Text

Reactor Cavity Neutron Dosimetry Program for Brunswick Unit 2
	ML20247G056
Person / Time
Site:	Brunswick
Issue date:	12/31/1986
From:	Shaun Anderson, Lippincott E, Sejvar J; GENERAL ELECTRIC CO.
To:	;
Shared Package
ML20247G053	List: ML20247G048; ML20247G056;
References
	WCAP-10903, NUDOCS 8904040112
	Download: ML20247G056 (78)
	v • d • e

. _- _

WESTINGHOUSE CLASS 3 l-1 WCAP-10903

,, CUSTOMER DESIGNATED DISTRIBUTION

. y l

REACTOR CAVITY NEUTRON DOSIMETRY PROGRAM FOR BRUNSWICK UNIT 2 S. L. Anderson E. P. Lippincott J. Sejvar j

l 1

WORK PERFORMED FOR CAROLINA POWER AND LIGHT COMPANY

.. 1 l DECEMBER 1986 l

Approved: I/ [au F. L. Lau, M,anager J Radiation and Systems Analysis l

Although the information contained in this report is nonproprietary, no j i

distribution shall be made outside Westinghouse or its Licensees without the '

customer's approval. ,

WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR ENERGY SYSTEMS

. g. P.O. BOX 355 PITTSBURGH, PENNSYLVANIA 15230 ,

>8904040112 89033o hDR ADOCK 05000324 PDC l

l L _________-

WESTINGHOUSE CLASS 3 TABLE OF CONTENTS j Page TABLE.OF-CONTENTS i-L LIST OF FIGURES ii LIST OF TABLES iv l

1.0 OVERVIEW OF THE PROGRAM 1-1

2.0 DESCRIPTION

OF SENSOR SETS 2-1 2.1 Capsule Description 2-1 2.2 Multiple Foil Sensor Sets 2-1

,', 2.3 Gradient Wires 2-5 2.4 As-Built Sensor Data 2-5 1

l 3.0 SENSOR PLACEMENT WITHIN THE REACTOR CAVITY 3-1 4.0 ANALYTICAL METHODOLOGY 4-1 i 4.1 Two-Dimensional Analysis 4-2 4.2 One-Dimensional Analysis 4-4 5.0 EVALUATION OF CYCLE 6 CAVITY DOSIMETRY 5-1 5.1 Measurement Results 5-1 5.2 Derivation of Neutron Exposure Parameters 5-16 6.0 DERIVATION OF PRESSURE VESSEL EXPOSURE 6-1 6,1 Evaluation of Cycle 6 Exposure 6-2 6.2 Uncertainties in Exposure Projections 6-5

7.0 CONCLUSION

S AND RECOMMENDATIONS 7-1

8.0 REFERENCES

8-1 i

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o WESTINGHOUSE CLASS 3-i I

1 l , .

, LIST OF FIGURES q Figure Title %

L 2 Sensor Set Holder for Cavity Dosimetry 2-2' 3-1 Azimuthal Location of Sensor Strings 3-2 I l

.3-2 Axial Location of Multiple Foil Sensor Sets 3-3 l 1:

4-1 Brunswick Unit 2 R, e Reactor Geometry 4-3 4-2 Brunswick Unit 2 Radial Power Distribution 4-5 l 5-1 Fast Neutron Flux (E >1.0 Mev) Derived from the 5-20 i Fe54 (N, P) Mn54 Reaction - Traverse 1 - 0' 5-2 Fast Neutron Flux (E >1.0 Mev) Derived from the 5-21 Fe54 (N, P) Mn54 Reaction - Traverse 2 - 45' 5-3 Fast Neutron Flux (E >1.0 Mev) Derived from the 5-22

- . Fe54 (N, P) Mn54 Reaction - Traverse 3 - 15' 4

. 5-4' Fast Neutron Flux-(E >1.0 Mev) Derived from the 5-23 Fe54 (N, P) Mn54 Reaction - Traverse 4 - 22.5*

5-5 Fast' Neutron Flux (E >1.0 Mev) Derived from the 5-24 Fe54 (N, P) Mn54 Reaction - Traverse 5 - 75' 5-6 Fast Neutron Flux (E >1.0 Mev) Derived from the 5-25 Fe54 (N, P) Mn54 Reaction - Traverse 6 - 135' 5-7 Fast Neutron Flux (E >1.0 Mev) Derived from the 5-26 Fe54 (N, P) Mn54 Reaction - Traverse 7 - 225*

5-8 Fast Neutron Flux (E >1.0 Mev) Derived from the 5-27 L Fe54 (N, P) Mn54 Reaction - Traverse 8 - 288' 5-9 Fast Neutron Flux (E >1.0 Mev) Derived from the 5-28 Fe54 (N, P) Mn54 Reaction - Traverse 9 - 315' 5-10 Fast Neutron Flux (E >1.0 Mev) Derived from the 5-29 i NiS8 (N, P) CoS8 Reaction - Traverse 1 - 0*

5-11 Fast Neutron Flux (E >1.0 Mev) Derived from the 5-30

- NiS8 (N, P) CoS8 Reaction - Traverse 2 - 45' 5-12 Fast Neutron Flux (E >1.0 Mev) Derived from the 5-31

-- NiS8 (N, P) CoS8 Reaction - Traverse 3 - 15* l 1

l ii l

WEST!NGHOUSE CLASS 3 LIST OF FIGURES (Cont.)

Title Page Figure Fast Neutron Flux (E >1.0 Mev) Derived from the 5-32 5-13 NiS8 (N, P) Co S8 Reaction - Traverse 4 - 22.5*

Fast Neutron Flux (E >1.0 Mev) Derived from the 5-33 5-14 NiS8 (N, P) CoS8 Reaction - Traverse 5 - 75*

Fast Neutron Flux (E >1.0 Hev) Derived from the 5-34 5-15 NiS8 (N, P) Co S8 Reaction - Traverse 6 - 135*

Fast Neutron Flux (E >1.0 Hev) Derived from the 5-35 5-16 NiS8 (N, P) CoS8 Reaction - Travarse 7 - 225*

Fast Neutron Flux (E >1.0 Mev) Derived from the 5-36 5-17 NiS8 (N, P) Co S8 Reaction - Traverse 8 - 288*

Fast Neutron Flux (E >1.0 Mev) Derived from the 5-37 5-18 ~'

NiS8 (N, P) Co S8 Reaction - Traverse 9 - 315*

Fast Neutron (E >1.0 Mev) Flux Projection Through 6-3 6-1 the Pressure Vessel Wall Iron Displacement Rate Projections Through the 6-4 6-2 Pressure Vessel Wall 4

iii

WESTINGHOUSE CLASS 3

,-, LIST OF TABLES Table Title Page 2-1 Neutron Flux Monitors Contained in Each Multiple Foil 2-3 Sensor Set 2-2' Multiple Foil Sensor Set Placement for the Cycle 6 2-6 Irradiation .

2-3 As-Built Data for Bare Iron Foils 2-7 l 2-4 Summary of Chemical Analysis of the Stainless Steel 2-10 Gradient Wires 3-1 Multiple Foil Sensor Set and Gradient Wire Locations 3-4 Within the Reactor Cavity

.. l

. 5-1 Fissionable Deposite Mass Calibration Checks 5-3 5-2 Irradiation History of Cycle 6 Reactor Cavity 5-4 Neutron Flux Monitors 5-3 Measured Activities in Multiple Foil Sensor Sets 5-5 Irradiated During Cycle 6 i 5-4 Measured Activities in Stainless Steel Gradient 5-6 Wires - Traverse.1 (0*)

5-5 Measured Activities in Stainless Steel Gradient 5-7 I Wires - Traverse 2 (45')

5-6 Measured Activities in Stainless Steel Gradient 5-8 i Wires - Traverse 3 (15')

5-7 Measured Activities in Stainless Steel Gradient 5-9 Wires - Traverse 4 (22.5')

5-8 Measured Activities in Stainless Steel Gradient 5-10 Wires - Traverse 5 (75')

- 5-9 Measured Activities in Stainless Steel Gradient 5-11 Wires - Traverse 6 (135*)

5-10 Measured Activities in Stainless Steel Gradient 5-12 Wires - Traverse 7 (225*)

iv l.

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WESTINGHOUSE CLASS 3 LIST OF TABLES Title Page Table 5-13 5-11 Measured Activities in Stainless Steel Gradient Wires - Traverse 8 (288')

5-14 5-12 Measured Activities in Stainless Steel Gradient Wires - Traverse 9 (315*)

5-15 5-13 Total Fissions Measured in Solid State Track Recorders.Irradested Druing Cycle 6 Spectrum Average Reaction Cross Sections and dPa/ Flux 5-18 5-14 Rates Within the Brunswick Unit a Reactor Cavity Summary of Reactor Cavity Neutron Desimetry Evaluations 5-19 5-15 at the Multiple Foil Sensor Set Locations

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WESTINGHOUSE CLASS 3 SECTION 1.0

. OVERVIEW 0F THE PROGRAM The reactor cavity neutron dosimetry program initiated at Brunswick Unit 2 in October of 1984 was designed to provide an experimental verification of the analytical predictions of the fast neutron exposure of the reactor pressure vessel and to establish a mechanism to enable long term monitoring of portions of the vessel that could experience radiation induced increases in reference nil ductility transition temperature (RTNDT) ver the plant lifetime. When used in conjunction with dosimetry from internal surveillance capsules and with the results of neutron transport calculations, the cavity dosimetry allows the projection of embrittlement gradients through the pressure vessel wall with a minimum uncertainty. Minimizing the uncertainty in the neutron

,' exposure projections will, in turn, help to assure that the reactor can be operated in the least restrictive mode possible with respect to 10CFR50

[, Appendix G pressure / temperature limit curves for normal heatup and cooldown of the reactor coolant system. In addition, an accurate measure of the neutron exposure of the pressure vessel can provide a sound basis for requalification of that component should operation of the plant beyond the current design and/or licensed lifetime prove to be desirable.

The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for implementation of tread curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through g the pressure vessel wall.

1-1

WEST!NGHOUSE CLASS 3 Because of this potential shif t away from a threshold fluence toward an energy '

dependent damage fun: tion for data correlation, ASTM Standard Practice E853

" Analysis and Interpretation of Light Water Reactor Surveillance Results",

recommends reporting displacements per iron atom (dPa) along with fluence (E > 1.0 MeV) to provide a data base for future reference. The energy dependent dPa function to be used for this evaluation'is specified in ASTM Standard Practice E693 " Characterizing Neutron Exposures in Ferritic Steel: in Terms of Displacements per Atom (dPa)."

