Application of solid state track recorders in light water reactor pressure vessel surveillance dosimetry

0191 -278X/83 $3.00 + .oO Pcrgamon Press Ltd.

NW/. Truck Vol. 1. Nor. i/2. pp. 63 - 77. 1983. Printed in Great Ehtain.

APPLICATION OF SOLID STATE TRACK RECORDERS IN LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY*

Westinghouse

Hanford

F. H. RUDDY, J. H. ROBERTS,R. GOLD and C. C. PRESTON Company, Hanford Engineering Development Laboratory, Richland,

WA 99352,

U.S.A.

Abstract-Advanced dosimetry techniques are being evaluated for surveillance of the pressure vessels of operating reactors. Accurate determination of the neutron fluence received by the pressure vessel can be used to determine the amount of radiation damage and predict the extent of embrittlement of the pressure vessel steel. Solid state track recorder techniques are included among these advanced dosimetry techniques. The development of these techniques under the auspices of the Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program is described.

1. INTRODUCTION

(PSF), a high-fluence metallurgical-dosimetry LWRPVS mock-up (McElroy et al., 1980; McElroy et al., 1982a); and on preparation of advanced SSTR dosimetry capsules for exposures at operating LWRs (McElroy et al., 1982b). These advanced SSTR dosimetry capsules will be used to obtain fission reaction rate information which can be used to derive neutron flux and dose information for the reactor pressure vessel.

THE ACCUMULATION

of neutron dose by the pressure vessel of an operating nuclear power plant results in damage in the form of steel embrittlement. In order to ascertain the safe operating lifetime of the reactor pressure vessel, dosimetric measurements must be made to evaluate the neutron dose to the pressure vessel and relate this dose to cumulative radiation damage. In 1977 the United States Nuclear Regulatory Commission established the Light Water Reactor Pressure Vessel Surveillance (LWR-PVS) Dosimetry Improvement Program. The objective of this program is to improve, standardize, and maintain dosimetry, damage correlation, and the associated reactor analysis procedures used for predicting the integrated effects of neutron exposure to LWR Pressure Vessels and support structures. (McElroy et a/., 1982b). Among the dosimetry methods being developed for LWR-PVS is the use of Solid State Track Recorders (SSTRs). The research under way will result in the validation of a new American Society for Testing and Materials Method for application and analysis of SSTR monitors for reactor vessel and support structure surveillance (Gold, Ruddy and Roberts, 1982a). Present SSTR research has concentrated on physics-dosimetric measurements at the Pool Critical Assembly (PCA) at ORNL, a low power mock-up of several thermal shield-pressure vessel surveillance configurations for LWRs (Kam, 1981); on measurements at the Poolside Facility

*This work

was supported

by the United

States Nuclear

AT THE POOL CRITICAL ASSEMBLY

2. MEASUREMENTS

The PCA was established as a pressure vessel wall benchmark facility to test and validate reactor physics calculations in simple commercial power reactor geometries. The PCA consists of a reactor core and ex-core components that are used to mock up pressure vessel surveillance configurations for LWRs. A pictorial representation of the PCA is shown in Fig. 1 (Kam, 1981). The ex-core components consist of a thermal shield, a pressure vessel simulator, and a void box which is a simulated reactor cavity. The pressure vessel simulator is a block of steel with a thickness similar to that of a reactor pressure vessel. The core and ex-core components are located at a depth of twenty feet in water and an air atmosphere is maintained inside the watertight void box. Experimental access tubes allow insertion of dosimetric measuring devices into locations in the PCA as shown in Fig. 2. Solid state track recorder fission rate measurements have been carried out (Ruddy et al., 1980; Ruddy et al.. 1981;

Regulatory

63

Commission

64

F. H. RUDDY,

J. H. ROBERTS,

R. GOLD and C. C. PRESTON

EXPERIMENTAL ACCESS

TUBES

PRESSURE STIMULATOR

FIG. 1. PCA Pressure

PRESSURE SIMULATOR

vessel wall Benchmark

VESSEL

THERMAL SHIELD ,

facility

PCA CORE

3

NOTES: INSIDE DIAM OF HOLES A-1THROUGH A-8 IS 1.834 IN. 14.6S8cm) INSIDE DIAM OF HOLE 8-115 2.469 IN. (6.271cm) TUBES A-l. A-2, A-3 ARE REMOVABLE DIMENSIONS ARE IN INCHES

FIG. 2. PCA Pressure vessel wall benchmark view; 817 configuration.

