Fast-neutron personnel dosimetry with solid-state track recorders

NUCLEAR INSTRUMENTS AND METHODS Ilfl (1974) , 383-399 ; ® NORTH-HOLLAND PUBLISHING CO. FAST-NEUTRON PERSONNEL-DOSIMETRY"WITH SOLID-STATE' TRACK RECORDERS* RAYMOND. GOLD Joint Centerfor Graduate; Study,`Richland,

Washington 99332,'

U .S.A .

ARMANI, GORDON K. RUSCH and THOMAS J . YULE Argonne National Laboratory, Argonne, Illinois 60439,- U.S.A .

ROLAND J.

Received 29 January 1974 Aluminum-covered ,and cadmium-covered Solid-State Track Recorders (SSTR) :consisting of pre-etched mica and asymptotically thick 2350 foils have been used to monitor the fastneutron dose received by personnel at the Zero-Power Reactor (ZPR) facilities (Argonne-East). Neutron spectra and absolute integral measurements were carried out at representative loca-

tions throughout the ZPR working environment. SSTR d~simeters were exposed at the same locations. Doses infcr,üd from the SSTR data are compared to those 'derived from the me,ssn-9 neutron spectra or to doses determined with portable rein ~~êacaq . On this basis, the applicability of the SSTR dosimetry method has been (conservatively) established.

1. Introduction

critical assembly. Consequently, the fast-neutron dose received by personnel working in the ZPR environs must be monitored. At the inception of plutonium cores in the ANL-East ZPR facility (1969), no satisfactory fast-neutron personnel dosimeter was available. Conventional film badges were not acceptable, sing they are almost completely insensitive relative tc allowed weekly personnel doses . While only a limited number of work environmentt ; comparable to the ZPR facilities are in existence today, transition to a national LMFBR power base augurs for a` proliferation of such environments . Moreover, one must also anticipate that future breeder-reactor fuel, fresh as well as spent, will possess even higher >pecific spontaneous-fission activity , and thereby cxa;,erbatc neutron-exposure problems . Faced with these expectations together with more stringent personnel exposure standards, the development of adequate fastneutron personnel dosimetry is imperative. f n fact, recent reductions in recommended personnel exposure levels not only imply reduced tolerance standards, but also° require that personnel be exposed to minimum practical levels. These new standards coupled with significantly, higher radiation levels from plutoniumloaded ZPR cores have created serious operational problems . Comparable problems may exist at many other installations and facilities which gmorate intense sources of high-energy neutrons . Consequently, methods devaed and experience gained in presently available fast-neutron work environments, such as the ZPR facilities, are therefore of both specific and general interest . The present effort has provided a basiF for an ade-

No panacea has been' advanced to date that can meet the requirement, currently demanded in fast-neutron personnel'dosimetry . In fact, problems in fast-neutron dosimetry exist at a sufficient number of levels to render questionable the :very concept of asingle unique solutiont. These problems'are both fundamental and applied, ranging' from the definition of quality factors on the one hand to more practical considerations, such as adequate methods of personnel monitoring in specific fast-neutron environments. A more complete exposition'of these views has already been advanced')_ The present effort, which falls into the latter cutegory, describes both the development ofand experience with a specific fast-neutron dosimetry service. The need for such a sere?ce arose at Argonne National Laboratory (ANL) where plutonium-fueled fastbreeder reactor cores are constructed as part of the national Liquid-Metal Fast-Breeder Reactor (LMFBR) development program. These reactors are operated at very low power, usually less. than a few hundred watts, and are customarily referred to as Zero-Power Reactors (ZPR) . The use of plutonium in these ZPR cores creates a neutron-dose hazard because of neutrons from spontaneous-fission decay (principally from saOPu and other even-even plutonium isotopes) and subsequent multiplication of these neutrons in the subWork performed under the auspices of the U.S . Atomic Energy Commission. t In analogy, it is now generally accepted that a single unique method of fast-neutron spectroscopy does not exist; for example, neutron spectroscopy in fast critical assemblies requires many methods to cover the entire energy region of interest. 383

384

R. GOLD

quate (conservative) fast-neutron dosimetry, service. However,- it must not be confused with a scholarly examination or systematic investigation in the field of fast-neutron personnel dosimetry. Support was not pro.idedbeyond theimmediate necessity ofestablishing a practical method of adequate sensitivity, which could, in turn, be applied to ZPR personnel . A preliminary oral account of this method has already been given'). This presentation revealed amuch broader level of interest than had been realized . The present publication has been written, in part, as a response to this demonstrated interest. The'applicability and specific choice of Solid-State Track Recorder (SSTR) are discussed in section 2 immediately below. Sections 3 and 4contain a description of the ZPR facilities atA fast-neutron measurements therein, respectively. SSTR measurements, data analysis and comparisons with fast-neutron measurements are given in section 5. Section 5 also presents some results generated since theinception'of the SSTR dosimetry service in 1969. The concluding remarks in section 6 attest to the applicability of SSTR dosimetry in the ZPR fast-neutron environment.

etat.