The dosimetry program designed for implementation at Brunswick Unit 2 contained sensor sets with spectral coverage sufficient to make an accurate assessment of the dPa parameter as well as of the conventional fast neutron fluence (E > 1.0 MeV). During the initial irradiation cycle for the Brunswick Unit 2 dosimetry (fuel cycle 6), comprehensive sensor packages including radiometric foils and solid state track recorders were installed at -

two locations within the cavity to characterize the neutron energy spectra ~~

near the beltline region of the reactor vessel. To achieve the goal of long

term monitoring of the reactor vessel, dosimetry packages consisting of a subset of the sensors used in the spectra determinations could be included in subsequent fuel cycles to provide integration of the neutron exposure throughout plant life. In all cases, whether for single cycle measurements or for long term monitoring, gradient wires would be used in conjunction with the encapsulated sensor sets to complete the mapping of the neutron environment over the entire beltline region.

' Analytical studies have indicated that, while changes in core power distributions can cause significant variations in the magnitude and spatial distribution of the neutron flux in the reactor cavity from fuel cycle to fuel cycle, the relative energy distribution of neutrons is controlled primarily by the reactor cavity geometry and is quite insensitive to changing fuel management schemes. Hence, the strategy of spectrum characterization in cycle 6 with emphasis in subsequent cycles placed on tracking neutron flux magnitude '~

and spatial distribution represents a valid approach. It should be noted, however, that even the senser sets employed for long term monitoring have ,,

1-2

WEST 1NGHOUSE CLASS 3

.- sufficient energy response to determine the neutron spectrum with reasonable accuracy. Thus, the analytical indications of.a constant spectrum would be

- verified by measurement.

In choosing sensor set locations for the cavity dosimetry program, advantage was taken of the quadrant symmetry characteristic of both core power distributions and reactor geometry. That is, an attempt war, made to concentrate measurement locations in a single 90' sector in order to obtain detailed azimuthal and axial flux distributions over the beltline region of the vessel. Verification of the degree o# nuadrant symmetry was provided by the inclusion of additional gradient wires at selected locations in each of the remaining three quadrants.

Placement of the two multiple foil sensor sets was such that spectra

. determinations were made at the 0* and 45' azimuths at an axial elevation characteristic of the peak in the fast neutron (E > 1.0 MeV) fluence with 1

the intent to measure changes in spectra caused by varying amounts of water l

located between the core and the pressure vessel. Due to the irregular shape l of the reactor core, water thickness varies significantly as a function of l azimuthal angle within a given quadrant. At each of the two azimuthal locations selected for spectrum measurements gradient wires extended over the' full twelve foot height of the reactor core. At all other azimuthal locations gradient wires alone spanned the axial extent of the core.

1 1-3

WESTINGHOUSE CLASS 3

)

,. SECTION 2.0

- DESCRIPTION OF SENSOR SETS l

2.1 CAPSULE DESCRIPTION l

The multiple foil sensor sets mounted on the O' and 45' axial traverses during the cycle 6 irradiation were retained within 3.5" x 1.0" x 0.5" rectangular aluminum 6061 capsules. A detailed description of these individual sensor' set holders is shown in Figure 2-1.-

As illustrated in Figure 2-1, each irradiation capsule contained three J compartments to hold the neutron sensors. The top compartment (position 1)

.. accommodated bare foil packages, whereas the two remaining compartments

, (positions 2 and 3) housed cadmium covered foil sets. The separation between

.. position 1 and position 2 was such that cadmium covers inserted into I position 2 did not introduce significant perturbations in the thermal flux within position 1.

Aluminum 6061 was selected for the irradiation capsules in order to minimize neutron flux perturbations at the sensor set locations as well as to limit the radiation levels associated with shipping and handling of the capsules following irradiation. Each capsule represented an aluminum mass of approximately 60 grams. Thus, each set of two capsules resulted in an increase in the aluminum inventory in the containment building of only 120 grams (~ 0.25 lb). An increase of this magnitude is insignificant relative to the total aluminum inventory already present in the containment.

2.2 MULTIPLE FOIL SENSOR SETS The multiple foil sensor sets used in the cycle 6 irradiation included the

.. materials listed in Table 2-1. Also given in Table 2-1 are the primary nuclear reactions of interest and the approximate response range of each 2-1 l

1

WEST!NGHOUSE CLASS 3 MATERIAL: 6061 ALUMINUM ,

^

FRONT VIEW O 1.00" W.W.W

,e iPOSITION) l POSITION !- POSITION l9 0.75"

' . s I.I2" l.00" ~ .75"-

=

3.50"

~ TWO HOLES FOR ATTACJ4 DENT TO WIRE' 8 HOLES - DRILL AND TAP FOR S-32 X i/4 FLAT FEAD .,

AI MACHINE SCREWS, COUNTER SITE COVER PLATES .

SIDE VIEW COVER PLATE e.

l<_____.4 _____.,lii" TOP iil u u_____.,w jii II

_.30" .2.

.38" POSITION I (TOP): BARE RADIOMETRIC FOILS AND SOLID STATE TRACK RECORDERS POSITION 2: CADMIW COVERED RADIOMETRIC FOILS POSITION 3 (BOTTOM): CADMIUM COVERED FISSION FOILS AND SOLIO STATE TRACK RECORDERS j

FIGURE 2-1 SENSOR SET HOLDER FOR CAVITY DOSIMETRY I

2-2

1 WESTINGHOUSE CLASS 3 1

TABLE 2-1

,..; NEUTRON FLUX MONITORS CONTAINED IN EACH MULTIPLE F0ll SENSOR SET Material- Reaction of Interest Response Range (MeV) l Copper CuS3(n a) Co60 E > 5.5 Titanium Ti46(n p) Sc46 E > 4.0 Iron Fe54(n,p)Mn54 E > 2.0 Nickel NiS8(n,p) CoS8 E > 2.0 Silver-Aluminum Ag109(n,r) Ag110M E < 0.0004 Cobalt-Aluminum CoS9(n,r)Co60 E < 0.015 SSTR U238(n,f) E > 1.5 SSTR U235(n,f) E < 0.002

~

SSTR- Np237(n,f) E > 0.2 -l SSTR Pu239(n,f) E < 0.0005 l

i 1 .

2-3

UEST1NGHOUSE CLASS 3 sensor. When packaged for irradiation, these multiple foil sensor sets were classified as either radiometric monitors (RM) or solid state track recorders (SSTR), where the term " radiometric monitor" was applied to all sensors for which the primary reaction of interest was a non-fission event. The acronyms .

RM and SSTR are used throughout this report when reference is made to the individual sensor sets.

During the cycle 6 irradiation, RM detectors were included in both the bare and cadmium covered positions. Iron, cobalt-aluminum, and silver-aluminum foils were contained in position 1 of each capsule, while cadmium covered iron, copper, nickel, titanium and cobalt-aluminum foils were located in petition 2.

In addition to the RM's, SSTR's were also included in both bare and cadmium covered positions for each irradiation capsule. These SSTR's provided the mechanism to obtain fission reaction data within the cavity. For the cycle 6 .

235 238 , pp 237 , and Pu 239 irradiation, track recorders employing U ,g fissionable deposits were included in cadmium covered holder position 3 and 235 239 were located in bare track recorders using deposits of U and Pu position 1.

There are several advantages to the use of SSTR's rather than fission foils to obtain fission reaction rates either in a single cycle irradiation or in a long term monitoring situation. The minute deposits of fissionable material required for the radiator not only minimize the inventory of necessary nuclear material, but also eliminate problems associated with the shipping and handling of larger quantities of fissile and fissionable isctopes. Since the track recorder readout is actually accomplished with the mica cover, the deposits themselves are not destroyed and may be used from cycle to cycle.

Furthermore, the mica recorders serve as a permanent record of the neutron exposure incurred during the irradiation.

W On 2-4 L __ - _________- ___ _ ___________ ____ _ _ _ ______ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

WESTINGHOUSE CLASS 3

.- 2.3 GRADIENT WIRES J- Along with the multiple foil sensor sets placed at the 0* and 45' peak neutron flux locations within the reactor cavity, gradient wires were employed to obtain axial variations of fast neutron exposure parameters along each of nine traverses. Subsequent to irradiation these gradient wires were removed from the cavity and segmented to provide neutron reaction rate measurements at one foot intervals over the height of the active fuel.

The gradient wires consisted of type 304 stainless steel of 0.040 inch diameter. When coupled with a chemical analysis, the stainless steel irradiation yielded activation results for the Fe54/.n,p) Mn54, NiS8(n,p)

CoS8, and CoS9(n,r)Co60 reactionc. The high purity iron, nickel, and cobalt-aluminum foils contained in the two multiple foil sensor sets provided

. a direct correlation with the measured reaction rates from these gradient wires; and established an overcheck on the chemical analysis of the type 304 stainless steel. These cross comparisons and overchecks permitted the use of the gradient measurements to derive neutron flux distributions in the cavity with a high level of confidence.

2.4 AS-BUILT SENSOR DATA The multiple foil sensor sets inserted into the reactor cavity for the cycle 6

~

irradiation were encapsulated in aluminum 6061 holders designated BA and BB.

Individual sensor set placements within these two capsules are specified in Table 2-2. From Table 2-2, it may be noted that the irradiation capsule identification was directly correlated with the identification number stamped on the bare iron foil contained in position 1 of each holder.

The bare iron foils designated BA and BB where purchased directly from Reactor Experiments Incorporated. All of the remaining multiple foil sensor sets including the SSTR's were obtained from the Department of Energy (DOE) through the Hanford Engineering Development Laboratory (HEDL). As-built data on each of these sensor packages is provided in Table 2-3.

2-5

{

WEST!NGH0VSE CLASS 3 TABLE 2-2 ,

MULTIPLE F0ll SENSOR SET PLACEMENT FOR THE CYCLE 6 IRRADIATION ,.

Sensor Set ID' Capsule Position Capsule BA Capsule BB Sensor Type 1 BA BB R/E IRON 1 B-1 B-2 HEDL RM 1 BK-1 BK-2 HEDL SSTR 2 Cd-1 Cd-2 HEDL RM(Cd)

.3 BK-3 BK-4 HEDL SSTR(Cd)

NOTE: Capsule positions 2 and 3 were cadmium covered. .' ,

O 4 6

2-6

WESTINGHOUSE CLASS 3

.- TABLE 2-3 l' AS-BUILT DATA FOR MULTIPLE F0ll SENSOR SETS I

Sensor Set ID Material Mass (mg) Purity BA Fe 115.6 99.98% Fe B-1 Fe 1211. 99.77% Fe B-1 Ag/A1 102.2 0.145% Ag B-1 Co/A1 110.1 0.506% Co

~9 BK-1 U-235 1.59x10

~9

. BK-1 Pu-239 1.77x10 Cd-1 Fe 1210. 99.77% Fe

)

Cd-1 Cu 421.1 99.99% Cu !

