

o NEDO-24793 2.

SUMMARY

The results of the analysis of the Browns Ferry Unit 3 in-vessel neutron-dosimeters indicate the following peak (1098 MWe) fast neutron flux densities at full power during core Cycle 1. Full-Power Flux Density (4 ) g Elevation t o Near Shroud Near Vessel (4 cm) (7 cm) >0.1 MeV >l MeV >0.1 MeV >l MeV 11 10 8 8 1.5x10 7.6x10 9.0x10 5.2x10 Core Midplane 3* 45' 10 10 8 8 9.4x10 5.0x10 4.7x10 2.7x10 Near Core Top 3* 45' 11 11 9 0 2.7x10 1.4x10 1.4x10 8.1x10 Core Hidplane 45* The extrapolation of the full-power flux densities (shown above) to the inner b' vessel wall using flux gradient wires which were located within the capsules J indicate near-vessel (7 cm)/ vessel fast flux density ratios of 1.5 (3' 45' ] azimuthal position) and 1.6 (45' position). The measurement method utilized radioactivity counting data from activatio.+ and fission detectors in combination with differential cross section information I and an engineering computer program, in accordance with the standard method. The significance of photofission effects, capsule assembly flux perturbation effects, localized power histories, and flux gradients, all incorporated into l the analysis of the generated neutron spectra and the resulting flux density results,- is discussed. Discovered anomalies in the full-power fluxes derived from dosimeter measure-l i l ments indicate that measurements of neutron flux / spectra in light water l reactors must use local power history (i.e., power as a function of time for [ a small region of the core near the dosimeters), instead of total reactor i l power. Otherwise, errors up to 120% may be introduced. 1 \\ 1 3

NEDO-24793 Photofission effects in pressure vessel dosimetry, if not corrected, can cause tiie service life of the vessel to be underestimated by as much as 35%. The . magnitude of these effects is not well known ( 25%). The above are critical problems in that pressure vessel dosimetry probably will be established.as the measurement method at the completion of the 8-year (1977-1984) NRC LVR-PV Surveillance Dosimetry Improvement Program. V' l 'l i .h L P 4

.NEDO-24793 3. EXPERIMENTAL Groups of neutron dosimeters were supported by four separate leak-tight carbon steel frames or assemblies (Figure 1). The capsules containing the dosimeters were assembled in groups of three; a near-shroud capsule (S), a near-pressure-vessel capsule (P), and one midway between them (C). Twin capsules. one bare and one thermally shielded, were positioned approximately 9 cm apart on a horizontal plane at several locations in the assemblies. Installations were made at the core-midplane and core-top-edge elevations at azimuthal positions where the core is most distant from and where it is nearest to the vessel wall. Thus, two dosimeter assemblies (designated G1 and G3) were attached to the outside of the core shroud at the azimuthal position of 3' 45' (4* nominal) and two (G2, G4) at 45*. These locations were selected as having.the least interference for installation, operation, and removal. At each azimuthal posi-tion there was one assembly at 288 in. (7.3 m) (G3 or G4), the elevation of the core midplane, and one at 350 in. (8.9 m) (G1 or G2), near the top edge of the active core. Figure 2 is a side-view schematic of the location of two of the assemblies. Figure 3 is the reactor core top view showing the in-vessel assembly locations. The pads on three support legs of each assembly were j welded to the shroud and the ass ably extended across the gap between the shroud and the reactor pressure vessel (PV) wall. To facilitate their removal, the l assemblies had built-in breakaway joints in their support legs. These joints l consisted of 1/8-in.-diameter stainless steel pins connecting the stainless steel pads, which were welded to the shroud, with the carbon steel support legs. I \\. The document identification (I.D.) numbers which include drawings for parts of lJ the neutron fluence assembly were given in Table 1. i l !3 The neutron dosimeters included: cxides of U-235, U-238, Np-237 and Th-232; salts of Ag, Sc, and U-235; and Fe, Ni, Cu, Ti, and A1Co wires (see Figure 4). The oxides and salts were individually contained in small quartz tubing.

One, two, or three wires were also contained 'in individual quartz tubes. The oxides 1

and wires were weighed prior to the sealing of the quartz tubes. The salts were added to the tubes as either 10A, 25A, or SOA aliquots from prepared stock solutions. The dosimeter weights are given in Table 2. l 5 l

NEDO-24793 Table 1-DOCUMENTS FOR NEUTRON FLUENCE ASSEMBLIESL . Document I.D. No. Name 767E809 Frame 767E818 Frame 131C9275 ~ Bracket 131C9276 Leg 166B8673 Bracket. 166B8674 Handle 166B8675 Pad 166B8676 Cap I 166B8677 Plug 166B8678 Tube '166B8679 Spacer 166B8680 Brace .166B8681 Pipe 213A6070 Pin 213A6108 Spring l' I!Il l l l l l-6

en e .M N) Table 2 NEUTRON DOSIMETER WEIGHTS Cap-Values are mg Radial Pos./ sule Np U-238 Th Cu Fe Co Ni Ti Ag Sc~ ~U-235 Elevation C3-13 10.4 55.98 55.40 102.0 10.41 2.55 2.21 0.00598 0.0061 2.30 Shroud G3-14 102.6 40.10 1.88 9.52 0.00598 0.0061-0.050f Midplane (4*) C4-15 14.2 52.57 56.04 101.7 11.27 2.08 1.94 0.00598 0.0061 1.91 Shroud C4-16 99.1 11.00 2.21 9.82 0.00598 0.0061 0.050j Midplane (45*) Cl-17 49.7 90.06 104.0 103.6 20.58 4.68 1.93 0.0209 0.040 2.45 Shroud C1-18 102.3 80.90 4.59 12.54 0.0209 0.0061 1.76 j Top (4*) G2-19 51.2 99.76 114.2 159.4 19.96 5.29 2.41 0.0209 0.040 1.89 Shroud 8 2M C2-20 151.9 21.47 5.04 16.60 0.0209 0.0061 1.96 f Top (45*) C3-21 122.5 153.2 929.0 93.52 47.4 10.02 0.1045 9.62, PV C3-22 923.2 372.0 49.9 0.1045 0.040 5.91 Midplane (4*) C4-23 126.8 163.0 924.8 93.22 49.0 9.98 0.1045 11.70 PV C4-24 922.1 93.56 43.7 0.1045 0.040 5.01 Midplane (45*) Cl-25 922.6 1125.0 104.7 PV (4*) G2-26 926.5 281.3 103.2 f Top (45*) C3-27 136.7 98.17 141.0 517.2 46.85 3.63 5.92 0.1045 11.28-Middle C3-28 513.2 47.44 10.86 44.38 0.1045 0.040 5.39 j Midplane (4*) C4-29 513.4 188.9 9.87 Midplane (45*) Cl-30 772.7 747.3 49.60 Middle (4*) C2-31 767.2 187.3 47.78 .[ Top.(45*) s e y --a

NEDO-24793 r- _ S C ,P Eh-3 1___ ET - L ,g _p - i-! i __- O-g RPV-WALL WAu g. \\p_ k /, f---b Figure 1. In-Vessel Neutron Dosimetry Assembly REACTOR RQ /f PRESSVE n / VESSEL / WALL __0000 3 [ l / ef / q,:.: / 360-5/16" ELEVATION / (TCP OF ACTIVE FUEL) s [d ' ' / 350" n / ELEVATION s f N ? // l / l // N / s N s' . / ~ 288" / ELEVATION / ~rr 4 / CORE SHROUD Figure 2. Typical Location of Neutron Dosimeter Assembly (Side View) 8

NEDO-24793 3 45' REF (," R E F I T 45'REF o 11 p . '~. s =.,

==.u l I I.. ! .. i., ,O .,/ \\ / 6 g ],._. SHROUD.. ll lv, w E..il i /f l/l .) i L l T V M.1 O_j -Tlj /l /{ l 1 ~i t j '/ [ _VESSE6 / W i 3 p'I.D.1 [ I 4. I.. _3 v ,f 270 REF-s i. i-. ! -- 9 0' REF / l l l 'I l l ( llui i I l 180'REF Figure 3. In-Vessel Assembly Locations in BF3 (Top View) 9

NED0-24793 Quality control or the dosimeters consisted of determining the elemental or isotopic quantities of vendor material as received and of prepared solutions for. concentration verification. Neutror, activation analysis (two irradiations in the GE Nuclear Test Reactor at Vallecitos Nuclear Cencer) or mass spectrometric techniques were used f.n the verification analysis. Cadmium oxide-copper thermal neutron flux shields were fabricated for use in this experiment (see Figure 4). These shields were in the form of long, thin-walled tubes with end plugs at each end. Cadmium oxide and copper metal powders were blended together in a spex blender for 20 minutes. The powders were obtained.from J. T. Baker Chemical Corp. A blend consisting of 75 mole % ..y Cdo and 25 mole % Cu was blended in this manner. The tubes and end plugs were pressed from the blended powders utilizing uniaxial pressing techniques in an annular die. The tubes were pressed at 3000 psi and the end plugs were pressed at 25000 psi. Thermal shield information is gi en in Table 3. Four flux monitor assemblies were installed between the.hroud and inside pressure vessel wall of the Browns Ferry 3 reactor prior o initial startup. The precise locations for capsules of assemblies G3, G4, aid G1 located in the w water gap between the shroud and the pressure vessel are shown in Figure 5. Browns Ferry 3 began operation August 19, 1976. Numerous tests and shutdowns occurred throughout the BF3 initial cycle, which ended September 9, 1978. Although approximately 60 shutdowns were evident from the supplied reactor daily power-time histograms, a combined 7-cycle power history could be gen-erated. In addition, localized power histories were generated by Science Applications, Inc., for assemblies G3, G4, and Gl. The significance of these m power histories on final flux / spectrum results will be discussed later in the report. The assemblies were removed October 4, 1978 during the first refueling outage. Assemblies (groups) 1, 2, and 4 (G1, G3, and G4) were removed with approximately 10

""y ~ _ g -- 7 ._. 7'

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O: Table 3 CADMit!M OXIDE-COPPER TIIERMAI. SitIEI.DS Tube Outer Inner. Assembly Assembly Weight Diameter Diameter. Ileight. Density Weight I.cngt h - Cadmium Part (g) (in.) (in.)' (in.) .(g/cm )- (g)

(in.).

(mils) 3 ^ P013 18.507 0.5525 'O.376 1.836. 4.76 .24.227 2.072 43.9 P015 18.492 0.5525 0.376 1.847 4.73 23.557 2.057 43.6 ~ P017 18.536 0.5525 0.376 .1.824 4.81 23.648 2.040 44.3 P019 18.283 0.5525 0.376 1.818 4.76 23.486-2.039 43.8 P021 18.743 0.5525 0.376 1.810 4.90 24.294' 2.045 45.1 $h .? P023 18.749 0.5525 0.376 1.838 4.93 24.106 2.066' 44.4 'Z P027 18.734 0.5525 0.376 1.834 4.83 24.554 2.058-44.5 w ( -M' 7

NEDO-24793 e g'h%*EWW'4'1alim e CdOCu

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e. _..g i Sc --W Ag US N U8 w l NP 's n Cu _w 'l p AICo Fe l e-Ni Ti ~ mlIllijlll! ll11l11lll111lllllll11llll111 IllllIll[ligsgggli 2 3 4 6 o Figure 4. BF3 In-Vessel Dosimetry Capsule 12 - - - - - -.. - ~. - - _.. _.... _ _ _.. _.. _ _ _, -

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'//h A1 830 l Al l425 4 N u -' . o.- . -n i.n ; ,n ) / p/ S ,u.. u--1 F- --i i-. o a- .,e. //f.,/,/,,5 s18 ' Al _...t. i Precise Capsule Locations in Water Cap Between Shroud and Vessel Figure 5. 13

NEDO-24793 4 100-lb pull. Assembly G2 was not in its as-built location and was found in I back of-the jet pump 8 -and 9 D/P lines. The measured dose rates were: R/h R/h Grour (cen::c:) at 3 ft 1 7 U.S 2 2 3 8 4 30 These assemblies were returned to the Vallecitos Nuclear Center (VNC) on "I ~ November 1, 1978. i .The four neutron fluence assemblies were sectioned in one of the hot cells at the Vallecitos remote handling facility with a high-speed abrasi.a cutoff wheel. After removal of the sections to a shielded hood, the capsules were separated from their carbon steel envelopes. Subsequently, the quartz tubes which contained the dosimeters were removed from each capsule, the position of each tube within a capsule being measured i j before removal. Each dosimeter position within a tube was also measured. Figure 6 shows the dosimeter locations within the capsules. The dosimeters withstood the handling, irradiation, and disassembly conditions very well; 119 out of 120 dosimeters were intact for analysis from three assem-blies (the fourth assembly, located at core top at 45', had detached during the irradiation and was not utilized). All tubes were found intact except the following: 1. tube tips were broken from (all were salvaged) Cu wires in capsules C4 #23, C3 #21, G1 #30; a. b. Fe wire in capsule Cl #30; c. Np 02 3 powder in capsule C3 #21; and 14

Z Qo w*w= 1 5V o P N, C i o N, c l oC A, e g o c e g u ueSc9 r, l C ,ehc o r A, CFUSA g7g nI WI ,e o A p7r d a ,e C 3r5 C B gs g21 ~ a* iam c I V.0 6 1 ( ~ ~ V 4 ) P ~ 2 2 2 d 5 ) 2 m D ~- e d bI. s u ^ r e o es g( Y,r-1, a s u n 4r f t + e s m a c ( 7 D. n n ( 0 o o i i 3 t t r a a 3 c c o o ^ e l l b s u r r e J t e e l ) t t u V z e e P t m m s O o ar t t p ds ds C C e u s s a h o o I S A o F, q l C A, C ( p r F'C F .e u* i o 1 s e 1 C n o T, A. T, N r I' F n e eI i o U[ C A T A i u1S { o h i CNU*An

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t t i i s Sc3l 1 P { 1 W 5 s V n t P o a i ^ i %)/. d o 2 s t m. -( e3r a n f ,= C e 0 Y i i m m o D .2 r c L t 3 h e re e t B e m o ,n i r T ,o. o a s r 7-D t c ep 1 1 S N, N. i o N, 6 C o ~e C O l I C A, A. A, u I r e i r eu g Fc T, eu y I e FC i r g* F gF + h S _~ r - ^ s ) r s D n o 2 ; 6 g r I 2f _~ 4 1 i _u G 1 s c 4 _[ _~. r 9 - r ^ r?- _l = N D^$, c - gc5e1 p DUA"5e 89 r F T,uA S "[g a n5 c s u A [' o ee qc y cs q tcS h ASUF,T, s US I, IIl ou u 3 g o f.l { CC C h l 1 S ) A 4 *s ) y 1 ce c4 c. t ( g g I

net 0-24793 '2. quartz tube in capsule G3 #27 containing Np 0 was broken; this 23 powder could not 'be salvaged. The intact non-wire-containing quartz tubes were cleaned with hcl and with acetone. These tubes containing powder and salt dosimeters were centered'and taped onto labeled-6-cm x 9-cm x 0.051-cm mounting cards. The wires were removed from the tubes, cleaned with acetone, and taped onto mounting cards. 7 The. dosimeters were non-destructively analyzed for radioactivity activation and q fission product content by gacna spectrometry. Gamma detectors utilized were 3-in. x 3-in. (7.6-ca x 7.6-cm) NaI(T1), 80-ce Ge(Li), and LS-ce Ge(L1) calibrated systems (instrument numbers 3564, 3561, and 3562, respectively). The Ge(Li) detectors were used in conjunction with a Nuclear Data 4420 computer / multichannel analyzer system. Reaction rates were subsequently calculated using slightly modified versions of engineering computer programs. Table 4 lists the nuclear parameters associated with the BF3 in-vessel dosimeters. Because the dosimeters in all capsules had been positioned perpendicular to the shroud / vessel walls, flux gradients within each capsule had to be measured ^ or extrapolated. For flux gradient analysis, iron, copper, and aluminum-cobalt wires taken from capsules G3 #13, G3 #14. G4 #15, G1 #17, G1 #18, l G3 #22, G4 #24, G1 #25, G4 #29, and G1 #30 were cut into sections; these sections were weighed and counted. The fast and thermal / intermediate flux gradients for all capsules were determined from these wires. Based on these flux gradient radioactivity measurements and earlier location

I measurements within the capsules, a short computer program was written to normalize all measured reaction rates to reference points. Three reference points in the water gap between the shroud and vessel were chosen as the measurement locations. These selected points were located near the center of each of the thermally-shielded capsules, i.e., 4 cm from the shroud. 30 cm from the pressure vessel, and 7 cm from the pressure vessel. These are designated as near-shroud, center, and near-vessel locations, respectively.