facility-plan

LIGHT

WATER

REACTOR

PRESSURE

Ruddy et al., 1982) in locations Al through A7 which are designated respectively as: Al, Thermal Shield Front (TSF); A2, Thermal Shield Back (TSB); A3, Pressure Vessel Front (PVF); A4, one-quarter of the thickness into the block (T/4); AS, one-half thickness (T/2); A6, three-quarter thickness (3T/4); and A7, the void box location (VB). Measurements have been made for the 8/7 configuration where the thermal shield to core face and thermal shield to pressure vessel simulator are approximately 8 cm and 7 cm, respectively (as shown in Fig. 2), and for the 12/13 configuration. Fission rate measurements have been made with l’*Th, *‘U, “*l-l, and “‘Np. Mica SSTRs were placed in contact with deposits of appropriate fissionable materials. For the “*Th thick (or asymptotic) and Z3sU measurements, sources were used. For “‘U, vacuum deposited UF, deposits on Al or stainless steel backings were used. The deposit thicknesses were 5- 100 ug cmm2 l”U on 5 mil AI and 36 ug cm+ ZJ’Np on 0.5 mil Ni, respectively. The deposits, surrounded by two 3/4-in. diameter mica SSTR wafers, were placed between two 10 mil cadmium wafers except in the case of the bare 23sU measurements, where iron wafers were used in the pressure vessel simulator locations and aluminum wafers were used in the void box. This sandwich was in turn centered in an iron capsule using iron spacers in the pressure vessel simulator positions and in an aluminium capsule using aluminium spacers in the void box position. A cadmium liner was used for the cadmium covered measurements, whereas iron or aluminum was used for the 235U bare measurements. In the water locations (TSF, TSB, PVF), the SSTR dosimeters were positioned by fastening to lucite supports with water-tight tape and Cd shielded measurements were not made. Midplane-only measurements were carried out for Z3’U and *“Np, while five axial locations at the 1/4T and 1/2T and three axial locations at the 3/4T and VB locations were sampled for *3ZTh and Z3BU.The exposed mica SSTRs were etched for 90 min in 49.2% HF at 22.7”C and manually scanned using optical microscopy to a statistical accuracy of about 3% (1 o) by at least two independent scanners. Track densities were in the range from 5 x IO’ to 4 x IO’ tracks crnmZ in all cases, resulting in negligible corrections due to background tracks and to track overlap (pileup). The reactor power time history was controlled by bringing the power up to 1% of the planned full

VESSEL

SURVEILLANCE

DOSIMETRY

power for each run, stabilizing at the 1% power level for one minute, and then approaching full power with a period of 20 s. At the end of each one-hour run during which the power was kept constant, the reactor was scrammed. The effective time duration of each run was, therefore, one hour plus the period or 3,630 s. Run-to-Run normalization was accomplished by means of a fission chamber The integrated power values were monitor. determined to a relative uncertainty of +I .5% (lo) and an absolute uncertainty of 4.1% (lo) (Fabry and McGarry, 1981). The measured mid-plane fission rates are listed in Table 1. The absolute accuracies tabulated were obtained by combining in quadrature the sources of error listed in Table 2. The absolute uncertainty contains the 4.1% uncertainty in the absolute core power. In general, these errors range from 2- 5% relative and 4-6% absolute. These fission rates are plotted as a function of radial distance from the core for the 8/7 and 12113 configurations in Figs 3 and 4, respectively. The fission rates for Z”Np, *“Th, and *‘U decrease with increasing radial distance from the core in a pseudoexponential manner as is expected for these threshold reactions. The radial distributions for bare and cadmium-covered 235U reflect the rapidly changing neutron spectrum in the thermal and epithermal energy ranges. The ‘lJU fission rate cadmium ratios are listed in Table 3. In general, the cadmium ratio decreases to close to unity with increasing penetration into the pressure vessel simulator block and then begins to increase as the void box is approached, reflecting the higher proportion of thermal neutrons in the void box. These thermal neutrons are produced in the water beyond the void box. Fission rates measured as a function of axial location are contained in Tables 4 and 5, respectively, and are shown plotted in Figs 5 and 6. The shape of the axial distribution obtained from Mol fission chamber traverses (McGarry and Fabry, 1981) is shown for comparison in both figures. Both the “*Th and lJ8U fission rates were normalized to the fission chamber response at midplane. For both *)*Th and *‘*U, the agreement is consistent within the expected uncertainties. A separate set of experiments were carried out in the PCA 12/13 configuration to establish the radial fission rate distributions for z17Np and “*U. In these cases, deposits were irradiated in all seven radial