Cd-1 Ni 297.1 99.66% Ni Cd-1 Ti 315.0 99.73% Ti Cd-1 Co/A1 118.4 0.506% Co BK-3 U-235 5.13x10

-8

-7 BK-3 Pu-239 1.68x10 BK-3 U-238 5.27x10

-6 BK-3 Np-237 9.72x10

-6 BB Fe 115.8 99.98% Fe B-2 Fe 781.7 99.77% Fe B-2 Ag/A1 110.5 0.145% Ag !

B-2 Co/A1 107.2 0.506% Co BK-2 U-235 1.39x10"9 BK-2 Pu-239 2.52x10

-9 i

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1 +

WESTINGHOUSE CLASS 3 l

TABLE 2-3 (Cont'd) -

. 1

, i AS-BUILT DATA FOR MULTIPLE FOIL SENSOR SETS . lj

?

l Material Mass (mg) Purity- l Sensor Set ID l

Fe 1213. 99.77% Fe :

Cd-2 Cu 384.1 99.99% Cu Cd-2 Ni 297.4 99.62% Ni :

l Cd-2 Ti 315.9 99.73% Ti f Cd-2 Co/Al- 118.4 0.506% Co Cd-2

-7 I BK-4 U-235 1.06x10 l

~7 BK-4 Pu-239 1.93x10 l

-6 BK-4 U-238 4.64x10

-6

.}

BK-4 Np-237 9.46x10 .

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WEST!NGHOUSE CLASS 3 l -

As-built data for the stainless steel gradient wires were obtained by conducting a series of inductively coupled plasma spectrometry measurements on selected samples to determine the iron, nickel, and cobalt content of the l wires. The results of those analyses are summarized in Table 2-4. As stated earlier, comparisons of reaction rates measured in the gradient wires with those measured in the high purity foil materials ~ provided additional confirmation of the veracity of these composition determinations.

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2-9

WESTINGHOUSE CLASS 3 TABLE 2-4 ,

SUMMARY

OF CHEMICAL ANALYSIS OF THE STAINLESS STEEL GRADIENT WIRES Elemental Weight Percent Cr Ni Cu Co Sample No. Fe 71.1 18.7 8.36 0.169 0.123 1

72.4 18.7 8.36 0.171 0.125 2

71.8 18.8 8.39 0.170 0.128 3

71.4 18.4 8.33 0.170 0.125 4

70.9 18.7 8.36 0.170 0.127 5

Average 71.5 18.7 8.36 0.'70 0.126 O

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I WESTINGHOUSE CLASS 3 1

SECTION 3.0

.~ l SENSOR PLACEMENT WITHIN THE REACTOR CAVITY l I

The two multiple foil irradiation capsules (BA and BB) and the nine stainless steel gradient wires were located in the annular gap between the pressure vessel mirror insulation and the sacrificial shield. The multiple foil sensor '

sets and gradient wires were supported in the reactor cavity by means of 0.040 inch diameter type 304 stainless steel wire. The sections of attachment wire opposite the beltline region of the reactor core served as the gradient ,

dosimetry discussed in Section 2.0.

The placement of the dosimetry strings within the Brunswick Unit 2 reactor cavity is illustrated in Figures 3-1 and 3-2 and tabulated in some detail in Table 3-1. In Figure 3-1, a plan view of the azimuthal locations of the nine I strings is shown while, in Figure 3-2, the axial extent of each of the sensor

~~

set strings is depicted along with the locations of the two multiple foil l capsules. Details of the sensor positions and attachment locations are provided in Table 3-1. The identification numbers given in Table 3-1 correlate with the sensor set summary listed in Table 2-2.

Attachment of the stainless steel support wire at the upper axial elevation was by means of #10 stainless steel sheet metal screws secured to the pressure !

vessel mirror insulation at either the feedwater nozzle elevation or the steamline elevation. No permanent attachment of the lower part of the stainless steel wire was possible due to access limitations. Therefore, steel weights were affixed to each wire to provide sufficient tension to allow the strings to hang plumb. Since only natural convection occurs in this area of the reactor cavity, the attachment wires were expected to remain plumb throughout the irradiation.

. In placing the sensor strings, exial spacing was assured by referencing fiducial markings on each wire to either the top of the sacrificial shield or 3-1

WEST!NGHOUSE CLASS 3 Plant North Sacrificial Shield o

0 15 0 M/ 22.5

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s.~. -

o f:" '

45 0

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288 C/L RPV -- 90

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' 225 Mirror Insulation 180

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FIGURE 3-1 AZIMUTHAL LOCATIONS OF SENSOR STRINGS f

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WESTINGHOUSE CLASS 3 y

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7 Sacrificial .* . -), s. ; -

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Shield x

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y y

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Insulation f p. .. A '>

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Nozzle -~~

Screw x 2

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e e FIGURE 3-2 AXIAL LOCATION OF MULTIPLE FOIL SENSOR SETS 1

1 3-3 1

WESTINGHOUSE CLASS 3 TABLE 3-1 ,

MULTIPLE FOIL SENSOR SET AND GRADIENT WIRE ,

LOCATIONS WITHIN THE REACTOR CAVITY String Azimuthal Attachment Elevation No. Location (deg.) Capsule ID Relative to Vessel "0" 1 0 BA 51' 1/4" l

I 2 45 BB. 38' - 5" 3 15 Gradient 51' 1/4" 4 22.5 Gradient 51' 1/4" 5 75 Gradient 49' 1/4" 6- 135 Gradient 37' 1/2" 7 225 Gradient 38' 1/2" ,,

8 288 Gradient 49' 1/2" 315 Gradient 37' 3/4" ,,

9 NOTE: Capsules are centered at 25' 1/2" relative to vessel "0".

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l WESTINGHOUSE CLASS 3

(

l

. to the elevation of the centerline of the steamlines. Azimuthal positioning-was determined by direct measurement relative to the location of either the 1

~~

feedwater lines or the main steamlines. The azimuthal and axial reference f dimensions listed in Table 3-1 are relative to the Brunswick Unit 2 0 = 0.0 and z = 0.0 as specified in General Electric drawings VPF 2478-22 and 197R602. It may be noted that the axial midplane of the active reactor core {

is located at 23' 1/2" above vessel zero. Thus, for the cycle 6 irradiation, the O' and 45' multiple foil sensor sets were positioned 2 ft.

above the core midplane. This axial elevation corresponded to the expected maximum in the axial neutron flux distribution. :

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1 WESTINGHOUSE CLASS 3 I

l b' SECTION 4.0 l

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ANALYTICAL METHODOLOGY

'In performing the fast neutron exposure evaluations for the Brunswick Unit 2 reactor pressure vessel, two distinct sets of neutron transport calculations L were carried out. The first computation, a two dimensional analysis in R,0 geometry, was utilized to provide baseline data characteristic of the average i

exposure for cycle 6. This baseline two-dimensional calculation provided the 3 l relative neutron energy spectra at key locations within the pressure vessel l

and the reactor cavity as well as relative radial distributions of exposure parameters and dosimeter reaction rates through the vessel wall.

l The neutron spectral information was necessary for the interpretation of the dosimetry withdrawn from the reactor cavity following the cycle 6 irradiation as well as for the-determination of important exposure parameter ratics; i.e.,

j dPa/e(E > 1.0 MeV), within the pressure vessel geometry. l l

l The second set of analyses performed for the Brunswick reactor consisted of a l series of one-dimensional calculations designed to investigate the impact of

)

several operational and geometric parameters on the interpretation of cavity i Q

r;eutron dosimetry and on the extrapolation of those measured results to the pressure vessel proper. In particular the one-dimensional studies were used to investigate the effect of void fraction in the reactor core, of the amount of steel present in the jet pump region of the downcomer, and of the amount of iron present in the sacrificial shield. Clearly each of these variables has a marked impact on the calculated neutron flux magnitude at the inner diameter and through the thickness of the pressure vessel shell. However, with cavity neutron dosimetry available to establish absolute magnitude, the calculations are required only to produce accurate relative neutron energy spectra and i

relative radial distributions. The intent of the one-dimensional studies was to confirm the insensitivity of this relative data to the key operating )

. parameters noted above.

-~

i The combination of the analyses described in this section and the measurement I results discussed in Section 5.0 allowed an evaluation of the absolute exposure 4-1

WESTINGHOUSE CLASS 3 of the pressure vessel and of the exposure gradients within the vessel wall with a minimum uncertainty. The uncertainties associated with these projections are discussed in further detail in Section 6.0 of this report. ,,

4.1 TWO-DIMENSIONAL ANALYSIS A plan view of the Brunswick Unit 2 reactor geometry at the core midplane is shown in Figure 4-1. Since the reactor exhibits 1/8th core symmetry only a O'

- 45' sector is depicted. In addition to the core, reactor internals, pressure vessel and sacrificial shield, the geometry also includes representations of the jet pumps located internal to the pressure vessel. '

Also shown in Figure 4-1 are the azmuthal locations of the reactor cavity neutron dosimetry strings.

Development of the geometric model shown in Figure 4-1 made use of nominal design dimensions throughout. The jet pumps were modelled as homogeneous ,l zones characteristic of the pump geometry opposite the axial midplane of the .

reactor core. For the purposes of this baseline neutron transport ..

calculation, the coolant void fraction associated with the reactor core was assumed to be 30%.

The small aluminum capsules containing the multiple foil sensor sets that were positioned along the O' and 45* axial traverses were not explicitly modeled in the two-dimensional transport analysis. These capsules were designed to minimize perturbations in the fast neutron flux and, thus, to provide free I field data at the measurement locations. This assumption of an insignificant perturbation in the fast neutron field due to the presence of the aluminum holders is supported by the close agreement between iron gradient measurements and iron measurements from foils contained within the irradiation capsules.

These iron measurements are reported in Section 5.0.

