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Nuclear Engineering and Design 241 (2011) 3306– 3316

Contents lists available at ScienceDirect

Nuclear Engineering and Design

jo u r n al hom epage : www.elsev ier .com/ locate /nucengdes

nternet accessible hot cell with gamma spectroscopy at the Missouri S&Tuclear reactor

dwin Grant, Gary Mueller ∗, Carlos Castano, Shoaib Usman, Arvind Kumaruclear Engineering, Missouri University of Science and Technology, 203 Fulton Hall, 300 W. 13th St., Rolla, MO 65409, United States

r t i c l e i n f o

rticle history:eceived 20 December 2010eceived in revised form 5 May 2011ccepted 6 May 2011

a b s t r a c t

A dual-chambered internet-accessible heavily shielded facility with pneumatic access to the Universityof Missouri Science and Technology (Missouri S&T) 200 kW Research Nuclear Reactor (MSTR) core hasbeen built and is currently available for irradiation and analysis of samples. The facility allows authorizeddistance users engaged in collaborative activities with Missouri S&T to remotely manipulate and analyzeneutron irradiated samples. The system consists of two shielded compartments, one for multiple samplestorage, and the other dedicated exclusively for radiation measurements and spectroscopy. The secondchamber has multiple detector ports, with graded shielding, and has the capability to support gammaspectroscopy using radiation detectors such as an HPGe detector. Both these chambers are connectedthough a rapid pneumatic system with access to the MSTR nuclear reactor core. This new internet-based

system complements the MSTR’s current bare pneumatic tube (BPT) and cadmium lined pneumatic tube(CPT) facilities. The total transportation time between the core and the hot cell, for samples weighing 10 g,irradiated in the MSTR core, is roughly 3.0 s. This work was funded by the DOE grant number DE-FG07-07ID14852 and expands the capabilities of teaching and research at the MSTR. It allows individuals whodo not have on-site access to a nuclear reactor facility to remotely participate in research and educationalactivities.

. Introduction

As the Internet matures, information services have becomeore available and applications that were not possible a few years

go are now commonplace on the World Wide Web. One suchpplication is the remote use of highly specialized experimentalquipment through participation in online scientific research andducation. Information technology (IT) has been able to resolve theecessary issues required to connect these devices to the Internethrough the form of virtual laboratories. Present virtual laborato-ies include, for example, configurations relating to nuclear reactorhysics laboratory education (Malkawi et al., 2010), nuclear fusionesearch (Yamamoto et al., 2010; Tsuda et al., 2008), electronics andomputer education (Azaklar and Korkmaz, 2008), medical imag-ng simulation (Dikshit et al., 2005) and a virtual nuclear magneticesonance facility (Keating et al., 2000). All of these facilities havene thing in common: they have very highly specialized and expen-ive experimental equipment at a particular location, and they are

urrently remotely accessible through a web-based application totudents, faculty and researchers.

∗ Corresponding author. Tel.: +1 573 341 4348; fax: +1 573 341 6309.E-mail address: gmueller@mst.edu (G. Mueller).

029-5493/$ – see front matter © 2011 Elsevier B.V. All rights reserved.oi:10.1016/j.nucengdes.2011.05.015

© 2011 Elsevier B.V. All rights reserved.

Typically, there are two main types of web-based laboratories:the first kind involves some sort of simulations or virtual labora-tory and the second kind involves a remote laboratory in which theuser can control and manipulate instruments and equipment andactually perform experiments in a real time setting (Azaklar andKorkmaz, 2008). It is the second kind of laboratory that has beenimplemented in this present study involving a university nuclearreactor and an internet-accessible heavily shielded facility.

The nuclear reactor used for this virtual laboratory is the Mis-souri S&T Reactor (MSTR) which is a pool type research nuclearreactor that first became critical in December 1961 with a maxi-mum operational power of 10 kW, detailed by license R-79 (Bonzer,2010). Later, in 1967, the MSTR was upgraded to a current max-imum operational power of 200 kW. As a part of the ReducedEnrichment for Research and Test Reactors (RERTR) program theoriginal highly enriched uranium (HEU) fuel was replaced with 20percent low enriched uranium-235 (LEU) fuel in July 1992 (Matos,1996). The reactor’s neutron chain reaction is controlled via 4 stain-less steel control rods. Three shim rods allow for coarse reactivitycontrol and the single regulating rod allows for fine reactivity con-trol. Each of the shim rods contain 1.5% by weight of natural boron,

while the regulating rod (control rod 4) contains zero boron and ismade of stainless steel 304 (Bonzer and Carroll, 2009). The reactorwas originally built with both a beam port and a thermal columnfacility, and in order to increase the number of research facilities

E. Grant et al. / Nuclear Engineering an

Table 1HCPT technical specifications.