The point-wise location of each dosimeter was arbitrarily taken to be the midpoint of the dosimeter. 16 t

.NEDO-24793 Table 4 DOSIMETER NUCLEAR PARAMETERS Half-Life Gamma-Ray Cac=as ver _'osize:ar-Radicauclide (7 ) _ Energy (MeV)-. Disintegration 9 Ag Ag-110m 249.9 d' O.8847 0.729 Co co-60 5.271 y 1.3325 0.9998 Cu Co-60 5.271 y 1.3325 0.9999 Fe Fe-59 44.6 d 1.2916 0.436 Mn-54 312.5 d 0.8348 0.9998 l-Ni Co-58 70.78 d 0.8108 0.9945 Sc Sc 83.83 d 1.1205 0.9999 Ti Sc-46 83.83 d 1.1205 0.9999 ~~ U-235, U-238, Np-237 Cs-137 30.17 y 0.6616 0.852 1 .() Zr-95 64.4 d 0.7242 0.442 j Ru-103 39.35 d 0.4971 0.895 Th-232 Cs-137 30.17 y 0.6616 0.852 Zr-95 64.4 d 0.7242 0.442 l Fission Yields Utilized Dosimeter Cs-137 Zr-95 _Ru-103 U-235 (thermal) 0.0622 0.0650 0.0304 'li U-235 (epithermal) 0.0615 0.0637 0.0328 U-238 0.0599 0.0510 0.0622 i_ Np-237 0.0650 0.0593 0.0589 Th-232 0.0593 0.0564 5 4 17

n. r: i NEDO-24793 - For each assembly, epithermal spectrum and fast spectrum reaction rate extrapolations were'made to the shroud (o.d.) surface and to the pressure vessel (i.d.) surface, respectively. The magnitudes of the flux gradients f' to these surfacca vare assacao to be :ne same as those ceasured across the near-shroud and near-vessel flux capsules, respectively. These are discussed later. Reaction rates were inserted into the computer program for neutron spectral analysis for the positions listed in Table 5. - The "X" designation refers to spectral analysis using full-spectrum or epithermal reaction rate data sets; the "0" designation refers to analysis based on relative reaction rates of a small (not full-spectrum) selected. set of reaction rates. Table 5 NEUTRON SPECTRAL ANALYSIS POSITIONS Capsule Assembly Spectral Analysis Positions ' Assembly Location Shroud Near-Shroud Center Near-Vessel Vessel [- 1G3 Core Midplane X X X X 0 3' 45' G1 Core Top X X 0 0 0 3' 45' t G4 Core Midplane X X 0 X 0 45' G2 Core Top 45' (could not be analyzed) i ;- \\ - I I I 18 i I I

m NEDO-24793 m, 4. RESULTS' ~.The measured disintegration rates and measured / extrapolated reaction rates (dps/ nucleus (saturated)) are presented in Appendix A.' The determined. full-power (1098 MWe) integral flux densities and actual' fluences at locations within the pressure vessel of the Browns Ferry Unit 3 nuclear power plant are presented ir Tables 6 and 7. respectively. : These are listed as flux densities and fluences above the energies of lx10'9 MeV (full- ~I'MeV (epicadmium), 0.1 and 1 MeV (fast), and 2 Nev, 3 Mev,. spectrum), 5x10 4 MeV, and 6 MeV. Although actual full-spectrum determinations were made at three near-shroud (4 cm), one center, and two near-vessel (7 cm) locations, extrapolation of some peak. flux densitias and fluences could be, and have - been, made to the shroud outer diameter (intermediate and fast spectrum) and to the pressure vessel inner diameter (fast spectrum only). - Reaction cross sections were taken from the ENDF/B-V dosimetry file. Localized power histories, rather than the reactor power history, have been l O uti11 sed to in, rove the accurac7 of the determinations. The fium densit7 and fluence data have been corrected for photofission (y,f) effects on the U-238, Np, and Th dosimeters using the latest data available.2 Corrections for the - i capsule assembly perturbation and spectrum effects in the fact (>0.1 HeV) energy region have been made2 (i.e., reported fast flux and fluence results are unperturbed). Thermal (<0.5 eV) and interme.diate (0.5 eV - 0.1 MeV) flux densities and fluences, although corrected for capsula perturbation, do not correct fer any changes in spectrum effectr caused by the assemblies. Ih The principal energy regions of interest for damage studies are the fast regions >0.1 MeV and >l MeV. Some pertinent information which can be obtained from Table 6 is presented below: 1. The highest full-power >l MeV flux densities at the shroud wall and pressure vessel vall were at 45' midplane and were determined to be 11 -2,,-1 and 5.0x10 n'em 8 -2,,-1 2.1x10 n*cm , respectively. l' 19

c NEDO-24793 Table 6 BROWNS FERRY UNIT 3 IN-VESSEL NEUTRON FLUX DENSITIES AT 1098 MWe Integral Flux Density Above Energy Energy Near-Shroud Center (30 cm Near-Vessel (MeV) Shroud (4 cm) from PV) (7 cm) Vessel .C3 Assembly - 3' 45' Core Midplane a. (nacm~ *s~ ) 1x10 ' 11 10 ~ 5.9x10 4.1x10 3.7x10' ~7 11 11 10 5::10 5.0x10 2.9x10 1.7x10 1.7x10' 11 11 10 8 8 O.1 2.5x10 1.5x10 1.1x10 9.0x10 6.1x10 11 10 9 8 8 1 1.2x10 7.6x10 5.8x10 5.2x10 3.5x10 10 0 9 8 8 2 7.2x10 4.7x10 3.5x10 3.8x10 2.6x10 3 10 10 9 8 8 3 4.3x10 2.9x10 2.2x10 2.8x10 1.9x10 10 10 4 2.7x10 1.8x10 1.4x10' 8 1.2x10 1.3x10 9 9 8 7 6 8.5x10 5.6x10 6.1x10 7.7x10 5.2x10 h b. G4 Assembly - 45' Core Midplane (n.cm'.s~1) ~9 2 0 9 1x10 1.1x10 9.4x10 6.5x10 ~7 11 11 10 9 5x10 7.9x10 5.2x10 3.9x10 2.6x10 11 11 10 9 8 O.1 4.2x10 2.7x10 2.0x10 1.4x10 8.5x10 11 11 10 8 8 1 2.1x10 1.4x10 1.0x10 8.1x10 5.0x10 11 10 9 8 8 2 1.2x10 8.1x10 6.0x10 5.8x10 3.5x10 2 10 0 3 6.8x10 4.8x10 3.6x10'9 8 8 4.1x10 2.5x10 0 10 4 3.8x10 2.7x10 2.2x10' 2.8x10 1.7x10 10 9 9 8 7 6 1.2x10 8.4x10 9.5x10 1.2x10 7.3x10 20

!k I f j!. ^' e [q, NEDO-24793 j, Mfd ,j&.. n Ql:

n

,9F Table 6 (+ 't 'L+ .f ).: BROWNS FERRY UNIT 3 IN-VESSEL NEUTn0N FLUX DENSITIES'.AT 1098 MWe (Continued) ];;. ;)- Integral Flux Density Above' Energy y Energy -Near-Shroud Center (30 cm ',', Ne a r-Ve s s el 4.. (HeV) Shroud (4 cm). from PV) T, (7 cm) Vessel l - 1;g .j; - s,p/Qg c. C1 Assembly - 3' 45' Near Core (Top i

3..

g (n*cm" +s~ ) ,.4,4 k 3 1x10-9 3.6x10 3.0x10 2.0x10 11 10 9 e 5x10"I 2.6x10 1.8x10 1.2x10 8.9x10 1 11 10 8 Elb 11 10 9 8 8 0.1 1.4x10 9.4x10 6.8x10 4.7x10 3.3x10 10 10 9 8 8 ll l' 7.4x10 5.0x10 3.4x10 2.7x10 1.9x10 f, 10 10 9 9 8 2 4.6x10 3.2x10 2.1x10 2.0x10 1.4x10 10 0 9 8 8 3 2.7x10 1.9x10 1.2x10 1.5x10 1.0x10 10 9 8 0 4 1.5x10 9.8x10 8.0x10 1.0x10 7.0x10 9 9 8 7 7 6 4.2x10 2.9x10 3.5x10 4.0x10 2.8x10 .S., .e t-5, i ,.I l' !PIg-21 m_

[p NEDO-24793 j_ j F Table 7 BROWNS FERRY UNIT 3 IN-VESSEL NEUTRON FLUENCES

f (AUGUST 19, 1976 - SEPTEHBER 9, 1978)

Integral Fluence Above Energy 'l Energy Nea r-Sh roud Center (30 cm Near Vessel (MeV) Shroud (4 cm) from PV) (7 cm) Vessel Y{ G3 Assembly - 3* 45' Core Hidplane a. ~ (nacm ) 2 1x10 2.7x10 1.9x10 1.7x10 9 I 17 6 5x10 _2.2x10 1.3x10 ' 8.0x10 7.9x10 ~ 0.1 1.2x10 ' 6.9x10 5.2x10 4.2x10 2.8x10 1 18 17 16 16 18 18 I7 6 16 1 5.7x10 3.5x10 2.7x10 2.4x10 1.6x10 1 10 i 2 3.4x10 2.2x10 1.6x10" 16 6 1.8x10 1.3x10 I6 18 18 16 15 3 2.1x10 1.4x10 1.0x10" 1.3x10 3.8x10 1 16 15 15 4 1.2x10 8.2x10" 6.6x10 8.9x10 6.0x10 O 3.9x10" 2.6x10 2.8x10 3.6x10 2.5x10 17 16 I 15 b. G4 Assembly - 45' Core Midplane (n*cm" ) -9 19 18 5.1x10 4.3x10 3.0x10" 1x10 19 19 3x10' 3.4x10 2.4x10 1.8x10 1.2x10" h 19 19 1 16 16 0.1 1.8x10 1.2x10 9.2x10 6.6x10 3.9x10 18 18 1 6 6 1 9.6x10 6.3x10 4.7x10 3.7x10 2.3x10 10 10 6 10 2 5.4x10 3.7x10 2.8x10" 2.7x10 1.7x10 8 3 3.2x10 2.2x10 1.7x10" 6 16 1.9x10 1.1x10 18 18 4 1.8x10 1.3x10 1.0x10" 16 15 1.3x10 7.7x10 6 5.5x10" 3.9x10" 16 15 15 4.4x10 5.5x10 3.3x10 22

,? .e< NEDO-24793 ' =;M mtf }id Table 7 b' ' L pi. h ig.Q BROWNS FERRY UNIT'3 IN-VESSEL NEUTRON FLUENCES (Continued) s bg; (AUGUST 19, 1976 - SEPTDGER 9, J1978) . }. a. a. ,M!El' c .[i:;{- Integral Fluence Above Energy .i , {?;yi Energy-Near-Shroud Center (30 cm);' Near Vessel A' (MeV) Shroud-(4 cm) frein PV) F (7 cm) _ Vessel l.t :T: ' I IA'; i-Tj; > @ Mi G1 Assembly - 3' 45' Near Core Top c.. . N -};if. - (n cm-2) i i hj 1x10 ' 19 8 16 ~ 1.7x10 1.4x10 9.2x10 'j I -7 19 18 17 10 5x10 1.1x10 8.3x10 5.5x10 4.2x10 18 18 1 16 16 0.1 6.4x10 4.4x10 3.1x10 2.2x10 1.5x10 } - l. 0 18 17 16 15 1 3.4x10 2.3x10 1.5x10 1.2x10 8.8x10 0 18 16 15 15 2 2.1xib 1.5x10 9.8x10 9.3x10 6.5x10 18 1 16 15 15 3 1.3x10 8.9x10 5.5x10 7.0x10 4.6x10 O4 6 6 to 4 5 1o 3 7 to 4 6 to 3 2 to 1 17 I0 15 6-2.0x10 1.4x10 1.6x10 1.9x10 1.3x1015 9 f ( ame t I 23

s. l r-NEDO-24793 r ll ^ 2. Ratios: G4/G3 (45' Core Midplane /3' 45' Core Midplane) ) Shroud Near-Shroud Center Near-Vessel >0.1 Mev, >l MeV 1.7 1.8 1.75 1.6 i, '[u 3. Ratios: G3/G1 (3' 45' Core Midplane /3' 45' Near Core Top) P. Shroud Near-Shroud Center Near-Vessel >0.1 MeV, >l MeV 1.7 1.6 1.65 1.9 4. Ratios: >0.1 MeV/>l MeV .g Shroud Near-Shroud Center Hear-Vessel G3, G4, C1 2.0 1.95 1.9 1.75 .c 5. Ratios >l MeV: Shroud /Near Shroud Near Vessel / Vessel Shroud / Vessel G3 1.6 1.5 345 G4 1.5 1.6 420 G1 1.5 1.5 390 Generated neutron differential flux density spectra from the computer program are given in Figures 7, 8, and 9 for the near-shroud, center, and near-vessel locations, respectively. Included in these figures are the energy ranges over which the flux dosimeters are sensitive (energies between which 90% of a the activity is produce.d). In general, the more dosimeters "covering" a region, the greater the accuracy in the resulting flux shape in that region. A dotted line indicates no flux monitor coverage (e.g., there is no dosimeter coverage of the thermal spectral region below lx10-8 MeV or of the fast spec-tral region above 12 HeV). Since the computer program represents the differ-ential flux density shape as a continuous analytical function, the amount of spectral structure which can be unfolded is minimal. Thus, if severe struc-ture effect exists in an energy region, the unfolded spectra vill not reflect this. 24

l 5 I _,. EM_ '_ ,.,j- ' ' '.~C^ i7 [~ " - 1927' 4 f N,,j,.- i \\ 0 10 %,.I[;.;Iltf I I!y!P N - l!!C5 m WMr" i E E - ]HE~a". ' "'i ~ "" f J^. !*i l l% ' i -S U**~' N@ ~_.[2 % Core Midplane 4)* (G4 Assetnbly) :- J2..h!I'*ff'-h*f/Td,,_3p;lM" l ik"5' 7 jh b. Core Midplane 3*45' (G3 Assembly) ii Tt" " l f -h.T"I;F'+ti _:.- f:.::r ~35~ "-l 5

/

e' ._-. 4. f : .f/ t. I. - / c. Near Core Top 3*45' (G1 Assetnbly) E 37 t - - + - u(n.r) Cs (Ca) m ttini i i iiinii ~ i iiiiii i tit -

d I@l;!*' ///; "*jinm4Dj[ji' Ii!!

