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1 II SANDIA REPORT -... ~.... kr a Self-Sustained ihem-type 9Mo......, ~---.E ~, Edward J. Par@%m ~ E. Vernon, Rihard L. Coats, Sandia National laboratories

2 IssuedbySandiaNationalLaboratories, operatedfortheunitedstatesdepart-merit of EnergybySandiaCorporation. NOTICE: Thisreportwaspreparedasanaountofworksponsoredby anagenyofthe United States Government. Neitherthe United StatesGovem-ment,nor any ageny thereof,nor anyof theiremployees,nor anyof theirontrators,subontrators, or their employees,make any warranty,expressor implied,or assumeany legal liabilityor responsibilityforthe auray,ompleteness, or usefulnessof anyinformation,apparatus, produt,or proessdislosed,or representhat its usewouldnot infringeprivatelyowned rights.referenehereinto anyspeifiommerialprodut,proess,or servieby trade name,trademark,manufaturer,or otherwise,doesnot neessarilyonstituteor implyits endorsementreommendation, orfavoringby theunitedstatesgovernment,anyageny thereof,or anyof theirontratorsor subontrators.theviewsandopinionsexpressed hereindonotneessarilystateorreflethoseoftheunitedstatesgovernment, anyageny thereof,oranyoftheirontrators. Printedin theunitedstatesof Ameria.Thisreporthasbeenreproduediretlyfromthe bestavailableopy. Availableto DOEandDOEontratorsfrom OffieofSientifiandTehnialInformation P.O.BOX 62 OakRidge,TN Priesavailablefrom(703) Website Availableto thepublifrom NationalTehnialInformationServie U.S.DepartmentofCommere 5285port Rd Springfield,VA NTISprieodes Printedopy A06 Mirofiheopy: AO1

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4 SAND Unlimited Release Printed April 1999 A Feasibility Study for a SeIf-Sustained Reator Fueled with Cintihem-type 9M0 Targets Daniel J. Romero Nulear Reator Failities Department Edward J. Parma and Milton E. Vernon Isotope Tehnology Department Rihard L. Coats Nulear Tehnology Programs Department Sandia National Laboratories P.O. BOX5800 Albuquerque, NM Robert D. Bush Department of Chemial& Nulear Engineering University ofnew Mexio Albuquerque, NM Abstrat This report presents a omprehensive feasibility study for a unique and modified onventional reator method to produe gmo, the preursor of the most widely used medial radioisotope, 9%T. Utilization of a typial nulear reator requires the irradiation of stainless steel tubes alled targets, oated with highly enrihed uranium, to produe Me, a fission produt of 235U. These targets are typially surrounded by a driver ore of fiel elements. The primary objetive was to determine if a ritial reator ould be onstruted solely of Cintihem-type targets without the aid of a driver ore. Additionally, this reator would be apable of produing the entire U.S. demand for 99M0 without ompromising the Cintihem proess, a hemial separation method used to extrat Mo from the irradiated uranium. Numerous ritial onilgurations were disovered, and additional reator neutroni studies are presented whih assess reativity losses. Furthermore, an eonomi study was ompleted to estimate initial onstrution osts; and to provide a reliable omparison among the onventional, target-fheled, and solution-fieled reator onepts in terms of waste dkposal osts. Based on the neutroni and eonomi analyses, this new reator onept appears quite feasible. Two reator systems were hosen for fi.u-ther onsideration as andidates in atual onstrution and experimental testing. 3

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6 CONTENTS EXECUTIVESUMMARY... L ISOTOPE PRODUCTION METEODS ~ODUnON... D.O.E. ISOTOPEPRODUCTIONPROJECT&HISTORY OF: NUCLEARFORMATION&MEDICAL USES OF9%10/99 TARGETIRRADIATION& EXTRACTIONPROCESSES... CURRENT lsotopeproductionmethods Conventional Reators Solution.Fueled Reators Aelerators... PRO~SED CONCEPT NEUTRONICS ANALYSIS MONTECARLOVS.DETE~STIC CODES MODERATORMATEWS P~~CmSnGATIONS mactordesignnsuts kfl Ver@ation Cases WaterMderator BeryIIium & BerylIium Oxide Moderators Grqhite Mderator Heavy Waterkbderator CONCLUSIONS TEMPERATURE& DENSITY EFFECTS ON KEFF HOTm COLDCO~~ONS WA~RMODEWTOR BER~mMODEWTOR Gw~MODEWTOR ~AVYWA~RMODEWTOR CONCLUSIONS FISSION PRODUCTS& NEUTRON POISONS BACKGRO~ momnon REACTIVITYLOSSESBY 135XE, 49SM AND6LIPRODUC 4.3 %, HEL~GAS, AND3H PRODUCTION CONCLUSIONS FUEL BURNUP ~LB~~S~TS CONCLUSIONS... 5

7 6. ECONOMIC ANALYSIS CONCLUSIONS CONCLUDING REMARKS REo~ATIONS FOR~WOK F~m CONCLUSIONS S. REFERENCES APPENDICES

8 Table 1. Table 2. Table 3. Table 4. Table 5. Table 6. Table 7. Table 8. Table 9. Table 10. Table 11. Table 12. Table 13. Table 14. Table 15. Table 16. Table 17. TABLES hffresults of MC~4AVetifiation Cases MC~Results for Water Moderated Reatiors...3l MCNP Results for Beryllium& Beryllium Oxide Moderated Reators MCNP Results for Low Enrihed Beryllium Moderated Reators...38 MCWResults fortiaphite Moderated Reatiors...4O MCNPResults for Heavy Water Moderated Reatiors...44 Parameters for Various Reator Materials at Cold& Hot Conditions Results of Reativity Losses for Water Moderated Systems Results of Reativity Losses for Beryllium Moderated Systems Results of Reativity Losses for Graphite Moderated Systems Results of Reativity Losses for Heavy Water Moderated Systems Results of Reativity Losses Due to 135Xe,149SWand Li SmMass and Fration of Saturation Results for 37 Target Cores, Ii, 3H, and He Conentrations for37 Target Cores Reativity Losses Due to 10% 235UBurnup for37 Target Cores Initial Constrution Cost Estimates of 37 Target Cores Comparison of Annual Costs for Various 99M0Prodution Methods....66