*‘Yl- Cd

*“Np “‘Th 215”

*“U -Cd 218”

2”Np *“Th 21s”

*,I”

Isotope

15.1

2.79

2.54 4.65 2.46 2.98 2.61 2.60 3.50 2.67 4.97

4.80 6.22 4.18 5.08 4.86 4.85 5.39 4.90

Position l/4 T % Accuracy (1a) Abs$ Relt

0.951 7.45 0.236 254.0 137.0 0.182 1.23 0.0467

Fission rate* 0.421 4.20 0.100 73.8 65.9 0.0781 0.645 0.0196 9.61 7.70

Fission rate 2.78 5.37 2.45 2.25 3.02 2.60 5.40 2.96 2.70 2.68 4.95 6.12 4.78 4.61 5.11 4.85 6.78 5.05 4.89 4.89

Position l/2 T % Accuracy (1o) Rel Abs

neutron

x 10”)

2.55 4.44 2.44 2.62 2.21 2.60 3.20 2.69 2.78 2.40

4.84 6.05 4.78 4.86 4.68 4.85 5.20 4.92 4.97 4.71

Position 3/4 T % Accuracy (1o) Rel Abs

0.178 2.05 0.0412 36.8 30.7 0.0336 0.346 0.00791 4.80 4.38

Fission rate

fission rates (fissions/atom/core

2.83 2.54 2.60 3.30 2.34 2.99 2.80

10.7 0.0101 0.101 0.00204 1.93 1.44

4.82 4.85 5.26 4.72 5.07 4.96

4.97

Void Box % Accuracy (I a) Rel Abs

0.0109

Fission rate

*All fission rates were corrected to a true core power of 9.75 kW at 10 kW. The *“U and -“‘Th fission rates were measured in November 1978. Corrections to these fission rates (due to slight mechanical modifications of the PCA support in June 1979) have been made using the factors from Table 7.2.1 of Fabry, Minsart, Kam and Baldwin, 1981. In the case of “‘U-bare and ‘“Th, the factors for *‘Yl -Cd and *3aU were used, respectively. The 314 T factors were used to correct the VB fission rates. The z’8U and *“Np fission rates were measured after June 1979. tThe relative accuracies shown are a combination of all uncertainties from sources listed in Table 2. $The absolute accuracies shown are a combination of these sources of uncertainty and the 4.1% uncertainty in the absolute power normalization (Fabry and McGarry, 1981).

8/l 8/l 81-l 8/l S/7 12113 12/13 12113 12113 12113

PCA Configuration

Table 1. PCA Midplane

LIGHT WATER Table

REACTOR 2. Compilation

PRESSURE

VESSEL

SURVEILLANCE

of possible uncertainties measurements

Source

in PCA

SSTR fission

rate

Magnitude

(“0)

of uncertainty

Fission track identification Measurement of amount of area scanned Chemical etching variations Deposit mass: 2’8U asymptotic sensitivity “‘Th asymptotic sensitivity “iU deposit mass “‘Np deposit mass Optical efficiency (applies to “‘U and ““Np only) Track statistics Run-to-run monitor variation of counts systematic dead time uncertainty PCA instruments (when used in place of run-to-run monitor) Time history

817 CONFIGURATIC

-I

DOSIMETRY

67

0.5 0. I 0.1 1.37 I .5

I 3 0.56 1.5 to 3.1 0.2 to 1.3 0.1 to 0.6 1.5 0.1

I

IN

I’

I

1

1

’

’

’

PCA 12/13 CONFlGURATlOr

%

I-

I-

PRESSURE lo-16

VESSEL II %T

/

SIMULATOR I,

I

%T RADIAL

FK.

3. Midplane

BLOCK

I %T

RESSURE

I

VB

‘%T

SIMULATOR II %T RADIAL

POSTION

fission rates vs. radial PCA 8/7 configuration.