M l l hilll5""%W ,"J. 1.'.7t "'#Yf.! 5?I!; - -[ h ] -'$.' 'N, l f 23Y 1 33 Ih ~ 8- / ' M 't* t ik ' - htttt' i 238 137 b. !X E* h [!Efl +o. r

,3

- c.ht. 1l235f ' t 1,, U rn, f)137T!!!!Tmd. Y t*?++t+- T }l..h.: g n!E,g _,- .i' -MN, II

==,4 ;M !...;T. U.. :::= UI"* fI. C5 Cs r.N' f "n / / 0 .I x =. ,4, ,gg.. "f t"'I 3 g O - t t " P' -+ * -t+<*-~*- 137 ] Ti![P p =(n,y-} 'Co 'IFrilllliT!T !!!!!?pilip;n?'"'i3 C" R1JJ,MT' Zy};f l =Co(n,Y! '+

b 1 ' " * :';

3, Co (Cd) . t " trerit ~k

W-- + t ! !+

58 Co. ]'tr'.n4Nh~58'II"*PI ~ '0 " ,:. f f. I (yJ}f, 4.. g -m, ._. r* ! '!!!!! ! "! U 45 l l f C

:liig! ith:45sc(n.Y)4*se M !

l g"- - + t i t] p~~ c 'I

P*

m 46 E--'- I F ~' ~ i t I 54dg ~ = [41niilll? M g 54 sc (n.H sc M 3n '11 ill 7j Eje - - -- " ,. ttRh-- "EWlinrm - r o m t!.-..:e - W.7 7 %u 46Ti(n,p)44 ? i r ...... = * - .-..,u.. m .., _. y;,, . a I-58 ,. 54 I ; r+ se +. 7 - b ] (( l lIly8l'fIlljlyy h._. 'N 0 C m g Q.];;*[Z p ~*.::*.'#47~I** 11 7 ;

i:~~
58 59

~ ~' I! PI' - 'i'" C"I" 'I C U WICill!,[RIC ' dLWITTIf;"igliiI : 24T ~ ' g' 7~;g, ygjjg>,gg L d F* I",u re (Ca)

- 8 {-- tl 9g gg n

r*' x pu:..___..,gt:::

tT 109,(n.u110m, r* 3 H. t= ~r, n.- x x

+rn-

- m'" "'m". i i : e / ' b"N" ~'^7.!'CdI"!ill!!!I'! : :mirir- ~1T ~ ITIII!!HMIEEEl stgupHUIish@_ NIL-r1 pp 'y$i!!!!~ l L wm- - ny{E.; gmivt= 1 ~

t=

iin M

a: I $.5 p 'm!!!!,!M_,. 'l,![lil"_N* UDINilIN-.= Y_2 $- -1 {Q,n%E%{IiCIQ Q,R,,:--- - J. = .m "

4. myre t t.- --m um -

g- -.

i

+m i

ii e--t im 9:t + N k:l_ " Ih[!TW WT.I{}$,,U__ISIllBan==M.._I r- ~- ti;. .T_. !@--l._ }lEIIII lll.!T_[N.. ll% II$5 Y i .. n i,i , tt ~. g i - m in. i.; W t i tiiJi ~ m lff@,W+ ' N MTIh!!IYT R m i -]:.rl@~-- ((I YWI N1[IEfW-PhMRlI Y% "jiN!!MM Q::m EIU IU t .--r--:q fr+i-I _= 4 t:::g_- r+g - i lini i ~ii iiiisi t ?ti't-_. _-m n.nJ

i

A liii i

lit - ~ iI-ms ili ili i i! i iiiiri !Iffi4--- io io io

io. s

...r m.. .s no. io.: ioo NEUTRON ENERGY (MEV) Figure 7. BF3 Differential Spectra - 4 cm from Shroud

l ~ 3. - ~, * " . ~- ~ t to" 5d 9 -t t.h:::- t l l !lf!N -~ h k kl N h:tlNl5 .ll!! M !!l.} !N-55l= i I NE l}N555EI!!N! SE k E II H i l~ T =t ::, ._f m,*t m r,m n -+ 1r rr. ,ois n. 1 u [g on n TFF ?{ {!% I"O p.W=} ..(

j. "::n]u ; fn';"

W I Hr N R v , n. ]lA[P " Core Mid,31ane 3*45' (G3 Assembly) !-- Uf""' f ttit' {

=n i

"H n a =: g:_ g P {m]Iljliil 7 / w di LJ9","I'"MJ".lT"--"-t-t "

t---q~tt*~-t i ti_

i j itiii t-"at ' t" ,, o m^ - 7~"-"" t ~t i o i in ]. T" j,tW"? "Ifp=m:.ni =fjip== .e _.., I'& "]e=+FfPg" "- j+fl? T H N rn !L m = r.t tt il-w-tT - = -pP~ = = !"t t / n*~ '-t t +Tt* -"t'*hM wt --- t t+titisi i ,,a ,a ~1-ILl ffIm plIT It!T"T'y illlllf" bn: T ]{ij'Pf /. - =

Il U

lm r . Mif ,.gf~.[h/Ql235. ...d k p h /'...: l 4 ,E 11-IE p tt t* t 3' M ll 137 ..s.1.:f,1 : ~/ %g: mg_gggy 235 g g gp"t llty.*y((hgg g 137 g g ,* Iim i Z . >gijr/rrN.:IF-)tr{.. a 5"

i t

im9 i sa...inic wsq tr.Ma.li.eug=nircium!!rmil'*z<o."'"Cs r u san.7 m,-> 1 t t m,.. 1 ,3 3 i h lI % 3 U N d d hlI ~~~ i II NN t-NIIIM ll!IM },{ f lIII_g Th (n, f) 137 232 m ~ Cs Z t' ' -f f Co (n,bCo l' ' f Co(n,'t) Co (Cos

~? i fN

~ f 5 '*q '..8 _._ F u ::* % P f3lillf MHillHfb T1.3,_;1.ilympg,4 T++ "~*w'. d Irpd ~ l 58 E 5 os .. m.n..;.. . t - 5 8

r. - -r

t..,. !!

O r4 % nitn.Pl .Co-L -u'- ++4-' -. 42 _-4. 59 58 59 L+- -t. I;;$_. f f'Uq, ; + ** ;*Jnr : -.- .~. llh 5 {hil l...:-. f " ;1::- ;. h. hI %

I-f

=- ~~~~ '.} re(n,p) Mn Z 't

~

:I r:-

h.p_ 110"g l ll!' Q I41NHH Fi-fifHi Q M - b T NIh~ 'ittitt "h."-lhN 109 fQ46 A4 0 7I s, ' ~i T ! Ag(n,Y) Ag (Cd) [ ; ~ :: f:: :hh..].:!N[rN,: -h- - r rp ft.;- ~ --~ %.IIyjdI-ii.:

.. = (m wrN*63

"'N 60 q:;;.1j N 'II'. - II I !lf!E -'I ~!. :a HM-2l-i I E 1. %=4llJiii2EDfIf!II FI}inCy* jl'f: r !lll.!!!.. g i@! ::;- git..u : :g:" gg : 1.qc.+: .e: y 1, l !!!!L.Al.fln: t4, M!Il!E' 111!!Ut"._.id11 2 -4fL Y lM I -A.-, yg"+---, - i. I ; m::::.==.. " t-_m. t: i

, a 2

i}-= e 4., j:':t* *-:=. _- - 4 g:::: -.

~:n

j e t-*Ptd - { 1 't t W~ v= L-e f' t

l*

30 m%me.. -o-vie ~~ * ~ ~ - .b~ w!~ , -. -. o erra eu -w' mrle we = b - h'M[#'~ K Q :. j jllrl ~= ~;' ' f

55-'

3. :s *I l-~Y.

F N f.

- - -=

i-J b#-.15.-,i...7..',M..1 1. 'i:

= --

m -n-a F 81: li 3M8*;" , h.!.2 3-S ..I .si .}13I M. - ' d - -I _n x y .'.I = T!f'

y'! gii g i i-itti

,ij, =E iO* ye ri i El I i.. eb i g'--'p qp!yggp qg': wm q:-on mp

- 1

3 a

_ _rc.V

=

iEE --E L -~i. 'ttnii F-ii i

i iiik

i iii N;ti l

i iii in c-ttti ---"- Mit: c - H 7 = ,, s -~F 30 80 ,o 30 8 IO 10 4 IO S 50 8 10-3 80-2 10 8 800 M J 80' NEUTRON ENERGY (MEV) Figure 8. BF3 Dif ferential Spectrum - 30 cm from Pressure Vessel Oe

j

si trM M 1

],_ l-- ~ - - 7 , s ::. m ( y,3 (v} tt io MfljlIlGjD jjjlERIMll M" M}l $E $ I !bW lbd!O l1. uti n1Mb MM6 3d T %d !!E Ml13 i.m . g..::n _1 7.;, - q pp q g 7 .p,_r m o E _. !j'::J]k {J.lOl!L3(( h II j W i7

! P Elj"

! a. C r-midplane 45' (G4 Assembly) y l~~ .._ _1 ;; NJ 5;s1h ]ll"j'b{ oi! b.". ,b. Core midplane 3*45' (G3 Assembly)*:@ .-g_+dl.l.{-{J3$. l IT *- fliPFM::.l.H'.=_- _5,.1l.i. Erih.l.!.b.d: 4~.gb=ifb, lI@t-"-"NE"~4"#~'.r E-l

g-.-. -. ;.

p j 77+g 7-+ {+.p_g.}7{p..; 7. {,,.. .a w... ; m ~- j j; ..,g j;

33 iois 7 9-

p g syff,.

.UI TI !!I! "$ lC (Cd) 3 = g f, m!N-O P IIIllITli ['f!!-E-U sil..l!!E=_.f_12 3 5U (n, f) 137 "!.'-b /.ti$ mN!!!{ 1.EW ' !!TE l -I / 't 1 Cs = 3.

.ggg..3_q 35

+ / -

O-
--t%

+v : t-.t i t' / / y^"';~g;[j jllgg _ g lJ);y (}llt,W:1d1 tiiut=' eft (f{t*-g

T.. jji./ [/ {'37~(n' f) Q*5

/

T A.

T-P O '8" (({g--t

t. ". II' ' gj

~.y { O

W *...J; Cs

- - 2#-

~Y *-4 II#^ 3' ~ r mt?" I-st g 237ne(a f)137c= t$m.__: f F. .U 4 U f 'T* !/fl f gE-e z utrid % gj }{ l d M hii-j {g 3MflffhN I j{,{._y_ - m tr t t-- t - t 11 -

t ~ e +---

e++-- +- - - f

,, [ I[

[ 'A 3 h t l t I ' 'N Sc (n,Y ) Sc

f i fb~ * ' A*

'I* 'fi Ififfi{

If3 -

1* U (n. f) Cs -5" -

j illl' -Pfm_t ::

.:r. :. nam:_n.tw. i [~D l l i s..' "' ' i.

'lj p Till!!iisp.l.l: lillie fij!!!!M-jm.fiIll? ^','e R

- fh aWjp 2..

m un. * - * " * ' *

- + * " - t ttL u

t" m:

r o.r.: z "1 ""' fl l -' it?"--

; - t :"

t ' ' t' 58 [

q

.4lNI!MI III. re (n.Yl' re "I*I] lll]{f'" ! lll:l:l *~ { l 9 g M f'. ;;fl_Hhjjjh] b:.:a::t_:Ql Ij,,, Ni (n,p) 58, Co l u ._r 79 y__ i. p p. . -,- : n ::.

..p.

~ m, yd m, ~ to" y

""W

'2 j4 c ss C gj%g, 3 =g nfr 54 y re(n,p)$4.Mn O

4 --".J 7 Co (n,Y ) 60Co f s 1 59 1 if rt.-

z 59 i.* Co (n,Y ) 60 . # t ut x Co (Cd)

, y,.

"L,=

- 4_.p.Z e

} e v (__1.1 {. ; -I + t.I!(r[_J o,oio ^ a satsch,_. j :{. f......arg h2= _- Q f. b.

3. N..

3,4 g r .. Am:::_.. ". f lo9 .. - jF: q:_ = -:-- Ag (n,Y ) 110 mag }. t l :1 U.,,.l H 8 MG -1 = t 63 60 3 2-

i4fc "r

2_g. Cu(n,al Co "Tirg n'.r} j jjjf!! j j f Ib'Ei ID" I Nlf.![ Il!9 I -^-~ MNm 5 5 ..j .E~ =m- -~ I II I t'~- _.1 ~" W;' 109^' '" " 110m

-+i

Ie"e N i+-+--

t litM9.ffiit-jd'i!l-Q.. _- * . f}. -M q_ i E-i~f r n::2 r- ^' '" *:1 r ' u@: a.t 11til!!e hh =' =z . 7p._...;_g:-k t:tzt.._.-.1.p.; q. ~ h..__

t. :;utur

. 7 ;g J. :.:... _ t,;;l::: : r.._ .p ,.y, 3, 7 .. g.. n T~._i l lQ$7MM l%Qll%-~S [Y[l 'fjf M-l hY fIjjE% Ef NM b I T j

~ * - * - * * * * + ~ - + "+)++ g 'fitL

~*III~ I 'IiT [tI"I 9ttt--" -E m' Ei T]!!!PYP" M}E q.m

g g-g 1 1[fM7i]!ffQt"{=U![MEM[]!h[

b Y: Siiq _: 2r--+ %mL* 'T3lk E M -, t +- + o :+ = ur-- - lii-i lii it i liid ii i:F i lii i i ii fini i 4 ios 80 80 10 90-8 so so e so-5 10 8 90-3 10 2 10 8 100 4 J 80' NEUTRON ENERGY (MEV) Figure 9. BF3 Differential Spectra - 7 cm from Pressure Vessel

q- [gl: NED0-24793 n w. d. The extrapolation of measured activities to the capsule centers and then to 4 the shroud (intermediate-and fast-neutron-induced activities) and pressure vessel (fast-neutron-induced activities only) was based on the analysis of flux gradient iron, copper, and aluminum-cobalt wires which were located within the capsules. Reactions utilized were Fe-54(n.p)Mn-54, Cu-63(n.o) Co-60, Co-59 (n,y)Co-60, Fe-58(n,y) Fe-59, and Fe-58 (n,y)Te-59 ! Co-59(n,y) Co-60. Flux gradients based on fast neutron reaction activities ranged from 1.0% per mm at the near-shroud capsule to 0.6% per mm at the near-vessel capsule. The intermediate flux gradients were similar to those of the f ast flux gradients. The thermal flux gradients at the center capsules (30 cm from PV) and at the near-vessel capsules were determined to be 1.3% per en and 0.7% per mm, respectively. The peak thermal flux within the watet gap appears to occur within the near-shroud capsule location at appre::imately 4 cm from the shroud. These flux gradients are summarized in Tabic 8. Thermal (<0.5 eV) flux densities using the cadmium-ratio method and generated by computer are presented in Table 9. The differences in the magnitude of the thermal flux densities between the two methods is small. This is not surprising since the algorithm written for the computer program for a thermal spectral region had been based on results using the cadmium-ratio-method data 'I of earlier experiments in the General Electric Test Reactor. Unfortunately, no known standard thermal neutron spectral shape is available to experimentally test both current measurement methods. The accuracy of integral thermal flux density determinations at the present time is probably no better than 30%. The 2200 meter per second (m/s) flux density reported in Table 9, column 3 is If'j Table 8 HEASURED FLUX CRADIENTS IN CAPSULES LOCATED BETWEEN SHROUD AND VESSEL Flux Density Gradient Within Capsules Near-Shroud Capsule Center Capsule Near-Vessel Capsule Flux Region (4 cm) (30 cm from PV) (7 cm) Thermal (1.3 0.1)%/mm (0.7 0.1)%/nm Fas t / Int e rmediat e (1.0 2 0.1)%/mm (0.7 ! 0.1)%/nm (0.6 2 0.1)%/mm

Peak thermal flux occurs at approximately 4-cm from shroud.