9 Figure 1. Figure 2. Figure 3. Figure 4. Figure 5. Figure 6. Figure 7. Figure 8. Figure 9. Figure 10. Figure 11. Figure 12. Figure 13. Figure 14. Figure 15. Figure 16. Figure 17. Figure 18. Figure 19. Figure 20. Figure 21. Figure 22. FIGURES Nulear Formation of99mo MoDeay Chain and Daughter Produts Shemati of a Typial Cintihem.t~e Target Cross Setional View of the Original& Modified ACRR Core Radial View of a Reator Fueled with Cintihem-type Targets...23 Radial and tilal Views of a Single Cintihem-type Target Modeled in MCNP Cross Setional Views of MCNP Modeled and Nominal Cintihem-type Targets k Results for Water Moderated Reators Using 85 Stainless Steel Clad Targets and Various Target Center Materials k Results for Water Moderated Reators Using 85 Zironium Clad Targets and Various Target Center Materials kff Results for Beryllium Moderated Reators Using 37 Stainless Steel Clad Targets and Various Target Center Materials km Results for Beryllium Moderated Reators Using 37 Zironium Clad Targets and Various Target Center Materials kff Results for Graphite Moderated Reators Using 37 Stainless Steel Clad Targets and Various Target Center Materials lgffresults for Graphite Moderated Reators Using 37 Zironium Clad Targets and Various Target Center Materials k Resuits for 37 Targets in a Heavy Water Moderated Reator with Various Target Pithes kff Results for 37 Targets in a Heavy Water Moderated Reator... Cooled with Heavy Water or Ordinary Water...46 Deay Chain and Nulear Formation of135xe Deay Chain and Nulea Formation of149sm Be (q2n) 8BeReation Nulear Formation of~i, He, and 3HfiomBeryllium Post Reator Shutdown 135Xefor 37 SS Targets with Polyethylene Plugs ina Begllium Moderator Post Reator Shutdown 135Xefor 37 Zr Targets with Polyethylene Plugs ina Be~llium Moderator Post Reator Shutdown 135Xefor 37 Zr Targets with Beryllium Plugs ina Gaphite Moderator

10 EXECUTIVE SUMMARY Over the past several years, numerous methods for the prodution of medial radioisotopes have been proposed, partiularly those relating to the prodution of gmo. The objetive of this researh is to evaluate the feasibility of a unique and alternative method for the prodution of gmo. There are three major methods urrently in use or under onsideration: the onventional reator method whih utilizes the irradiation of targets;. the solution reator method whih relies on a ontinuous bath proess with on-line extration apabilities; and. target bombardment with the use of aelerators. At Sandia National Laboratories, the Annular Core Researh Reator (ACRR) has been tasked by the U.S. Department of Energy to serve as the domesti soure for 99M0prodution. This involves irradiating highly enrihed uranium targets in the entral region of the reator using a driver ore of fiel elements, and subsequently extrating the 99M0 fission produt from the uranium oxide oating inside the targets via the Cintihem proess. The proposed tehnique utilizes the ACRR but does not require the use of a driver ore. Instead, the reator ore onsists solely of targets whih serve the dual purpose of a fbel element and 99M0prodution soure. This researh is to determine if a ritial reator ore ould be onstruted with a reasonable number of targets that would still be able to meet the U.S. demand for gmo. Los Alamos National LaboratoV s Monte Carlo N-Partile transport ode alled MCNP (Version 4A) was utilized as the major analysis tool to model and evaluate the neutronis of thk reator. The projet onsisted of a neutronis and eonomi analysis omposed of several major omponents: 1. onstrution of optimized ombinations of moderator materials and target numbers whih yield a ritial system, and a parametri study on the variables whih impat the system ritiality suh as target pith ladding materials, 235Umass loading, and the number of targets in the ore; 2. an analysis of temperature effets on kff and assoiated reativity losses due to hanges in moderator temperature and density as well as fbel temperature; 3. investigation of the effets of fission produt poisoning from 135Xeand 49Sq neutron poisoning from ~i prodution in beryllium-moderated systems, and assoiated reativity losses; iiel bumup analysis based on operation time and ore size; and an analysis of the eonomi viability of the proposed design with a omparison to other urrent and proposed methods in terms of onstrution and waste disposal osts. The results of this investigation prove there are signifiant advantages assoiated with this reator onept, The results demonstrate that the proposed system is a unique design that is both a neutronially and an eonomially viable option to 9?v10prodution. This fiuther validates that the reator an meet or exeed the urrent 99M0 demand for the United States while preserving the Cintihem proess, thereby eliminating the need for modifiation of the urrent target design and extration proess. In additio~ this onept is an improvement of the urrent onventional reator method as the need to use a driver ore, the large amount spent fiel waste generated, and the assoiated osts of spent fiel storage and disposal are eliminated. 9