VESSEL

,I

position

for the

FIG. 4. Midplane

I

I

BLOCK /

1

%T

II

/

1

VB

POSTION

fission rates vs. radial PCA 12113 configuration.

position

for the

F. H. RUDDY,

J. H. ROBERTS,

Table 3. PCA Configuration 8/7 12/13

Position

l/4

T

1.857 2 0.074

“‘U

Position I.121 I.248

R. GOLD and C. C. PRESTON

fission rate cadmium

I/2

T

-c 0.042 + 0.048

locations simultaneously so that uncertainties from run-to-run power normalization would not enter into the comparison of locations. The extremely wide range of sensitivity of the SSTR method enabled simultaneous measurements from the TSF to the VB positions even though a range of almost a factor or 10’ in absolute fission rate is involved. The experimental results of these radial traverses are contained in Table 6. These data are plotted in Fig. 7 for 13’Np and Fig. 8 for *?J. These fission rates display the pseudologarithmic decrease as a function of distance within the PVS block that is characteristic of threshold reactions. The departure of the “‘Np fission rates in Fig. 7 from linearity in the water locations is due to contributions to the fission rate from sub-threshold fission. The cross section for neutron induced 2)7Np fission shows resonances in the epithermal energy range and the relative number of epithermal neutrons increases as the core is approached. In the case of the “*U data plotted in Fig. 8, a straight line with a slope slightly less than the slope in the PVS is obtained in the water positions. This decrease in slope reflects the decrease in effective neutron cross in the water vs. the PVS. These lines intersect at the PVS - HZ0 boundary. The contribution to the measured fission rate from the 6 ppm *‘W in the Z18U foils is appreciable in the water positions. A 14.6% correction was required in the PVF position and a 30% correction was required at the TSB location. The thermal fission correction resulted in an overall uncertainty of 15% for the TSB *3sU fission rate, and although this point has been plotted in Fig. 8, it has been omitted from Table 6 because of its large uncertainty. In the TSF position, the z3BUfission rate could not be accurately measured even with *W deposits containing as little as 6 ppm *W due to the extremely high thermal to fast The relative neutron ratio at this location. uncertainties (lo) have been obtained by combining the sources of error tabulated in Table 2 in quadrature. Uncertainties in power normalization do not enter into the calculation of the relative

ratios

Position 3/4 T

Void Box

1.201 zk 0.042 1.096 2 0.040

1.339 -t 0.056

uncertainties, since a single run was used for ‘W or “‘Np. To obtain the absolute uncertainties from the relative uncertainties of Table 6, the 4.1% uncertainty in the absolute power normalization must be combined in quadrature with the tabulated values. The absolute uncertainties in these data are generally 5% (lo) or less. The results of the SSTR fission rate measurements may be compared with the fission chamber measurements made at PCA (McGarry and Fabry, 1981). The comparisons for Z’sU and *)‘Np are contained in Tables 7 and 8, respectively. In both cases, the fission chamber results are higher than the SSTR results by about 6%. The overall SSTR to fission chamber ratio for the six “‘Np results is 0.940 -+ 0.031 and for the six ZIW results is 0.940 -1-0.016. The smaller standard deviation for the *W ratio reflects the fact that the SSTR “W fission rates have lower uncertainties than the *3’Np fission rates. Although a 6% discrepancy is within the range of the relative uncertainties of the SSTR and fission chamber measurements, the SSTR to fission chamber ratios are consistently about 0.94 for all of the measurements to date suggesting that there is a systematic bias between the two types of measurements. A possible explanation for this bias is the fact that the void introduced by the fission chamber causes some uncertainty as to the effective position of the fission rate measurement. The fact that the fission chamber measurements are consistently higher would indicate that the fission chamber measurements correspond to a position closer to the core side of the void rather than the assigned central position. The results of additional SSTR measurements in the S/7 and 12/13 configurations will be compared to existing SSTR and fission chamber measurements to attempt to resolve this matter. Also, the fissionable deposits used at PCA have been exposed to the same NBS Standard Neutron Fields that were used to benchmark the fissionable deposits used in the fission chamber measurements. Direct comparisons between

+75 0 -75 -130 + 150 +75 0 - 75 - 130

2.46 2.27 2.46 2.10 2.60 2.73 2.67 2.69 2.69

3.72 4.21 4.67 4.64 3.94

4.85 4.89 4.90 4.94 4.91

4.79 4.69 4.78 4.60

Position I /4 7 % Accuracy (lo) Relt Abs:

18.3 21.6 23.6 23.0

Fission rate* 7.93 9.61 9.99 9.68 9.12 1.56 1.72 1.95 1.85 1.76

2.48 2.41 2.45 2.45 2.46 2.97 2.67 2.96 2.95 2.66

4.81 4.77 4.78 4.77 4.80 5.06 4.89 5.05 5.06 4.89

Position I/2 T Fission % Accuracy (I o) rate Rel Abs

x

10”

0.791 0.173

3.19 4.12 3.78

2.69 2.70

2.46 2.44 2.45

4.92 4.89

4.78 4.78 4.78

Position 3/4 7 Fission % Accuracy (I o) rate Abs Rel

neutron)