absolute 'record of induced' fissions, consequently absolute neutronfluencereceived by exposedpersonnel can be estimated. In analogy to current practice with film badges, SSTRs arepermanent records, and there fore can be indefinitely stored for future reference. Consequently, when better analysis techniques are developed, it will be a simple matter to re-evaluate exposure doses more accurately . The SSTR method has been put to extensive use (previously to this dosimetry application) in diverse experiments at Argonne requiring absolute fission-rate measurements 3~. Trackformation in recorders such as mica, glass, and polycarbonate resins arises at radiation-damage sites upon exposure of . the SSTR to fission fragments . The transparent dielectric track recorder is subsequently . etched to enlarge these radiation-damage sites . Thesefission tracks (oretch pits) are counted manually wi«I an optical microscope (-200x). Asymptotically thick sources provide the' highest sensitivity. (An asymptotically thick foil has a thickness greater than the, longest fission-fragment range in the foil material .) This technique has been_ used for fission-process experiments with many, heavy elements , ). 2. SSTR dosimeters It is possible to determine the doses received from A neutron dosimeter which could be immediately shielded,.-partially shielded, or bare -ZPR cores by used by ZPR persornel was mandatory. Theoverriding utilizing a dosimeter assembly consisting of two consideration was to have a dosimeter of adequate SSTRs, each having a unique spectral response. Two sensitivity, yet simple enough to be rapidly, andreliably constants typical of thecore andthe SSTRs, alongwith implemented. Dosimeters satisfying these requirements the spectrum of the shielded core and the spectrum of were not available at the inception of plutonium the bare core provide all oftheinformation required to loadings in the ZPR critical assemblies . In fact, low translate SSTR derived data to dose,information. The sensitivity is undoubtedly th,: major limitation of many constants are readily obtained by exposing a SSTR fast-neutron personnel dosimeters in current use. Furthermore, it must be stressed that adequate sensitivity is a necessary (although not sufficient) condition foraccurate dosimetry. In other words, while accurate dosimetry is not possible withouthigh enough sensitivity, having satisfied this sensitivityrequirementwith SSTR or other dosimeters does not imply sufficiency (i.e., does not automatically :guarantee accurate dosimetry) . SSTR possesssufficiently high sensitivity forapplication in fast-neutron personnel dosimetry (approximately 1 trackjcm2 per mrem in typical fast critical-' mblyspectra). Theuse of suitable SSTRs not only overcomes this sensitivity problem, but provides agamma-insensitive, reasonably accurate,simple method of measuring personnel dose. Small SSTR detection gackages can be easily assembled, whicharejust as convenicntty worn as are pocket ionization chambers or Fig. 1.Photograph of the SSTR fast-neutron personnel dosiUm, teto-mover;' the' SSTR provides " a direct meter used at the ZPR facilities (ANL-East).

FAST-NEUTRON PERSONNEL DOSIMETRY

385

dosimeter package to the neutron flux from the bare nent. Approximations' may, however, be useful to core and anotherto theneutron flux from . the shielded. reduce the number of SSTRs required for complex core.Thespectramay be obtained from calculations or fields. from measurements. ` Although this method is, in principle, sufficiently In the case of ZPR facilities, it appears that :for general to encompassthe use. of various types of dosimetry purposes, combinations of these two spectra fissionablematerials andshielding materials (to modify adequately describe the, fields encountered . For situa- the SSTR spectral response), this presentation will tions where two spectra- cannot -be combined to'ade- emphasize the MRS used in the ZPR dosimetey quatelydescribe a resultantfield, it maybe necessary to service, viz. pre-etched mica SSTR together with determine thespectrum of each unique component and asymptotically thick sU foils (93.1 °9aenriched) . Fach to employ an additional SSTR'with asuitable, unique personnel dosimeter consists of two such SSTRs spectral response for each such contributing compo- mounted on a cardboard card, one enclosed in aluminum andthe other in cadmium. A photograph of the dosimeter is shown in fig. 1. This arrangentent caaVes FOIL (eaax,) the SSTR dosimeter to be worn in a manner analorous to the use of film badges. Fig. 2 shows a detailed view of each individual SSTR assembly. Each assembly contains a pre-etched mica SSTR together with an asymptotically thick 1' SU foil, 5 mils thick á,!í3 .1 ®9n enriched) . The twoassemblies on the card differ only in the case material - one `assembly has an aluminum case and the other a cadmium case . The mica is preetched to enable one to distinguish fossil tracks - tracks Fig . 2. Detailed view of the SSTR assembly containing asymp. totically thick -U and pre-etched mica. that hadbeen formed in the mica from thespontaneous 23

11

Fig. 3. Microphotograph of etched fission-fragment tracks in mica. This mica specim.n has been. pre-etched in order to distinguish fossil fission tracks. The large diamond (arrow) is a fossil track.