1 l The two-dimensional calculation for the reactor model shown in Figure 4-1 was ,,

carried out in R,0 geometry using the 00T two-dimensional discrete ordinates code [1] and the SAILOR cross-section library (2]. The SAILOR library is a 47 ,;

group ENDFB-IV based data set produced specifically for light water reactor applications. In the Brunswick Unit 2 analysis anisotropic scattering was 4-2 l

-WESTINGHOUSE CLASS 3

.- FIGURE 4-1 l

.~- BRUNSWICK UNIT 2 R,0 REACTOR GEOMETRY l

l

,1 1

i I l

l l.

t 1

- L l I*' -

.3 ;

L '

or a .

1

J i

~

Insulation '3

. Pressure Vessel !

Jet Pumps

, s.

. . Shroud f g , f o,.

A N .*

L n'*.

l .

\ % b O f J

1 V

. - s Core / s l

g, 4 A1 ag,

. / N A .',,

~, ,

fa n 4 f ,j i a 0 50 100 150 200 250 ' 350 400 Radius (cm) e #

t.

l- 4-3

(

l L _ _ - _ - _ _ _ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _

WESTINGHOUSE CLASS 3 treated with a P expansion of the cross-sections and the angular ;

3 -

l rder of angular quadrature.

discretization was modelled with an S8

~

Plant specific reactor core power distributions for use in the neutron transport analyses were provided by Carolina Power and Light Company. For the cycle 6, R,0 analyses the assemblywise power distribution representative of mid cycle was employed. Relative assembly powers for this point in the cycle 6 burnup history are illustrated in Figure 4-2.

In addition to the cycle 6 relative assembly power fractions, axial distributions of both core void fraction and relative power density were also supplieu by Carolina Power and Light Company. An examination of these data indicated that the use of an axial peaking factor of 1.20 was consistent with the assumption of a 30% core void fraction in the peripheral fuel assemblies '

and, further that this calculational model should be representative of an -

axial location near core midplane. Thus, an axial peaking factor of 1.20 was ~~

utilized to adjust the R,0 analytical results to core midplane.

In order to compute the absolute magnitude of neutron radiation levels within the pressure vessel geometry, a design basis core power level must be chosen.

For the Brunswick Unit 2 reactor, all data both analytical and experimental were referenced to a reactor core power level of 2436 MWt.

4.2 ONE-DIMENSIONAL ANALYSIS The one-dimensional discrete ordinates neutron transport calculations were carried out using the ANISN SN transport code [3]. In performing the one-dimensional analyses, the reactor core was modeled as an equivalent volume cylinder with the remaining reactor and cavity components modelled as cylindrical annuli as illustrated in Figure 4-1. The presence of jet pumps was approximated by including stainless steel as a homogeneous mixture in the downcomer region. All calculations were carried out in 47 neutron energy .

groups using the SAILOR library with a P3 scattering cross-section approximation and an S8 rder of angular quadrature. ,,

, )

l l

4-4

l WESTINGHOUSE CLASS 3 f FIGURE.4-2

. BRUNSWICK UNIT 2 RADIAL POWER DISTRIBUTION

/

/ j

/ !

.442 l

.726 .671 .340 l

1.067 .910 .875 .485

' l 1.217 1.158 1.085 .987 .532 !

1.178 1.149 1.230 1.125 1.057 .911 .578

.794 1.116 1.276 1.027 .980 1.091 .970 .805 .388

.780 .834 1.224 1.234 1.019 .972 1.146 1.090 .948 .535 l 1.273 1.175 1.247 1.136 1.326 1.232 1.228 1.152 1.129 .982 .560 i 1.360 1.243 1 364 1.258 1.301

. 1.177 1.236 1.127 1.472 1.124 .996 .583

. 887 1.290 1.384 1.119 1.060 1.216 1.258 .862 .791 1.077 1.129 .991 .597 e 4 4-5

WESTINGHOUSE CLASS 3

.e To complete the parameter evaluation of core void fraction, stainless stee l in l

the jet pump regien, and iron content in the sacrificial shield the following analytical cases were run. _

Core Void Downcomer SS Iron in Shield Case ID %

0.0 0.0 0.0 A

30.0 0.0 0.0 B

60.0 0.0 0.0 C

0.0 0.0

-D 90.0 30.0 10.0 0.0 E

30.0 20.0 0.0 F

30.0 30.0 0.0 G

30.0 0.0 15.0 H

I 30.0 0.0 30.0 ,}

From an interpretive viewpoint, the key outputs from these one-dimensional ,

computations are the neutron spectra in the reactor cavity that are used in the derivation of exposure parameters (v(E > 1.0 MeV) and dPa/sec ) from the measured reaction rates and the ratios of exposure parameters at the vessel inner radius to those at the cavity dosimetry locations that are employed to relate cavity measurements to the exposure of the vessel itself.

Absolute comparisons of calculation and measurement were not an intended goal of these one-dimensional studies.

6 4-6

l WESTINGHOUSE CLASS 3 y SECTION 5.0

. EVALUATION OF CYCLE 6 CAVITY DOSIMETRY 1

1 l

l 5.1 MEASUREMENT RESULTS l Following irradiation during cycle 6 (10/84 through 11/85), the multiple foil j sensor sets and gradient wires were removed from the reactor cavity and i shipped to Westinghouse for analysis. Analysis of all radiometric foils and gradient wires was performed at the Westinghouse Analytical Services Laboratory while the evaluation of the solid state track recorders (SSTRs) was carried out at the Westinghouse R&D Center.

.. , The specific activity of each of the radiometric monitors was determined using

.. established ASTM procedures. Following sample preparation and weighing, the

. activity of each monitor vas determined by means of a lithium-drifted l .. germanium, Ge(Li), gamma spectrometer. In the case of the multiple foil sensor sets contained in irradiation capsules BA and BB, the analysis was performed by direct counting of each of the high purity foils. For the nine l stainless steel gradient wires, individual monitors were obtained by cutting the wires into a series of segments to provide data points at approximately l one foot intervals over an axial span encompassing +7 feet relative to the l

reactor core midplane. I i

Following preparation of the mica discs, all SSTRs were scanned either 1

manually or with the Westinghouse R&D Automated Track Scanner to determine the number of fissions that occurred during the course of the irradiation.

1 Examination of the SSTRs revealed that in all cases the fission tracks were confined to a 0.250" diameter area corresponding to the active area of the l'.. fissionable. deposit. The edges of the active area were sharply defined and the drop-off in track density at the edges indicated excellent signal to

'.- background ratios for the measurements. Deposit uniformities were consistent with previous experience and presented no difficulties for track scanning, t

5-1

__ j

WEST 1NGHOUSE CLASS 3 Since the fissionable material deposits associated with the SSTRs were intended to be suitable for reuse, they were also examined for possible damage during exposure and handling. No evidence of damage was observed. Those .

deposits with measurable decay rates were subjected to mass analysis to check for possible loss of deposit material during the course of the reactor exposure. These mass analyses and the corresponding HEDL masses are given in Table 5-1. In all cases, no evidence of mass loss (or miscalibration) was evident. On the basis of the mass s'ampling of 6 of the 12 deposits exposed at Brunswick, it is recommended that all of the deposits are suitable for reuse.

The irradiation history of the Brunswick Unit 2 reactor during cycle 6 is given in Table 5-2. In Table 5-2, the reference power level for the reactor core is taken to be 2436 MWt and the decay time for each monthly irradiation period has been referenced to the counting date of the dosimeters. The irradiation history data was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report" for the applicable period. }

Results of the radiometric evaluations of the multiple foil sensor sets ,

contained in irradiation capsules BA and BB are presented in Table 5-3 and evaluations of the gradient wires are given in Tables 5-4 through a-12. In each table data are listed in terms of measured specific activity as well as irradiation history corrected full power reaction rates. The monitor specific activities are referenced to 12:00 noon on March 10, 1986 and are represented as disintegrations per second (dPS) per unit mass of dosimeter, either foil or wire. Reaction rates referenced to 2436 MWt are presented in terms of

- reactions per second (RPS) per target atom. Results of the analysis of the twelve SSTRs contained in capsules BA and BB are given in Table 5-13. Here the measured data is presented in terms of fissions per target atom integrated over the cycle 6 irradiation period as well as in terms of full power reaction rates normalized to 2436 MWt.

e e 5-2 I

WESTINGHOUSE CLASS 3 i A TABLE 5-1 1

FISSIONABLE DEPOSIT MASS CALIBRATION CHECKS Mass Prior to Mass After Ratio Deposit Exposure (7/84) Exposure (3/86) After/

Label Isotope (nanograms) (nanograms) Before

-3

1230 239Pu(236Pu) 1.77x10 1.71x10-3 (+3.0%)* 0.966

-3

1231 239Pu(236Pu) 2.52x10 2.52x10-3 ( 3.0%)* 1.000 239 -1

1321 Pu 1.68x10 1.66x10~1 ( 0.59%) 0.988 239 -1

1323 Pu 1.93x10 1.92x10~1 ( 0.93%) 0.995 237 l #1297 Np 9.72 9.68 (+0.48%)** 0.996 237

. #1298 Np 9.46 9.45 (10.87%)** 0.999 Average 0.991 1 0.012

Spectral corrections and decay corrections were made using HEDL data of 6/85.

l ** Alpha decay rates required corrections for 238 Pu ingrowth during the neutron exposure.

l l

a, 5-3

i WESTINGHOUSE CLASS 3-l TABLE 5-2 ', ;

IRRADIATION HISTORY OF CYCLE 6 REACTOR CAVITY- -

NEUTRON FLUX MONITORS Irradiation Decay Time Time

( Irradiation Pavg Pavo Pref (days) -(days)

Period (MWt) 73 .030 7 495 10/84 628 .258 30 465 11/84 960 .394 31 434 12/84 2297 .943 31 403 1/85 2353 .966 28 375 -

2/85 -

1410 .579 31 344 3/85 4/85 1705 .700 30 314 283

'}

I 5/85 2392 .982 31 ~~

2231 .916 30 253 6/85 2105 .864 31 222 7/85 2273 .933 31 191 8/85 1812 .744 30 161 9/85 1113 .457 31 130 10/85 1907 .783 30 100 11/85 NOTES: 1 - Reference core power = 2436 MWt 2 - Total. irradiation time = 0.791 EFPY 3 - Decay time is referenced to the counting date of the flux monitors (3/10/85) e W e 5-4

l i lI ll 7 7 5 7 7 8 8 9 5 1 I 1 1 1

) - - - - - - 1 - I 1-m 0 0 0 0 0 0 0 0 0 n o 1 1 1 1 1 1 1 1 1 o t i e a x x x x x x x x x B t t / _

B c a S 8 8 2 4 2 4 6 2 4 _

a R P 4 3 3 3 2 6 5 0 4 e e R _

l R ( 2 2 3 2 3 5 5 4 2 u

s p _

a c y)

C i t m f i g 6 8 5 0 2 0 2 9 7 0 i v / 9 7 0 5 8 2 0 7 4 4 _

c i S 3 1 1 1 0 6 8 7 2 8 e t P 5 5 3 6 5 1 1 1 1 p c d 1 6 1 S A ( _

0 P l T T _

i D - -

oI 2 2 2 2 2 2 S F B 2 2 2 D D D D D 0 -

T B B B 8 C C C C C C .