0.186 0.204 0.201

0.95 I 1.09 1.06

Fission rate

2.36 2.34 2.35

2.20 2.83 1.96

4.73 4.72 4.13

4.64 4.97 4.53

Void Box % Accuracy (lo) Abs Rel

*All fission rates were corrected to a true core power of 9.75 kW at 10 kW. All *“Th fission rates were measured in November 1978. Corrections to these fissions rates (due to slight mechanical modifications of the PCA support in June 1979) were made using the *“U factors in Table 7.2.1 of Fabry, Minsart, Kam, and Baldwin, 1981. The 3/4 T factors were used to correct the VB fission rates. TThe relative accuracies shown are a combination of all uncertainties from sources listed in Table 2. $The absolute accuracies shown are a combination of these sources of uncertainty and the 4.1% uncertainty in the absolute power normalization (Fabry and McGarry, 1981).

+ 150

a/7

(mm)

Axial Position

8/l 817 8/l 8/7 12113 12/13 12/13 12/13 12113

PCA Configuration

Table 4. PCA >‘?h fission rates (fissions/atom/core

F. H. RUDDY,

70

J. H. ROBERTS,

Table 5. PCA ?‘“U fission PCA Configuration

Axial position

u/7

817 U/7 x/7 U/7

Fission

Position 114 T WI Accuracy

R. GOLD and C. C. PRESTON

rates (fis,ions/atom/core

(la)

(mm)

rate*

Reli

AbsS

Fission rate

+ 150 + 75 0 - 75 ~ 130

1.44 8.78 9.51 9.24 8.50

2.56 2.54 2.54 2.57 2.54

4.82 4.83 4.80 4.83 4.82

3.37 4.03 4.21 3.90 3.51

neutron)

x IO”

Position l/2 T % Accuracy (1 IS) Rd Abs

2.61 2.63 2.78 2.53 2.61

F&ion rate

4.86 4.90 4.95 4.78 4.86

Position 314 T ‘?I Accuracy ( I o) Rd Aba

I .64

2.56 2.55 2.59

I .78 1.75

4.83 4.84 4.85

*All fission rates were corrected to a true core power of 9.75 kW at IO kW and coincide with post-June 1979 PCA specifications. IThe relative accuracies shown are a combination of all uncertainties from sources listed in Table 2. $The absolute accuracies shown are a combination of these sources of uncertainty and the 4.1% uncertainty in the absolute power normalization (Fabry and McGarry, 1981).

=rll’FISSION RATES’ 1.0

c”

IN THE PCA St7 AND CONFIGURATIONS

2%

12113

1.0

if

0.9

0.8

0.7

a i

0

%T. 817

v

'hT.817 Y’T. 817 %T. 12113 HT.12113 %T. 12114

0 0

v 0

0.8

FISSION

0.5

p

NORMALIZATION

P ;

RATES

NORMALIZATIOP

0.9

,4

2 $

FISSION

IN THE PCA St7 CONFIGURATION

0.8

z ? F

0.7

4 I;’

0.8

l

%T

0

'hT %T

0 CHAMBER

FISSION

-

CHAMBER

0.5

-300

-200 DISTANCE

-100

TO MIDPLANE

FIG. 5. Axial l’>Th fission rate distributions 8/7 and l2/13 configurations.

-300

100

0

(MM)

FIG. 6. Axial

for the PCA

Table 6. SSTR fission rates measured

TSF TSB PVF l/4 T I/2 T 3/4 T VB *Distance

12.0 23.8 29.7 39.5 44.7 SO. I 59. I

from inner face of core aluminium

in the PCA

-100

x x x x x x

1.01 x

rate neutron)

2’“”

IO-‘O lr 3.2% IO-” + 3.2% IO-” + 3.2% IO-” ? 3.5% IO-” -t 5.4% IO-” +- 3.2%

6.75 1.82 7.81 3.36

lo-”

1.01 x

simulator

I

100

3.3%

(or window)

x x x x

lo-” IO-” IO-” IO-”

200

(MM)

12/ 13 configuration

“‘NP 8.23 7.78 3.23 1.23 6.45 3.46

0

TO MIDPLANE

>‘*U fission rate distributions 817 configuration.

Fission f:atom/(core

Distance from core (cm)*

Location

-200

DISTANCE

-+ 4.0% 5 2.6% I 2.6% 5 2.6%

IO-” k 2.6%

for the PCA

LIGHT

WATER

REACTOR

RADIAL FISSION RATE DISTRIBUTION 237Np IN THE PCA 12/13 CONFIGURATION I ‘@a

I,

(

PRESSURE

r

I

7.