11 1. ISOTOPE PRODUCTION METHODS 1.1 Introdution This researh involves a omprehensive parametri analysis to determine th~ prodution with a unique and alternative reator method. This report will obtained from parametri studies performed with the Monte Carlo N-Par MCNP (Version 4A) whih was used to model a series of plausibl onfigurations with Cintihem-type targets. An overview of the urrent state-of-the-art prodution methods for medi given in Chapter 1. This inludes an explanation of the nulear formation n urrent prodution methods suh as onventional reators, solution-ii aelerators; and a desription of the proposed onept. Chapter 2 explahs ode was hosen as the primary analysis tool, the parameters investigated, obtained from the optimization of speifi reator onfigurations. Temp effets on &fi in terms of reativity loss are summarized in Chapter 3 for ritial onfigurations hosen from the optimized reator onfigurations pres( Based on the ritial or@urations analyzed for temperature and de] onfigurations were hosen from this list of andidates for fiuther analysis evaluates the prodution of fission produts and poisons. Primary fission ] here are 135Xeand 149Sm. For onfigurations ontaining beryllium as the mo~ the prodution of Li, helium gas, and tritium are also summarized. An anal! losses due to 235Uburnup for a given operation time is presented in Cha quanti~ the neutronis results fkom an eonomi perspetive, Chapter 6 de! analysis inluding the approximate osts of onstrution and waste dispos onept. These estimates are ompared to estimates in the literature for methods. Appendies ontain sample hand alulations, and a sample opy and output file from a representative reator onfiguration. 1.2 D.O.E. Isotope Prodution Projet& History of 9M( Medial isotope prodution was pioneered in the United States. The first g was developed by Brookhaven National Laborato~ in the 1950s. Unti government was the main produer of 9gMo using reators at Brookhax National Laboratories. At that time, the private setor developed the a isotopes on a ommerial sale (Bragg, 1996); however, government reator as bak-up prodution soures. Domesti private setor prodution was initi General Eletri Test Reator (GETR) in Pleasanton, California and the Cir in Tuxedo, New York. Prodution eased at the GETR in 1977 wh deommissioned, leaving the United States with only one major prodution fi 10

12 Prodution ativities were suspended in 1989at Cintihem when itwas determined that ostly reator upgrades would have to be done to meet requirements by the Nulear Regulatory Commission (NRC). Beause of this and the reator s age, the deommissioning option was subsequently hosen by Cintihem and began in Deember With the losure of the Cintihem reator, the U.S. ontinues to rely on the National Researh Universal (NW) reator in Chalk River, Ontario, Canada (operated by Nordlon International) to provide approximately 60?40of the supply of 99M0to pharmaeutial ompanies for distribution. The remaining 40?40is supplied by a European soure. Due to an impendkg vulnerability to the U.S. supply of 9%40 and its reliane on foreign soures; Congress tasked the United States Department of Energy (DOE) to establish the Isotope Prodution and Distribution Program (IPDP). The mission of the IPDP is to establish a domesti 99M0prodution apability, ensure a reliable supply (up to 100VO of the U.S. demand), and pursue eventual privatization of prodution (Carroll, 1997). In November 1991, the DOE purhased the rights to the Cintihem tehnology to ensure ontinuity and stability in the prodution proess, and to redue the amount of time required for FDA approval. A omprehensive feasibility study was onduted to determine the most feasible DOE faility to serve as the prodution site. On ompletion of the study, the Annular Core Researh Reator and adjoining Hot Cell Faility (HCF) at Sandia National Laboratories were hosen as the reator and proessing site; and the Chemistry and Metallurgy Researh Complex at Los Alamos National Laboratory was hosen to manufature the irradiation targets. The Reord of Deision (ROD) was issued by the DOE in Otober 1996, and physial modifiations to the ACRR were ompleted a year later. However, physial modifiations to the HCF are urrently in progress and prodution ativities are not slated to begin until summer 1999 at the earliest. The NRU reator has ontinued to supply the U. S. market for *Mo despite two labor disputes in June 1997 and May 1998 whih temporarily shutdown the reator and dkrupted the supply. These events have reated heightened awareness of the vulnerable state of the supply. Atomi Energy of Canad~ Ltd. (AECL) of Canada has proposed and begun onstrution of two Maple reators to take over the prodution horn the NRU reator when it is shutdown and deommissioned in the year However, this does not mitigate the possibility of fiture disruptions to the supply as these new reators still require the U.S. to rely on a foreign soure. It is hoped that the IPDP s efforts to establish a domesti prodution apability will be i?uitfid in providing an immediate prodution soure with the ACRR and enourage the reestablishment of a private setor prodution faility. 1.3 Nulear Formation & Medial Uses of 9M0 / ggmt ~o is the preursor to its daughter isotope, 9bT whih k the most widely used isotope for nulear medial proedures in the United States today. About 36,000 nulear diagnosti proedures are perl?orrned daily in addition to approximately 50,000 mediinal therapies as well as 100,000,000 lab tests per year with 9fiT. This aounts for between 80 and 90% of all nulear mediine proedures done annually in the U.S. (Parm< 1995). 11