Specifications

Number of samples 11Sample size 2.22 cm (7/8′′) O.D. by 5.72 cm (2-1/4′′)Sample weight 16.9 gTypes of samples Solids, liquids,a and gasesa

Irradiation times Seconds to 8 hDetection times Seconds to daysMaximum reactor power 200 kWMaximum dose on shieldsurface

1 mSv/h

Maximum sample activityfor 0.5 MeV

28.66 ± 1.2 GBq (774.5 ± 32.5 mCi)

Maximum sample activityfor 0.75 MeV

13.47 ± 0.5 GBq (364.2 ± 13.6 mC)

Maximum sample activityfor 1.0 MeV

8.51 ± 0.27 GBq (230.1 ± 7.4 mCi)

Maximum sample activityfor 1.5 MeV

4.23 ± 0.11 GBq (114.2 ± 3.1 mCi)

Maximum sample activityfor 2.0 MeV

2.75 ± 0.07 GBq (74.2 ± 1.8 mCi)

Maximum sample activityfor 2.5 MeV

2.05 ± 0.05 GBq (55.4 ± 1.3 mCi)

Maximum power by fueledexperiment

<1 Wt

Transportation time 3 sMaximum reactivityinsertion

0.4% �k/k

Thermal neutron flux 2.94E12 n/cm2 s

btwMlsttadpg

assdia2a8tbi2fl

2

tsfes

Fast neutron flux 2.65E12 n/cm2 s

a Special preparations and precautions needed.

are and cadmium lined pneumatic systems were added throughhe years. Before the new hot cell pneumatic tube (HCPT) systemas built, researchers either had to be physically present at theSTR to do their experiments or face the time consuming and

aborious process of having their samples sent in and their resultshipped back. Users can now remotely access their samples in realime, manipulate their samples in the presence of radiation detec-ion equipment and save their spectral data from their own desksnywhere in the world. This novel hot cell is a unique, one-of-a-kindesign available only at MSTR and it has the capability to controlost-irradiated experiments with data collection from across thelobe via the Internet.

Presently, the HCPT facility has been completed and tested, ands a result it is meaningful to inform the academic and researchcientific communities of both its educational and research pos-ibilities and its capabilities, which are listed in Table 1. Variousiverse potential samples can be irradiated with the system. This

ncludes fueled experiments, however, these fueled experimentsre limited to a thermal power of 1 watt (Bonzer and Carroll,009). Presently, the time limit a sample can be irradiated is 8 ht full power due to the fact the MSTR facility is only open for

h daily during the week and is closed on weekends. In addi-ion, the maximum reactivity insertion for any experiment muste <0.4% �k/k (Bonzer, 2008). Kulage (2010) recently character-

zed the thermal and fast neutron flux to be 2.94E+12 n/cm−2 s and.65E+12 n/cm−2 s, respectively and thus giving a thermal to fastux ratio of 1.11.

. Materials and method

Initially, a rigorous pre-design analysis of the new hot cell sys-em was performed in order to ensure that it would incorporate

elected key systems. The pre-design analysis also allowed for care-ul consideration and selection of components that would be robustnough to support the needs of the project. First, the proposedystem needed to utilize and accommodate multiple samples, in

d Design 241 (2011) 3306– 3316 3307

contrast with the existing pneumatic system, which can only useone sample and has no storage capacity. Having the ability to placemultiple samples in predetermined storage locations allows highlyactive short-lived isotopes to safely decay away without humanintervention; therefore unmasking any long lived isotopes.