Radial

I

3. MEASUREMENTS

fission rate distribution PCA 12113 configuration.

for z”Np

in the

RADIAL FISSION RATE DISTRIBUTION FOR 238U IN THE PCA 12/13 CONFIGURATION 10-B

DOSIMETRY

71 fluxes in the

‘*I

i I

FIG.

SURVElLLANCE

the benchmark referenced fission equivalent should also help to resolve the discrepancy measured PCA fission rates.

FOR

I

VESSEL

,

(

,

I

,

I

I

I

-HzO-

30 40 50 NCE FROM CORE Icm)

I

60

i

FIG. 8. Radial fission rate distribution for 2’8U in the PCA 12/13 configuration.

AT THE PSF

Whereas the PCA is used for low fluence dosimetry intercomparisons and verification of transport calculations, dosimetry for high fluence metallurgical studies is carried out at PSF. The PSF at the Oak Ridge Reactor (McElroy ef a/., 1980; McElroy et al., 1982) is being used to irradiate capsules filled with metallurgical test specimens of vessel and support structure steels as well as a variety of dosimetry sensors in a simulated thermal shield, pressure vessel wall, vessel cavity assembly identical to that used at PCA. SSTR capsules have been included with the advanced dosimetry sensor capsules as shown in Fig. 0. The SSTR subcapsules contain fissionable deposits of “‘Np, *18U, and “‘U electroplated onto nickel backing wafers placed in firm contact with natural quartz crystal SSTR as shown in Fig. 10. The required fissionable deposits have mass densities in the picogram to nanogram per square centimeter range and active diameters of less than 3 mm. The fissionable deposit masses used in a typical PSF capsule are listed in Table 7. The extremely low masses required for these deposits result from the upper limit on scannable SSTR track density and the anticipated high neutron fluences in the PSF irradiations. The production and quantification of these deposits required the development of special low mass electroplating and spiking techniques. Even with such low deposit masses, the resultant track densities will be in the range from (2-5) x 10’ tracks cm-‘, requiring the use of high track density scanning methods currently under development (Gold, Roberts and Ruddy, 1982b; Gold et al., 1982b; Gold, Roberts and Ruddy, 1980; Gold, Ruddy and Roberts, 1980). Natural quartz crystal SSTRs were used in the high temperature environs of the PSF because of the higher annealing temperature of this material relative to mica which is used in most SSTR dosimetry (Roberts et a/., 1980; Roberts et al., 1982). SSTR capsules for a separate PSF perturbation experiment were irradiated for a much shorter length of time allowing mica SSTR to be used. In selected locations, Exploratory SSTR capsules were included. These capsules contained quartz SSTRs with deposits of the fissionable isotopes Z14U, *‘OTh, and **6Ra which are

72

F. H. RUDDY, J. H. ROBERTS, R. GOLD and C. C. PRESTON Table 7. Contents Capsule Label Sl s2 s3 s4 S5 S6 s7 S8 s9 SlO

of SSTR dosimetry

*‘J”

Location* 2.26 2.37 2.39 2.43 2.53 2.72 2.75 5.32 1.83 1.98

Surveillance capsule Surveillance capsule Surveillance capsule Surveillance capsule Surface of pressure vessel Surface of pressure vessel T/4 T/2 Behind void box Behind void box

Table 8. SSTR dosimetry Reactor type*

Location

BWR PWR PWR PWR PWR PWR PWR PWR

Browns Ferry (3) (U.S) McCuire (2) (U.S.) Tihange (I) (Belgium) H. B. Robinson (2) (U.S.) Maine Yankee (U.S.) Fort Calhoun (I) (U.S.) Crystal River (3) (U.S.) Oconee (U.S.) *BWR

= Boiling

Water

Reactor;

PWR

capsules

x x x x x x x x x x

capsule

Isotopic

exposed

and Mass Density 238”