. R. GOLD et'-t

4

FAST-NEUTRON PERSONNEL" DOSIMETRY

3S+7

fission of trace impurities of uranium . Fig. 3 is aIM Alt of these considerations clearly imply that the microphotograph' of fission-fragment tracks in mica absolute fast-neutron spectrum is a prime requisite in after the mica has been chenticaily':etched . `The; large theanalysis and reduction of SSTR data to absorbed diamond is a fossil track. The smaller diamonds are dose . Consequently,`the next two sections describe the neutron-induced fission tracks that were created after ZPRfacilities and'measurements carried out to delinc the pre-etching. this fast-neutron environment.° The choice of mica SSTR and asymptotically thick uranium foils yields a very high sensitivity indeed, 3. ZPR facility ï0 í9atoms/cm 2 . In addition,-this asymptotic sensiZPR-6 is a split-table type critical' assembly contivity has been accurately measured- (to - 1.5% sisting of one stationary and one moveable table as error) 3).'Bare and cadmium-covered trick`densities are shown in fig . 4. A 45 row by 45 column matrix of 2 used to infer upper limits of the dose received by the square 'stainless-steel tubes 4-ft . lone; is stacked wearer. The actual analysis of track density data can be horizontally on each table. Reactor materials, such its found in section 5. clad plutonium and steel plates are loaded into dnmers In the last few, years other approaches: have been which are inserted from between the halves into the proposed . in personnel neutron dosimetry . An AEC matrix tubes . The control and safety rods are shown task force exists which is charged with developing a attached to the back plates. suitable - personnel dosimeter for fast neutrons' , '). Fig. 5 is atop view of the ZPR-6cell. The table . :" Therehasbeen a considerable effort to developthermo- shown 5 ft. apart, which is the customaryconfjuratie luminescent dosimeters to detect "albedo neutrons" when the critical assembly is not at power. The crossthat have been slowed either by the body of the wearer hatched rectangles represent personnel shields, which or by moderator surrounding the thermoluminescent are used to decrease the, neutron and gamma doses. phosphor. To date no satisfactory design has been These shields are moveable and consist of three seefound"). More recently, interest has shifted to the tionst for each table ; they extend the height of the development of dosimeters which employ SSTRs 10, t t). matrix. The x's indicate positions at which dosimetry These newly proposed dosimeters differ from the measurements were made . These measurements were SSTR dosimeter described herein in two important made 5 ft . abovethe floor, except for those betweenthe respects . Plastic rack-recorder materials are usually table halves which were made at the height of the core used rattler tnan mica because tracks m plastic can be center. The circled x's are locations at which more counted automatically by sparking techniques and detailed studies were carried out and only the results thorium or neptunium replaces 235U as the fissionable obtained at these specific positions are used in the material. subsequent dosimetry analysis. Both of these alternatives had been independently The dosimetry studies carried out in the ZPR-6 considered t2) prior to our choice of asymptotically cell were made with assembly 7, a large, single-zone thick 235U and pre-etched mica SSTR for the fast- loading. The core had a length-to-dianieter ratio close neutron dosimetry service. None of these alternatives to unity and was surrounded by a depleted-uranium matched the simplicity, reliability, and sensitivity of blanket. The core composition was typical of an lids combination. Consequently, this combination LMFBR fueled with plutonium oxide. The fuel volume ffords more accurate dose information Indeed, the fraction was about 33%, the sodium volume fraction ultimate uncertainty in dose depends generally upon about42%, and the steel volume fraction about 20% . errors in the fission cross section, the fast-neutron This assembly had a core volume of approximately spectral data, thequalityfactors, andfinally theinherent 31001. In an actual LMFBR, therewould be additional counting statistics 3) of . the SSTR data . It follcws, steel reflectors and various biological shields. therefore, that the asymptotically thick 23 'J-mica SSTR combination will produce, in principle, dose 4. Absolute neutron measurements information of higher accuracy .* The analysis and interpretation of SSTR data ' An ideal dosimeter would simultaneously possess adequate require adequate knowledge of the neutron environsensitivity and proper energy response. Unfortunately, ideal ment. In this section the techniques used to obtain dosimeters do not, at present, exist . In fact SSTR dosimeters, this information are described and the data presented. and for that matter all other fast-neutron personnel dosimeters in current use, do not possess a simply interpreted response . This characteristic creates additional limitations in the accuracy cfthedosimetrymethod which will become apparent in sequel .

t

Each section is 8' thick and contains (starting from the core face) I' 'oral, I° lead, I ° steel and 6' benelex .

388

R. GOLD ei

The absolute neutron measurements include spectrum mapping with proton-recoil proportional counters, integral measurements with bare and cadmiumcovered HF3 proportional counters and surveys with a portable fast-neutron counter. Dose mappings were made with a neutron rem meter. An oral : account of these neutron measurements hasalready been given' 3). 4.1 . PROTON-RECOIL NEUTRON SPECTRUM MEASUREMENTS

4.1 .1 . 5 chniaaes Proton-recoil proportional counters have been used for some time now to measure in-core fast-neutron spectra in plutoarum-loaded ZPR cores at ANL14). Speciali .md electronics are required, and spectrum measurements axe rather time consuming, usually requiring a minimun. of a hours to obtain one spectrum. The applicability of this method for defining environmental fast-neutron spectra has already been

al .

demonstrated 15). Detectors- used : for these measurements were larger than those normally used . forin-core'. measurements. The proton-recoil spectrum above 100 keV was measured with a 1" diam. methane-filled' detector. possessing: a 3" -sensitive length,' filled to 90 psia. The proton-recoil spectrum below 100keV was measured with a1" diam. hydrogen-filled. detector possessing a6" sensitive length, filled to 80 psia.Fig. 6, shows such a detector and preamplifier mountedin a thin-walled stainless-steel drawer. The drawer was' held in position by alight aluminum stand. The proton-recoil method allows an absolute' determination of neutron flux . To convert the proton-' recoil spectrum to an absolute neutron spectrum, all that is needed is then-pscattering cross section (which is we äi known) and the number of hydrogen atoms in' the sensitive volume of the counter (which can be determined). Extensive investigations have been made of the sources of systematic errors in this type of

ZPR M CELL

Fig. 5, Top view of theZPR-6 cell showing the positions, x, where dosirnetry measurements were conducted .

FAST-NEUTRON PERSONNEL DOSIMETRY'

The relative 'error of these-menmeasurement' surements is about10%'overmostof theenergyrange. The-measured neutron, spectrum can be used to determine he close by applying the energy-dependent flux-to-dose: conversion curve.*; Because the measured spectrum necessarily possesses afinite energy range, it is necessary to extrapolate this spectrum -beyond the upper- and loNvtr-energy bounds of: observation. At lower energies the equivalent IIEspectrum (see section 4.2)was used to determine thecontribution to the total dose from neutrons with energies between the cadmium cut-off and about 5 keV. The measured thermal flux (see section4.2) wasused to determine the thermalcon6.17)

' To obtain theflux-to-dose

conversion curve, linear interpolation was used between values tabulated at various neutron energies in ref. 18 . Some additional approximations are necessary and these arebriefly discussed in section 4.3 .

tribution. For none .of the measured spectra did ci contribution amount to more than 3%, or,the comNd tion to more than 5%, of the total doses The extrapolation at higher energies (above 3 McV) assumct ., spectral shape which was that of : the fission spectrum. The existence-of this °shape at higher 'energies was confirmed by calculation. The absolute intensity o¬tsar. high-energy part of thespectrum wits determined I,I, continuous extrapolation of the- measured spectnam . The relative error associated with this extrapolation may-be ;as high as 30%. However, the contribution of the extrapolated region to thetotal dose neverexoedeO 3% . Thus, more than 92% of the dose arises fram neutrons with energies within the finite sensitivity range of proportional-counter spectroscopy. 4.1 .2. Measured spectra Fig. 7 shows the spectrum observed between tlne

Fig. 6. Proton-recoil proportional counter and preamplifier used for fast-neutron spectral measurements.