E _

S _

R _

O S 7 7 5 7 I 8 0 9 5 _

N 1 1 1 1 I 1 I I 1 E ) - - - - - - - - -

S m 0 0 0 0 0 0 0 0 0 n o 1 1 1 1 1 1 1 1 1 l 6 o t l i e a x x x x x x x x x 3 0 E A t t /

F L B c a S 6 8 1 5 1 4 4 9 3 S C a R P 3 2 5 2 8 2 3 4 7 S E Y e e R A L C l R ( 1 1 2 1 1 3 3 2 1 L P u C 3 I G s

- T N p E 5 L I a c y) 5 S U R C i t m -

U E M U f i g 0 8 9 0 9 0 2 7 3 0 5 O L D i v / 5 7 6 5 2 3 3 6 5 9 H B N c i S 9 7 0 1 7 6 0 0 1 3 G A I D e t P 2 2 2 2 2 4 1 1 8 N T E p c d 1 3 I S T S A ( .

T E A d S I I e E T D r W I A e V R E F v I R l S S o T I i D - - c C o I 1 1 1 I 1 1 A F A 1 1 1 D D D D D D m B B B B C C C C C C u D i E m R d U a S c A M E 0 s M 1 i 4 4 1 0 4 8 6 0 0 0 5 5 g6 5 S 4 4 6 6 r n

o n n A o n o c c o o o MM C M C S S C C t i ) i t

c

)

p p

)

r, )

r,

)

p p p p

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a,

)

r, n

o a , , n , , , .

m e n n ( n n n( n ( n (n n R ( (9 ( ( ( ( e 4 4 0 9 4 8 6 6 3 9 h 5 5 1 S 5 S 4 4 6 S t e e g o e i i i u o F F A C F N T T C C t a

h t

l s a e i ) t r d o e

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a

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l / / / d i e e g o e i i i u o C o F F A C F N T T C C (

F *

,I lll 'll ll

WESTINGHOUSE CLASS 3 TABLE 5-4 '

MEASURED ACTIVITIES IN STAINLESS STEEL GRADIENT WIRES - TRAVERSE I (0')

Axial

Position Activity (dPS/gm) Reaction kate (RPS/ atom) 54 D0 60 p,54(n,P) Ni58(n,P) cod 9(n ,1r )

(ft) Mn Co Co 6.5 550 830 2450 3.53 (-18)** 5.19 (-18) 2.05 (-15) 5.5 1140 1550 3360 7.31(-18) 9.70 (-18) 2.81 (-15) 4.5 1420 2030 3820 9.10 (-18) 1.27 (-17) 3.19(-15) 3.5 1790 2460 3450 1.15(-17) 1.54 (-17) 2.88 (-15) 2.5 1950 2640 4120 1.25(-17) 1.65 (-17) 3.44 (-15) 1.5 2040 2830 2840 1.31 (-17) 1.77 (-17) 2.37 (-15) ,l 0.5 2290 3120 4050 1.47 (-17) 1.95 (-17) 3.38 (-15)

-0.5 2260 3110 4460 1.45(-17) 1.95 (-17) 3.72 (-15) ,,

-1.5 2040 2780 3540 1.31(-17) 1.74 (-17) 2.96(-15)

-2.5 2010 2760 2920 1.29(-17) 1.73 (-17) 2.44 (-15)

-3.17 1740 2440 4550 1.12 (-17) 1.53 (-17) 3.80 (-15) l

Axial dimensions are referenced to the active core midplane.

-18

- Read 3.53 (-18) as 3.53 x 10 5-6

{-

WESTINGHOUSE CLASS 3-l ,'- TABLE 5-5 1

5' MEASURED ACTIVITIES IN STAINLESS STEEL GRADIENT WIRES - TRAVERSE 2 (45')

l Axial

Position Activity (dPS/gm) Reaction Rate (RPS/ atom)

D4 58 60 cow (n,r)

(ft) Mn Co Co p,54(n,P) Ni58(n,P) l 6.5 1120 1620 1910 7.18 (-18)** 1.01 (-17) 1.59 (-15) 5.5 1990 2760 2570 1.28(-17) 1.73 (-17) 2.15 (-15) 4.5 2820 3760 3070 1.81 (-17) 2.35 (-17) 2.56 (-15) 3.5 . 3490 4580 3500 2.24 (-17) 2.87 (-17) 2.92 (-15)

. 2.5 3520 4860 3930 2.26(-17) 3.04-(-17) 3.28 (-15)'

1.5 3870 5310 4760 2.48 (-17) 3.32 (-17) 3.97 (-15) 2.37(-17) 3.13(-17)

O.5 3690 5010 4870 4.07 (-15) ]

-0.5 3870 4970 5120 2.48(-17) 3.11-(-17) 4.28 (-15)

-2.5 3660 4730 4950 2.35(-17) 2.96 (-17) 4.13 (-15)

-3.5 3020 3890 4520 1.94 (-17) 2.43 (-17) 3.77(-15)

-4.5 2490- 3180 3740 1.60(-17) 1.99 (-17) 3.12 (-15) !

-5.5 1860 2460 2940 ' t9 (-17)

- 1.55 (-17) 2.46 (-15)

-6.5 1080 1470 1830 6.92 (-18) 9.20(-18) 1.53 (-15)

-7.5 540 750 1090 3.46(-18) 4.69(-18) 9.10(-16)

Axial dimensions are referenced to the active core midplane.

-18

- Read 7.18 (-18) as 7.18 x 10 e

9 .

e 5-7

i WESTINGHOUSE CLASS 3 TABLE 5-6 MEASURED ACTIVITIES IN STAINLESS STEEL GRADIENT ,,

WIRES-TRAVERSE 3(15*)

Axial

Position Activity (dPS/gm) Reaction Rate (RPS/ atom) 54 58 60 p,54(n,P) Ni58(n,P) Co59(n,r)

(ft) Mn Co Co 6.5 850 1200 1620 5.45 (-18)** 7.51 (-18) 1.35 (-15) 5.5 1400 1980 2070 8.98 (-18) 1.24 (-17) 1.73 (-15) 4.5 2040 2890 2530 1.31 (-17) 1.81 (-17) 2.11 (-15) 3.5 2410 3360 2830 1.55(-17) 2.10 (-17) 2.36(-15) 2.5 2710 3700 3280 1.74 (-17). 2.31(-17) 2.74 (-15) 1.5 2830 3850 3820 1.81(-17) 2.41(-17) 3.19 (-15) 0.5 2730 3710 3970 1.75 (-17) 2.32(-17) 3.32(-15) '}

-0.5 2700 3690 3980 1.73(-17) 2.31(-17) 3.32(-15)

-2.5 2680 3630 3770 1.72(-17) 2.27(-17) 3.15 (-15)

-3.5 2220 2970 3350 1.42 (-17) 1.86(-17) 2.80(-15)

-4.5 1910 2550 2990 1.22 (-17) 1.60(-17) 2.50(-15)

-5.5 1540 2070 2580 9.87 (-18) 1.30(-17) 2.15 (-15)

-6.5 910 1220 1820 5.83 (-18) 7.63(-18) 1.52 (-15)

-7.5 460 640 1120 2.95 (-18) 4.00 (-18) 9.35(-16)

Axial dimensions are referenced to the active core midplane.

-18

- Read 5.45 (-18) as 5.45 x 10 5-8

WESTINGHOUSE CLASS 3

g. TABLE 5-7

{- MEASURED ACTIVITIES IN STAINLESS STEEL GRADIENT WIRES - TRAVERSE 4 (22.5') i Axial

Position Activity (dPS/gm) , Reaction Rate (RPS/ atom) 54 58 60 (ft) Mn Cc Co p,54(n,P) Ni58(n,P) Co59(n,r) 6.5 820 1130 2900 5.26 (-18)** 7.07 (-18) 2.42 (-15) 5.5 1360 1900 3950 8.72 (-18) 1.19(-17) 3.30 (-15) 4.5 1770 2500 4350 1.13(-17) 1.56 (-17) 3.63 (-15) 3.5 2260 3070 3240 1.45(-17) 1.92 (-17) 2.71 (-15)

. 2.5 2440 3350 3700 1.56 (-17) 2.10 (-17) 3.09 (-15)

.- 1.5 2620 3470 6220 1.68(-17) 2.17 (-17) 5.19 (-15) 0.5 ~2870 3740 6970 1.84(-17) 2.34 (-17) 5.82 (-15)

-0.5 2500 3390 6670 1.60 (-l'.') 2.12 (-17) 5.57 (-15)

-1.5 2340 3100 6170 1.50(-17) 1.94 (-17) 5.15(-15)

-2.5 2170 2900 578C 1.39(-17) 1.81 (-17) 4.83 (-15)

-3.5 1660 2290 4750 1.06 (-17) 1.43(-17) 3.97 (-15)

-4.5 1320 1670 3060 8.46(-18) 1.04 (-17) 2.56 (-15)

-5.5 810 1050 1620 5.19(-18) 6.57 (-18) 1.35 (-15)

-6.5 360 490 1160 2.31(-18) 3.07 (-18) 9.69 (-16)

Axial dimensions are referenced to the active core midplane.

-18

- Read 5.45 (-18) as 5.45 x 10

'a .