28

l(

~ 3 :- 3 NEDO-24793 g :b - Di WO Table 9 ,;y },@' DETERMINED THERMAL FLUX DENSITIES 1.i fy g' Thermal Flux Density ~ Conventional 6 t,, - 2200 m/sa <0.5 eV iih* (Cd-Ratio (Cd-Ratio Location Reaction Method) Method)b Determined i hk 4 cm from Shroud Fe-58(n,y)Fe-59 2.0x10 3.1x10 11 11 hh 7],N1dl*"* Sc-45 (n,y) Sc-46 2.1x10" P 4 3.3x10 1 (G3 No. 13,14) U-235(n f)Cs-137 1.9x10 2.9x10" ~ 2.0x1011 (Av) 3.1x1011 (Av) 3.0x10 11 ?- 4 cm from Shroud Co-59(n,y)Co-60 3.8x10 5.9x10 11 11 'y leMidplane Fe-58(n,y)Fe-59 3.6x1011 11 5.6x10 1 1 (G4 No. 15,16) Sc-45 (n,y) Sc-46 4.5x10 7.0x10 11 11 U-235 (n, f) Cs-137 3.9x10 6.0x10 3.9x10" (Av) 6.1x10l1 (Av) 5.8x1011 4 cm from Shroud Co-59(n,y)Co-60 1.4x10 2.2x10 11 11

- 1. Core Top 1

1 Fe-58(n,y)Fe-59 1.2x10 1.9x10 11 11 C No. 17,18) Sc-45(n,y)Sc-46 1.2x10 1.9x10 11 U-235(n.f)Cs-137 1.1x10 1.8x10" 1.2x1011 (Av) 1.9x10 (Av) 1.8x10 1 30 cm from PV Co-59 (n,y) Co-60 1.7x10 2.6x10 10 10 !], Midplane Fe-58(n,y)Fe-59 1.6x10 2.5x10 10 10 10 10 (G3 No. 27,28) U-235 (n. f)Cs-137 1.tx10 3.0x10 1.7x1010 (Av) 2.7x10 0 (Av) 2.3x1010 7 cm from PV Co-59 (n,y )Co-60 1.5x10 2.3x10 9 9 !*3, Midplane U-235(n.f)Cs-137 1.4x10 2.1x10 I 9 9 ( C3 Ho. 21,22) 1.4x10' (Av) 2.2x109 (Av) 2.0x10 7 cm from PV Co-59(n,y)Co-60 2.7x10 4.2x10 9 9 9 9 5 U-235 (n, f)Cs-137 2.7x10 4.2x10 (G4, No. 23,24) 2.7x10' (Av) 4.2x109 (Av) 3.9x109 a. $0" ~ 0 b. $<0.5 eV = [(RR)(1 - 1/Cd)]/[o (4T/nTf! ] = (RR)(1 - 1/Cd)(1.56)/ 0 0 (from Reference 4) l 29 l

i NEDO-24793 <r defined as that conventional flux density, 4, calculated using a measured 0 cadmium-ratio correction to the BNL-325 2200 m/s crons section, o.3 Dosimeters o utilized for calculating thernal flux density by the cadmium-ratio method were Co, Fe, Sc, and U-235. If a thermal flux density within a power reactor is desired, caution must be taken when using a cobalt (or other) sole flux monitor because of the uncertainty of an effective cross section, s. Calculated >0.1 HeV and >l HeV fast cross sections for the six spectra deter-mined by this spectrum unfolding method for the Fe-54(n.p)Mn-54, Ni-58(n.p) Co-58, Cu-63(n,a)Co-60, Np-237(n,f)Cs-137, and U-238(n,f)Cs-137 reactions are given in Table 10. A cross section is calculated by dividing a measured reaction rate (see Appendix A) by the appropriate integral fast flux density ,,..I ' given in Table 6. 24 27 .F This quotient multiplied by 1x10 or lx10 gives the ' j cross section in barns (b) or millibarns (mb), respectively. Yariations in cross sections with shroud-to-vessel locations are large for three common dosimeters, Fe, Ni, and Cu. For example, as shown in Table 10, for the 3' 45' core midplane C3 assembly, the >l HeV a) Fe(n p) cross section increases from 140 mb at the near-shroud location to 210 mb at the near-vessel location, b) Ni(n.p) cross section increases from 190 to 280 mb, c) Cu(n.a) cross section increases from 1.8 to 3.7 mb. The U-238(n,f) and Np-237(n,f) cross sections, however, remain fairly constant for near-shroud and near-vessel determinations: 0.54 0.05 b for U-238 and 4 2.05 1 0.20 b for Np-237(>1 MeV). l Because BWR's currently irradiate only Fe and Cu dosimeters, the far-right column of Table 10 gives measured Fe-54(n,p)/Cu-63(n,a) reaction rate ratios as a tool for dosimetrists to estimate a fast cross section for a measured i Fe-54 (n.p)Mn-54 reaction. ) 30

^; ._ M NEDO-24793 i. 4' i[n

.p" Table 10 h;h d (;

-DETERMINED FAST CROSS SECTIONS y% Fast Cross Section l' hh (barns) Ratio: kd{ ' Location Reaction >0.1 MeV >l MeV Fe RR/Cu RR ,e {,tj' 4.cm from Shroud Fe-54(n p)Mn-54 0.070 0.14 82.4 Core Midplane 3'45' Ni-58(n.p)Co-58 0.095 0.19 .;y;p Cu-63(n.a)Co-60 1 0.0018 [I Np-237(n f)Cs-137 1.05 2.1 !n -- U-238 (n, f)Cs-137 0.26 ; 0.50 l[. l, 4 cm from Shroud 'Fe-54(n,p)Mn-54 0.061 0.12 86.7 o Midplane 45'

.y.gp Ni-58(n,p)Co-58..

0.082 0.16 .Cu-63(n.a)Co-60 0.0014 Np-237(n f)Cs-137 1.14 2.25 U-238(n,f)Cs-137 0.28 0.51 4 cm from Shroud Fe-54(n.p)Mn-54 0.070 0.13 83.8 Core Top 3'45' N1-58(n.p)Co-58 0.091 0.17 r~ Cu-63(n.a)Co-60 0.0016 L Np-237 (n, f)Cs-137 1.13 2.1 U-238(n,f)Cs-137 0.28 0.52

[

30 cm from PV Fe-54(n.p)Mn-54 0.075 0.15 59.4 re Midplane 3'45 Ni-58 (n.p) Co-58 0.10 0.20 Cu-63(n a)Co-60 0.0024 U-238 (n, f) Cs-137 0.23 O.43 1 i' f' 7 cm from PV Fe-54 (n.p)Mn-54 0.12 0.21 55.7 h, Midplane 3'45' Ni-58 (n,p) Co-58 0.16 0.28 Cu-63 (n.a) Co-60 0.0037 Np-237(n,f)Cs-137 1.05 1.8 U-238 (n. f) Cs-137 0.34 0.59 7 cm from PV Fe-54(n,p)Mn-54 0.12 0.21 54.6 Midplane 45' N1-58(n.p)Co-58 0.15 0.26 Cu-63 (n.a) Co-60 0.0038 Np-237(n,f)Cs-137 1.1 1.9 U-238(n,f)Cs-137 0.31 0.54 31/32

g m ih,(, NEDO-24793 ~ mi4 5. DISCUSSION Q t y{ 't -n" h ~ An important phenomenon has been discovered and another experimentally tested t ,i which should cause concern to the experimentalist who strives for high-n, f' accuracy flux / spectra determinations in light water reactors (and specifically-j [ in BWR's). i One severe problem appears to be the inability to use the normally provided 1 M reactor power-time history for accurate spectrum and' flux density determina- 't' tions at all in-vessel locations. Using the reactor power histograms in the I. calculation of dosimeter reaction rates for this in-vessel experiment, anoma-lies in the measurements of reaction rates for the fission dosimeters U-235, e U-238, Np-237, and Th-232 appeared. '} If activity measurements are accurately performed and decay corrections 7 adequately made, if accurate fission yields are used, and if a known irradia-tion history is incorporated, the measured reaction rates for a fission ] dosimeter should be independent of the fission product analyzed. Thus, when ~ measuring reaction rates from two fission products for a given fission dosimeter, consistent measurements should yield a reaction rate ratio of one, within experimentai' error. 1 Up to three fission product nuclides (30-y) Cs-137, (64-d) Zr-95, and (39-d) + Ru-103 were measured in the U-235, U-238, Np-237, and Th-232 dosimeters which were contained in up to six in-vessel flux capsules at near-shroud, near-center, and near-vessel locations at azimuthal angles 3'45' and 45*. Reaction rate ratios designated Cs/Zr and Cs/Ru were calculated and appear in Table 11 in six colue:ns under the heading "Reactor Power." For the G3 assembly, the ratios are close to unity (1.027 av.). For the G4 and G1 assemblies, however, the ratios are approximately 1.16 and 0.86, respectively. Since the experi-mental uncertainties in the activity measurements / fission yields are 17%, the ratios for C4 and G1 are not attributed to experimental error. 33

l' i _ 'EM7, ,idf b ff. Table 11' FISSION DOSIMETER REACTION RATE RATIO COMPARISONS:. REACTOR POWER VERSUS LOCALIZED POWER - 3F3 CYCLE 1-Reaction Rate Ratios C3 Assembly G4 Assembly Cl Assembly-Core Hidplane 3*45' Core Midplane 45' Near Core Top 3'45' Fission eac r wer ca ze per eactor Pwer kcanzed Pwer Reactor P wer Loca hzed Pwer Dosimeter and Location Cs/Zr Cs/Ru Cs/Zr Cs/Ru Cs/Zr Cs/Ru Cs/Zr Cs/Ru Cs/Zr Cs/Ru Cs/Zr Cs/Ru US (S) (B) 1.05 1.03 1.03 1.02 1.24 1.19 1.13 -1.03 0.87 0.89 1.09 1.12-(S) 1.04 1.03 1.02 1.03 1.16 1.19 1.06 1.08 0.84 0.89 1.06 1.09 (C) (B) 1.02 0.97 1.00 0.97 (C) 1.03 1.03 1.02 1.02 z. (P) (B) 1.02 1.00 1.00 1.00 1.17 1.12 1.07 1.02 E ?- (P) 1.00 1.05 0.99 1.04 1.15 1.18 1.05 1.07 w G-U8 (S) 1.03 1.00 1.01 0.99 1.17 1.15 1.07 1.05 0.86 0.86 1.08 1.09 (C) 0.98 1.01 0.97 1.00 (P) 1.00 1.09 0.99 1.08 1.11 1.16 1.01 1.05 7 Np (S) 1.07 1.03 1.05 1.02 1.19 1.16 1.08 1.05 0.90 0.88 1.12 1.10 (P) 1.06 1.09 1.04 1.09 1.14 1.20 1.04 1.08 ~ ~ Th (S) 0.99 0.97 1.12 1.02 0.82 1.03 (C) 1.03 1.01 Key: US = U-235 (S) = Shroud - 4 cm U8 = U-238 (C) = center (pressure vessel - 30 cm) Np = Np-237 (P) = Pressure vessel - 7 cm Th = Th-232 (B) = Bare Cs - Cs-137 Others are Cd covered Zr = Zr-95 as - Ru-103

.y,, y T. NEDO-24793

t}i OQ. ~.

jh This trend'was investigated by Science Applications,.Inc. (SAI).I The local c 'I a function of time was generated by analyzing relative power data power as ,g supplied by Browns Ferry. The method of investigation, and results and (b conclusions are provided in a letter received from SAI which appears as Appendix B in this report. .p }; Using renormalized localized power history calculational predictions provided by SAI, fission dosimeter reaction rates were recalculated and appear in Table 11 in the columns under the heading "Localized Pover." Considerable improvement should be apparent. The new average Cs/Zr and Cs/Ru ratios for the G3, G4, and G1 assemblies are 1.015, 1.059, and 1.086, respectively. 4 Restating two important conclusions of SA1:

a. "errors of up to 120% may be introduced by the use of reactor power history *

.ead of local power history for the calculation of flux from activity measurements;"

b. "specimens should be located in positions where flux gradients and sensitivity to small changes in control rod configurations are minimized."

A rf) j Table 12 gives a. the BF3 Cycle 1 power history as evaluated from provided daily histograms, and b. the renormalized localized power history calculational predictions of SAI. Full power is 1098 MWe, The comparisons of reactor power versus localized power for the three assembly locations are shown graphically in Figure 10. Table 13 lists the adjustments (multiplicative factors) to all measured reac-tion rates which were made as a result of utilization of localized power his-h tories. For the 3* 45' midplane G3 assembly, adjustments of only 2% or less were made; for the 45' midplane G4 assembly, up to 9 to 10% (x1.1) adjustments were needed for the shorter-lived (40-85d) nuclides; for the 3' 45' top G1 ~ assembly, multiplicative factors of approximately 0.80 were made to the reac-tor power history-calculated reaction rates of the short-lived nuclides mentioned. The net result of these adjustments on integral flux density and fluence results will be discussed later in this section. 35

.o . ; rr G. 1 h NEDO-24793

ij t'g

};q(p. g ' Table 12 jl[( BF3 POWER HISTORY (AUGUST A9',1976 - SEPIDiBER 9,1978) t q l (M:j M.! Evaluated from BF3 Reactor Power Daily Histograms a. . 9 -[ j;;!-^ W. ' Date Irradiation Days Days Between Full-Power-Irradiation Cycle Irradiated DCycles Fractiona F.@ [' 8/19/76 - 4/25/77 1 249.5 0.645 'L 5.5 3: ' ll 5/1/17 - 7/7/77 2 67 0.738 9 7/16/77 - 3/16/78 3 243 0.824 6 3/22/78 - 4/16/78 4 25 0.789 6 4/22/78 - 7/12/78 5 81 0.867 3 7/15/78 - 8/18/78 6 34 0.765 7 I,8/25/78-9/9/78 7 15 0.682 b. Localized Power. History calculational Predictions of Science Applications, Inc. I Full-Power Fraction" Days Period G3 G4 G1 107 1 0.402 0.438 0.211 'i 59 2 0.868 0.922 0.417 ' I 49 3 0.847 0.879 0.600 50 4 0.852 0.879 0.534 63 5 0.756 0.7'18 0.489 H, 54 6 0.724 0.766 0.820 49 7 0.724 0.766 0.804 31 8 1.125 1.120 1.436 67 9 0.600 0.616 1.056 51 10 0.702 0.675 0.922 52 11 0.793 0.702 0.938 56 12 0.799 0.799 1.029 62 13 0.627 0.547 0.793 a. Full power is 1098 MWe, 36

g. iss- . i - -{ xq.3 f:_giQ=fy{'f ~~',$@:j:4,liIl.(('[;he,dh;;.. h.y Osm 1-Table 13 ADJUSTMENTS

TO REACTION RATES FOR ALL DOSIMETER RADIONUCLIDES USING LOCALIZED POWER HISTORIES - BF3 CYCLE 1 U-235 U-238, Dosimeter Np-237, Th-232 Co Cu.

Fe Ag Sc Ti-Ni Fe Radionuclide Cs-137 Zr-95 Ru-103 Co-60 Co-60 Mn-54 Ag-110m Sc-46 Sc-46 Co-58 Fe-59 Assembly Location C3 3*45' Midplane 1.00 1.017 1.006 1.001 1.001 1.015 1.009 1.020 1.020 1.018 1.010 G4 45* Midplane 1.00 1.096 1.104 1.001 1.001 1.034 1.034 1.087 1.087 1.093 1.103' G1 3*45' Top 0.998 0.799 0.792 0.980 0.980 0.902 0.877 0.805_0.805 0.801 0.794 x a. Multiplicative factor to reactor power history-calculated reaction rates ?.' w ~ n bwW 4** e I 4 a

a s

+-er e

i 1 NEDO-24793 .w g, 'd ~ ~ - ~ G3 locallaed Power ~ ,y 1.00 7 --- Reactor Power ~ .c 1"~~t 3- .._ 4.=.- -j O.75 1-= I***:*l W 'A >t ~-- -^ 3=f'd ; ..1 '*l 4 g [' I..~*--4 j~' l ~ o, 0.50 Y~ It} 1. ~ ._ O g. 1 -. i I 1 ' h1M - ?j} }siM i A

1f,

-:?i ^ni-l w 0 25 E - mlhW{ 9% 'I -.i G O ll 'I tu o i 1.00 G4 Localized Power" I --- Reactor Power ~ t1:___ g- -. _ -mm 8 0.75 2 '-PI , L.-4 .a tj _ i .- M.. +- U . g,jf. -J. ;; E es ..lf- ' J43 t A g 0.50 -il l! i 1-4

la g

t i -_. d 1- . f! g .;-4 w !- .il.. I g5 ,5 - y U l- ' @1. -)' 'i ' I g, 0,25 .I: 1 ~ 8);!..< l - b-- [-- ~ ~ m}t

i

+1 ,j i ; _.. o

ii E

~,i ~ li " t--P} ~ ~ G1 Incalized Power' --- Reactor Power 1.00 )- r --- -}f{*& '-~~ m-- g O.75 ,=y..

l u

= m '~' A i i_ ; ? iif*,e _L A 1 -. t 'l f L'pd, - ~ ?! I '6 r - Q' 0.50 f .j ' !' - IM ,3., 11- 't : 0 _:1 I[.~. 4 0.25 - g<. {' ,j .j ~ j j).< g ~, PL -il' ;g - L u 0 100 200 300 400 500 600 700 \\ f 8/19/76 750 Days (Dr3 Cycle 1) 9/9/78 Figure 10. Power llistory Comparisons: Localized Power Versus Reactor Pove:r - BF3 Cycle 1 38

W j@ NEDO-24793 b !p& h '6i The phenomenon of photo contamination with the application of pressure vessel .; 4 Eg dosimetry.is relatively recent. Currently, the prac'tical dosimeters which y gn' activate in the important 1 HeV to 3 HeV neutron energy region are the fission -r W ng dosimeters Np-237, U-238, and Th-232. These had been considered quite ideal y! since they could be used for a large range of irradiation durations (seconds ,4 to years). Any one of several measurable fission products induced with half j ^~ lives up to 30 years could be utilized. However, recent calculations 5 and i measurements / calculations performed 6 have shown that photofission (y,f) reac-tions can significantly compete with neutron-fission (n.f) reactions, result-ing in erroneous reaction rate data. Vnen such fission dosimeters are applied g without any adjustment for the photofission effect, an overprediction of the neutron flux density can result at locations where the relative gasua flux is high (i.e., as the distance f rom the reactor core is increased and the ratio of high-energy gamma rays to fast neutrons increases, the gamma-ray effects begin to become important). Although photocontamination at any particular point is unique for each reactor, the vicinity of the pressure vessel experi-ences the highest ratio of gamma rays to neutrons. This effect at prer,sure ( vessels is more pronounced in BWR's than in PWR's. Appendix C gives photofraction data from SA1 for U-238, Np-237, and Th-232 at all capsule locations.2 These are one-dimensional calculations using the ~ ANISN transport code and incorporating the effects of void fractions, loca-lized power distributions, and equivalent core radii. The data are expressed as percent photofraction (pf) for each reaction, i.e.,1-% pf/100 is the cor-rection factor to be applied to a measured reaction rate. The photofraction corrections are shown to be small (1 to 4-1/2%) at the near-shroud locations of capsules 17, 13, and 15, moderate (5 to 20%) at the center locations of capsules 30, 27, and 29, and large (20 to 54%) at the near-vessel locations of capsules 25, 21, and 23. For example, the photofractions for the U-238 (n f) reaction at the near-vessel capsules are about 30-40%, while for Th-232(n.f) approximately one-half of the fission product activity is due to photofission. These near-vessel numbers are reported with an uncertainty of 125%. The photofission effect used for all fast reactions for this BF3 experiment a"e given (as multiplicative factors) in Table 14, column 6. 39

u.