R. GOLD et al.

halves with the shields removal.* Typical measurerunts extend from a few keV to several-MeV. The depicted ctror represents only statistical uncertainty. The spectrum is that,of.a degraded fission source. Dips in the spectrum caused by oxygen resonances at about 450 keV and l MeV and by iron resonances at about 30 and 90 kcV are clearly seen.The solid line is the energy-&V.ndent flux-to-dose conversion curve: the ordinate is the number of n/cma s necessary to deliver 5 mrem/hr dose. The fast-neutron environment in theZPR cell evidently extends over an energy region of rapid variatiot: in the dose conversion curve. Thespectrum measured between thehalves when the shields are in place is shown in fig. 8. Fast-neutron intensity is substantially reduced. Thespectral peak has moved to higher energy, which can be attributed to the ability of the higher-energy neutrons to penetrate the shields more easily. Furthermore, - the low-energy tail has become a -more significant spectral component because of the down-scattering in the shield . Fig. 9 shows the measured spectrum at ROOM-1, a position in the cell near the assembly (cf. fig. 5). Although the' spectral'peak has shifted to lower energy, almost all the dose still' comes from highThis particular spectrum was measured with a loading of OPu fuel than all subsequent measurements. lower-content $4 Alimited set of proton-recoil measurements between thehalves 10 I0 with the higher-content a4DFu fuel demonstrated that theshape . of thespectrum didnot change, butthat theintensity increased by ao% .

Fig. 7. Neutron spectrum observed between the halves with the shields vmoved. The solid line is the energy-dependent flux w'- conversion curve, i.e., the neutron flux required to produce , toa, dose of 5 ,cmlhr.

energy neutrons . ; This spectrum is representative of those found in other locations in the; cell, especially ; where the- fluxes were too low to - measure absolute spectrawith the same statistical confidence. To assess the effect of a human torso on the; spectrum, measurements were made with and: without a phantom simulation of the human torso. Fig. 10 shows-the; spectra measured in front, of a,gap in the shields with and without a,phantom present. (During these measurements, the shields on the opposite,core CENTER BETWEEN THE NALVES SHIELDS IN PLACE

- .~ 11,

130

Y W

20

W 10

0 L . .I IOe

I

I IIIII'

10 1

I

1

I I~III~ .

1 ENERGY, k .V

I

I I IIInI

I

1 1

3

Fig. 8 . Neutron spectrum observed between thehalves with the shields in place.

Fig. 9. Neutron spectrum observed at position ROOM-1 .

EAST-NEUTRON PERSONNEL DOSIMETRY face were in place .) The detector was placed on the chest 3f the phantom and is represented by the solid rectangle in fig . 10.' Although the relative error has not beendepicted for these spectra; it is about the same as that shown in . previous spectra. The dose is only slightly changed by the presence of the phantom from 314 mrem/hr without the phantom tc - 340 mrem/hr with the phantom . Scattering from the : phantom substantially increases the number of low-energy neutrons. This result shows that any dosimeter that is sensitive to low-- or : intermediate-energy neutrons must be calibrated with a phantom: 4 .2.' INTEGRAL MEASUREMENTS 4.2.1. Techniques 4 .2.1 a. Boron reaction rates Boron reaction rates were measured with bare and cadmium-covered BF, proportional counters . The counter had a 1 " diam., a 10" sensitive length, and was filled to approximately 40 cm Hg. The boron reaction rates may be used to determine the thermal flux and an equivalent 1/E intermediate-energy flux. Because of uncertainties in detection efficiency, volume of the sensitive region, and filling pressure, it was necessary to calibrate the counter-in a known thermal flux. The count rate for the bare counter, RB , minus the count rate for the cadmium-covered counter, R, is given by

where s is the probability of detecting an ionizing 04nt in the proportional counter. NT is the number of 1eB atoms in the sensitive volume, a,ha is the thermal " °B(n,a) cross section, vu, is the thermal velocity, and z, is the average velocity (which in the case of a N? axwellBóltzmannénergy distribution is 1 .128 Vt . at _20 °Q and ¢,b is thee thermal flux. Calibration was obtained by observing count rates in a thermal flux that had been measured absolutely with a SSTR package containing a natural-uranium foil and a mica track recorder"'). The quantity ::,Va was determined using' eq. (1). The BF3 counter could then be'used to measure ¢, h in other locations . f13 statistical error of the SSTR calibration produced m} uncertainty of roughly 10% in the thermal-neulrou calibration factor ; an accuracy which is sufficient in this dosimetry application. An equivalent' 1/E intermediate-energy flux is determined from the cadmium-covered reaction rate, which is simply R,d =

NT

&

/' 10 keV i

J E..

aa(E) .P(E)dE,

(2)

(1)

where the cadmium cut-off energy, E.d, is assumed to be 0.5 eV, oB(E) is the 1011(n,a) cross section, and ¢(E) is the flux . The 10 keV upper limit is chosen to provide adequate overlap with proton-recoil measurements (see section 4.1). Assuming that aa (E) may be represented by

Fig. 10. Neutron spectra observed in front of a gap in the shields with and without a phantom .

and that O(E) is proportional to 1/E (flux per unit lethargy is constant), the integral is easily evaluated and one may determine for a given R d the intensity of the 1/E flux that would give that rate . Thus, one can obtain an approximate representation of the, intermediate-energy flux in terms of an equivalent I /E flux . This is a useful approximation for determining the dose contribution from neutrons with intermediate energies (less than 10 keV). In using the rate, Rtd, to estimate the intermediate-energy flux, it is important that any contribution to the count rate from higher-energy neutrons (above 10 keV) be small, since the spectrum at higher energies is several orders of magnitude greater than the equivalent 1/E value. However, calculations reveal that any contribution to the cadmium-coveredboron reaction rate from neutrons above 10 keV is less than 5% in all cases . One can choose another detector for the intermediate-energy region, fir example, one using cadmiumcovered 115 U. An equivalent I/E flux would then be

RB-Red

=ENT

ath0Vth V

~h,

aB(E) = 0.611/E¢,

(3)