5-9 L_

WESTINGHOUSE CLASS 3 TABLE 5-8 ,

. 3 MEASURED ACTIVITIES IN STAINLESS STEEL GRADIENT I WIRES - TRAVERSE 5 (75')

Axial

Activity (dPS/cm) Reaction Rate (RPS/ atom)

Position cod 9(n,r) 64 S8 60 peD4(n.P) NiUO(n,P)

Co (ft) Mn Co 670 970 1630 4.30 (-18)** 6.07 (-18) 1.36 (-15) 6.5 1220 1750 2010 7.82 (-18) 1.09(-17) 1.68 (-15) 5.5 1700 2240 2370 1.09 (-17) 1.40(-17) 1.98 (-15) 4.5 2040 2910 2660 1.31 (-17) 1.82 (-17) 2.22 (-15) 3.5 2330 3250 3000 1.49 (-17) 2.03(-17) 2.51 (-15) 2.5 ~

2410 3400 3360 1.55 (-17) 2.13 (-17) 2.81 (-15) 1.5 ~'

2670 3670 3780 1.71 (-17) 2.30(-17) 3.16 (-15) 0.5 ,

3220 3720 1.52 (-17) 2.01 (-17) 3.11 (-15)

-0.5 2370 ,,

3270 3740 1.58 (-17) 2.05(-17) 3.12 (-15)

-1.5 2470 2820 3420 1.34 (-17) 1.76(-17) 2.86 (-15)

-2.5 2090 2360 2980 1.13 (-17) 1.48(-17) 2.49 (-15)

-3.5 1760 1910 2360 8.91 (-18) 1.19 (-17) 1.97 (-15)

-4.5 1390 1150 1580 5.45 (-18) 7.19 (-18) 1.32 (-15)

-5.5 850 440 610 980 2.82 (-18) 3.82(-18) 8.18 (-16)

-6.5

Axial dimensions are referenced to the active core midplane. .

-18 .

- Read 4.30 (-18) as 4.30 x 10

- s' 5-10

i WESTINGHOUSE CLASS 3 l

.l- TABLE 5-9 MEASURED ACTIVITIES IN STAINLESS STEEL GRADIENT WIRES - TRAVERSE 6 (135')

l l

1 Axial

Position Activity (dPS/gm) Reaction Rate (RPS/ atom) l 54 D0 60 (ft) Mn Co Co 7,54(n,F) Ni58(n,P) cod 9(n,r) 6.5 920 1290 1900 5.90 (-18)** 8.07 (-18) 1.59(-15) 5.5 1650 2410 2460 1.06 (-17) 1.51 (-17) 2.05(-15) 4.5 2430 3410 3110 1.56 (-17) 2.13(-17) 2.60(-15) 3.5 3140 4330 3310 2.01 (-17) 2.71(-17) 2.76(-15) ;

. 2.5 3500 4800 3620 2.24 (-17) 3.00 (-17) 3.02 (-15)

. . 1.5 3690 5010 4040 2.37 (-17) 3.13(-17) 3.37(-15) ,

0.5 3890 5200 4300 2.49 ( '.7) 3.25(-17) 3.59(-15) l

-0.5 3540 4680 4300 2.27 (-17) 2.93(-17) 3.59(-15)

-1.5 3430 4300 4100 2.20 (-17) 2.69(-17) 3.42(-15)

-2.5 3080 3940 3940 1.97 (-17) 2.47 (-17) 3.29(-15)

-3.5' 2350 3060 3290 1.51 (-17) 1.91(-17) 2.75(-15)

-4.5 1830 2450 2760 1.17 (-17) 1.53 (-17) 2.30(-15)

-5.5 1190 1530 1970 7.62 (-18) 9.57(-18) 1.65(-15)

-6.5 600 780 1180 3.85 (-18) 4.88 (-18) 9.85(-16)

,

Axial dimensions are referenced to the active core midplane.

-18

- Read 5.90 (-18) as 5.90 x 10 l

1.

e 4-l l

l 5-11 i

WEST 8NGHOUSE CLASS 3 TABLE 5-10 ,

MEASURED ACTIVITIES IN STAINLESS STEEL GRADIENT .

WIRES - TRAVERSE 7 (225*)

Axial

Activity (dPS/gm) Reaction Rate (RPS/ atom)

Position D4 58 60 p.,54(n,P) Ni38(n,P) Co59(n,r)

(ft) Mn Co Co 6.5 1130 1580 2480 7.24 (-18)** 9.89 (-18) 2.07 (-15) .

5.5 1510 2120 2900 9.68 (-18) 1.33 (-17) 2.42 (-15) 4.5 2140 3010 3210 1.37 (-17) 1.88 (-17) 2.68(-15) 3.5 2710 3720 3250 1.74 (-17) 2.33(-17) 2.71 (-15)

?.5 2730 3690 4230 1.75(-17) 2.31(-17) 3.53 (-15) 1.5 2790 3930 7260 1.79(-17) 2.46 (-17) 6.06 (-15) [

0.5 3090 4200 7790 1.98 (-17) 2.63(-17) 6.51 (-15)

-0.5 2960 3910 7430 1.90 (-17) 2.45(-17) 6.20 (-15) ,,

-1.5 2810 3650 6900 1.80 (-17) 2.28(-17) 5.76 (-15)

-2.5 2620 3520 6580 1.68(-17) 2.20(-17) 5.49 ( 15)

-3.5 2010 2630 5380 1.29 (-17) 1.65(-17) 4.49 (-15) 1950 3410 9.68 (-18) 1.22 (-17) 2.85 (-15;

-4.5 1510

-5.5 920 1210 1830 5.90 (-18) 7.57 (-18) 1.53 (-15) ;

-

Axial dimensions are referenced to the active core midplane.

-10

- Read 7.24 (-18) as 7.24 x 10

.: l

O 5-12

WEST!NGHOUSE CLASS 3 i TABLE 5-11 MEASURED ACTIVITIES IN STAINLESS STEEL GRADIENT WIRES - TRAVERSE 8 (288')

Axial

Position Activity (dPS/gm) Reaction Rate (RPS/ atom)

D4 58 60 (ft) Mn Co Co 7,54(n,P) Ni58(n,P) CoM(n,r) )

6.5 720 1000 2240 4.62 (-18)** 6.26 (-18) '1.87 (-15) 5.5 1330 1860 2890 8.53(-18) 1.16 (-17) 2.41(-15) 4.5 1770 2530 3280 1.13 (-17) 1.58(-17) 2.74 (-15) 3.5 2220 3070 3100 1.42 (-17) 1.92(-17) 2.59 (-15) 2.5 2660 3570 3440 1.71(-17) 2.23(-17) 2.87 (-15) 1.5 2690 3680 4520 1.72(-17) 2.30(-17) 3.77 (-15) 0.5 2770 3770 5180 1.78 (-17) 2.36(-17) 4.33(-15)

-0.5 2930 3740 5210 1.88(-17) 2.34(-17) 4.35 (-15)- l

-1.5 2610 3530 4620 1.67(-17) 2.21(-17) 3.86 (-15) ;

-2.5 2380 3110 3840 1.53 (-17) 1.95(-17) 3.21 (-15)

-3.5 2140 2670 2970 1.37(-17) 1.67(-17) 2.48 (-15)

-4.5 1560 2000 2320 1.00(-17) 1.25(-17) 1.94 (-15) .)

-5.5 960 1270 1690 6.15(-18) 7.95(-18) 1.41 (-15) !

-6.5 440 60') 1050 2.82 (-18) 3.75(-18) 8.77 (-16)

Axial dimensions are referenced to the active core midplane.

-18

- Read 4.62 (-18) as 4.62 x 10 5-13

WESTINGHOUSE CLASS 3 TABLE 5-12 ,

MEASURED ACTIVITIES IN STAINLESS STEEL GRADIENT WIRES - TRAVERSE 9 (315*)

Axial

Position Activity (dPS/gm) Reaction Rate (RPS/ atom) 54 58 60 p,54(n,P) Ni58(n,P) Co59(n,r)

(ft) Mn Co Co 6.5 1270 1780 2440 8.14 (-18)** 1.11 (-17) 2.04 (-15) 5.5 2060 2860 3010 1.32 (-17) 1.79 (-17) 2.51 (-15) 4.5 2980 4080 3590 1.91 (-17) 2.55 (-17) 3.00 (-15) 3.5 3480 4820 3760 2.23(-17) 3.02 (-17) 3.14 (-15) 2.5 3680 .5020 4080 2.36(-17) 3.14 (-17) 3.41 (-15) '

1.5 3970 5330 4650 2.55(-17) 3.33 (-17) 3.88 (-15) ..

O.5 3930 5210 4870 2.52(-17) 3.26 (-17) 4.07 (-15)

-0.5 3800 5020 5100 2.44 (-17) 3.14 (-17) 4.26 (-15) ..

l

-1.5 3680 4650 5030 2.36 (-17) 2.91 (-17) 4.20(-15)

-2.5 2970 3850 4310 1.90 (-17) 2.41 (-17) 3.60(-15)

-3.5 2450 3160 3690 1.57 (-17) 1.98 (-17) 3.08 (-15)

-4.5 1820 2360 2760 1.17 (-17) 1.48(-17) 2.30(-15)

-5.5 1010 1320 1780 6.48 (-18) 8.26 (-18) 1.49(-15)

-6.5 510 700 1130 3.27 (-18) 4.38 (-18) 9.44 (-16) 4

Axial dimensions are referenced to the active core midplane.

-18

- Read 8.14 (-18) as 8.14 x 10 D

s a 5-14 L

1 l WEST 1NGHOUSE CLASS 3

.- . TABLE 5-13 l

i TOTAL FISSIONS NEASURED IN SOLID STATE TRACK -

RECORDERS IRRADIATED DURING CYCLE 6 l Capsule BA Capsule BB Deposit Fissions Deposit Fissions Reaction ID Atom RPS/ Atom ID Atom RPS/ Atom 1

-14 2.58x10 -14 U235(n,f) ~7 ~7 1149 4.98x10 2.00x10 1151 6.43x10 Pu239(n,f) 1230 6.14x10-6 2.46x10 -13 1231 7.78x10

-6 3.12x10

-13 U235(n,f)* -7 -15 ~7 -15 1206 1.85x10 7.41x10 1148 2.44x10 9.77x10 Pu239(n,f)* -7 -15 ~7 -14

, 1321 1.85x10 7.41x10 1323 2.69x10 1.08x10

~9 -17 -16 U238(n,f)* 3.61x10 ~9

.. 885 1.93x10 7.73x10 878- 1.45x10

-8 -16 2.20x10 -8 8.81x10 -16.

Np237(n,f)* 1297 1.31x10 5.25x10 1298

Denotes that the SSTR is cadmium covered. '

l.

..n e 4 5-15

WESTINGHOUSE CLASS 3-5.2 DERIVATION OF NEUTRON EXPOSURE PARAMETERS Having the measured reaction rates presented in Section 5.1 and the caculated ,'

neutron energy spectra at the measurement locations, the derivation of the pertinent neutron exposure parameters proceeded as follows. From the two-dimensional R,0 multigroup transport calculation, spectrum averaged reaction cross-sections were defined relative to a 1.0 MeV threshold as j"og(E) # (E) dE 3=

4 I " # (E) dE 1.0 MeV where: og = The spectrum average cross-section for the ith dosimetry reaction.

o$(E) = The multigroup dosimetry reaction cross-sections for the ith reaction from the SAILOR library.