It was determined that within the detection chamber there mustbe multiple detection ports of different sizes for different types ofradiation detectors to be used. Each chamber not only needs toproperly shield the reactor staff, students, and public but also needsto shield samples from external sources, including backgroundradiation. The key advantage of having two separate chambers isthat they allow for isolating one sample’s spectrum from the rest ofthe samples within the storage cell. Another important require-ment within the detection chamber is to incorporate shieldingmaterials with decreasing atomic numbers or graded shielding toremove low energy K shell � and � X-rays from the shield itself(Wolbarst, 2005). The overall hot cell system consists of the fol-lowing key components:

1. Pneumatic transportation system2. Radioactive sample manipulation system3. Radiation detection and spectroscopy system

The pneumatic transport system, radioactive sample manipu-lation system, and radiation detection with spectroscopy systemwere all required to be fully integrated into a remotely accessibleInternet based computer system.

Once the guidelines of the design were identified, it became pos-sible to research different materials and equipment. First, the shielditself was an important design parameter that was fulfilled withthe best overall design of the G-16-S Roll Top Counting Shield fromGamma Products, Inc. (Gamma Products, Inc., 2009). As a countingshield, the standard G-16-S incorporates no options for detectorports; however Gamma Products, Inc. (2009) has the capability toadd custom multiple ports.

The top rail mount design was convenient for the easy inspec-tion of key mechanical and pneumatic devices which were placedwithin the cell. Other possible designs were based on sliding doorsthat locked together. Those designs were not selected due to theconvenience and availability of the shields by Gamma Products Inc.(2009).

Two shielded cells were selected, one for storage and one forspectroscopy. The basement of the reactor was selected to housethe HCPT shielded cells to facilitate in providing distance betweenirradiated samples and the reactor bay. Each shield weighs in excessof 1315 kg (2900 pounds) which required the transport of theshields into the MSTR basement so that the cells could be disman-tled (Gamma Products, Inc., 2009). As a result, the custom 10.16 cm(4 in.) thick G-16-S cells were designed to have the capability tobe disassembled and reassembled. Also, having the flexibility to bedisassembled allows for possible future cell modifications. GammaProducts Inc. (2009) offers both a lead and an A36 steel option forshielding material. The A36 steel option was selected due to the factthat lead had the potential concern of toxicity during machiningon the Missouri S&T campus. In addition, at the time of purchasethe cost for the lead model, with machining and manufacturingincluded, was roughly twice as expensive as the steel option. Fig. 1indicates how each cell is fully constructed with the detailed layoutof detector port locations.

The current design allows for up to three High Purity Germa-nium (HPGe) detectors to be located in the center of parts 2, 4, and7, as seen in Fig. 1. The HPGe port geometry allows for many exper-

iments, including coincidence testing with surfaces 2 and 4. Fouradditional smaller 2.54 cm (1 in.) diameter ports are located on part7 to permit cabling into the shield for mechanical manipulation. Thesmaller ports surround the HPGe port in a square formation. When

3308 E. Grant et al. / Nuclear Engineering and Design 241 (2011) 3306– 3316

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n use, the smaller ports are positioned where lead blocks can belaced within the cells for shielding the open ports. In addition toabling, the smaller ports are positioned so they can be used foreiger-Müller detector tubes. Geiger-Müller detector tubes will besed to verify count rates inside the shield. Installed video mon-

toring equipment within the nuclear reactor facility can visuallyerify count rates from portable ion chambers and dose rates ifhe spectroscopy system should fail to operate properly. Parts 1nd 3 provide space for the pneumatic transfer piping entering andeaving the shield. Stainless steel 316 sleeves were machined tot between the pneumatic ports and the smaller pneumatic tubes.he 7.62 cm (3 in.) diameter pneumatic port holes allow for differ-nt sized pneumatic tubes in the case of future additions to thehields. Additional ports and equipment can be easily added toccommodate future needs as they become necessary for a specificxperiment.

.1. Pneumatic transportation system

The medium used to transport samples had to exhibit low neu-ron activation properties to minimize the concern of contaminatedas, and containers holding the specimens were required to alsoxhibit low neutron activation properties. If the specimen vial wereo become activated during irradiation it would highly affect laterample spectroscopy. In addition, the size of the container neededo be large enough to hold adequately sized samples while beingir tight. The transit time of the sample from the core to the hotell must be kept at a minimum. This short transit time guarantees

low potential dose to nearby materials and the reactor staff. Theneumatic system had to be able to move samples with a minimumass of 5 g. To safely exhaust pneumatic gas the system incorpo-

ated high efficiency particulate air (HEPA) filters. To make certainases do not escape down near the core a custom in-core assemblyas needed.