. ;. w. a _

x. --

a g, _ =

-u-

^ m. - f X,e,:~r.

ni.m7.3, - : m.amunre.yz =lG=..c5:

+ ~ .. 7 ege f, ~. w p.y ..,q .g g s,,. ,. + + - v~ w +" Table 14 EFFECTS OF PHOT 0 FISSION, ASSDfBLY PERTURBATION, AND LOCALIZED POWER HISTORY - ON MEASURED FAST NEUTRON REACTION RATES - BF3 CYCLE 1. a. C3 Assembly 3* 45' N!dplane Multiplicative Factor Loca11 red Assembly / No Photofission Perturbetton Power Mistory Combleed Dostmeter Radionwc11de _ Capsule Location Effects

Effect Effect ~

Effect Effectsb

p Cs-137 C3 No. 13 Shroud-4 cm 1.0 0.990 1.064.

1.M 1.053 Co-137 C3 No. 21 FV-7.ca 1.0 0.794 1.010 1.09 0.002 Zr-95 C3 No. 13 Shrood-4 cm 1.0 0.990 '1.064 1.017 1.071 Zr-35 C3 Mo. 21 FV-7 cm 1.0 0.794 1.010 1.017 0.816 11-238 Cs-137 C3 No. 13 Shroud-4 ca 1.0 0.982 1.053 .1.00 1.034 Co-137 C3 No, 27 FV-30 cm 1.0 0.926 1.031 1.00 ' O.953 Co-137 C3 No. 21 FV-? cm 1.0 0.718 1.042 1.00 0.748 Zr-95 C3 No. 13 Shroud-4 cm 1.0 0.982 1.053 1.017 1.05 Zr-95 C3 No. 27 FV-30 ca 1.0 0.924 1.031 1.017 0.96 h 3 Zr-95 C3 No. 21 FV-7 cm 1.0 0.718 1.042 1.017 0.761 y o M Th Cs-137 C3 No.13 Shrowd-4 cm 1.0 0.963 1.075 1.00 1.035 Cs-137 C3 No. 27 FY-30 ca 1.0 0.858 1.124 1.00 0.964 ww Zr-95 C3 No. 13 Shrewd-4 cm 1.0 0.963 1.075 1.017 1.053 2r-95 C3 No. 27 FT-30 ca 1.0 0.858 1.126 1.017 0.981 Ns co-58 C3 No. 13 shread-4 ca 1.0 1.075 1.018 1.094 Co-58 C3 No. 27 FY-30 on 1.0 1.064 1.018 1.083 Co-58 .C3 No. 21 FV-7 cm 1.0 1.00 1.018 1.018 m, re Ma-54 C3 No. 13.14 shroud-4 cm 1.0 1.075 1.015 1.090 Ma-54 C3 No. 27.28 FY-30 ca 1.0 1.064 1.015 1.080 Mn-54 C3 No. 21.22 FV-7 cm

1. 0 1.00 1.015 1.015 71 Sc-46 C3 No. 16 shrood-4 cm 1.0 1.075 1.020 1.096 Sc-46 C3 No. 28 FV-30 cm 1.0 1.064 1.020 1.083 -

Ce Co-60 C3 No.13.14 Shroud-4 cm 1.0 1.075 1.001 1.076 Co-60 C3 No. 27.28 - FV-30 cm 1.0 1.064 1.001 1.065 Co-60 C3 No. 23.22 FV-7 cm 1.0 1.00 1.001 1.001 - Normalized to unity m. b. Multiplicattwe factor to tre ap;plisd to measured reacties rates obtalmed from reector power history. ,.u-a ,y-w yy .-g. +,

-3 g ,g 7 w. ,a y.. ~ f.. +..;;_y,j g g (q, i 3~= Table 14 EFFECTS OF PHOT 0 FISSION, ASSEMBLY PERTURBATION, AND LOCALIZEI) POWER HISTORY ON MEASURED FAST NEUTRON REACTION RATES - BF3 CYCLE 1 (Continued) b. C4 Assembly 45' Midplane Multiplicative Factor Localtred Assembly / No ' Photofission Pertur'stion Fouer History - Combined Dostmeter Radionuclide Capsule -Location Effects

Effect Effect Effect Effectob Np Cs-137 C4 No. 15 shroud-4 cm 1.0 0.989 1.064

.1.00 1.052 Cs-137 C4 No. 23 FV-7 cm 1.0 0.722 1.010 1.00 - 0.729 - 2r-95 C4 No. 15 shroud-4 cm 1.0 0.989 1.064.- 1.096 1.153 Zr-95 C4 No. 23 FT-7 cm 1.0 0.722 1.010 1.096~ 0.799 V-238 Cs-137 C4 No. 15 Shroud-4 cm 1.0 0.977 1.053 1.00 1.029 Cs-137 C4 No. 23 Fv-7 cm 1.0 0.623 1.042 1.00 0.649 Zr-95 C4 No. 15 shroud-4 cm 1.0 0.977 1.053 - 1.096' 1.128 Zr-95 C4 No. 23 FT-7 cm 1.0 0.623 1.042 1.096 0.711 Th Co-137 C4 No. 15 Shroud-4 cm 1.0 0.955 1.075 1.00 1.027 ma. s Zr-95 C4 No. 15 Shroud-4 cm 1.0 0.955 1.075 1.096 1.125 U -w N1 Co-58 C4 No. 15 Shroud-4 cm 1.0 1.075 1.093 1.175 Co-58 C4 No. 23 FT-7 cm 1.0 1.00 1.093 1.093 Fe Pte-54 C4 No. 15.16 Shroud-4 cm 1.0 1.075 1.034 1.112. Mn-54 C4 No. 29 Fv-30 cm 1.0 1.064 1.034 1.00 1.034 -1.100 Mn-54 C4 No. 23.24 FT-7 cm 1.0 1.0 34 ft Sc-46 C4 No. 16 Shroud-4 cm 1.0 1.075 1.087 '1.169 Cu Co-60 C4 No. 15.16 shroud-4 cm 1.0 1.075 1.001 1.076 Co-60 C4 No. 29 FV-30 cm 1.0 1.064 1.001' 1.065 Co-60 C4 No. 23.24 Fv-7 cm 1.0 1.00 1.001 1.001 a. Normeltred to eatty Multipitcative facter to be applied to measured reaction rates obtained from reactor power history. b. m [ ,#m em & M ""

C .I " 'Bf ;._ g jgpy .yif t:7, S;p_z.7.2 _..7. O M: x.-. -- [w ..i. ..-,.m_.... e-y m.. ..m g n Table 14 EFFECTS OF PHOT 0 FISSION, ASSDiBLY PERTURBATION, AND LOCALIZED POWER-HISTORY ON MEASURED FAST NEUTRON REACTION RATES - BF3 CYCLE 1 (Continued) c. C1 Assembly 3' 45' Top Multiplicative Factor Localised Assemblyf No Photofission ' Perturbation Power History Combined Desteeter Radionuclide capsule location Effects

Effect Effect Effect Effectsb Np Cs-131 C1 No. 17 Shroud-4 cm

1. 0 -

0.990 1.064 - 1.00 1.053 2r-95 C1 No. 17 Shroud-4 cm 1.0 0.990 1.064 0.799 .0.842 U-238 Co-137 C1 No. 17 $hroud-4 cm 1.0 0.982 1.055 1.00 1.054 2r-95 C1 No. 17 Shroud-4 cm 1.0 0.982 1.053 0.799 0.826 4 Th Cs-137 C1 No. 17 Shroud-4 cm 1.0 0.963 1.075 .1.00 1.035 2r-95 C1 No. 17 shroud-4 cm 1.0 0.963 1.075 0.799 0.827 Ni Ce-58 C1 No. 17 Shroud-4 cm 1.0 1.075 0.801 0.861 g Fe Mn-54 C1 No. 17,18 Shroud-4 cm 1.0 1.075 0.902 0.970 4e Mn-54 C1 No. 30 Pv-30 cm 1.0 1.064 0.902 0.960 Mn-54 C1 No. 25 PV-7 cm 1.0 1.00 0.902 .0.901 T1 Sc-46 C1 No. 18 Shroud-4 cm 1.0 1.075 - 0.805 0.865 Cu Co-60 C1 No. 17.18 Shroud-4 cm 1.0 1.075 0.980 . 1.054 Co-40 C1 No. 30 PV-30 am 1.0 1.064.. ,c. 0.900 1.043 - Co-40 C1 Mo. 25 Fv-7 cm 1.0 1.00 " ~ ~ ' O.980 'O.980 a. Normalized to unity b. Multiplicattee fatter to be opplied to measured reaction rates obtained from reactor power history. { l I

J NEDO-24793 siij!'

j;;; Table 15 [ .g SIGNIFICANT ADJUSTMENTS MADE TO MEASURED FAST-NEUTRON REACTION RATES .;N k Combined-Effects [;p-Reaction Assembly Location Factora L .U-238(n.f)Cs-137 G3,G4,G1 PV-7 ca 0.65 to 0.80 77 Np-237(n.f)Zr-95 f) -U-238 Np-237 (n,f)Zr-95 G4 Shroud-4 cm 1.125 to 1.15 Th-232 G1 Shroud-4 cm 0.83 g Licf Ni-58(n.p)Co-58 G3 Shroud-4 cm, Center 1.09 ,;,l -. G4 Shroud-4 cm 1.175 G4 PV-7 ca 1.09 G1 Shroud-4 ca 0.86 Ti-46(n.p)Sc-46 G3 Shroud-4 cm, Center 1.09 G4 Shroud-4 ca 1.17 G1 Shroud-4 ca 0.86 Fe-54(n.p)Mn-54 G3,G4 Shroud-4 cm, Center 1.08 to 1.11 G1 PV-7 ca 0.90 Cu-63 (n.a)Co-60 G3,G4 Shroud-4 cm, Center 1.07 4 1 4 43

. p 'l 2 i. 7 ' '1C * *; k 7 { 3 ,p @l The presence of the 1/2-in.-diameter capsule assembly tubes which consisted p of 200-mil wall carbon steel and solid aluminum spacers led to a calculational !.:(, study of the possibility of spectral shape changes and perturbation effects .x y, affecting this spectrum-unfolding experiment in the fast neutron energy g,,o region. This work was also undertaken by SA1.2 y j Appendix C contains the reported perturbation data expressed as correction factors to be applied to (divided into) measured reaction rates for fast-neutron-induced reactions. These effects are shown to be relatively small. For the shroud capsules (267 cm radius), center capsules (289 cm), and near-i..g vessel capsules (313 cm), the correction f actors to be applied to U-238., fI Np-237, Th-232, and >l HeV activant (Ni, Fe, Ti, Cu) reaction rates average approximately 0.94, 0.94, and 0.98, respectively. For example, the measured reaction rate of Fe-54(n.p) when divided by 0.93 gives the effective reaction rate of this dosimeter reaction had it been situated, by itself, 4 cm from I.13%. the shroud. The error assigned to these Appendix C with assembly /no assembly numbers is 4 The perturbation effects used for all fast reactions are given (as multiplica-tive factors) in Table 14, column 7. The effects of photofission, assembly perturbation, and localized power his-tory, individually and combined, for all fast dosimeter reactions used in j this experiment are given in Table 14. These effects are expressed as multi- ) plicative factors which were applied to the measured reaction rates. It is apparent that, in many cases, significant adjustments to the measured reaction rates had to be made. These are summarized in Table 15. All other corrective factors for fast reactions are within 17%. The effects on determined integral fast flux densities as a function of photofission, perturbation, and localized power history effects for six in-vessel locations using both Cs-137 and Ir-95 are shown in Table 16. The >0.1 HeV, >l MeV, and >3 HeV integral numbers are the output numbers resulting from individual computer runs (expressed as three significant figures for comparisons). 44

w. - x -. -? l .,h2=' i' .} ~ N;c Q_ hQ o Table 16~ INTECRAL FAST FLUX DENSITY RESULTSa AS A FUNCTION OF PHOr0 FISSION, ASSEMBLY PERTURBATION, AND LOCALIZED POWLR HISTORY EFFECTS - BF3 CYCLE 1 b a. For Fission Product Cs-137 Integral Fast Flux Density Effects Incorporated Into Results All" Assembly Photofission (Photofission Assembly and Energy and Assembly Parturbation, and _ Capsule Location _(MeV) None Photofission Perturbation Localized Power History) 1 11 11 11 C3 #13,14 4 cm from Shroud, >0.1 1.40x10 1.40x10 1.40x10 1.48x10 10 10 10 10 3* 45' Midplane >l 7.14x10 7.09x10 7.39x10 7.56x10 10 10 10 10 >3 2.73x10 2.68x10 2.92x10 2.93x10 { l 1 l' 3.09x10 f 2.95x1C G4 #15,16 4 cm from Shroud, > 0.1 3.10x10 2.68x10 1 45' Midplane >l 1.44x10 1.38x10 1.47x10 1.35x10 .w 10 10 10 10 >3 3.80x10 3.84x10 4.03x10 4.76x10 g 10 10 10 10 0 C1 #17,18 4 cm from Shroud >0.1 9.31x10 8.89x10 9.80x10 9.38x10 0 10 10 10 3* 45' Top >l 4.86x10 4.75x10 5.12x10 5.00x10 10 10 10 10 >3 1.97x10 1.98x10 2.04x10 1.92x10 10 10 10 10 G3 #27,28 30 cm from PV >0.1 1.14x10 1.0$x10 1.09x10 1.11x10 9 9 9 9 3* 45' Midplane >l 6.03x10 5.45x10 5.70x10 5.84x10 >3 2.06x10' 2.01x10 2.16x10 2.16x10 9 9 9 9 0 8 8 C3 #21,22 7 cm from PV >0.1 1.36x10 8.89x10 8.92x10 9.03x10 8 8 8 8 3* 45' Midplane >l 6.79x10 5.15x10 5.18x10 5.21x10 8 0 0 >3 2.57x10 2.74x10 2.76x10 2.77x10 2.36x10'9 C4 #23,24 7 cm from PV >0.1 9 1.39x10 1.49x10' 1.42x10' 0 8 45* Midplane >l 1.10x10 8.07x10 8.12x10 8.06x10 8 8 0 8 >3 3.99x10 3.83x10 3.79x10 4.11x10 'Using an engineering computer program; third significant figures for comparisons only U-235,U-238, Np-237 (except C3 #27,28), Th-232 (except G3 #21.22 and C4 #23,24) "This report (Table 6) ) e n--3 g ,,-ii-7- y-w