13 The major reasons for its widespread use are due to its many valene states, low energy gamma (143 kev), and short half-life (6.01 hr). This redues the amount of time the isotope resides in the body and minimizes the radiation dose to other parts of the body unaffeted by the disease or ailment the isotope is being used to image. Medial imaging studies with 99mTare done on the brain, bone, liver, spleen, and kidneys, as well as in blood flow studies. 99M0is shipped daily to pharmaeutial houses where 99mTgenerators are produed. 99mTis milked fi-om the generator as the 99M0deays and is subsequently injeted into the patient in various hemial forms. Compared to the time sale of long-lived isotopes, the half-lives of ~o and 99mTare sho~ about 66.0 hours and 6.01 hours, respetively. Thus it is imperative to ontinually produe the 99M0preursor to offset the loss of produt due to radioative deay, most of whih ours during proessing and transportation. The urrent demand for the United States is approximately 19,700 Ci/wk at the soure (inluding losses based a 30-hour proessing and shipping time), or about 3,000 6-day Ci/wk at the pharmaeutial house dok (Parm~ 1997). A 6-day Ci is defined as the amount of radioative material reeived at the pharmaeutial house dok with a alulated deay time of six days. The soure refers to the loation where 99M0is produed. The world demand is about twie the U. S. demand. 99M0 an be~n-edued in a nulear reator either through the fission of 235Uor by neutron ativation of Mo. Fission prodution of 99M0requires the interation of a neutron with a 235U nuleus to indue the formation of fission fkagments. The lighter of the two fission fragments is typially on the order of 95 AMU and an be up to 99 AMU whih would orrespond to a 99M0 nuleus. In ontrast neutron ativation requires a neutron to interat with a 98M0 nuleus subsequently induing an exitation reation to form a 99M0 nuleus. Figure 1 illustrates the nulear formation of 99M0and the physial differenes between fission prodution and neutron ativation. The formation of a 9~o nuleus ours in approximately 6. l% of all possible fission events. One a 99M0 nuleus is formed via 235Ufission or neutron ativation two forms of 99T are produed via beta emission. The most probable of these is the metastable isotope 9hT, and its emission of a low energy gamma whih are of primary interest to pharmaeutial mediine. After deaying with a short half-life of about 6 hours, the metastable state of 99mTdeays to a longer-lived form of 99T and then ultimately to the stable isotope of 99Ru. Figure 2 is an illustration of the 99M0deay hain and its daughter produts. 12

14 Fission: *% (n,f) 99M0 19,700Ci Mo = 2.5xl@0atoms 9M0= 0.041g Gamma light fission fraglwrlt peak -95 amu Neutron Z7qJ Atom,f 2 f 2.5 Neutrons z w - 7: V heavy fission fragnwrt peak -136 amu Ativ*on: *Mo(n,y) Mo *Mo= /ia natural abundana Neutron - Gamnw # J 0 o 9M0 Atom Figure 1. Nulear Formation of 9?M0. 13

15 m -1 IT (143 kev y) m Figure 2. 9?M0 Deay Chain and Daughter Produts. There are several major reasons why the fission prodution of 99M0 is preferred over neutron ativation. The major disadvantages of neutron ativation are:. the low speifi ativity of 99Mogenerated;. the requirement of large 99mTgenerators; the generation of large amounts of spent fiel; and. the need to use enrihed *Mo. In the ACRR for example, irradiation at a neutron flux level of 2 x 1013n/m2-se generates a speifi ativity for fission produed 99M0 of 75,000 Ci/g, whereas the speifi ativity generated through neutron ativation is only 0.39 Ci/g, a differene in speifi ativity of about 200, ). This illustrates why a high power reator with a large irradiation spae would be needed to obtain a speifi ativity omparable to that of fission produed Me. Furthermore, suh a large, high-power reator would onsequently produe large amounts of spent fiel. For a prodution level of 16,500 Ci/w~ approximately 47.7 kg of filly enrihed 9*M0would need to be irradiated (l%rrna, 1995). In ompariso~ 37 targets (eah with a mass loading of 20 grams of 235U)ould be imadiatei at a low reator power level to meet 113 XOof the total U.S. demand (krrn~ 1997). However, there are also disadvantages with using the fission proess to produe gmo. First, highly enrihed uranium (HEU) on the order of 93 /0 235Umust be used in the targets. This generates large amounts of low-level waste (LLW) whih need to be reproessed or disposed, in omparison to neutron ativation whih does not use uranium targets. Seond, hemial proessing must be used to separate the 99M0from the uranium whereas neutron ativation does not require a hemial separation proess. Unfortunately, a hemial extration proess an generate relatively large amounts of waste. However, unlike other extration proesses, the Cintihem proess does not generate any mixed waste.

16 There are also trade-offs with the fission proess whih enhane the prodution of gmo, but simultaneously ompliate the waste disposition proess. First, the power level in the tar et an $ be inreased by an inrease in the 235Umass loading. For example, by adding 5 grams of 35Uto a 20 gram target, the power level in the target is inreased by a fator of Consequently, this inreases the amount of LLW that must be disposed o~ with a orresponding inrease in the target oating thikness whih is limited by the eletroplating proess and adherene issues (l%rma, 1995). Seond, the stainless steel ladding of the target dereases the fission prodution by Up to 30Yo. To redue this figure, a redesign of the urrent target would have to be done inluding hanging the ladding to a less-absorbing material suh as zironium. Researh is urrently being done at Argonne National Laboratory in Illinois to develop targets whih utilize low enrihed uranium (LEU) in the form of a thin foil (-125 ~m thik) sandwihed between two slightly tapered inner and outer tubes of aluminum and zironiurq respetively (Hofinan et al., 1997). To date, only thermal tests have been done with this onept, and it is unlear whether this would be a feasible alternative to the urrent HEU targets. Proliferation onerns have been expressed onerning the use of HEU to produe gmo. However, there are some drawbaks to the use of LEU targets. First, it has not been determined whether the U.S. demand for 99M0an be met with the use of LEU targets. Seond, the United States Food and Drug Administration (FDA) has not given approval to the method of 99M0prodution with LEU targets. Beause there is more prodution of 23??uthrough (~y) reations with the large amounts of 23*Uin an LEU target, this presents a major problem in obtaining radionulidi purity of the final produt as required by the FDA and the pharmaeutial ompanies. 1.4 Target Irradiation & Extration Proesses Several riteria must be met for Mo to be produed by the fission proess. First, the prodution. faility must be able to handle and store relatively large quantities of Speial Nulear Material (SNM) in the form of targets. Seond, the faility must have a reator dediated to prodution ativities. Third, a hot ell must be present to proess the targets, prepare the final produt for shipment, and temporarily store solidified liquid waste. Fourt~ a waste disposal site must be hosen to store radioative waste generated. And finally, a distribution site must be nearby to provide timely aess to transportation (Coats and Parma, 1994). Target irradiation is typially the proess used to produe 99M0 in nulear reators. The Cintihem-type target whih will be used in the ACRR onsists of a stainless steel tube 18 inhes (45.72 m) long and 1.25 inhes (3. 18 m) in diameter. The thikness of the tube is about 30 roils ( m), and is oated by eletroplating the inside with a thin layer of approximately 20 grams of 235Uin the form of 93 /0 enrihed U02 (Parm~ 1998a). The top and bottom of the tube is sealed at both ends with welded end fittings, and the inside is bak-filled with helium gas. The top end fitting allows the target to be handled remotely when removing it from the reator, and the bottom end fitting has a hollow ylinder inside that ats as a ather for any oating that flakes off flom the inside of the tube. Figure 3 is a diagram of a typial Cintihem-type target with the appropriate dimensions. Eah target is irradiated in the reator ore for 7 days (24 hrs/day). After the irradiatio~ several targets are removed fi-omthe reator and are transferred to a hot ell for proessing and preparation for shipping. 15