Nitrogen was selected as the medium of the transportationystem due to concerns about the possibility of neutron activa-ion of contaminates in generic compressed air. Fig. 2 details theransportation of samples throughout the rapid pneumatic sys-

of each shielded cell.

tem. Sealable low density polyethylene (LDPE) vials (“rabbits”)(Becker, 1999) are used to contain the specimens during transporta-tion because they exhibit low activation after sample irradiationsand are used at other facilities (Lindstrom et al., 2008). The phys-ical dimensions of the vials are 2.22 cm (7/8 in.) OD by 5.72 cm(2-1/4 in.) overall height, leading to a total volume of 16.9 ml (Cole-Parmer, 2010). The physical dimensions are ample enough to holddifferent sized specimens that have a total mass of 5 g. To guar-antee proper transit, in the HCPT, a maximum sample mass withvial is set to approximately 16.9 g. The total transit time from thereactor core to the inside of the shield cells is approximately 3.0 s.During the 3.0 s the sample travels through 25.9 m (85 ft) of pneu-matic tubing with an average velocity of 8.63 m/s (28.33 ft/s). Thishigh speed allows for any external dose to be limited during rapidsample transportation.

Furthermore, the custom in-core assembly needed to be flex-ible to allow for multiple core locations for any future changesto the core configurations. Aluminum 6061 was chosen as the in-core material due to its lower neutron absorption cross section ascompared to other steel alloys (Baum et al., 2002). Other researchreactors have used aluminum 6061 for in-core components due toits good low temperature properties (Farrell, 2001; Weeks et al.,1988). Furthermore, the in-core assembly was designed to allowthe centerline of the sample vial to stop at the midpoint of thereactor fuel assemblies to maximize the possible neutron flux. Theposition of the HCPT within the fuel grid plate is position F6 asshown in Fig. 3. The position of the HCPT can and most likely willchange with different core configurations and with that in mindthe assembly was built to insure no interference with surroundinggrid positions.

2.2. Radioactive sample manipulation system

In order to make the sample manipulation system robust, sev-

eral levels of automation were needed. The hardware selectedfor the system was standard engineering equipment which canbe easy updated if required. The automation development is ableto incorporate both the pneumatic transfer of a sample and the

E. Grant et al. / Nuclear Engineering and Design 241 (2011) 3306– 3316 3309

neum

miccdttne

Fig. 2. Mapped rapid p

anipulation of multiple samples which can be supervised by annternet user. Also, the automation is able to link directly with aomputer and have a graphical user interface that someone canontrol. The manipulation system is capable of consistently repro-ucing experiments without need for human intervention. All of

he manipulation equipment had to fit within the sealed pneumaticransportation system. Some type of containment structure waseeded to conform around the manipulation equipment to prop-rly seal and hold the pneumatic gas since it becomes activated in

Fig. 3. MSTR core configu

atic transport system.

the reactor the core region and therefore should not leak out of thesystem.

A Programmable Logic Controller (PLC) was selected and it isessential to make multiple processes and control them in a preciseand accurate sequence. The DPY 50601 from Anaheim Automation

Inc. (2005) allows for eight inputs, eight outputs, and one singleaxis step motor for control (Anaheim Automation Inc., 2005). Eachinput and output has a time resolution control of one millisecondallowing ultra fine control of the system. Also the PLC has a built-in

ration 120 mode W.

3 ing and Design 241 (2011) 3306– 3316

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Table 2ORTEC GEM10P4-70-PLUS specification.

Resolution 1.6 keVa FWHM 571 eVb FWHMPeak to Compton ratio 45:1a

Relative efficiency 10%a

Peak shape 1.8 FWTM/FWHM

Detector location

in Table 2 and Fig. 6, where it has been placed into one ofthe seven detection ports on the detection cell. Furthermore,some important Ortec High-Purity Germanium (HPGe) detector

310 E. Grant et al. / Nuclear Engineer

tepping motor indexer which allows up to 12800 steps per revolu-ion or rotational resolution of 4.909E−4 radians per step (0.028125egrees per step).