392

R. GOLD et al.

defined by CR-1 = 4S .ha~h , (4) ~uajaa whereCR is thecadmium ratio, which is theratio of the bare to cadmium-covered fission rate, 0,e is thethermal flux, a,ts is the thermal 235U fission cross section, 0 t ,x the equivalent 1/E flux, and l_ the 235U resonance fission integral for infinite dilution. Unfortunately, 0or these fast spectra more than 50% of the cadmium-covered fission rate may come from neutrons with energies greater than 10 keV. Consequently, this spectral index is of less applicability in thefast-neutron environments which were encountered. However, when there were a significant number of intermediate-energy neutrons, the equivalent 1/E flux determined by these two methods were in good agreement. ó.2.1b. Portable neutron counter A portable fast-neutron detector* was used to determine the fast-neutron flux, which is obviously an important integral quantity because almost all of the dose is induced by fast neutrons . At locations in the ZPR-6 cell where proton-recoil measurements are not available, thefast-neutron flux, together with dose-rate information, can be used to infer the hardness of the spectrum . This detector consists of a BF3 counter surrounded by a paraffin moderator, all enclosed in a cadmium shield. Since the resultingenergy response is not an ideal step function, the accuracy of these flux measurements will depend on the shape of the neutron spectrum . The counter was calibrated at the position between the halves, using the proton-recoil result for the fast-neutron flux (energy greater than 100 keV). From comparisons with proton-recoil measurements at other positions, the error in the fast-neutron flux determined with this counter may be as high as 30%. An independent calibration with a Pu-Be source (average energy about 4.5 MeV) gave a calibration factor about 18% lower. 4.2 .1 c. Portable neutron rem meters

Neutron dose rates were measured with a portable rem meter developed by Andersson and Braun") known as "Snoopy ".t Snoopy is cylindrical in shape and uses a BF3 counter. Polyethylene is . employed around the counter to provide moderation and " Model PNC-1, Eberline Instrument Corp .. Santa Fe, New Mexico. 'p' A.N,PDR-î0 (SNOOPY NP-2), LFE Corp., Tmpelo Div., WA&=,, M busetts :'

absorption in order to obtain an appropriate rem response. One of'-Snoopy's main disadvantages is .its directionality.At 0.5 MeV, the anisotropy ratio(rightangle to on-end response) is as high as 1.4 . In the present set of measurements ; directionality was taken into accountfor measurements between thehalves The measured functional form of thedirectional dependence ' (see fig . 9 of ref. 20) was used with an anisotropy ratio -' of 1.35, which was assumed independent of `energy. Source neutrons were assumed to be uniformly distributed across thecore face ; and a numerical integration was performed. In this manner, the average sensitivity was found to be 0.88 of the right-angle response. For measurements elsewhere in the cell the right-angle response was used, since thedetector axis wasplaced at right angles to the line between the detector and the core center (which is the approximate location of the neutron source). Snoopy wascalibrated with a 2 SZCf source of known intensity. Since thefission spectrum is known, the dose rate could also be determined . Because of the uncertainties introduced by directionality effects, energy response, and calibration errors, an error of at least 30% must be assigned for the 'absolute dose-rate determinations with this portable rem meter. The agreement betweenthe doses inferred from theprotonrecoil results and the portable neutron rem meter were usually within 20%. 4.2.2. Measured integral quanti!ies

The dose rate, thermal flux, equivalent 1/Eflux, and fast flux observed at the positions circled in fig. 5 are given in table 1 (phantom present) and table 2 (no phantom). The dose rates listee in these tables were determined with Snoopy except for locations at which proton-recoil data were available. In thecenter location, the plane of the phantom was either perpendicular (facing one of the halves) or parallel to the reactor axis. For the other measurements, the phantom was lookingat the center of the core and thedetectors were placed near the chest. Thedose rate varies from several 'hundred mrem/hr between the halves with the shields absent to less than t mrem/hr out in the celi with the shields in place. The thermal-neutron flux does not vary significantly throughout thecell, but is afactor of ten higher between the halves. Comparison of tables I and 2 reveals that with a phantom present a high thermal-neutron flux is produced : between the halves due to down-scattering in the phantom . In addition, with shields between the halves slowing down in the benelex also produces a significant number of thermal neutrons . Theintermediate-energy neutronflux between

:393

FAST-NEUTRON PERSONNEL DOSIMETRY TABLE 1 Neutron dosimetrystudies Location

Center Phantom 1 to axis Phantom // to axis ` Phantom :L to:axis shields in place Frontof gap Room no, 1 . Room no . 2 No shields Shields in place Room no . 3

Dose rate mrem/hr

witha phantom.

Thermal flux nlcm2 s

1/Eflux n/cm2 sper unit lethargy

307 431

360 250

70 55

3170 3260

7.4 340 11 .1

104 560 40

5.7 100 8.1 6.4 1.4 5.1

7.4 0.77 3.2

34 11 24

Fast flux- ` Dose rate n/cm2sthermalflux

Dose rate 9/E flux

Dose rata fast flux

0.85 1.7

4.4 7.8

0.097 0.130

93 3880 200

0.07 0.61 0.28

1 .3 3.4 1 .4

0.080 0.098 0.055

130 25 75

0.22 0.07 0.13

1 .2 0.6 0.6

0.057 0.031 0.043

s Neutrons with energy > 100keV. TABLE 2 Neutron dosimetry studies without a phantom. Location

Center No shields Shields in place Front of gap Room no. 1 Room no. 2 No shields Shields in place Room no. 3

Dose rate mrem/hr

460 13 .7 314 11 .7 8.1 6.85 4.2

Thermal flux n/cm2 s

17 95 180 23 21 5.3 17

I/Eflux n/cm2s per unit lethargy

Fast fluxn/cm2 s

Dose rate thermal flux

Dose rate RIEflux

Dose rate

15 6.7 34 6.3

5080 120 2790 200

27 0.14 1 .7 0.51

31 2.0 9.2 1 .8

0.090 0.114 0.112 0.058

5.4 1 .2 4.6

140 23 80

0.38 0.16 0.25

1 .5 0.7 0.9

0.058 0.037 0.053

fast flux

s Neutrons with energy >100 keV.