(E) = The multigroup neutron energy spectrum at the dosimeter location from the R,0 transport calculation.

i These spectrum averaged cross-sections were then used with the measured reaction rates to determine the average fast neutron flux (E > 1.0 MeV) at the dosimetry locations based on the following relationship.

- n (Eq. 5-1)

(E>1.0MeV)=fIi=1R /H $ $

1 the measured reaction rate for the ith dosimetry reaction where: R$=

3= the spectrum averaged reaction cross-section for the ^

4 ith dosimetry reaction .. ,

n = the number of foil reactions at the sensor set locations 5-16 i

WEST 1NGH0VSE CLASS 3 <

J. Having the derived threshold fast neutron flux (E > 1.0 MeV) at each dosimetry l

l location, the derived neutron exposures in terms of iron displacements per atom were determined as follows.

d5 = i (E > 1.0 MeV) (, (E (Eq. 5-2)

.0 Mev) )

1 1

, where: d5 = the derived iron displacement rate at the dosimeter I location l p(E > 1.0 MeV) = the derived fast neutron flux (E > 1.0 MeV) at the l l dosimeter location 1

f(E > MeV)

= the calculated dPa to fast neutron flux ratio at the dosimeter location.

. Appropriate spectrum averaged reaction cross sections and dPa/$ (E > 1.0 MeV) h ,. ratios for use in Equations 5-1 and 5-2 are summarized in Table 5-14 for the l multiple foil sensor set locations at O' and 45' as well as for each of the i

gradient were traverses. As can be seen from Table 5-14 the variations in l L neutron spectrum as a function of azmuthal position in the reactor cavity are quite small with changes of less than 6% in the spectrum averaged reaction cross-sections observed over the entire azmuthal span of the cavity.

I Using the spectrum averaged reaction cross-sections listed in Table 5-14 along with the measured sensor reaction rates given in Section 5.1, Equations 5-1 and 5-2 were employed to derive fast neutron exposure parameters at the dosimeter locations within the reactor cavity. The derived fast neutron fi m j (E = > 1.0 MeV) and iron atom displacement rate at the multiple foil sensor 1

locations are summarized in Table 5-15. The results of the evaluation of the stainless steel gradient wires are illustrated in Figures 5-1 through 5-18.

4 l

ea s l

1 5-17

WESTINGHOUSE CLASS 3 L

TABLE 5-14 SPECTRUM AVERAGED REACTION CROSS-SECTIONS AND dPa/ FLUX RATIOS WITHIN THE BRUNSWICK UNIT 2 REACTOR CAVITY ,

3 (barns) 0* 15.(a) 288* 22.5* 45.(b)

REACTION Cu63(n,a)Co60 0.00182 0.00181 0.00178 0.00178 0.00172 Ti46(n,P)Sc46 0.0240 0.0240 0.0237 0.0239 0.0233 Fe54(n,P)Mn54 0.110 0.110 0.109 0.110 0.109 NiS8(n,P)CoS8 0.139 0.140 0.139 0.140 0.138 U238(n,f) 0.390 0.392 0.389 0.391 0.389 .

NP237(n,f) 3.34 3.28 3.24 3.28 3.21 --

-21 -21 -21 -21 -21 2.22x10 2.20x10 2.22x10 2.15x10 dPa/$(E>1.0 MeV) 2.26x10 (a) data are also applicable at the 75* location (b) data are also applicable at the 135*, 225*, and 315* locations e O 5-18

WEST!NGHOUSE CLASS 3

;. TABLE 5-15 l l-

SUMMARY

OF REACTOR CAVITY NEUTRON DOSIMETRY EVALUATIONS l AT THE MULTIPLE FOIL SENSOR SET LOCATIONS .

1 REACTION RATE e (E > 1.0 MeV) 2 REACTION (RPS/ NUCLEUS) (n/cm - see)

O' CAVITY LOCATION Cu63(n,c)Co60 2.49 x 10 -19 1.37 x 10 8 Ti46(n.p)Sc46 3.29 x 10 -18 1.37 x 10 8 l Fe54(n,p)Mn54 1.29 x 10 -17 1.17 x 10 8 NiS8(np)CoS8 1.81 x 10 -17 1.30 x 10 8 ;

U238(n,f) 6.97 x 10 ~17* 1.79 x 10 8

- . NP237(n,f) 5.25 x 10 -16 1.57 x 10 8 J

. 1 i(E>1.0MeV) 1.43 x 10 8 3.23 x10 -13 i

dPa/Sec 45' CAVITT LOCATION Cu63(n,a)Co60 4.02 x 10 ~19 2.34 x 108 Ti46(n.p)Sc46 5.60 x 10 -18 2.20 x 10 8 Fe54(n,p)Mn54 2.40 x 10 -17 2.20 x 108 NiS8(n p)CoS8 3.22 x 10 -17 2.33 x 10 8 U238(n f) 1.30 x 10 -16* 3.34 x 10 8 NP237(n.f) 8.81 x 10 -16 2.74 x 10 8 i(E>1.0Mev) 2.57 x 10 8 dPa/Sec 5.53 x 10 -13

Measured values have been adjusted to account for photofission occuring durir the irradiation period.

5-19

FIGURE 5-1 FAST NEUTRON FLUX (E > 1.0 MeV) DERIVED FROM THE Fe54 (n, p) Mn 54 REACTION TRAVERSE 1 - 0

.. 1 9

.10 '

I 9 I B !

' i l 7

~ ~ '

'= l~~~ I 6

~

22 ~~ 'NF5' ~~i

.= :_::: a 2:= ; :. .. ..::: :. . :: 3.. 3..; __:3 , 5 . _.: 5 y : 3 4 .: : :. :::: - .- -::

- - - - ---- - - - - ~ - -

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j .

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-4 -3 -2 -1 0 3 4 5 6 7

-7 -6 -5 1 2 DISTANCE FROM CORE MIDPLANE (ft.)

5-20

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FIGURE 5-3 FAST NEUTRON FLUX (E > 1.0 MeV) DERIVED FROM THE Fe54 (n,p) Mn 54 REACTION TRAVERSE 3 - 15

~.

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-7 -6 -5 -4 -3 -2 -1 0 1 2 3 DISTANCE FROM CORE MIDPLANE (ft.)

5-22

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L C D8NO

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i FIGURE 5-10 FAST NEUTRON FLUX (E > 1.0 MeV) DERIVED

. FROM THE Ni58(n,p)Co58 REACTION i

TRAVERSE 1 - 0 1

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5-29

. I FIGURE 5-11 l FAST NEUTRON FLUX (E > 1.0 MeV) DERIVED

{

FROM THE Ni58 (n.p) Co58 REACTION i TRAVERSE 2 - 45 4

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l FROM THE Ni58(n,p)Co58 REACTION TRAVERSE 6 - 135 ,

1 9

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-7 -6 DISTANCE FROM CORE MIDPLANE (ft.)

5-34

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FIGURE 5-18 FAST NEUTRON FLUX (E > 1.0 MeV) DERIVED l ,.

FROM THE NiS8 (n.p) CoS8 REACTION l TRAVERSE 9 - 315 1 \

l e 1

J l

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-7 -6 -5 -4 -3 -2 -1 0 1 2 3 4 5 6 7 DISTANCE FROM CORE MIDPLANE (ft.)

5-37

WESTINGHOUSE CLASS 3 Based on the data presented in Table 5-15 the peak fast neutron fluxes at the ~~

8 ~

O' and 45' reactor cavity positions were determined to be 1.43 x 10 n/cm2 -sec and 2.57 x 108 n/cm2 -sec, respectively. Corresponding values .

-13 -13 -

for the iron displacement rates were 3.23 x 10 dPa/sec. and 5.53 x 10 dPa/see at the O' and 45'. positions. Recall that these measurements derived from.the multiple foil sensor sets represent data characteristic of an axial location 2 feet above core midplane at the azimuthal minimum (0') and maximum (45') neutron exposure positions.

The gradient data provided in Figures 5-1 through 5-18 are useful in the establishment of axial and azimuthal distributions over the beltline region of the reactor. The flux levels provided in Figures 5-1 through 5-9 were based on the Fe54 (n,P) Mn 54 measurements from the stainless steel wires while the flux values depicted in Figures 5-10 through 5-18 were derived from the NiS8 (n,P) CoS8 reaction. Both sets of neutron flux information show '

consistent results for both the azimuthal and axial traverses. Furthermore, 54 as illustrated on Figures 5-1 and 5-2 for the Fe reaction and on Figures }

5-10 and 5-11 for the Ni reaction the results of the gradient wire S8 evaluations were in full agreement with the iron and nickel measurements from the multiple foil sensor sets.

One further point of interest regarding the results of the gradient evaluations is illustrated by a comparison of the measurements obtained in each of the four reactor quadrants.

- ~The following derived neutron flux values were obtained at an axial elevation of +0.5 ft. relative to the core midplane along the 45', 135', 225*, and 315*

axial traverses. These locations represent the four symmetric maxima in the azimuthal neutron flux distribution.

m 0@

5-38

WESTINGHOUSE CLASS 3

. # (E > 1.0 MeV)

Fe54 (n,P) Mn54 NiS8 (n,P) CoS8

~

45' 2.15 x 10 88 2.25 x 10 88 135* 2.25 x 10 8 2.32 x 10 8 225* 1.80 x 10 1.90 x 10 315* 2.26 x 10 8 2.35 x 10 8 An examination of this data indicates that both the iron and nickel measurements exhibit symmetric behavior at the 45*, 135*, and 315' positions, but that the 225' measurements are approximately 20% lower than expected.

This apparent non-symmetric behavior may be due either to operational variables such as core power distribution differences or water density variations or, on the other hand, may be caused by differences in as-built vs.

design dimansions for the reactor internals, reactor vessel, or sacrificial

. shield. In any event this non-symmetry was observed over the entire 225'

. . axial traverse; and whether the behavior may be due to operational variables or to physical makeup of the reactor could be determined only through further measurements.

i 0

W mN 5-39

1 WESTINGHOUSE CLASS 3 L ;. SECTION 6.0 e DERIVATION OF PRESSURE VESSEL EXPOSURE In order to davelop accurate neutron exposure profiles at the inner diameter and through the thickness of the pressure vessel wall, the experimental data described in Section 5.0 must be combined with analytically determined spatial gradients obtained from the calculations discussed in Section 4.0. For the l Brunswick Unit 2 cycle 6 evaluation, spatial gradients of fast neutron flux (E > 1.0 MeV) and dPa were taken from the results of the two-dimensional R, e calculation and used with reactor cavity measurements to develop projections of exposure rate gradients within the vessel wall from the following relationship.