A stepping motor setup was included as a component of theurable radioactive sample manipulation system in order to helpransfer up to eleven different samples. The model 8718L step-ing motor from Lin Engineering (2010) was selected for theesign because of its high torque, accuracy, and resolution (Linngineering, 2010). The motor has a holding torque of 10.1 N m7.45 lbf ft) which is sufficient considering that the revolver onlyas a radius of 10.16 cm (4 in.) and houses samples that are <16.9 g.lso, the motor has a full step angle of 0.0314 radians (1.8◦) butith the built-in indexer it is possible to use its micro stepping

apability. The stepping motor is set to operate at a maximum of.5 rpm when selecting a given sample. A slow rotation guaran-ees no missed steps since holding torque is directly proportionalo operating speed. The stepping motor receives control signalsrom the PLC and rotates the storage revolver until the cho-en sample is loaded and operates in the same way a revolverype pistol loads a bullet into the chamber. This expands theystem so that it is able to handle more than a single samplend allows up to eleven different research samples to be activeuring any given time. There are twelve different slots in theevolver but the first position has been designated as an openosition and is the zero reference point for the PLC. By keep-

ng the first slot open there is no chance for possible streaminghrough the pipe during storage. A small aluminum 6061 contain-

ent structure was constructed around the revolver to contain alladioactive material. The containment box also allows the com-ressed pneumatic gases to easily transport samples without anyajor loss of gas pressure that could hinder sample transporta-

ion into the core. Aluminum 6061 was selected as the materialor the revolver and containment building material due to its hightrength, low cost, excellent corrosion resistance, and low neutronctivation cross section, which are all superior benefits for longerm use.

In order to control the pneumatic valves relay coils, transform-rs, and power supplies were added to make the needed voltagesompatible. Fig. 4 details the wiring scheme needed to correctlyntegrate all electronic components to the PLC.

.3. Radiation measurement and spectroscopy system

The design of the radiation measurement and spectroscopy sys-em was one of the most challenging parts of the configuration. Thisas primarily due to the fact that there are available engineer-

ng computer automatic pneumatic and hydraulic systems but aetup that incorporates internet-accessible radiation spectroscopyith sample automation is not common. The spectroscopy systemeeded to allow for a multitude of different options for users athe MSTR or the Internet. Some of the options for users are gainontrol, upper and lower decimation, voltage, and counting dataollection time. One major goal was that the system would allowultiple users to watch and record live data. Another aspect of the

pectroscopy system is detector geometry and orientation design.he design allows for multiple types of detectors, of different sizes,nd allows for unique detection techniques such as coincidenceeasurement.A direct online server system was needed to allow an internet-

ased option for collecting spectroscopy data. The Canberra Lynx2011) system was selected and it connects directly to an internet

onnection and has its own unique web address (www.rad-ynx2.device.mst.edu) so that users can locate it with an internetrowser, as seen in Fig. 5. A brief introduction and tutorial forecoming familiar with using the Internet facility is available for

a 1.33 MeV 60Co.b 122 keV 57Co.

new users.1 In order for users to control the system, they mustuse a virtual network connection (VNC) to be granted accessto the secure Lynx (Canberra Lynx, 2011) server. A usernameand password will be sent to each approved user for access tomanipulate the specimen and collect the spectrum data. Usersare able to acquire data directly from the HPGe detector withinthe shielded cell. In fact, the Lynx (Canberra Lynx, 2011) systemallows a multitude of other detectors including NaI(Tl) and Lan-thanum Halide. Standard techniques are also available to users,such as selecting regions of interest, changing rough and finegain control, a multichannel analyzer (MCA), a multichannel scalar(MCS), a single channel analyzer (SCA), energy calibration, highvoltage power supply (HVPS) change and saving collected data.The Lynx (Canberra Lynx, 2011) Internet accessible software isa web based Java program which features capabilities from Can-berra’s native Genie 2000 basic spectroscopy software with mostof its available inputs. To properly view all of the functional-ity of the web based Lynx system users must have Java 6.0 orgreater installed on their computer. The software also allows fora bias-voltage shutdown input from the HPGe detector. The bias-voltage shutdown will turn off the high voltage to the detectorif the detector starts to warm up. This guarantees no detec-tor damage if HVPS is turned on accidentally by a web-baseduser.