L.. 4 i, ' MM.. }-' U2h,) .i%,7;"\\ h: Table 16 (Continued) D b. For Fission Product Zr-95 -Integral Fast Flux Density Effects Incorpqtated Into Results All" Assembly Photofission (Photofission, Assembly and Energy and Assembly Perturbation, and Capsule Imcation (MeV) None. Photofission Perturbation I.ocalized Power History) 11 11 11 C3 #13,14 4 cm fron Shroud, >0.1 1.40x10 1.32x10 1.41x10 1.44x10 10 10 10 10 3* 45' Midplane >l 7.04x10 6.90x10 7.30x10 7.46x10 10 10 10 10 >3 2.67x10 2.69x10 2.87x10 2.92x10 11 11 11 11 C4 415,16 4 cm fron Shroud, >0.1 2.50x10 2.40x10 2.26x10 2.54x10 11 11 11 45* Midplane >l 1.22x10 1.17x10 1.19x10 1.28x10 10 10 10 10 >3 3.83x10 3.87x10 4.56x10 4.66x10 g y 11 11 11 10 C1 #17,18 4 cm from Shroud, >0.1 1.06x10 1.08x10 1.08x10 9.04x10 10 10 10 10 3* 45' top >l 5.55x10 5.53x10 5.73x10 4.68x10 g 10 10 10 10 >3 2.05x10 2.00x10 2.19x10 1.78x10 g 10 10 10 10 C3 #27,28 30 cm from PV, >0.1 1.17x10 1.02x10 1.11x10 1.13x10 9 9 9 9 3* 45' Midplane >l 5.94x10 5.35x10 5.86x10 5.97x10 9 9 9 9 >3 2.07x10 2.02x10 2.14x10 2.17x10 9 8 8 8 C3 #21.22 7 cm from PV, >0.1 1.33x10 8.62x10 8.78x10 8.91x10 9 8 8 8-3* 45' Midplane >l 6.84x10 5.02x10 5.07x10 5.15x10 8 8 8 8 >3 2.68x10 2.70x10 2.70x10 2.74x10 9 8 9 C4 #23,24 7 cm from PV, >0.1 2.26x10 1.18x10 1.32x10 1.46x10 8 8 8 45' Midplane >1 9.63x10 >3 3.91x10 3.'86x108 8 8 8 3.65x10 .3.94x10

ns JI NEDO-24793 M'- a iV jjf[ From Table 16 it is apparent that for fast flux density determinations using ,on 01? Sid the spectra unfolding method with the dosimeter sets used in this experiment: aw ,.d-?. T!p a. no significant differences occurred whether Cs-137 or Zr-95 was thiU used as the measured fission product provided that all effects gc were incorporated into the final reaction' rate measurements. N m U[c b. for localized power history effects, no significant differences S I '. ' occurred at the near-shroud locations of the three assemblies G3, df G4, and G1 when Cs-137 was used; conversely, if Zr-95 was used,- [ G4 and G1 gave results 10% low and 20% high, respectively, without localized power history adjustments. at the G3 and G4 near-vessel locations, the use of the large photo-c. fission. corrections for U-238 and Np-237 (Th-232 was not included in the near-vessel sets) reduced the >0.1 MeV and the >l MeV (no-ef fects) integral values approximately 60% and 35%, respectively. ( \\ ,-p ). 'ee' d. Zr-95 could be effectively used as a fission product ac the G4 near-vessel location if the localized power history was utilized (10% change), t -1 ' Also calculated for the six capsules were integral thermal (<0.5 eV) and ( integral intermediate (0.5 eV-0.1 MeV) flux densities as a function of loca-lized power history / photofission effects. Again, no significant differences occurred using either Cs-137 or Zr-95 as long as localized power / photofission ,1 adjustments were made. For the G3 assembly, these adjustments had no signi-ficant effect on any integral thermal values for either Cs-137 or Zr-95. The integral intermediate values remained unchanged at the G3 near-shroud capsule but decreased by 20% at the near-vessel capsule, due to neutron spectral changes from the photofission adjustments to U-238 and Np-237. Uncertainties in the fast neutron spectra have direct ifnear correlations with uncertainties in the predicted reactor pressure vessel lifetimes. Photo-fission effects, in particular, are not only large near the reactor vessel, but the uncertainties in the photofractions (125%) are still too high for 47

e NEDO-24793

9.

I '! I. ,[ accurate integral fast' flux determinations using the multiple dosimeter pn spectrum unfolding technique. The resulting spectral shape (and subsequent

p

>l MeV values) in the 1.to 4 MeV energy region is exclusively dependent on Myj the fission dosimeters. For example, it can be shown that a further increase p; of 20-25% in the photoiractions of U-238 and Np-237 at the near-vessel loca- 'A

tions would not only cause an unrealistic minimum (valley) t', occur within h-the 1 to 4 MeV. spectral region, but the >l MeV integral va'.ue would also i.

decrease 20-25%. Future photofission measurements, especially, should be performed in light water power reactor environments to reLuce the uncertainty j3 .b.) in the neutron spectra 'ncident upon and throughouc the pressure vessel. pb w. Si Error analysis of the ir "egral flux densities and fluences of Tables 6 and 7 [ involves the incorporation of several effects An uncertainty in the number of neutrons above 1 MeV for one in-vessel location may be quite different from the uncertainty at another location. The 20 ras errors in several - energy groups have been calculated based on estimated errors in dps/g, gradients, power history, photofission, perturbation, exitapolation, and unfolding. These are given in Table 17. The energy groups are: total spec-trum, thermal, intermediate, >0.1 MeV, >l MeV, and >3 MeV. For the >l MeV energy group, for example, the estimated uncertainties range from il5-20% at the shroud and center locations to 130% at the pressure vessel locations. Table 18 gives the estimated 2a uncertainties in the >0.1 MeV and >l MeV cross sections for the five fast reactians of Table 10 {Fe-54(n.p), Ni-58 (n.p), Cu-63(n,a), Np-237(n,f), U-238(n,f)]. 0 1 48

bI /" NEDO-24793

u.

e-L ty - R 9." r i f3,

l '.

Table 17 lM .l] ESTIMATED 20 UNCERTAINTIES IN FLUX DENSITIES AND FLUENCES jj (SEE TABLES 6, 7, AND 9) k)fi i Near-Shroud Center: Near Vessel Shroud (4 cm) (30 cm from PV)' (7 cm) Vessel 1: 'J; 4 G3 Assembly - 3 45' Core Midplane a. 1 Total Spectrum i40% 140% 140% -) -(>1x10-9 MeV) Thermal i40% 135%- 130% 7 (<0.5 eV)- Inte rmedia te !65% 60% 170% 70% (0.5 ev-0.1 Mev) >0.1 HeV 25% 20% 120% 35% i35% .s. >l MeV - 120% 115% 115% 30% 130% >3 MeV d20% 115% 115% il5% 120% -s; b. G4 Assembly - 45 Core Midplane TotalSgectrum 40% 145% 40% (>1x10 _HeV) 7g-Thermal l h' 30% 40% 35% (<0,5 eV) - In te rmedia te ~ 65% 60% 170% i60% 1 (0.5 eV-0.1 MeV) >0.1 HeV i25% 25% ~25% 135% i35% >l MeV 20% i20% 120% 30% 130% o >3 MeV' 20% 15% t20% il5% 120% G1 Assembly - 3*45' Hear Core Top c. -Total Spectrum 140% 45% i40% ifi (>1x10-9 MeV) Thermal i30% 40% 30% (<0,5 eV) Intermediate 175% 70% 170% i70% (0.5 eV-0.1 MeV) >0.1 MeV !30% 125% 125% 35% 40% >l MeV-125% !20% 20% 130% 35% >3 MeV 120% 15% 20% t20% t20% 49

,m +q [, 1 ~ ,NEDO-24793-li;;., m p' :k - Table 18 ,(. ESTIMATED 20 ' UNCERTAINTIES IN FAST CROSS SECTIONS IN TABLE 10 0 db 1!m. . Location Dosimeter

>0.1 HeV

>l HeV Fe RR/Cu RR tl 1.,/ t. _G3 Shroud-4'em Fe 120% 115% 10%- Ni 20% il5% Cu 125%_ Np 120% 15% 4 q U-238 120% 115% -l G4 Shroud-4 cm' Fe i25% 20% .112% ,; lI '30% 125% Ni. i Cu 30% I' 14> 25% 25% U-238 i20% i20% ' G1 Shroud-4 cm Fe 25%' i20% 12% Ni 30% 125% Cu 30% Np i25% i20% U-238 25% 20% h G3 Center Fe 20% 15% i10% Ni 120% il5% 25% Cu U-238 125% 120% r

r G3 PV-7 cm' Fe 135%

130% i10% Ni 135% 130% 135% Cu Np 145% 140% U-238 145% 40% - G4 PV-7 cm Fe 35% i30% 110% Ni 135%- i30% E Cu 35% Np 45% i40% U-238 145% 40% 50

HEDO-24793 3 .Lt, Ik REFERENCES J!.%] 1 b y,, 1. C ~ L. Simmons and W. K. Hagan, Science Applications, Inc., Private

[.

Communication (September 1979). 1;: y 2. C. L. Simmons and W. K. Hagan, Science Applications, Inc., Private Ip' Communication (June 1980). .o L 3. "Neutron Cross Sections Volume 1 Resonance Parameters," BNL-325, Third f Edition, June 1973. y !6 4.- "Nautron Fluence Measurements," Technical Reports Series No. 107, International Atomic Energy Agency, Vienna,1970. 5. ~ C..D. Bowman, C. M. Eisenhauer, and D. M. Gilliam, "Photofission Effects

~q in Reactor Vessel Dosimetry," 2nd ASIM-Euratom Symposium on Reactor Dosimetry, Palo Alto, CA, October 1977.

6. "Measurement and Analysic of Gamma Ray Induced Contamination of Neutron Dosimetry Procedures Used for Reactor Pressure Vessel Applications," EPRI.NP-1056 Project 827-1, April 1979. O I' ~ t 51/52

O NEDO-24793 o, ehi .~ I' y< J- ' i t' e i. .y.4 APPDIDIX A P% g],V 7' DISINTEGRATION RATES AND REACTION RATES nr3 - cyctE 1 53/54

1 NEDO-24793 Q: m ! '. h

1. :

DF5/NUCLtL'S (54T!*RAf tD) s .g 1strapolated' to g,,,,g, g. (Raference I.acation) and l' In-Vessel (Dostmeter DF5/8 4 cm 30 ca 7 en Doetmeter or Capevis to Ret, at t ece free free peaction Ident. C Locatten Location) 9/9/78 Measured shrewd Shroud PV FY FY g r As-109(n.))As-llon C3 #13 C ~ arewd 0.676 9.98+s 5.90-13 f.58-13 4.88-13 8 C3 #14 3 Shrewd 0.818. 2.19+ 9 1.29-12 1.55-12 C3 827 C Center D.683

2. 6 7+ 7 1.58-14 1.39 14 i

C3 824 8 Center 0.564 1.76+8 1.04-13 8.62-14 C3 #21 C PV 0.873

2. 9 9+ 6 1.77 15 2.00-15 C3 #22 8 PV 0.754 1.0947 6.46-15 7.34-15 C4 #15 C shroud 0.541 2.04+9 1.24 12 1.56-12 1.08-12 C4 #16 8 Shroud 0.705 4.0249 2.43-12 2.76-12

~ C4 #23 C FV 0.227 5.46+6 3.31-15 3.17-15 C4 #24 5 FV 0.787 1.80+7 1.09-14 1.28-14 C1 #17 C Shrewd 0.644 7.26+8 3.23-13 4.66-13 3.20-13 5 C1 #18 8 shrewd 0.621 1.67+9 8.57-13 8.96-13 Co-59(n.v)Co-60 C3 #13 C Shroud -0.758 9.61+8 5.54-13 1.06-12 6.84-13 l-C3 f27 C Center -0.066 4.27+7 2.46-14 2.49-14 C3 #28 5 Center 0.502 1.32+9 7.62-13

6. 4 b-13 C3 #28 C FY 0.504 5.34+6 3.07-15 i:

3.30-15 C3 #22 B FY 0.631 9.24*7 5.32-14 C3 #22C3 8 FY 0.663

9. 2 5+ 7 5.32-14 C3 #22C4 5 PV 0.199 9.3067 5.36-14 5.93-14 C4 #15 C Skreed

-0.622 1.97+9 1.14-12 1.93-12 1.32-12 ,/ C4 ill 8 Shrood 0.149 2.6)*10 1.52-11 1.53-11 r$d Cf f J B Center 0.295 2.2949 1.32-12 1.46-12 f C4 723 C FY 0.477 9.4144 5.42-15 4.98-15 i C4 f24 3 FY 0.100 1.80*8 1.05-13 1.06-13 C1 #17 C Skrood -0.668 6.7248 3.79-13 C1 f!7MI C Shrood -0.233 6.99+8 3.94 13 C1 #17d2 C Shrewd -0.608 6.60+8 3.72-13 6.54-13 4.48-13 C1 #18 8 Shreve 0.683 9.40+9 $.30 12 5.62 12 Cl f)0 8 Center 0.502 1.00+9 5.65-13 4.82-13 C1 #25 & FY 0.287 5.56+7 3.14-14 3.28-14 '8 = Sareg C = Cadalas-covered. blocallred power histories utt11:ed (see Table 13) 'locerporated into entrapolated reaction rates ares 1. Phote!!aelen effects for floales acatters: pp-237. U-238. Th-232 (see Table 14) 4 2. Assembly perturb.*iees for all feat reactionst p (n.f). U-238(e.f) N!(s.p) Fe(n.p). Ti(n.p). Cv(n.e). p 111 Th(n.f) (see Table 14) 3. Localised power histories (see Table 13) 4. Westron Flus Crediente 8 8eef 9.94:10 and 5.9010'I3 -,.o SS

.[t] r ..) t mb - NEDO-24793-1 17 ' Ap_ 1 ! s ;l /: tfi.* Oa e; E!- DPS/NUCl.ElfS (SATU8Af ED) .E o ut,.,si.te r to o- ';0' oe ra-a t-a i-> - C.,s.ie th!h! .nd s' In-Vessel - (Deelmeter DPS/8 4 cm 30 em 7 ee Q, t ' Dosteeter OR Careene -to sef. et I '" ' I"" I" b neeetten Ident. ,C toestion tocacInc) 9/9/78 Measured '. 5hrevd - Shroud PV E fy JH Cu-4)(n.e)Co-60 - C3 fl3 ~C Shroud -0.606 ' l.12+5 9.96-17 r.,, W _ C3 f!3M1 C Shrs=4 -0.538.- ~1.!!+5 9.90-17 t 'C3 #13K2 C '$hroud -0.664 1.08+5 9.59-17 1.94-16 1.25-16 4. C3 fie .8 Sh resd 0.674 1.595 1.36-16 [ C3 #144 8. Shroud 0.567

1. 5 )* 5 1.36-16 C3 #148

'8

Sh r'es d

'.390 1.5t+5 1.35-16 1.96 16 1.29-16 0 C3 #27 -C Center ' -0.129 1.49+4 1.32 17 C3 #27Al C.Cente* -0.379

1. 4 )+ 6 1.27-17 f.

e C3 827A2 C Centet 0.Ill 1.64*4 1.28-17 1.41-17 h t.) c'28 8 Center 0.314 1.64+4' l.66-If C3 +28DI 4 Center 0.064 - 1.49+4 1.33-17 C3 #2802 IL : Center 0.564 1.6844 1.49-17 1.43-17

i.

C3 #21 C FV G.280 2.08+3 3.85-18 1.92-14 1.30-18 C3 #22 8 PV 0.25%

2. 05+ 3 1.02-18 1.89-18 1.28-18.