'-~

'R,0 (E > 1.0 MeV) = ed (E > 1.0 MeV) { '8) (Eq. 6-1) d

~~

where: W R,0 (E > MeV) = projected fast neutron flux at position R,0 within the pressure vessel wall ed (E > MeV) = measured fast neutron flux at the multiple foil sensor set locations (0* or 45')

'R,0 = calculated fast neutron flux at position R,0 within the vessel wall I

ed = calculated fast neutron flux at the multiple foil sensor set ]

locations In essence, the usa of Equation 6-1 accepts that the measured value of the

- exposure rate represents the best available neutron flux data for the irradiation period in question and, further assumes that the neutron transport results provide an accurate representation of the spatial gradients that exist between the measurement point and the point of interest within the pressure vessel.

6-1

WEST!NGHOUSE CLASS 3 As written, Equation 6-1 establishes a basis for the determination of the fast ,

neutron exposure internal to the pressure vessel in terms of threshold neutron flux (E > 1.0 MeV). .,

However, the same relationship using the measured dPa values from Section 5.0 and calculated dPa gradients suffices to provide evaluations of that energy dependent exposure parameter throughout the vessel.

6.1 EVALUATION OF CYCLE 6 EXPOSURE From the results of the R, e calculation discussed in Section 4.0 and the multiple foil measurements presented in Section 5.0, the following experimental / analytical comparisons of exposure parameters can be made for the O' and 45* cavity locations.

e (E > 1.0 MeV) ,,

O' 45' .

8 8 Measured 1.43 x 10 2.57 x 10 8 8 Calculated 1.46 x 10 2.29 x 10 C/E 1.02 0.89 Thus, on an absolute baris the reference calculation for fuel cycle 6 agrees with the measurement results within + 10%. Using these normalization ratios (0.98 at O' and 1.12 at 45') and the calculated radial distributions through the vessel wall, best estimates of the fast neutron flux (E > 1.0 MeV) at te O' and 45* azimuthal locations were computed using Equation 6-1. The results of that evaluation are depicted in Figure 6-1. Again, based on calculated dPa/$ ((E > 1.0 MeV) ratios the data shown in Figure 6-1 was converted to displacements per iron atom. The dPa based exposure rates for cycle 6 are illustrated in Figure 6-2. .

O d W

6-2

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WEST 1h'. 4USE CLASS 3

. 6.2 UNCERTAINTIES IN EXPOSURE PROJECTIONS r The various uncertainties and biases associated with the exposure projections for the Brunswick Unit 2 reactor depend on the radiation transport analysts ability to define appropriate neutron source distributions within the reactor core, on the adequacy of the cross-section data base used in the analysis, on the applicability of the geometric modelling of the reactor; and because of the normalization procedure employed in the evaluations, on the accuracy of the neutron flux measurements in the reactor cavity.

To estimate the uncertainties associated with the measured values of neutron flux (E > 1.0 MeV) and iron displacement rates within the reactor cavity the FERRET [4] least squares adjustment code was employed. Based on the FERRET analysis which employed multiple foil reaction rates, dosimetry cross sections and the calculated neutron spectra in the adjustment procedure the estimated

. - lo uncertainties in the cavity measurements were as follows.

la UNCERTAINTY !

0* 45' e (E > 1.0 MeV) i 11% ! 11%

dPa/see i 15% 1 14%

The projection of neutron exposure within the pressure vessel wall has an ;

additional uncertainty associated with the calculated slope through the steel wall. For the Brunswick Unit 2 reactor geometry, this additional uncertainty in exposure projection is estimated to be 'on the order of 5-7% based on the analysis of the PCA benchmark mockup [5).

One further point to note regarding projections to the pressure vessel inner radius is that, since neutron transport calculations tend to slightly over-predict neutron attenuation through steel, calculations normalized to measurements in one reactor cavity would tend to over estimate the flux levels at the pressure vessel inner radius. Thus, errors associated with vessel fluence projections based on cavity data would tend to produce conservative results.

6-5 L _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . .

WEST!NGHOUSE CLASS 3 As stated earlier in this report, a series of one-dimensional sensitivity ,

~~

I studies were performed to verify the adequacy of the reference two-dimensional neutron transport calculation for use in the interpretation of cavity ,

l dosimetry and for the extrapolation of the dosimetry results through the pressure vessel wall. The results of these studies indicated that the use of cavity dosimetry together with a reference transport calculation provides an accurate technique for the determination of axial, azimuthal, and radial exposure gradients within the pressure vessel wall of boiling water reactors.

The approach can yield more accurate results than could be achieved through analysis alone.

1 I

From an interpretive viewpoint, with the exception of the 20% and 30%

downcomer steel cases, the variation.in derived flux at the dosimeter locations was 15% for all cases investigated. Further, the variation in the extrapolated flux at the vessel inner radius was likewise excellent at 16%. }

The 20% and 30% downcomer steel cases showed slightly higher devie.tions from -

I 1

the two-dimensional case, but they both represent unrealistically high steel ,

content in the jet pump region. ,,

4

W 6-6

WESTINGHOUSE CLASS 3

- SECTION 7.0

'.' CONCLUSIONS AND RECOMMENDATIONS Based on extensive fast neutron dosimetry irradiated in the reactor cavity, an evaluation of the exposure gradients through the wall of the Brunswick Unit 2 reactor vessel was completed. The evaluations applicable to fuel cycle 6 were performed in terms of both fast neutron fluence (E > 1.0 MeV) and iron displacements per atom (dPa).

~

Procedures were described that permit the evaluation of any potential future cavity dosimetry using spectrum averaged cross-sections to derive the neutron exposure parameters at the measurement locations. Also, radial distributions

. of neutron fluence (E > 1.0 MeV) and dPa were established to permit

,. projection of the measured exposure parameters through the pressure vessel wall at the 0* and 45* azimuthal locations.

From the results of the cycle 6 evaluations and analytical / experimental comparisons, the following observations were made. j

1. The projected fast neutron flux (E > 1.0 MeV) and iron displacement rates at the pressure vessel clad / base metal interface averaged over cycle 6 operation were 0* 45' 9 (E > 1.0 MeV) 8.98 x 108 + 17% 1.92 x 109 + 17% I dPa/see 1.49 x 10 -12 1 21% 3.11 x 10 -12 1 20%

These values are characteristic of the peak in the axial exposure distribution (~ 2 f t. above core midplane).

l *

~

2. Based on a cycle length of 0.791 effective full power years, the

, integrated exposure of the Brunswick Unit 2 pressure vessel over fuel cycle 6 was 7-1

WEST!NGHOUSE CLASS 3 1

0* 45* l 16 16

(E > 1.0 MeV) 2.24 x 10 4.79 x 10 %

-5 -5 dPa 3.72 x 10 7.76 x 10 These values are also characteristic of the peak in the axial exposure distribution (~ 2 ft. above core midplane).

r

3. Based on the one-dimensional sensitivity studies, interpretation of cavity dosimetry using spectrum averaged reaction cross-sections as well as projection of important exposure parameters through the pressure vessel wall are relatively insensitive to operational variables such as core void fractions and power distribution variations. Therefore, the use of cavity dosimetry in conjunction 1 with a single reference calculation represents a viable accurate '

method to establish pressure vessel exposure. Furthermore, this approach; i.e., normalization to cavity measurements, mitigates the ~l need for modelling complexities in BWR geometries with the attendant ,,

inaccuracies introduced into the transport analyses.

In view of the importance of the cavity dosimetry program in the overall normalization procedure for the pressure vessel exposure projections, it is recommended that the cavity dosimetry program be continued for future cycles.

Long term data from an in place measurement program would provide an accurate data base for use in future 10CFR50 Appendix G updates as well as for any life extension initiatives. The dosimetry, even when analyzed at relatively infrequent intervals, would provide an accurate integration of the pressure vessel exposure including the effects of changing core power distributions and other operational variables. Furthermore, any non-symmetric behavior such as that observed at the 225* azimuthal location would at least be properly accounted for and perhaps resolved by implementation of a long term monitoring program. .

I as es 7-2

i 8

WESTINGHOUSE CLASS 3 i

I In summary, the recommended exposure parameters for the Brunswick Unit 2-

~ pressure vessel based on the'results of the present evaluation applicable to fuel cycle 6 (0.79) EFPY) are as follows.

4 (E > 1.0 MeV) 0* 45'

.I OT 2.2 4 x 10 16 4.79 x 10 16 1/4T 1.61 x 10 16 3.51 x 10 16 1/2T 1.04 x 10 16 2.27 x 10 16 3/4T 6.29 x 10 15 1.30 x 10 15 iT 3.85 x 10 15 7.64 x 10 15 dPa

, O' 45*

OT 3.72 x 10 -5 7.76 x 10 -5 1/4T 2.70 x 10 -5 5.63 x 10 -5 1/2T 1.90 x 10

-5 3.88 x 10 -5 3/4T 1.26 x 10

-5 2.50 x 10 -5 IT 8.56 x 10

-6 1.55 x 10 -5 3

4 4

7-3

L WESTINGHOUSE CLASS 3

?. SECTION 8.0

' b* REFERENCES

1. Soltesz, R. G, et. al., " Nuclear Rocket Shielding Methods, Modification, Updating, and Input Data Preparation - Volume 5 - Two-Dimensional Discrete OrdinatesTransportTechnique",WANL-PR-(LL)-034, August 1970.

2. SAILOR RSIC DATA LIBRARY COLLECTION DLC-76, " Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray,3P , Cross-Section. Library for Light Water Reactors".

l l

3. Soltesz, R. G. et. al., " Nuclear Rocket Shielding Methods, Modification, Updating, and Input Data Preparation - Volume 4 - One-Dimensional Discrete a= Ordinates Transport Technique", WANL-PR-(LL)-034, August 1970.

l'

' ~

4. Schmittroth, F. A., " FERRET Data Analysis Code", HEDL-TME-79-40, September 1979.

5. Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology -

To be published.

I l

MO

\

s

8-1 1

j

ML20247G056

Contents

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2.0 DESCRIPTION

7.0 CONCLUSION

8.0 REFERENCES

SUMMARY

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ML20247G056

Text

2.0 DESCRIPTION

7.0 CONCLUSION

8.0 REFERENCES

SUMMARY

SUMMARY

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