Additionally, the system has three levels of security: VNCwith optional IP address verification; two locally controlled,normally closed shutoff valves; and one locked sample stopvalve to guarantee cyber security and stop unwanted coreaccess.

An Ortec GEM10P4-70-Plus HPGe (Ortec, 2008) detector iscurrently used to collect the needed data but again the sys-tem is flexible enough to handle different detectors. The OrtecHPGe detector (Ortec, 2008) was selected because it possessesmany useful options. The detector uses an ultra-high count ratepreamplifier for counts up to one million one MeV gamma raysprocessed every second (Ortec, 2008). Also, the detector modelhas an operating voltage of positive 3500 V. A 10 percent effi-ciency was selected which is sufficient since highly active samplesare expected. These high activity samples can easily be countedwith a ten percent efficiency detector. The detector can be seen

1 Inquires about experimental research along with HCPT tutorials can be made bysending a request to Dr. Gary Mueller at +15733414348 or gmueller@mst.edu.

E. Grant et al. / Nuclear Engineering and Design 241 (2011) 3306– 3316 3311

matic

stdttjce

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Fig. 4. Wiring sche

pecifications (Ortec, 2008), such as resolution and peak to Comp-on ratio, can be seen in Table 2. The technology behind HPGeetectors uses liquid nitrogen to cool the semiconductor crys-al down to cryogenic temperatures which is a way to removehermal excitation of electrons and guarantee electron band gapumping due to gamma ray interactions. This allows a directorrelation between electron excitation energy and gamma raynergy.

. MCNP modeling

A MCNP (X-5 Monte Carlo Team, 2008) computer model wassed to facilitate the responsiveness of the operational limits of theTPR before and during construction. The physics card on the MCNP

imulations is set with generation of secondary electrons allowedo be produced for thick-target bremsstrahlung modeling whichccounts for bremsstrahlung interactions and possible X-ray fluo-escence whose interactions can contribute to the total dose rate.nother selection to the physics card includes coherent photoncattering. While the cross-section for coherent photon scattering

s several orders of magnitude lower compared to the photoelec-ric cross-section there is no significant computational penalty forncluding coherent scattering. No photo neutron interactions arellowed with this model. Since the typical energies encountered

for PLC electronics.

with photo neutron interactions are much higher than the modeledkinetic energies of 2.5 MeV, or lower, it will not need to be tracked.Also, Doppler energy broadening for photons was also included tohelp carry out a more precise simulation for low energy transportand much like coherent scattering, no large computational penaltywas introduced.

Knowing the maximum specimen activity is highly importantto insure the facility stays within 10 CFR 20 (1991) limitations.Fig. 7 shows both the full MCNP geometric model with the backwall made transparent from the three-dimensional (3-D) plot andthe actual picture of the assembly. The overall MCNP geometricmodel incorporates the hot cells, pneumatic assembly, and sam-ple manipulation equipment. Monte Carlo techniques are used topredict the path of particles by a random-walk and with a largenumber of particles an accurate behavior can be observed (Shultisand Faw, 2000).

Fig. 8 shows the MCNP modeled sample revolver and step motorassembly, with the pressure box made translucent and a photo-graph of the actual revolver and step motor assembly in the storagecell. The fabrication of the revolver and its components was mod-

eled from the exact dimensions of the plans. The main design factorwithin the revolver is to prevent samples from continuing past therevolver and pressure box while at the same time letting pneumaticgas continue to the HEPA filter system.

3312 E. Grant et al. / Nuclear Engineering and Design 241 (2011) 3306– 3316

ectros

4

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FG

Fig. 5. Online Gamma sp

. Results

With the hot cell installation finished, a maximum sample activ-ty characterization, with MCNP, of the facility has been completed.

his provides the facility with a guideline to begin experimentalnalysis. A sample experiment exhibiting the capabilities of the hotell is included. The experiment shows the decay of an irradiatedn-56 sample which was remotely monitored and transported

ig. 6. Hot cell system with HPGe detector: A – storage cell, B – detection cell, C – glovebo – HPGe detector, H – HEPA filtration pump.

copy for Co-60 example.

from the storage hot cell to the detection cell several times overa period of time.