C4 81581 C Shrood -0.272

1. 8 )+ 5 1.43-16 C4 8158234 C shrewd

-0.397 1.7845 1.59-16 C4 #1581-4 C Shrevd -0.335 1.8145 1.61-16 2.74 16 1.88-16 C4 e16 - 8 shrewd 0.111 2.0)*) 1.81-16 2.76-16 1.89-14 C4 e29 8 Center 0.014 2.40+4 1.16-17 2.28-17 C4 821 C FV t'.289 3.59+3 3.33 3.04-18 1.85-10 C4 ele 3 PV 0.225 3.33+3 2.96-18 3.08-18 1.84-18 Cl *l7 C Shrewd -0.t 4 8 7.1144 6.19-17 Cl '17C1 C Shrewd -0. 7.'1 7.28*4 6.34-17 Cl elFC2 C threwd -0.606 6.99*4 6.08-17 . 36 16 7.83-17 Cl "AGF2 8 Snrewd 0.68)

9. 76+4 8.50-17 Cl *18F21 3

Shrewd 0.633

9. 5 78 a 8.33 Cl s18F22 8 shsewd 0.727 9.8944 8.61-17 Cl *187) 8 Shroud 0.667 9.41+4 8.19 17 Cl 810F4 8 Shrevd

.0.621 9.29+4 4.09-17 ,4 Cl 818F5 4 Shrewd - 0.433

9..'l

4 - 8.10 17 C1 #10F51

.8-threwd 0.378 9.12+4 7.94-17 Cl 818752 3 Shrewd 0.496

9. 3 5+4 8.14-17 1.12 16 7.71-17 Cl e rJ48 8 Center 0.533

9. 37+ 3 8.16-18 Cl *30A2 8 Center 0.220 9.02+3 7.86-18 7.96-18 Cl #25 -

8 PV n.14 e 1.12+3

9. Pi 18 1.02 18 7.16-19 F e-5. ( n.,1.**e-5 4 C3 el3GI C skrevd

-0.758 2.9 96 7.73-15 C3 833Cll C Shrewd -0.728 2.98+6 7.79-15 C1 813Cl2 C Shrewd -0.788 2.9246 7.65-15 C3 81) C Shroud -0.768

2. 9 5+ 6 7.71-15 1.60-14 1.03-14 C3 814A 8

Shrowd -0.157 3.61*6 9.46-15 C3 fl&d 8 Shrowd 0.253 6.06+6 1.06-14 C3 el4Cl 8 Shrewd 0.788 4.72*6 1.24-14 C3 e34cg3 8 Shread 0.813 4.70+6 1.23-84 C3 el4Cl2 8 Shroud 0.763

4. 5b 6

1. 2% I 6 56

..w N NED0-24793 1 4 r* 4 k lll::i w .m. 3 f DFS/NUCLit?$ ($ATUkAf tD) y[. ira ~ istre,.ieted' i. (**'""" '"' 8a) --Ce,e.ie m '. ..d i, i.-ve.sei o.oei-to opus 4-30 - F e.

-4 1 Sesetten

' Doelmeter er Cepeale to net. et free from free g Ident. C ' Loc e t ten laestion) 9/9/78 Messered $hroud Shroud PV PY PV ..i 'Pe-54(n.p)Mn-54 C3 #14D 8 ~Shrowd 0.723 4.64+6 1.22-14 (Contlawed) g( C3 #14D1 8 Shrewd 0.753 4.42+4 1.21-14 C3 #1402 3 %rewd 0.693 - 4.54+6 1.19 14. 1.62 14 1.06-14 C3 f27 C Cea.ter -0.066 2.97+5 7.76-16 8.37 16 C3 #28 8 Center 0.439 3.2945 8.62-16 8.48-16 C5 #21 C PV 0.438 3.85+4 1.01-16 1.07-16 7.26-17 C3 #22A1-8- PV -0.056 4.14+4 1.08-16 C3 #22 A2 3 PV 0.131 4.1t*4 1.08-16 C3 #22A3 5 PY -0.119 4.1 H 4 1.08-16 C3 #22A4 3 PV 0.037 4.14+4 1.08-16 C3 #22C B PV 0.693

3. 70+4 9.49-17 C3 f22C1 8 PV 0.849 3.64+4 9.52-17

.m C3 #22C2 8 PV 0.537

3. 8 3+4 1.00-16 C3 #22D 5 PV 0.443 3.91*4 1.02-16

. C3 #221 8 PV 0.491 3.72+4 9.75-17 1.08-16 7.32-17 Ce f15 C Shrevd -0.497

5. 0 H6 1.54-14 2.17-14 1.63-14 8

64 #16 B Shrewd 0.024 5.7H5 1.54-14 2.40-14 1.64-14 C4 #29A B Center -0.330 5.94+5 1.59-15 C4 #298 8 center -0.111 5.84+5 1.56-15 C4 82941 8 center -0.267 5.91+5 1.58-15 C4 fl982 5 Center -0.080

5. 7 H 5 1.54-15 C4 82953 8 Center 0.076

5. 5H 5 1.44-15 C4 #19C1 B Center 0.342 5.3t+5 1.42-15 4

C4 f29E 5 Center 0.287 5.40+5 1.44-15 1.60-15 C4 #23 C PV 0.415

6. 69+ 4 1.79-16 1.66-16 1.01 16 C4 f2451 1 PV 0.070

6. 32+ 4 1.49-16 C4 #248]A S PV

-0.283

6. 6 6+ 4 1.78-16 C4 f24818 8 PY

-0.064 6.4144 1.71-16 C4 #2451C 3 PV 0.370 6.16+ e 1.64-16 C4 #24310 3 PV 0.420

5. 46+ 4 1.56-16 1.69-16 1.03-16 C1 flhi C % rowd

-0.79) 2.20+6 5.12-15 C1 #170 C Shrewd 0.824 2.1t+6 5.10-15 Cl fl?u2 C shrowd -0.742 2.12+6 4.93-15 9.58-15 6.56 15 5 Cl fl8A B $hrewd -0.098

2. 7 3+6 6.34-15 fl #18A1 9 Shroud

-0.145

2. 6 6+ 6 6.23-15 Cl fl8A2 B Shrowd

-0.051 2.7646 6.44-15 C1 #188 5 threwd -0.284 2.55+6 5.92-15 4 C1 #1831 8 Shroud -0.339 2.50+6 5.81-15 C1 #1882 8 Shrewd -0.245 2.58+6 6.00-15 C1 #18P1 3 Shroud 0.683 3.26+6 7.57-15 Cl fl8P11 2 Shrewd 0.636 3.1 H6 7.43-15 C1 f!8F12 B Shroud 0.730 3.22+6 7.49-15 Cl 8186 8 Shrewd 0.683 3.23+6 7.51-15 1.00-14 6.87-15 Cl #30B 8 Center -0.154 2.0745 4.82-16 Cl #30C & Center -0.030 2.09+5 4.87-16 Cl #3001 8 Center 0.408 2 2HS 5.30-16 57

~.... 4 I i q' _ -NEDO-24793 a-by. r DF5/ItUCtJUS (SAft' RATED) f.' rapolated" to f"' 00'r 4 ~ t~ai-> ~ C.,.uie and l' le-Vesset.(Doetmeter DP5/g 30 cm 7 cm r lli Destmeter- ' or capsule to Raf. at face from f ree, l Reaction ^1 dent. 1 _1.acation Location) 9/9/78 Measured $hroud. Shroud FV PV FV 4 Fe-54(n.p)Mn-54 C1 #30D2 8 - Center 0.158

2. l t+ 5 5.10-16

~j (Continued) C1 f)of. 3 Center 0.564

2. 34+ 5 5.45-16 5.25-16 Cl #25C B-FV 0.631 2.11+4 4.91-17 C1 #250 t FV 0.443 2.16+4 5.03-17 C1 #25r a FV

-0.463 2.43+4 5.66-17 5.33-17 3.73-17 Fe-58(n.i)Fe-59 C3 #13G1 C Shroud -0.758

1. 8 3+ 5 8.25-15 1.59-14 1.02-14 11 C3 #14A B Shroud

-0.157 4.996 2.24-13 If C3 #149 5 Shroud 0.253 5.0t+6 2.31-13 h C3 #14C1 B Shroud 0.788 4.77+6 2.16-13 T' C3 f14C11 B Shroud 0.813 4.65+6 2.11-13' C3 #14C12 B Shroud 0.763 4.6H6 2.11 13 C3 #140 B Shroud 0.723 4.80+6 2.18-13 C3 #1401 3 Shroud 0.753 4.7046 2.13-13 C3 #14D2 3 Shroud 0.69) 4.76+6 2.16-13 2.42-13 .[ C3 827 C Center -0.066 7.95+3 3.60-16 3.65-16 C3 #28 8 Center 0.439 4.80+5 2.17-14 1.88-14 C3 f22A1 B FV -0.057 3.82+4 1.73-15 C3 722A2 8 FV 0.131 1.74-1) C3 #22A) - 5 PV -0.119 1.74-15 C3 822A4 B PV 0.037 A 2.70-15 C3 f22C 8 FV 0.493 . /u4 1.50-15 - C3 822C1 8 FY 0.849 3.31+4 1.50-15 I C3 f22C2 B FV 0.537 3.45+4 1.56-15 I'3 C3 822D B FV 0.443 3.50+4 1.59-15 C3 8221 B FV - 0.691 3.42+4 1.55-15 1.73-15 l-C4 #15 C Shroud -0.497 3.33+5 1.65-14 2.71-14 1.86-14 C4 f16 8 Shroud a'4 8.82+6 4.34-13 4.34-13 C4 829A B Center 'J

9. 66+ 5 4.80-14 G4 #293 8 Center

-v.111 9.08+5 4.49-14 C4 f2981 8 Center -0.267 9.37+5 4.64-14 C4 f2982 8 Center -0.000 8.91+5 4.41-14 l-g C4 82983 5 Center 0.076 8.42+5 4.17-14 C4 829CI S Center 0.342 7.76+5 3.85-14 C4 829t 8 Center 0.217 8.06+5 3.99-14 4.29-94 C4 82451 S FV 0.070 5.894 2.90-15 C4 7248tA B FV -0.283 6.21+4 3.07-15 ~ C4 #24318 3 FV -0.064 6.02+4 2.98-15 C4 #2481C 5 FV 0.170 5.70+4 2.82-15 C4 72431D 8 PV 0.420 5.39+4 2.67-15 2.92-15 C1 7184 B Shroud -0.098 4.0t+6 1.45-13 Cl el8Al B Shroud -0.145 3.99+6 1.42-13 Cl fl8A2 8 Shroud -0.051 4.01+6 1.43-13 Cl #198 8 Shroud -0.286 4.00+6 1.43-13 C1 el881 B Shroud 0.331 3.93*6 1.40-13 Cl silti B Shroud -0.245 4.00+6 1.43-13 Cl 81stl 8 shroud 0.683 3.86+6 1.38-13 C1 #18Fil a Shroud 0.636 3.84*6 1.37-13 L 58

til: DO-24 /93 tPS/IrtfC1.EUS (5ATURAftD) J Estrepolated' to ug,; u...r.n.e t .u-> c.,,,,, and In-Yes se l (Dostneter DFS/S 4 cm 30 ca 7 cm Doetaster or Cape *1a. to Rat. et from from from Seattles ideat. C Location tec.a t ion) 9/9/78 Maasured $hroud threwd FV FV FY Fe-59(n.i)Fe-59 C1 #18712 8 Shroud 0.730

3. 6 H4 1.31-13

'I***#I C1 f18C 3 Shroud 0.483 '3.95+6 1.41-13 1.45-13 ~!5 C1 flht C Shroud =0.793 1.595 5.37 15 9.48-15 6.49-15 C1 #308 8 Center -0.154 3.4S+5 1.2 ble C1 #30C 8 Center =0.030

3. 5 7+ 5 1.27 14 C1 f30D1 8 Center 0.408 4.0H3 1.46-14 C1 f3002 3 Canter 0.158 4.00+5 1.43-14 C1 # 30E 8 Center 0.564 4.41+5 1.57-14 1.31-14 C1 #25C 8 FV 0.631 2.1544 7.66-16 C1 #25D 8 FY 0.443 2.21+4 7.89-16 C1 #23F 8

FV -0.443

2. 5 6+ 4 9.12-16 8.46-14 N1-58(e.p)Co-58 C3 #13 C Shroud

-0.758

5. 2 S+ 7 1.0ble 2.17-14 1.40-14

!,T! C3 f 27 C Center -0.064 5.27+4 1.06-15 1.14-15 C3 #21 C FY 0.438 6.7845 1.36-16 1.44-16 9.80-17 4 C4 #15 C Shrewd -0.622 8.22+7 1.77-14 3.22-14 2.21-14 + C4 #23 C FV 0.415 1.06+4 2.27-16 2.11-16 1.29-16 C1 #17 C Shrewd -0.437 4.3147 6.80-15 1.24-14 8.51-15 Sc-45(n.1)Sc-46 C3 #13 C Shroud 0.707 1.02 +9 1.06-13 1.37-13 8.8ble C3 #14 8 skrowd 0.818 4.5S+10 4.80-12 5.74-12 C3 #28 8 Coster A66? 4.66+9 4.%8-13 3.97-13 C3 f22 8 FV Q. 90

2. 9 )+ 8 3 14-14 3.72-14 C4 #15 C Shroud 0.720 2.34+9 2.6.-13 3.21-15 2.20-13 C4 #16 8 Shrowd 0.893 8.68+10 9.74-12 1.20-11 C4 f24 8 FY 0.850 5.97+4 6.70-14 7.99-14 C1 #17 C thr md 0.394 1.01+9 4.41-14 1.12-13 7.65-14 C1 #18 8 Shroud O. N 4 3.62+10 3.01-12 3.24-12 fi-46(e.p)sc-46 C3 #14 8 shroud

-0.157 1.4 H6 1.95-15 3.32-15 2.18-15 C3 f 28D2 8 Centes 0.439 1.67+5 2.18-16 C3 f 28D3 8 Center 0.564 1.56+5 2.07-16 2.07-16 C4 #16 8 shrt f 0.111 2.16+4 3.00-15 4.59-15 3.15-15 C1 #14A2 8 shroud -0.098 1.16+4 1.20-15 C1 #18A1 8 shroud -0.098 1.16+4 1.20-15 19bl5 1.32-15 U-2 35(n. f) Co-13 7 C3 #13 C Shroud 0.343 1.77+7 3.40-12 4.76-12 3.07-12 , 'f' C3 #27 C Center 0.465 6.4 95 1.24-13 1.14-13 O 5 C3 #21 C FY 0.498 8.01+4 1.54-14 1.65-14 _ h C4 f15 C throud 0.315 3.25+7 6.24 12 8.44-12 5.79-12 ' "j C4 #23 C FY 0.415 1.1H3 2.2ble 2.07-16 C1 #17 C shroud 0.332 1.20+7 2.31-12 3.11-12 2.13-12 U-235(s.f)Co-137 C3 #14 8 shroud 0.840 4.74+4 9.09-11 1.12-10 '[0.0622 C3 #28 8 Center 0.314 6.51+7 1.24-11 1.12-11. C3 822 8 FY 0.412 4.00+4 7.60-13 8.14-13 C4 #16 8 throud 0.830 I IT+1 1.9 b10 2.31-10 C4 f24 8 FY 0.413

7. 6 H4 1.46-12 1.59-12 C1 #18 8 Skreud 0.371 3.60+8 6.77-11 4.06-11 59

r... "3, -j NED0- 24793 'l. nt. '1 '3 j ?. 1 ? Ors /:tUCLEUS (547URATED3 I, Eastapelated" to: ( Capsule II'I' I'II'* c -and

8' In-Vessel

'Dostmeter er Capsule '(poeleeter Dr5/8 - a ca 30 ca 7 en to Ref, at tI'" 'f" f" h L' Reaction-Ident. C Location Location) - 9/9/18 Measured Shrewd ' Shrood PV PT PY U-235(n.f)fr 95' ^ C) #13 - C. Shrewd 0.363 3.8248 3.34-12 4.64-12 3.01 12 ' .,k I, '7((dI C3 #27 C-Center 0.465 1.39+7 1.22-13 1.12 37 C3 821 C PV 0.498 1.78+4 1.56-14 ~ 1.67-14 C4 #15 ~ C Shrewd ~ 0.315 6.26+8-5.90-12 7.94-12 5.47-12

i

'C4 #23' C-PV 0.415 2.26+6 2.1)-14 1.94-14 ,.#\\ C1 #17 C Shroud 0.3 32 ' ' 3.16+8 2,17-12 2.92-12 2.00-12 1 t% 2 35(n. flIr-9 5 'C3 #14 8 Shrewd 0.880 1.0 3+ 10 B.8)-18 1.09-10 'f 0.065h. C3 824 8 Center-0.384 1.45+9 1.25 11 1.12 11 i, C3 822 -8 PY 0.412 8.45+7 7.58-13 8.12-13 C4 816 8 Shrowd 0.8 11 1.84+10 1.70-10 2.04-10 ,d ' ' C4 8'24 8 PV 0.483 -. l.49+8 3.37-12 1.49-12 !* l[c . L-233tn.ftte 103 C3 *13 C Shroud ' O.363 1 88+8 3.29-12 4.62 12 2.97 12 Cl +18 - 8 Shroud 0.371 9.26+4 6.23-18 6.31 11 3 I h#'7h C3 822 ~ C. Center 0.465 6.94+6 1.22-13 1.12-13 y C3 21 C PV 0.495 8.42+5 1.47-14 1.58-14 ,l C4 815 'C Shrewd '). 315 + 00+8 5.76-12 7.79-12 5.34-12 C4 2 3 C PV 0.415