IO-” IO-” IO-” IO-” lo-” lo-” IO-” IO-” IO-” IO-“’

exposures

2.76 3.07 3.15 3.55 3.66 3.78 4.31 1.00 2.48 2.59

December, 1978 -August, 1979 February, 1981 -February, 1982 January, l980- January, 1981 August, 1982-(January, 1984) November, 1982 -(April, 1984) Planned for January, 1983 Planned for 1983 Planned for 1983 Water

not conventionally used in SSTR dosimetry in addition to z”U, ?J, and *“Np. The development of these isotopes for SSTR dosimetry will provide information on additional regions of the neutron spectrum. For instance, =16Ra has a threshold of 3 MeV and may be useful in defining the extremely high energy region of the neutron spectrum. A typical Exploratory SSTR Dosimetry Capsule is shown in Fig. 11. SSTR dosimetry packages of design similar to those shown in Figs 10 and 11 can now be made available for use in long term LWR pressure vessel and support structure surveillance at operating reactors. 4. LWR-PVS DOSIMETRY CAPSULES Advanced dosimetry capsules similar in design to those shown in Figure 9 have been prepared for irradiation at several different operating LWRs. A summary of the irradiations carried out or in progress is contained in Table 8. Analysis of the SSTRs and other dosimetry included in the advanced

x x x x x x x x x x

in operating

Time of exposure

= Pressurized

in the PSF

IO-’ 10-p 10-p 10-g 10“ IO-’ 10-q 10-a IO‘* IO-’

(g cm-*) l”Np 2.60 3.83 4.26 4.57 4.60 4.94 5.15 7.93 2.20 6.00

x x x x x x x x x x

lo-” lo-‘0 IO-‘O IO-10 IO-‘O lo-‘0 lo-‘0 lo-‘0 IO-9 IO+

LWRs

Isotope

included

_

*‘*Th Z”“, 131” ‘,‘Np >19Pu *XT,: Z”“, 131”: 211Np’ 21,” >“‘” . *“NP “‘“~ *“” “‘NP ’ 23’Np l”“, “‘U, *“” 7 *‘*” 3 “‘NP *““, *“” ‘“NP

=‘“, 2’8”,’ 237Np

Reactor

capsules coupled with the calibration and calculational experience gained at PCA and PSF will allow more accurate determinations of the radiation doses to the pressure vessels of operating LWRs and allow more precise predictions of the safe operating limits for power plants. 5. CONCLUSIONS In addition to radiometric and SSTR dosimetry sets, helium accumulation fluence monitors, damage monitors are being monitors, and temperature studied. The ideal dosimetry set would monitor neutron fluence, damage, and temperature with as few materials as possible in order to reduce costs and required space. It is hoped that materials such as quartz SSTR and sapphire damage monitors can be developed as multipurpose materials. Sapphire for instance, might be used as a combined fluence, temperature and damage monitor (for example, analyzed for helium accumulation, Np2” fissions, and direct neutron damage).

LIGHT WATER

REACTOR

FIG. 9. Hanford

PRESSURE

Engineering

Development

VESSEL

SURVEILLANCE

Laboratory

advanced dosimetry

DOSIMETRY

capsule.

73

74

F. H. RUDDY,

FIG. 10. Hanford

Engineering

J. H. ROBERTS,

Development

R. GOLD and C. C. PRESTON

Laboratory capsule.

solid state track

recorder

advanced

dosimetry

LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE

DOSIMETRY

FIG. 11. Hanford Engineering Development Laboratory solid state track recorder exploratory dosimetry capsule.

75

LIGHT

WATER

REACTOR

PRESSURE

Continuing research will result in the optimization of dosimetry packages for use in long term surveillance of LWR Pressure Vessels, at operating power plants.

REFERENCES Fabry

A. M. and McGarry E. D. (1981) Run-to-run monitoring and absolute normalization of :he PCA benchmark facility experiments. In L WR Pressure

Vessel Surveillance Dosimetry Improvement Program: PCA Experiments and Blind Test. (Edited 1.3-I - 1.3-6. by McElroy W. N.) pp. NUREG/CR-1861. Fabry A., Minsart G., Kam F. B. K. and Baldwin C. A. (1981) Implications of PCA blind test results. In

L WR Pressure Vessel Surveillance Dosimetry Improvement Program: PCA Experiments and Blind Test. (Edited by McElroy W. N.) p. 7.2-3.

VESSEL

SURVEILLANCE

DOSIMETRY

McElroy W. N. et al. LWR pressure vessel surveillance dosimetry improvement program annual report, Nuclear Regulatory Commission eighth LWR safety research information meeting, Metallurgy and Materials Research Branch, NUREG/CR-1747, pp. 25 - 52. McElroy W. N., Kam F. B. K., Grundl J. A., McCarry E. D, and Fabry A. (1982a) LWR pressure vessel surveillance dosimetry improvement program 1982 annual report. NUREG/CR-2805, Vol. 3, pp. 2-20 - 2-23. McElroy W. N. et al. (1982b) Surveillance dosimetry of operating power plants. In Radiation Metrology

Techniques, Data Bases and Standardization: Proc. Fourth ASTM-Euratom Symposium on Reactor Dosimetry, March 22 - 26, 1982, Gaithersburg, MD (Edited by Kam F. B. K.) pp. 3-44. NUREG/CP-0029 CONF-820321. McGarry E. D. and Fabry A. (1981) Fission chamber measurements. In L WR Pressure Vessel Surveillance

Dosimetry Improvement Program: PCA Experiments and blind test. (Edited by McElroy W. N.) pp.