the halves is also significantly affected by the presence of the phantom. When the shields are absent, the increase due to the phantom is about a factor of four. The number of fast neutrons is more or less independent of the presence of the phantom except for the shielding effects of the phantom. The last three columns in tables 1 and 2 show the total dose rate divided by the thermal flux, I/E flux, and fast-neutron flux, respectively. In this form the data indicate the utility of a dosimeter which would be sensitive to each of these different flux components, i .e., what might be expected if a dosimeter with a thermal, 1/E, or fast response were used. The largest variation occurs in the ratio of the total dose rate to the thermal-

neutronflux, which ranges from 1.7 to 0.07- afactor of 25 . Thereis less variation in the ratio of total dose rate to 1 /E flux, and theleast in tharatio oftotal dose rate to fast-neutron flux, whichvaries from 0.130 to 0.037 - a factor of four. These spectral and integral measurements give a rather complete description of the neutron environment in the ZPR cell . The general variations found in spectra are probably similar to those which might encountered around plutonium processing facilities or in the environment of fast-breeder power reactors . These observations demonstrate that variations in spectral shape occur throughout the cell, as well as with various shielding configurations . It is clear that

394

R. GOLD

changes in the spectral shape exist which significantly alter the total dose per fast neutron . Thus, even if one simply has a (scalar) dosimeter whose response is proportional to thefast-neutron intensity alone, sizeable uncertainties in dose must result. This dilemma can only be resolved thr'ugh systematic research anddevelopment of an ideal dosimeter, which wouldsimultaneously possess adequate sensit:,, ity and proper energy response . 4.3 . DOSE CALCULATION

The discussion which follows may be considered as an aside on problems that are encountered if one wants to exactly determine the dose equivalent from a measured neutron spectrum. In previous sections we assumed that the relationship between flux and dose is easily obtained by employing a tabulated dose conversion curve. This procedure is not elstirely correct and to see why, let us briefly consider huw the dose conversion curve was derived'') . A broad monoenergetic beam of neutrons was assumed to impinge on a cylindrical phantom 60 cm high and 30 cm diam . of tissue equivalent material . Using Monte-Carlo techniques the absorbed dose, D, in various volume elements was obtained and linear energy transfer distributions, LET, generated. The dose equivalent, DE, for each volume was calculated using recommended values of the quality factor, QF. The QF relates DE to D at a given LET. The QF is a non-measurable quantity, being legislated by such groups as the National Council on Radiation Protection and Measurements in light of experimental data on animal and cell survival, etc. Since the relationship between DE and D varies from one volume element to the other, one must choose which volume element to use to determine the DE corresponding to a given flux. The procedure for radiation protection is to use thehighest average qualityfactor, TF. Forneutrons less than 5 MeV, the highest QF occurs near the surface. The point to be made is that for environments in which the beam is not parallel, adifferent dose conversion curve should be used. For neutronenergies less than 5 McV, the parallel beam geometry leads to the lowest neutron flux-to-dose ratio and thus is aconservative value to use. The situationis furtherconfused by the fact that the International Commission on Radiation Units and Measurements has recently defined the dose equivalent index at a pointas the maximum dose equivalent within a 30 cm diam . sphere centered at the point and consisting of tissue-equivalent material"). Thus, the use of a dose conversion curve for a parallel beam-cyllindrieW:' phantom geometry leads to an

et al.

inaccurate determination of the dose equivalent . The difference is probably small, but one' should be aware of it. 5; Analysis and reduction of SSTR data 5.1 . DOSE DETERMINATION FROM TRACK DENSITY SSTR dosimeters provide two observables, namely, bare and cadmium-covered track density, which will be denoted by Te and Td , respectively.* Cadmiumcovered SSTRs respond to both intermediate-energy neutrons, as well as fast neutrons, whereas aluminumcovered (bare) SSTRsalso respondto thermal neutrons, which can dominate the response . It is convenient to introduce the cadmium ratio CR definedby : . CR == Te/Tcd

(5)

as aspectral index, which in I very crude sense provides a measure of the hardness of the spectrum . These quantities have beer. used in two independent approaches relating dose to track density.Oneapproach is based on an analytical model of the neutron environment") ; the alternate approach is empirical. These two approaches complement each other and, at the same time, produce numerical results whichare in good agreement. Consequently, a limited discussion of both approaches is justified . In the analytical model, one generally assumes' the existence of a source of energetic neutrons. Shielding provides both attenuation and moderation of these energetic source neutrons. In addition, neutrons that undergo a sufficient number of scattering events will eventually be thermalized (i.e., so-called room return). An adequate assessment of neutron dose can, therefore, be obtained by arbitrarily decomposing the neutronflux into three components, via., fast (orsource) neutrons, intermediate (or moderated) neutrons, and thermal neutrons. These three flux components will be designated by O f , ¢i and tß,s, respectively . In the ZPR-6 environment, of would be the neutron spectrum arising from unshielded core faces, whereas O; corresponds to the neutron spectrum from fully shielded faces of the critical assembly.In this analytical model it is assumed that the shape of each of these threecomponents remains the same throughout the environment and only the relative intensities of these components may vary . Let Of , ¢ and ¢,n be unit normalized. The dose rate per unit flux, ßf , bi and b,n, corresponding to these " Track density will possess customary units, i .e . tracks/cm-1, throughout.

395

FAST-NEUTRON' PERSONNEL DOSIMETRY

three components canbe written as: ! -

and -

Q(E) Of(E)dE+

(6a)

Q(E) Od (E) dÉ,

(6b)

and where Q(E) is the` flux-to-dose conversion factor, )FF, whichdepends on neutron: energy E. Hence, the total dose rate, b, is given by the linear superposition

Fcd =,I, i af(E). .Pd(E)dE = 1, 

where
rt5 = IfISr+IIrti,+I,hU,r+

where If, Ii, and I,b are the corresponding absolute intensities of #f , ¢i, and,¢,h, respectively. The absolute intensities If , Ii , and Ith are related to absolute fission rates measured by the bare and cadmium-covered SSTR exposures. To this end, let the fission rate per atom of the bare and cadmiumcovered SSTR exposures be given by

(l t b)


al Tod is . .