4.1. MCNP results

Dose rate field plots are used for indicating how the dosechanges throughout the modeled hot cell. Fig. 9 shows the X–Y axisalong the centerline axis of a single energy 2.5 MeV gamma ray

x, D – glovebox HEPA filter, E – lead pig for removal, F – pneumatic transport tube,

E. Grant et al. / Nuclear Engineering and Design 241 (2011) 3306– 3316 3313

F anspo

sar3mdi

dlaabwtf

ig. 7. MCNP geometric model: A – storage cell, B – detection cell, F – pneumatic tr

ource. The units have been normalized to a dose rate per samplectivity, using the ICRP-21 Gamma Flux to Dose Tables. In order toenormalize the dose rate, a FACTOR card was used with a value of.7E+10 within the MESH tally to create the dose field in terms ofSv/h per MBq (mrem/h per milliCurie) of activity. The maximum

ose stated by 10 CFR 20.1601 (1991) for a given radioactive samples 1 mSv/h (100 mrem/h) 30 cm away from the source.

Fig. 9 illustrates the most noticeable issue, which is with theose within the transfer pipe. The dose rate is much higher in this

ocation due to the fact that there is no material other than air tottenuate the radiation. There is no shielding to attenuate the radi-tion within the counter cell piping besides closing the ball valves,

ut even when the valves are closed the pneumatic gas relief holeill still allow streaming through the tube. However, the inside of

he tubing is still considered as a part of the shield and will not allowor direct streaming. The light blue dose rate around the tube is then

rt tube, I – ball valve, J – sample revolver box, and K – revolver stepping motor.

the limiting factor due to the fact that it represents dose rates out-side of the shield. Another key point is that from the center of theshield to the outside walls is 30.48 cm (12 in.). The 30.48 cm (12 in.)matches well with the needed 30 cm (11.8 in.) defined by 10 CFR20.1601 (1991).

In order to quantify the results harvested with MCNP, a nomen-clature must be used to separate dose rates at different locationsalong the counter shield walls. Location � is located on the centerpoint of the outer surface of the main three inch diameter detec-tion port. Location � is at the center line of the pneumatic transfertube at a location 30 cm (11.8 in.) away from the source stoppingpoint, within the detection cell. Location � and � are on the outer

surface of the detection cell half way between the center and outeredge of the pneumatic transportation piping sleeve, facing the glovebox. Fig. 10 shows the differences between the different locationsalong the counter shield boundaries. Around the pneumatic tube

3314 E. Grant et al. / Nuclear Engineering and Design 241 (2011) 3306– 3316

Ft

tos

ag

Fig. 9. X–Y axis dose rate plot of counter shield with 2.5 MeV gamma source: F –pneumatic transport tube, I – ball valves, N – 7.62 cm (3 in.) detector port and hatched

Fi

ig. 8. Modeled sample revolver and step motor assembly: F – pneumatic transportube, J – sample revolver box, K – revolver stepping motor, and L – sample revolver.

wo locations were picked: the first mesh cell on the top and bottomf the aluminum 6061 pneumatic tube and the results did match

ymmetric expectations.

When the normalized dose rates along the counter shield wallsre known, it is possible to find the maximum activity for a givenamma ray energy. All that is needed to be done is to take the

ig. 10. Counter shield dose rate location identification: A – storage cell, B – detection ces located (see Fig. 6), and ˛, ˇ, � , ı are dose calculation locations.

lines-cross sectional area of detection cell B and M – 7.62 cm (3 in.) detector port,detector G (see Fig. 6).

maximum dose rate as stated by the 10 CFR 20.1601 (1991) anddivide by the normalized dose rates per MBq (milliCurie). This willresult in the maximum activity at different energies which corre-spond to different locations along the counter shield wall. Resultsfor the maximum activity are seen in Table 3. Random numberprocesses such as the Monte Carlo computational techniques yielderror within each result. With the MCNP code, one consistent wayto reduce the error in the results is by increasing the amount ofsource particles (X-5 Monte Carlo Team, 2008).