17 T18 in. (45.7 m) + Top End Fitting Stainless Steel Tubing roils ( mm) <. Bottom End Fitting in. (3.18 Top End Fitting ) SS Welded Plug Figure 3. Shemati of a Typial Cintihem-type Target 16

18 Target irradiation in general requires a hemial extration proess to dissolve the uranium from the target and to preipitate the 99M0fission produt from the uranium. There are several methods for hemially extrating 99M0produed from the fission proess. The first is alled the Cintihem proess whih was previously used by CintiheW In. to produe 99M0and will be used by Sandia in its prodution ativities. It is a omplex but robust tehnique whih relies on an aid oktail to dissolve the uranium oating from the inside of a target and subsequently uses a preipitation proess to extrat the 99M0. The waste fission produt-uranium solutioq target, glassware, and assoiated equipment beome radioative waste whih must be disposed. Overall, the Cintihem proess is very effiient and has some areas whih an be fbrther improved upon, but it does not generate the large amounts of waste assoiated with the UAI (uranium aluminide-aluminum dispersion) method used by Nordion to produe 99M0 from irradiated targets (l?arma, 1995). The seond method for fission-produed 99M0is a fission fragment deposition method. This entails diret fission fragment deposition into the reator moderator liquid or oolant gas that is eventually proessed. A third method is similar to the seond as it involves diret fission fragment deposition, exept this proess relies on the deposition to our in the fbel whih is at a high temperature and over time sublimes the leakkg fission produts into the oolant gas whih is later proessed. 1.5 Current Isotope Prodution Methods The three major prodution methods desribed in the following setions are onventional reators, solution-fbeled reators, and aelerator appliations of 99M0 prodution. General attributes of these methods as well as the advantages and disadvantages assoiated with eah are disussed. These methods are ompared to the proposed onept desribed at the end of this hapter Conventional Reators The use of a onventional nulear reator typially requires the irradiation of targets in a ore of iiel elements whih fimtion as a driver ore. The primary fimtion of the driver ore is to provide a seondary neutron soure to the targets to aid in the prodution of 99M0via the fission proess of 235U. For example, the entral region of the ACRR ore will aommodate up to 37 targets for irradiation. Figure 4 illustrates the ACRR ore prior to modifiation for 99M0 produtio~ and the modified ore arrangement of targets and fhel elements. The NRU reator in Canada uses the target irradiation method to produe Me. There are other reators throughout the United States whih are also apable of produing 9?kfo. Among them are the ATR (Advaned Test Reator) at the Idaho National Engineering & Environmental Laborato~ in Idaho, the High Flux Isotope Reator (HFIR) at Oak Ridge National Laboratory in Tennessee, and the Mksouri University Reator (MU RR) at the University of Missouri- Columbia. These also use the target irradiation proess, but instead either use modified targets whih are slightly different than the typial Cintihem-type targets or irradiate fiel elements as in the ase of MURR. 17

19 Current ACRR Configuration -236 Fuel Elements Water +, Refletor 00000( 2. Nikel Plate,-,, @ Fuel Element Control Element - Fuel Followed Safety Element - Fuel Transient Element Nikel Elements Aluminum Elements Empty Grid Loation o Cintihem Target Nikel Plate - (FREC Attahment) 129 Fuel Element Baseline Core -37 Targets Water Fuel Element Control Element - Fuel Followed Safety Element - Fuel Followed Transient Element Nikel Elements O Cintihem Target Figure 4. Cross Setional View of the Original& Modified ACRR Core. 18