(12b)

FB -Fod

= TO-Ted,

(12c)

a,

Is« a,h

Fb`= Tb/(tsm),

(8a)

where a,, is the thermal ZssU fission crass section . Dividing eq. (12c) by eq . (l2b) and introducing the definition: of the cadmiumratio, viz., eq. (5), provides a relation between ori and CR of theform

Fcd = Tod/(tsw) .

(8b)

where

and

af = 1+J-SCR,

(13)

respectively.-Here s. is the known asymptotic sensiS = (II/Ith)(
396 b=

R. GOLD ei al.

r(( 1+S)15r- SJ)t+
+

(_~t

\~ai> .

time r, eq. (19` can also be written in the form

< .>,)

-

Al S CR] ts,'

far)r/

(20)

Dssm(T,+T2) = DmTit(Tl)+Dssra(T2).

It follows immediately from eq. (20) that DsSTR is a linear function of track density*. There exist only two independent' variables which can enter into this linearfunction. Ifone chooses Ted and CR as the independent variables, then the expression for Dsy-TR must be of the form

T~d +

(16) Evaluation of the coefficients in eq. (16) can be carried out in the manner described above . For the environs of the ZPR-6 plutonium-loaded fast critical assembly, one finds the numerical form 15 = [0.50-0.015(CR)] (TId lt),

DssTR =

(17)

or D = bt = [0,50-0,015(CR)] Kd . (18) In aq. (18), Dwa denotes dose (in mrem units) explicitly deduced from SSTR measurements, i. e., wherein T~d is the observed cadmium-covered track density and CR is the observed cadmium ratio. Let us now examine the alternate approach, which is strictly empirical. The dose representation function R must satisfy the following distributive rule: DssTR(tt+tz) = ÜSSTR(tt)+DssrR(tz),

(21)

[a - b (CR)] Tm 1

where a and b are constants. In this approach, the constants a and b are empirically determined by comparing measured doses and track densities at representative locations . In this manner, the numerical values a = 0.' and b = 0.011 were determined, in close agreement with numerical results provided by the analytical model, cf. eq . (18):

(19)

from the obvious fact that doses received under the same conditions in consecutive time intervals tt and t z must be additive . Since track density T (either bare or cadmium-covered) is proportional to exposure which follows

5.2. COMPARISON OF SSTR AND CALCULA'T'ED DOSE A summary ofSSTR dosimetry measurements can be found in table 3. These exposures were conducted with a phantom at representative positions in the ZPR-6 cell. SSTR dosimeters were placed on the chest of the phantom and the phantom was oriented in a manner identical with that used for the absolute neutron measurements . The first three columns contain track data obtained from these exposures . The fourth column ° This is true provided that Dssrn lies in the class of continuous functions.

TABLE 3

SSTR studies with a phantom. Location

Center Phantom 1 to axis Phantom /Í to axis Phantom .L to axis shields in place Front of gap Room no. 1 Room no. 2 No shield Shields in place

Room no. 3

TO tracks/ cm2 hr

CR

5107 3159

820 725

6.2 4.4

1424

7584 561

44 1251 76

33 6.1 7.4

430 121 348

53 11 30

8.1 11 12

& ToW dose rate, D, is that given in table 1 . b

T.3/t

tracks/ cm2 hr

-10 .54-0.011((:~R)l (Tealt) .

l) ., mrem/hr

_D

(Telt)

15 (Tedlt)

307 431

0.060 0.136

0.37 0.59

1 .2 0.82

7.4 340 11 .1

0.005 0.045 0.020

0.17 0.27 0.15

1 .0 1 .7 3.1

0.017 0.006 0.009

0.14 0.07 0.11

3.2 5.9 3.8

7.4 0.77 3.2

_áss*a b

i)

FAST-NEUTRON PERSONNEL DOSIMETRY

gives the totaldose rate, d, as calculated from the absolute neutron measurements, ;which has already been cited in table 1 . Thenext twocolumns provide the ratio ofthe total dose. rate, b, to theobserved bare and cadmium-covered track density rates, Tb/t and Tedlt, respectively. Thevariation in theratio bl(Tblt) is about the same as found for the ratio of total dose rate to thermal flux (see table 1) . The ratio bl(Tdlt) varies from 0.59 to 0.07. or about a factor` of eighty This is hardly surprising since the dose' per fast neutron was found to vary by about a factor of four and the dose per intermediate-energy neutron varied by about a factor of thirteen. The last column in table 3 contains the ratio of the dose rate deduced from SSTR data to the actual total dose rate, i.e., J5ssrx/& SSTR dosimeters do an acceptable job between the halves, where the neutron intensity is substantial and the spectrum is hard . At more remote locations in the cell, where the spectrum is soft and exposure rates are several orders of magnitude less than those between the halves, SSTR dosimeters may overestimate the fast-neutron dose by as much as a factor of six. These results are shown graphically in fig. 11, which displays the ratio D/Ted as a function of CR . The solid straight line is the representation of SSTR dose given in eq. (21) for the empirically determined values, a= 0.54 and b =0.011. Sincethe SSTR dose bounds the

397

actual dose from above, i.e., DSSTR >_ D over theentire CR range of physical, interest, the SSTR fast-neutron dosimetry,method is clearly established as conservative throughout the environs of the ZPR-b cell. Unfortunately, this dosimetry service is toeconservative and modifications are under consideration which will reduce present overestimates . For example, the dashed line in fig. 11 represents an improvement in DSSTR which would accrue if the CENTER-SHIELDS point at CR -32 could be shifted to the ROOM-3 point at CR - 10 . One way this might be accomplished would be to place perforated cadmium sheets over the personnel shields, where the perforation allows for some suitable adjustment of thermal-neutron flux, In this event, DssTR would exceed the measured dose by only about a factor of two in the worst case. Statistical data obtained over many months of experience with the SSTR dosimetry service further emphasize the need to reduce overly conservative estimates provided by SSTR personnel dosimeters. A frequency histogram of cadmium ratios obtained from 174case histories in aten-month interval is displayed in fig . 12 . This distribution reveals that cadmium ratios found in actual practice fall in a region which exacerbates SSTR overestimates (cf. fig. 11). The cadmium-ratio frequency histogram given in fig. 12 does not include all SSTR dosimeter histories for the ten-month inter ">W. Only those SSTR results