Fig. 11 shows a large maximum activity at the energy of 0.5 MeVfor location �. This is due to the fact that the photons have the great-est thickness of steel to travel through and the mass attenuation

coefficient of iron increases almost exponentially below 0.5 MeV(Lamarsh and Baratta, 2001). Location � was omitted from Fig. 11because its location is physically within the shield and would not

ll, F – pneumatic transport tube, M – 7.62 cm (3 in.) detector port where detector G

E. Grant et al. / Nuclear Engineering and Design 241 (2011) 3306– 3316 3315

Table 3Maximum activity GBq (mCi) with 1 mSv/h (100 mREM/h) at shield wall.

Location 0.5 MeV 1.0 MeV 1.5 MeV 2.0 MeV 2.5 MeV

106.46 (2877.35) 12.40 (335.03) 5.09 (137.48) 3.35 (90.64) 2.35 (63.44)ˇ 29.97 (810.04) 8.63 (233.14) 4.47 (120.87) 2.85 (77.05) 2.06 (55.77)� 1.04 (28.21) 0.52 (13.92) 0.37 (9.98) 0.30 (8.03) 0.25 (6.89)ı 28.66 (774.47) 8.51 (230.10) 4.23 (114.23) 2.75 (74.21) 2.05 (55.39)

ity ar

fma

iesss

Ft

A small 0.308 ± 0.002 g sample of Manganese was irradiatedat 1 kW in the hot cell rabbit tube located in the MSTR core for

Fig. 11. Maximum activ

all under 10 CFR 20.1601 (1991) guidelines. To clarify the maxi-um activity values for all photon energies greater than 1.0 MeV,

n expanded view is shown within Fig. 11.The maximum activity possible for each given gamma energy

s useful for different applications but for safety the lowest value,xcluding the inside of the pneumatic pipe, for each energy waselected. The lowest of the maximum activity values will yield the

afest result and best measurement for a factor of safety. Fig. 12hows the results of the maximum activity for the entire radiation

ig. 12. Maximum activity GBq (mCi) for 1 mSv/h (100 mrem/h) dose rate on detec-ion cell surface.

ound the detection cell.

detection shield. A 0.75 MeV case was added to better show themaximum activity at lower photon energies.

4.2. Mn-56 irradiation

two minutes. After irradiation, the sample was returned and stored

Fig. 13. Remote detection and storage for the radioactive decay of Mn-56.

3 ing an

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316 E. Grant et al. / Nuclear Engineer

nside the storage cell, in storage position one, for 10 min. Then,rom across the Missouri S&T campus, the sample was remotelyransported into the detection cell for 10 min. This was done byoading one file into the “send to detector” sample position on theLC software which executes the operation sequence. Then theample was sent back to storage for ten additional minutes. Thisas done by loading one file into the “send to storage” sample posi-

ion on the PLC software which executes the operation sequence.his cycle was repeated remotely multiple times. The entire events captured live with the Lynx digital signal processing system andhe data is shown in Fig. 13. Each data point represents one sec-nd of gross counting. Therefore, it can be seen how the Mn-56pecimen decays over time. Thus, this example demonstrates that

remote user can have access to manipulate samples and collectxperimental nuclear data.

. Conclusion

A dual-chambered hot cell internet-accessible shielded pneu-atic facility at the MSTR has been built and is currently available

or sample irradiation and analysis. The facility allows remote userserforming collaborative work with Missouri S&T researchers toemotely manipulate and analyze neutron irradiated samples. Theystem uses computer automation with user feedback to eliminateuman exposure throughout the entire irradiation and measure-ent process. The potential to analyze multiple samples without

uman intervention allows the MSTR to concurrently accommo-ate several researchers with different needs. Finally, as the needsf Missouri S&T and its collaborators change, it will be possibleor the facility to expand with the addition of detection chambers,nsuring that the MSTR hot cell pneumatic facility will continue toe an important research and educational tool for years to come.

cknowledgments

This work was made possible by a DOE grant DE-FG07-7ID14852 and Missouri S&T resources. The authors would also

ike to acknowledge the Missouri S&T reactor staff for their sup-ort in carrying out the experimental work at the reactor facilitynd S&T McNutt Hall Machine shop for the fabrication needs.

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