NUREG/CR-1861. Gold R., Roberts J. H., Preston C. C., McNeece J. P. and Ruddy F. H. (1982) Computer controlled scanning systems for quantitative track measurements. In

23-l-23-24. NUREG/CR-1861. Roberts J. H., Gold R. and Ruddy F. H. (1980) Thermal annealing studies in moscovite and in quartz. In Solid

Radiation Metrology Techniques, Data Bases, and Standardization. Proc. Fourth ASTM-Euratom Symposium on Reactor Dosimetry, March 22-26,

State Nuclear Track Detectors, Proc. 10th International Conference, July 2 - 6, 1979, Lyon, France (Edited by Francois H. et a/.) Pergamon,

1982, Gaithersbura. MD (Edited bv Kam F. B. K.) pp. 281 - 292. NURECXP-0029 CbNF-820321. Gold R., Roberts J. H. and Ruddy F. H. (1982a) Standard

Oxford. pp. 177 - 190. Roberts J. H., Gold R. and Ruddy F. H. (1982) Selected etching and annealing properties of Brazilian quartz crystals for solid state track recorder applications. In

Method for Application and Analysis of Solid State Track Recorder Monitors for Reactor Surveillance. American Society for Testing and Materials Standard E854-81, ASTM 1982 Book of Standards, pp. 1 - 14. Gold R., Roberts J. H. and Ruddy F. H. (1982b) Buffon Needle method of track counting. In Solid State

Nuclear Track Detectors, Proc. I Ith International Conference, September 7 - 12, 1982, Bristol, England. Clapham).

(Edited by P. H. Fowler and V. M. pp. 891- 898. Pergamon Press, Oxford.

Gold R., Roberts J. H. and Ruddy F. H. (1980) Advances in SSTR techniques for dosimetry and radiation damage measurements. In Dosimetry Methods for

Fuels, Cladding and Structural Materials: Proc. Third ASTM-Euratom Symposium on Reactor Dosimetry, October I - 5, 1979, lspra, Italy (Edited by Rottger H.) pp. I l72- 1187. EUR 6813. Gold R., Ruddy F. H. and Roberts J. H. (1980) Applications of solid state track recorders in United States nuclear reactor energy programs. In Solid State Nuclear Track Detectors, Proc. 10th Int. Conf., July 2-6, 1979, Lyon, France (Edited by Francois H. et al.) pp. 533 - 548. Pergamon Press, -Oxford. Kam F. B. K. (1981) Description of experimental facility. In

L WR Pressure Vessel Surieillance Dosim-etry Improvement Program: PCA Experiments and Blind Test (Edited by McElroy W. N.). NUREG/CR-1861.

Solid Slate Nuclear Track Detectors, Proc. Ilth International Conference, September 7 - 12, 1982, Bristol, England. (Edited by Fowler P. H. and Clapham V. M.) pp. 417 -420. Pergamon Press, dxford. Ruddy F. H., Gold R. and Roberts J. H. (1980) Solid state track recorder measurements in the pool critical assembly. In Dosimetry Methods for Fuel, Cladding,

and Structural Materials: Proc. Third ASTMEuratom Symposium on Reactor Dosimeter, October l-5, 1979, lspra, Italy (Edited by Rottger H.) pp. r( 1069- 1075. EUR6813. Ruddy F. H., Gold R. and Roberts J. H. (198 ) Solid state track recorder measurements. In L WR Pressure

Vessel Surveillance’ Dosimetry Improvement Program: PCA Experiments and Blind Test (Edited by McElroy W. N.) pp. 2.5-l -2.5-13. NUREGICR-1861. Ruddy, F. H., Gold R. and Roberts J. H. (1982) Light water reactor pressure vessel neutron spectrometry with solid state track recorders. In Radiation

Metrology Techniques, Data Bases and Standardization, Proc. Fourth ASTM-Euratom Symposium on Reactor Dosimetry, March 22 - 26, 1982, Gaithersburg, MD (Edited by Kam F. B. K.) pp. 293 - 301. NUREGICP-0029 CONF-820321.