0.6 N

E H

Eá

É '04 tu W N O O

0.2

ROOM-1

OOROOM-2-NO SHIELDS Q ROOM-3 O RQOM-2-SHIELDS

CENTER- __ OSHIELDS

16

CADMIUM RATIO Fig. 11 . The ratio D/Tea as a function of CR. The solid line is the result of the empirical formulation of SSTR dose, whereas dashed line represents an improvement which couldpossibly accrue in SSTR dose estimates by judicious useofcadmiumsheet on the surface of themovable personnel shields (see text) .

39E

R. Gnr .y :

al. 350 HISTORIES IN A ns 120

10 MONT14 INTERVAL 53 CASES>0 .5-/QUARTER)

too

r U W ÓW K 4

to

.100 .200 : 300 400 500. 1000 SSTR DOSE PER 2 WEEK INTERVAL, mrem .

Fig. 13 . Frequency histogram of SSTR dose, D- observed for ZPR personnel in a ten-month interval .

analysis of the SSTR dosimetry method is not justified at this time.In spite of this deficiency, the SSTR dosimetry method has-been clearly established for the ZPR environment in that it provides adequate sensitivity and conservative monitoring of fast-neutron exposures whichsatisfy DssTR > 50 mrem (in a two-week interval) f(TZPR personnel . Because of simplicity,convenience, have bee-it used. This comprises about half of the total and sensitivity, t.is dosimetry system warrants serious case histories as can be seen from fig. 13, which presents consideration for general fast-neutron personnel a frequency histogram of DssTR results over the saliiie dosimetry. ten-monlb interval . data reveal that the average chase received by ZPR personnel is no more than References 60 mrem/wk. However, more than ten percent of the t) R. Gold, Fact, fashion and/or fantasy in fast neutron per. sonnel dosimetry, Health Phys. 26 (1974) 366 . c histories exceeded 1.5 rem/quarter implying an Gold, R . J . Armani and G . K . Rusch, Health Phys. 23 average exposure in excess of the recommended 100 z) R. (1972) 404. mrem/wlt radiation protection guideline for these a) R. Gold, R . J . Armani and J. H. Roberts, Nucl. Sci . Engng . cases . Consequently, adequate fast-neutron personnel 34 (1968) 13. dosimetry in the environs of the plutonium-loaded 4) J. H . Roberts, S. T . Huang, R . J . Armani and R. Gold, Nucl. Applic . 5 (1968) 247 . ZPR critical facilities is mandatory. Fig. 12. Frequency histogram of cadmium ratios obtained in a ten-month interval from the SSTR dosimetry service for ZPR personnel .

6. C This limited study reaffirmsanticipated inadequacies not only for this particular design SSTR fast-neutron dosimeter, butfor all other scalar fast-neutron dosimeters in current use. It has been demonstrated that the accuracy of such dosimeters is severely curtailed by spectral variation of the neutron energy distribution thr cswt thegiven working environment. In view of this fundarla^nisl lïmimion ; a quantitative error

s) R. J: Armand, R . Gold and J. H. Roberts, Fission process phenomena observed with solid-state track recorders, abstracts of papers, NUCL 41, !58th ACS National Meeting (New York, Sept. 1969). s) C. M Unruh (Ed.), AEC workshop on personnel neutron dosimetry, Bethesda, MD, BNWL-1340 (1969). 7) C. M. Unruh (Ed.), Second AEC workshop on personnel neutron dosimetry, New York, NY, BNWL-1616 (1971). s) D . E . Hankins, Factors affecting the design of albedoneutron dosimeters containing lithium fluoride themtoluminescent dosimeters, LA-4832 (1972). s) K . Becker, Health Phys. 23 (1971) 729. to) K . Becker, Trans . Am . Nucl . Soc . 15 (1972) 116.

FAST-NEUTRON PERSONNEL DOSIMETRY tt)

~-)

-~)

14)

is) 16)

399

K. Becker, Dosimetric applications of nuclear track etching, 17) E. F. Bennett and T . J. Yule, Nucl. Instr. and Moth . 98 6t: Progress in radiation dosimetry (ed. F. H. Attix, Academic te) (1972)393. Protection againsi neutron radiation, NCRP report no. 38 Press, New York, NY, 1972). (National Council on Radiation -Protection and MeasureR. Gold, R . J. Armani . and G. K . Rusch, Fast neutron perments, Washington, 1971). ZPR-6 and ZPR-9, in: Applied Physics sonnet dosimetry in ts) J. H . Roberts, F., J . Congel, R. J . Armani, R. Gold, J . Division Annual Report, ANL-7910 (July 1970 to June 1971)' Kastner and B. J. Oltman, Trans . Am. Nucl . Soc. 1 3 (1970) . p. 422 119. T . J. Yule, Health Phys. 23 (1972) 403 . 20) I. O . Andersson and J . Braun, Nukleonik 6 (1964) 237. E . F . Bennett, R . Gold and R . J. Huber, Proc . intern. 21) Radiation quantities and units, ICRU report no. 19 (interconf. on Fast critical experiments and their analysts, ANLnational Commission on Radiation Units and Measurements. 7320 (1966). Washington, 1971). 22) Evaluated Nuclear Data File-8 (ENDF/B), National Neutron R . Gold, Phys . Rev. 165 (1968) 1406. (1968) Cross Section Center, Brookhaven National Laboratory, and E. F. Bennett . Nucl. Instr, and Meth . 63 R. Gold BNL-50066. 285.