20 In general, target irradiation is the most ommonly used and doumented method for fissionprodued 99M0today; however, the major problem. with this method is the vast amounts of waste generated in prodution ativities. The power level at whih the reator is operated, the number of fiel elements in the driver ore, and the resulting fiel burnup determines the ore lifetime and ultimately the amount of spent fiel generated. k additioq LLW is also generated from the proessing of targets. Unfortunately, the large amounts of waste generated with onventional reators pose a problem from an eonomi and storage standpoint. This is the main argument against the viability of the target irradiation method whih has prompted the searh for alternative methods Solution-Fueled Reators Solution-fheled reators were among the first types of reator systems investigated for power prodution purposes ikom the late 1940s through the 1960s. Among the first was the SUPO reator at Los Alamos, a 25 kw reator used to experimentally measure radiolyti deomposition of water. Reent proposals have been disussed and submitted regarding the use of this type of system to produe gmo. To date, only small sale tests have been done with this type of reator to determine its feasibility (Glenn et al., 1996); however, large sale tests have not been onduted to evaluate if it an handle fill sale prodution. Advoates of this proess laim the use of solution reators an ompete with or even outperform the target irradiation methods used in onventional reators. Babok & Wilox (B&W) Environmental Servies, In. has proposed a passively ooled, solution reator dediated to 99M0 prodution alled the MIPR (Medial Isotope Prodution Reator) (Ball, 1995). The reator onsists of a Ziraloy ylindrial vessel filled with a lowenrihed (-1 9 /0235U) aqueous solution of uranyl nitrate tie], and is immersed in a pool ofwater whih fimtions as a heat si~ refletor, and radiation shield. A ower level of 200 kw has been 8 proposed and estimated to produe approximately 2,000 Ci of Mo daily. The reator an be shutdown for about 1 hour daily to remove fiel for proessing and extration. The solution is passed through an ion exhange olumn to extrat the gmo, and the exess uranium and fission produts are routed to a dump tank. This remaining exess solution is pumped bak to the reator vessel where it is reused for another prodution yle. Among the advantages ited by B&W for this reator are its inherent safety harateristis (large negative power oeffiient of reativity), the ability to proess and extrat the 99M0 online, ontinuous operation at a lower reator power level than a onventional reator, and a redued inventory of fissionable material generation (Chopelas and Ball, 1993). The major advantage laimed by this onept is a signifiant redution in the amount of waste generated beause the fhel an be reused for additional prodution yles as all of the fission produt deposition ours in the I%el. This eliminates the need to manufature, purhase, and dispose of targets and fbel elements and their assoiated waste. 19

21 A seond type of solution reator proposed for Mo prodution is based on a irulating solution reator onept developed by Los Alamos National Laboratory alled the KING reator (Hills and Heger, 1997). The original intention of this reator onept was to serve as an intense neutron soure as it operates at a very high reator power, approximately 25 MW. The reator onsists of an irradiation region filled with water in the entral region of the ore, the fbel region is ontained in zironium ladding outside the water regio~ and the outside of the ore is surrounded by a beryllium refletor. The fhel onsists of a 93.5 /0 235Uenrihed solution of uranyl sulfate, U02S04 with a fiel onentration varying between 86 to 220 g 235U/L. It is estimated that if the reator were operated at the nominal power of 25 MW that 15,000 Cl of 99M0 ould be produed in approximately 1.6 hrs (HiIls and Heger, 1997). The major reason given for the apability to operate this type of a solution reator at a muh higher power as ompared to the MIPR is the inherent heat apaity in a rapidly moving fiel solution. A seond advantage ited is the elimination of an internal heat exhanger beause the fbel an be ooled outside the reator in a subritial state. In general, solution reators seem to have an advantage over onventional reators as they do not require the use of or generate waste with solid fiel assemblies and irradiation targets. However, there are several important issues whih need to be investigated fiu-ther with this prodution onept. Among these are heat removal, operating temperatures, and power level duration and limitations. There are urrently no solution reators operating on a regular basis in the U.S. in the tens of kw range. In omparison to the power levels proposed for the previously mentioned reator onepts, the SUPO reator s peak power was only 25 kw. A solution reator in Taiwan has been identified as operating at a power of 100 kw but this has not been onfirmed (Varley, 1994). Other issues of onern are the reombination of 02 and H2 due to the radiolyti deomposition of water in the fuel, and the orrosive nature of the fiel solution on reator omponents. Reombination systems are neessary with a solution reator due to the large prodution amounts and the explosive nature of 02 and H2. In pressurized systems, orrosion problems have been found to our at -225 C and phase stability problems at 300 C (Lane et al., 1958). Assuming suessfi.d solutions are found to the previously mentioned tehnial and safety problems, this type of prodution method still requires FDA approval whih ould prove to be lengthy and umbersome Aelerators In addition to onventional and solution fbeled reators, aelerator systems an also be used to produe gmo. Aelerators primarily rely on spallation neutron reations and harged partile interations to ahieve results equivalent to fission and neutron ativation proesses used in a reator. There are numerous possible reation ombinations, but only a handful are feasible as their effetiveness is determined by ross setion and threshold energy. Among these reation mehanisms are neutron and proton ativation, deuteron interation and fission reations suh as 9*MO(Ry)wMo, 9*Mo(d,p)9%40, 100Mo(p,2n)9wT, 1 0Mo(p,2p)WMo, and 23*U(p,f)99M0(US DOE, 1995). 20

22 The Los Alamos Meson Physis Faility (LAMPF) aelerator at LANL has been used to ondut experimental 99M0 prodution studies by bombarding natural uranium targets via a 238U(p,f)reation with high energy proton beams. Studies in the late 1980s showed that the aelerator was apable of produing 1,250 6-day Ci of 99M0per week by bombarding a natural uranium target with an 800 MeV proton beam at a urrent of 100 VA (Howe et al., 1996). Subsequent studies were onduted with lower energy beams of 200 and 400 MeV to determine if 99M0prodution is feasible with spallation and low-energy neutron interations. Tests with a 200 MeV proton beam having a urrent of 1 VA inident on a target for 24 hours produed Ci of 99Mo(Waters and Wilso~ 1996). In additio~ ylotrons have also been used in studies with natural uranium targets. Cylotrons aelerate protons to GeV energy levels. Researh in this area has been done at the University of California-Davis (UCD) and test results have onluded that a 99M0 yield of 17 Ci/hr is obtainable with proton-indued fission in 23*Utargets (Lagunas-Solar et al., 1996a). The same group at UCD has pro osed an alternative to the use of generators used with reator-based methods by produing 9L T diretly with aelerators. This method uses a (p,2n) reation with enrihed l*omotargets to produe the 9*T d~etly without its preursor Me. Test results from this investigation have shown that a 1 hour bombardment of a target with a 1mA proton beam yields 23 Ci of 9tiT (Lagunas-Solar et al., 1996b). However, there are two major disadvantages with the diret prodution method. First, diret prodution of 99mTis not very pratial due to its very short halflife of 6.01 hours, whih presents transportation and prodution level limitations even with reator methods. Seond, the produt was found to ontain tehnetium-96 (4.35 days), a ontaminant whose prodution level is diretly proportional to proton energy. In theory, this method is possible; however, it ould pose serious problems with obtaining FDA approval sine the proess falls short of meeting urrent U.S. prodution level demands and requires subsequent steps in the radiohemial proessing to redue produt ontamination. Two other important drawbaks to aelerator produed Mo are assoiated with the limitations of urrent aelerator tehnology. First, isotopes generated through nulear transmutation generally have very low speifi ativities. Seond, target heating beomes a serious problem with beam urrents above several hundred wa. To meet 100 /0of the U.S. demand for gmo, beam urrents would have to be several orders of magnitude higher than those doumented here (US DOE, 1995). Consequently, aelerator produed 99M0 has been deemed for now to be neither tehnially feasible nor eonomial for meeting the urrent U.S. 99M0demand. 21