1. 09+4. 2.09-14 1.94-14 Cl.687 C

Shroud 0.332 1.54+8 2.12 12 2.85-12 1.95 12 t-2 3 5!. f ~. a. - 10 3 C3 -14 8 Sheeud - 0.889 4.71+1 8.90-81 1.10-40 C 3 _18 8 Cea ur 0.314 6.764 L 28-11 1.15-11 o,o yg; C3 22 8 FY 0.412 '4.04+7 7.6)-13 8.14-11' C4 'l4 8 Sh r ewd 0.8 31 8.64+9 1.79-10 2.34-10 c4 14 8 PV 0.413 - 6.9047 1.43-12 1.56-12 Cl 'Id B shroud 0.373 4.04+9 6.02-11 6.10-18 t 238(n.f tc6137 ~ C3 =13 C Shroud 0.238 1.96+5 3.93 14 5.90-14 3.80-14 + ' ' I *I*

O' N C3 '27 C Center 0.121' l.35+4 2.70-15 2.52-15.
i j

C3 a ll C PV 0.404 1.92+3 3.85-16 3.05 16 2.06-16 [ (lf C4 *l5 - C ' Shroud 0.259 3.56+5 7.09-14 1.00-13 6.86-14 f C4 21. C PV 0.352 3.61 + 1 7.20-l6 4.38-16 2.67-16 ..f Cl 't? C. Shroud 0.051 . 3.27+5 2.52-14 3.76-14 2.56-14 6 218tn.l Rr-H C3 13 C, Shroud 0.238 3.53+6 3.91-14 5.88-14 3.76-14 "*hE ~ C3 27 C Center 0.121 2.52*5 2.79-15 2.60-15 C3 '24 C. FY 0.404

3. 5 3+4 3.91-16 3.09-16 2.10-16 g

C4 'Il C Shroud ' O.259 5.5646 6.62-14 9.35-14 6.41 14 s 6 21 f FY

0. 152 5.96+4 7.l*-16 4.33-16 2.64-16 Cl *17 C'

Shrowd 0.051 2.68+6 2.23 84 3.48-14 2.38-14 . -4 e f s f s n -101 C3 all C Shroud 0.238 4.2246 3.95-14 5.94 14 3.42 14 F.Y. M22 g3 27 C Center 0.121 2.88*) 2.69 15 2.51-15 c1 21 C PV 0.404 3.81+4 3.57-16 2.87-16 1.91-16 t.4 ell C Sh r sv J 0.2b9 6.60+6 6.77-14 9.55-14 6.55-14 C4 -23 C PV 0.132 6.68+4 6.86-16 4.18-16 2.55-16 49 al? C Shrewd 0.051 3.14+6 2.31-14 3.45-14 2.36-14 60 I

NEDO-24793 =., DFS/NUC12US ($ATURAf tD) titrepelated' to tu t.r.... t-.iio.) ce,s le and 1 l' to-Tessel (Doetmeter DFS/g 4 cm 30 cm 7 cm - Dos Lae t e r. or Capsule to Ra'. at i '" t#** i b tasetton Ident. 1 Location Location) t/9/78 Measured $hroud Shrewd FV FV Fv up 237(s.f)Co-137 C3 #13 C Shroud 0.332 4.84+5 1.62-13 2.42-13 1.5b13 F.T. - 0.0450 .C) illa C FY -0.221 6.41+3 1.1k!5 t C3 #218 C FV 0.185 6.66+3 1.2 bl5 9.4b/5 6.41 16 C4 ill C Shroud 6.415 1.75+6 3.20-13 4.45-13 3.05-13 c4 #23 C ~ PV 0.415 1.24+4 2.26-15 1.5)-15 f.34-16 C1 #17 C shroud 0.332 5.34+5 1.09-13 1.54-13 1.06-13 up-2 37(e. f)2 r-t $ C3 #13 C shroud 0.332 1.63+7 1.54-13 2.30-13 1.4bl 3 f.1. = 0.0H 3 C3 #21A C FV -0.221 1.2H3 1.14-15 C3 #218 C FY 0.185

1. 2 )+ 5 1.1415 9.16-16 6.17-16

^ C4 #15 C skreed 0.415 2.90+7 2 96-13 4.12-13 2.42-13 C4 f23 C FY - 0.415 2.12+5 2 17 15 1.47-15 8.94-14 C1 #17 C Shroud 0.332 1.31+7 9.7bl4

1. )b13 9.4ble e

II.] up-2 3 7(e.f) h-103 C3 f13 C shrewd 0.332 1.62+7 1.59-13 2.37-13 1.5)-13 F.T. = 0.0H9 C3 #21A C FV -0.221 1.1)+5 1.11-15 8.64-16 5.85-16 C4 #15 C Shrowd 0.415

2. 8 )* )

3.05-13 4.24-13 2.tb13 c.4 f/3 C 77 0.415 1.95+5 2.14 15 1.42-15 8.6bl6 C1 #17 C Sisoud 0.332 1.2t+7 9.96-14 1.41-13 1.44-14 l' Th-2 32(e. f)Co-13 7 C3 #13 C shroud 0.394 5.26+4 1.04-14 1.49-14 9.6&l5 F.T. - 0.0H 3 C3 #27 C Center 0.277

4. 3 )* 3 8.55-16 7.83-16

.p C4 #13 C shrowl 0.353 9.194 1.79-14 2.46-14 1.69-14 ( C1 217 C shrood 0.332 3.14+4 6.16-15 0.39-15 ' 5.8bl5 / Th-232(e.f)2r-95 C3 #13 C Shroud 0.394 1.C++6 1.06-14

1. 5 bl4 f.85-15 F.. = 0.M64 C3 #27 C Center 0.277 8.64+4 8.42-16 7.71-16 C4 #15 C Shroud 0.353 1.66+6 1 76-14 2.42-14 1.66-14 C1 #17 C Shroud 0.332 7.42+5 5.96-15 8.34-15 5.72 15 b.'

lUf 61/62

NEDO-24793 WPENDIX B LOCALIZED POWER HISTORY CALCULATIONS BY SCIENCE APPLICATIONS, INC. Measurements of neutron flux have been made at the Browns Ferry Unit 3 reactor by personnel from the General Electric Company. The technique used involved placing fissionable and activation specimens between the core shroud and the pressure vessel at three different positions (see Figures B-1 and B-2). Neutrons leaving the core cause these specimens to activate or fission (giving rise to the production of fission products). At the end of the irradiation period (750 days in this case) the specimens were withdrawn from the reactor i and the decay activity of certain fission / activation nuclides was measured. f 'h This measurement, coupled with. knowledge of the reactor power history, m. i macroscopic production cross sections, and half-lives of the nwasured nuclides, y, allows the calculation of the full-power flux, (f, at the specimen location. Knowing the full-power flux and the reactor power history allows the calcula-c tion of the fluence at the specimen location. This fluence can then be used in pressure vessel (PV) dosimetry studies and PV lifetime evaluation. Anomalies in the full-power fluxes derived from the measurements described above have resulted in an effort at Science Applications, Inc., (SAI) to make independent predictions of these measurements and in so doing understand ,- l these anomalies. The conclusion of this preliminary evaluation is that i measurements of this type must use the local power history (i.e., power as a function of time for a small region of the core near the detector or i l l', specimen), instead of total reactor power. Even this refinement is not l ( precisely correct. Obviously, the correct data to use are quantities e proportioned to the fluxes as a function of time at the specimen location. j This information is seldom, if ever, available (except by detailed calcula-tion). However, local power appears to be a reasonable alternative to use in lieu of any information. B.1 THE PROBLEM The measurements at Browns Ferry consisted of measuring the activity of Cs-137, Zr-95, and Ru-103 in the fission specimens at the end of a 750-day 63 1

V HEDO-24793 . -k ) /,p'E. t ,b yw,m .t,9; !jf '. I ) "', G hi i SPECIMEN HOLDER i, 4- - PRESSUR E VE S3 E L SHROUD c... k ....."'P!'SE,_,,_ G 3, SPECIMEN HOLDER f T ff Tf Figure B-1. Vertical Section Showing Specimen Holding Locations

//////////////

PR ESSUR E VESSEL CONTAINMENT wwww y en.. SPECIMEN G1,G3 HOLDERS ,,L,,,, 'llO $HR000 MB E wh% ~ \\ VAW/GVA pygg t N N l Figure B-2. Horizontal Section Showing Specimen Holder Locations. Shaded Regions in Core are Regions Used'to Generate Local Power Histories 64

q s HEDO-24793 .. p. pg

l i

, :s. n ,,m ,;:Q [%((;. A' ^ irradiation period in the re tor. The full-power flux was then calculated ,h according to the following equation: .. c :- i t a;, a %Er AN Mf $g(j)K ~ -A at ' 3 = (B-1) 4 {, 1 -A)(t-t ) n j f g h, Fj b i, y -. i= 1 a;;

N-where 3.r s.

j subscript for Cs-137, Zr-95, or Ru-103 = y I, (f(j) full-power flux as determined by a measurement of nuclide = j activity ,,4 i A) decay constant for nuclide j = i N) number density of nuclide j at time t = 4 I macroscopic fission cross section = p Y) fission yield for nuclide j = full-power fraction for the reactor for the ith P = time interval we number of time intervals in the irradiation period n = th at length of i time interval = f length of irradiation period T = Hf' K proportional 1*y constant = bh " 'l,. The values of the fission cross section and yields depend on the specimen being measured, but this analysis will concentrate on only one specimen, U-235 , b without a cadmium cover. The measured quantity in Equation B-1 is the activity at time t, A)N). 1 j. To check the reproducibility of this type of full-power flux measurement, three l-nuclides instead of one were measured and ratios of the determined fluxes were calculated. Thus, l l 65 1.

\\ HEDO-24793 ' E Y b d.. ijIg,2 Y

i !'

n A "j k { # c Y j i k,1 (3) K- = (B-2) '{i (f(k) n-AN O kkji1 1 j,1 3p : }I vhere 1 i ~ -A)6 t " -A)(T-t ) t y G 1-c e = 3,1 (B-3) L i /. [j Consistent measurements should yield flux ratios ofi one. Experimental results ~ (shown below) indicate that this fs not the case. 3 G1 G3-G4 '$ (Ca) f i 0.87 1.05 1.24 (g(Zr) c $g(Cs) 0.89 1.03 1.19 $g(Ru) The experimental uncertainties are estimated to be about i?%, thus the observed trend is not attributable to' experiment error. B.2 RESOLUTION OF ANOMALIES h. in order to make calculational predictions of these flux ratios it was necessary to calculate the measured activities at the end of the irradiation ~ period, A)N). This was accomplished by assuming that the flux at the specimen location was proportional to the local power near the specimen. The local power as a function of time was generated by analyzing relative power data supplied by Browns Ferry. This relative power data took the form of BWR simulator code results for relative power in each bundle for three different control rod configurations and also data for relative power in coarse radial and axial core regions for 66

h NEDO-24793 Y b"

U' 0[

all of the control rod configuration used during the irradiation period. These [ >3' ~ coarse data were generated by a process computer using actual in-core measure- -ments as a data base. Local power histories were generated by considering enly f the shaded core regions shown in Figure B-2 and assuming that the power in these regions could be factored into an azimuthal / radial component and an 4 ?? axial component. The simulator results were used to supply the az!muthal/ radial part and the process data provided the axial part..The power histories for all a , [' three specimen locations are shown in Table B-1. The power is normalized at each position so that, integra: ed with time, they add to one. e The measured activity can be calculated as 5.,,, a n f$ A N) IY f p P{G3g (B-4) = y i=1 L t where P' ic the local full-power fraction (as opposed to the F which is the 1 reactor full-power fraction). It should also be pointed out that the $g in this equation is independent of j because this is not a measurement. Sub-stitution of Equation (B-4) into Equation (B-2) yields f (j) j,1 i k,1 i f $g(k) (B-5) n n k,i i j,i _ "1 . "1 Note that the 4 cancelledontherighthandside.AlsonotethatifP{=P f 1 for all 1, then indeed the flux ratio would be one. The fact thatP{/P1 for all i can be ascribed to local control rod movements which affect local power, l but not overall reactor power. Peripheral assemblies are especially sus-ceptible to these variations. 1 67

a gI l NEDO-24793 m

]' Table B-1 NORMALIZED POWER HISTORY AT THE THREE LOCATIONS Normalized Power / Day Irradiation .ll Period Time (days) G1 G3 G4 1~ 0-107

3. 94 (-4) a 7.51(-4) 8.18(-4) i 2

107-166 7.79(-4) 1.62(-3) 1.72(-3) 3 166-215 1.12(-3) 1.58(-3) 1.64(-3) 4 215-265 9.96(-4) 1.59(-3) 1.64(-3) 5 265-328 9.13(-4) 1.41(-3) 1.34(-3) a-6 328-382 1.53(-3) 1.35(-3) 1.43(-3) 7 382-431 1.50(-3) 1.35(-3) 1.43(-3) 8 431-462 2.68(-3) 2.10(-3) 2.09(-3) 9 462-529 1.97(-3) 1.12(-3) 1.15(-3) 10 529-580 1.72(-3) 1.31(-3) 1.26(-3) 11 580-632 1.75(-3) 1.48(-3) 1.31(-3) 12 632-688 1.92(-3) 1.49(-3) 1.49(-3) A 13 688-750 '1.48(-3) 1.17(-3) 1.02(-3) ~4 "dead 3.9 x 10 The results of Equation (B-5) are shown in Table B-2 along with the measured values. The agreement between measurement and calculation, in light of the limited detail of the data available for the calculations, is very good. It is anticipated that if detailed power distributions were available for each time interval in the irradiation period, the agreement would improve significantly. Table B-2 does illustrate the conclusion, however, that deviations from a value of one in the full-power flux ratios are due to variations in power level near the specimen locations. 68

NEDO-24793 (

c.. -

jf., Table B-2 !! - F ~ .g ~ CALCULATED AND HEASURED FULL-POWER FLUX LATIOS U k, .)4 J, y i, G1-G3 G4 i 4

h Calc Meas Calc Heas.

' Calc Heas T . g'. (g(Cs). i. O.85 0.87 1.07-1.05 1.15 1.24 -(g(Z r) (g(Cs) 0.85 0.89 1.08 1.03 1.18 1.19 L, (g(Ru) ? HALF-LIVES * ..l; IO9-Cs-137 30.17 years Zr-95 63.98 days Ru-103 39.35 days

From ORNL/NUREG/TM-102, D. G. Kocher, August 1977.

M B.3 CONCLUSIONS Fluence measurements of the type described here are valuable tools in the analysis of pressure vessel dosimetry, but errors of up to 120% may be intro-duced by the use of reactor power history instead of local power history for the calculation of flux from activity measurements. In other words, some information of power or flux near the specimen tocation as a function of j time must be available for the analysis of the measurements. A secondary conclusion of this work is that if the reactor power history is the only information available, specimens should be located in positions where flux gradients and sensitivity to small changes in control rod con-figuratione are minimized. The G3 position in this work fulfills these requirements and indeed the flux ratios at this location are the closest to 69 i

NEDO-24793 l 4 ^ one.- In most cases, it is reasonable to assume that specimens located in the middle of the flats at the core midplane will produce accurate results when analyzed with the simple core averaged reactor power history. 9 I 70 l

g, NEDO-24793 ~ APPENDIX C PHOT 0 FRACTION AND ASSEMBLY PERTURBATION 'l CALCULATIONS BY SCIENCE APPLICATIONS,~1nc.

6 PHOTOFRACTION'AT CENTER OF CAPSULE j

l At G1 U-238 Np-237 Th-232 i Capsule 17 1.8 1.0 3.7 i 30 7.6 4.7 14.2 l 25 28.2 20.6 43.6 - !" l At G3 Capsule 13 1.8 1.0 3.7 27 7.6 4.7 14.2 l~'r^' 21 28.2 20.6 43.6 At c4 Capsule 15 2.3 1.1 4.5 29 10.5 6.1 18.8 23 37.7 27.8 53.8 WITH ASSEMBLY /NO ASSEMBLY UN00LLIDED Radius (cm)* U-238 Np-237 Th-232 $(1100 kev) ((11 HeV) 267 0.95 0.94 0.93 0.96 0.93 289 0.97 0.94 0.89 0.94 0.94 ,j 313 0.96 0.99 0.95 1.01 1.00

Radius (cm)

Capsules 1 267 17,18,15,16,13,14 289 30,29,27,28 313 25,23,24,21,22 i 71/72 t -.}}

ML20147D670

Text

SUMMARY

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