23 1.6 Proposed Conept The three major methods presented previously have ertain harateristis whih make them feasible options for 99M0prodution. However, eah has some disadvantages whih limit their apability to produe 99M0 on the ommerial sale required by pharmaeutial needs. Consequently, other methods ontinue to be proposed. Conventional reator methods are the most ommonly used tehniques for 99M0prodution, and thk researh fouses on eliminating the large volume of spent fhel and its assoiated osts for disposal. The main objetive of this researh was to determine if a unique alternative to the onventional reator method ould be devised whih would be able to meet or exeed the U.S. demand for gmo. The proposed onept is a modified version of the onventional reator method with the major modifiation being use of the targets as the ore replaing the usual driver ore of fhel elements. For the reator onept to be a viable prodution option, several riteria had to be met:. the ore must onsist of a reasonable number of targets to maintain ritiality and prodution levels; reator ritiality must be maintained over its lifetime despite fiel burnup and poison prodution; and. the system must be onstruted and operated eonomially. The proposed reator ore is omposed of Cintihem-type targets arranged in a triangular pith submerged in a moderator and oolant material. The exterior of the ore is surrounded radially and axially with a refletor to redue neutron leakage. Figure 5 illustrates a radial view of the reator ore of targets, and the refletor region. One of the major disadvantages of the onventional reator is the large amount of waste generated by spent fbel elements. This translates into a signifiant volume of physial waste over the lifetime of a reator, and a ostly endeavor with regard to the manufature, purhase, storage and ultimate disposal of the spent fuel elements. This proposed onept eliminates this waste volume beause there is less uranium in the target than in a typial fbel element, and eah target serves a dual fimtion as a fiel element and a prodution soure. The next hapter will detail the neutronis analysis and the results of the numerous parametri studies onduted to determine the optimized reator system. 22

24 Moderator between Targets / Crntihem~ al Figure 5. Radial View of a Reator Fueled with Cintihem-type Targets. 23

25 2. NEUTRONICS ANALYSIS This hapter summarizes the neutronis analysis, and explains the results obta parametri studies onduted to determine optimized reator onfigurations. T1 explains why the Monte Carlo approah was hosen over the deterministi odes analyze numerous reator onfigurations. Seond, the moderator materials whh with various target onfigurations are desribed in terms of their neutroni advantages. Third, the various parameters investigated with eah reator or disussed. And finally, the onstrution and analysis methodologies of the onfi their respetive IGffresults are presented along with the major differenes between 2.1 Monte Carlo vs. Deterministi Codes Numerous algorithms have been developed over the ourse of several dead{ partile transport in various media. The speifi type of partile transport for this ~ of neutrons in a fissile system. The primary parameter of interest for measuring fissile system is the effetive multipliation fator (k&). bff is defined as tl neutron prodution gains relative to the losses inurred through absorption and aount for these proesses, the Boltzmann transport equation is used to de; transport. Consequently, to obtain a result for b, this equation is solved as problem. Two major methodologies used in neutronis analyses to simulate neutron tra Monte Carlo and the deterministi methods. Several major differenes are evidel methods determine the Lff of a fissile system.. the algorithm they use to obtain a result for lgff;. differenes in ross setion data libraries;. apabilities for modeling omplex geometries; and omputer time required to omplete a given alulation. Deterministi odes solve the Boltzmann transport equation diretly to obtain an ~ bff using the disrete ordinates method. This algorithm divides the spae throl partiles travel into small boxes. lgffis alulated by measuring the neutron bl differential portion of spae. In ontrast, Monte Carlo methods do not solve equation diretly; rather, they use a statistial approah to obtain a result for ~ff the stohasti behavior of individual partiles and reording various aspets of this time. A Monte Carlo ode utilizes a set of random numbers to generate a statisl eah partile through what is alled a random walk analysis. Random numbers a interation to determine what proess the neutron will undergo: fission, absorption As eah neutron is traked throughout a media, the amount of energy lost or neutron, how many new neutrons are reated from fission, and the diretion of interation are among the measured parameters. 24

26 hotherdifferene istheformat ofneutron ross setiion datalibrafies utilized in both odes. A neutron ross setion is the probability that a neutron will undergo a ertain interation in its lifetime suh as fissio~ sattering, or absorption. The random numbers generated by the ode are diretly orrelated to the ross setions in the data library and are used in the alulation to determine how a neutron will behave in various materials. Deterministi odes generally rely on multigroup ross setions whereas Monte Carlo odes typially utilize ontinuous energy ross setion data. Some Monte Carlo odes are also apable of using multigroup data. Multigroup ross setions average the probability of interation with respet to a flux weighted spetrum over a speifi range of energy bins. Consequently, the flux spetrum of the modeled system must math the spetrum of the multigroup data set being used for the alulation to be orret. In ompariso~ ontinuous energy ross setions are not averaged aording to a spetrum but are instead point-wise ontinuous over a wide range of energies and thus an be applied to a wider variety of problems. A major advantage of Monte Carlo odes is their ability to model and analyze omplex, heterogeneous geometries. For a deterministi ode to analyze the same geomet~, the geometv must be simplified or homogenized. However, a disadvantage of Monte Carlo odes is the amount of omputer time required for a alulation. This is generally a fimtion of the omplexity and omposition of the modeled geometry, the number of neutron generations simulated, and the number of neutrons per generation the ode is required to trak. Deterministi methods are generally quiker beause they are limited to simplified geometries, utilize simpler ross setion data sets, and solve a mathematial equation diretly rather than simulating neutron behavior. The main tool used in the neutronis analysis for this projet was Los Alamos National Laboratory s Monte Carlo N-Prutile transport ode alled MCNP (Version 4A) (13riesmeister, 1993). MCNP is a robust and versatile ode that is widely used for neutronis analyse~ in reent years, its use has expanded worldwide. There are several reasons why MCNP was hosen over deterministi odes to onstrut and simulate this reator onept. Firs~ utilization of ontinuous energy ross setions proved more advantageous than using mukigroup data beause the neutron spetra of the modeled onfigurations were not expliitly determined. Using thk type of ross setion data redued the possibility of modeling a system with a different spetrum than that of the ross setion library. Seond, modeling a heterogeneous system of targets and moderator with a deterministi ode requires physial and alulational homogenization of the ore. With the numerous parametri studies onduted, this would introdue addhional error with respet to ensuring the ore was properly homogenized and the orresponding atom densities were used. Finally, the required omputer time for a Monte Carlo alulation has been signifiantly redued with the advent of faster omputers in reent years. It has beome a more viable option onsidering the reent advanes in omputer speed and tehnology. 25

27 2.2 Moderator Materials Reators generally require the use of a speial material alled a moderator t down fission neutrons. In some reators, the moderator also serves a dual material to ensure proper and effiient removal of fission generated heat fr material to be onsidered a viable hoie for a moderator, it must possess se~ a low mass number, a small absorption ross setion, and a large sattl However, no speifi material possesses all of these traits so some trade-offs when hoosing a moderator for the reator design. For the proposed moderators were tested in the parametri analyses with MCNP: water; b oxide, graphite, and heavy water. Most reators are moderated and ooled with ordinary water. It is relatiw exellent slowing down properties, has a small migration length for therm also be used as the reator oolant. However, for a system to attain and enrihed uranium fbel must be used. The use of enrihed fiel provides for due to the small thermal neutron migration length. Beryllium has a low abs but a reasonable sattering ross setion. A major neutroni advantage form with beryllium is its ability to produe (q2n) reations. However, if properties, low thermal ondutivity at high temperatures, and high ost to fa generally unfavorable. Beryllium metal is very brittle and this is enhaned : due to the formation of helium gas in the voids of the metal through (~x) rl oxide has neutroni properties very similar to berylliu~ is easier to fabriat( expensive. Utiortunately, it vaporizes rapidly in the presene of moisture z and leads to the formation of beryllium hydroxide. A major drawbak wil ompounds are their extreme toxiity to the body when inhaled. Graphite is the most widely used of the solid moderators. It has good neut not as good as those of beryllium. It is less expensive to fabriate than be] mehanial and thermal properties. It also has good fission produt gas whih make it omparable to the ontainment harateristis of typial fix Utiortunately, graphite has a tendeny to ontrat and expand at high irra due to high neutron fluene and the anisotropi nature of its atomi struture. Heavy water has relatively good slowing-down properties, a very low abso] and a reasonable sattering ross setion. This enhanes the resonane est thermal utilization of the fbel. Another major advantage of heavy water is 1 an be used as fiel, enrihed fiel is not required for operation as it is in reator. However, heavy water is expensive, rather limited in distrib pressurization if it is used as a oolant in addition to a moderator. 26

28 2.3 Parametri Investigations To reate a feasible reator onfiguration, it was neessary to perform parametri studies to determine optimum ombinations oftargets and moderator. Thetargets andreator vessel were modeled with MCNP in three dimensions. Eah target onsisted of a series of onentri ylinders with planes at the top and bottom to give the target a finite height. A reator ore was onstruted by enasing the targets in a hexagonal lattie to arrange them on a triangular pith. Figure 6 is a two-dimensional shemati showing a radial and axial view of the target with the moderator surrounding the target s omponents as modeled in MCNP. The thiknesses of the UOZlayer and stainless steel ladding in Figure 6 are thiker than those in an atual target and exaggerated here for larity. L Target ladding Void \ Moderator - Radial View \ U02 layer Axial View Figure 6. Radial and Axial Views of a Single Cintihem-type Target Modeled in MCNP. 27

29 MCNP Target Model Atual Cintihem-type Target Figure 7. Cross Setional Views of MCNP Modeled and Nominal Cintihem-type Targets. Figure 6 also illustrates an axial view of the target with the top and bottom end aps omitted from the model. These end aps were omitted assuming the &ff results would not be ompromised by a gross over-simplifiation of the geometry. Figure 7 is a ross setional view illustrating the differenes between the modeled target and an atual target. It was determined that adding the end aps to the targets would have reated a more omplex model. In additio~ the resulting lg~ would have been less onservative beause the additional stainless steel reates more neutron absorption. To gain a omprehensive understanding of the dynamis whih affet this type of syste~ several major parameters were investigated in the analysis. Among these were: number of targets that would yield a ritial system;. optimum moderator material; 235Utarget mass loading; target pith or spaing; and a omparison of the nominal stainless steel ladding to a material with less neutron absorption suh aszironium. The baseline number of targets was 37, whih orresponds to the maximum number of targets that will fit into the entral region of the ACRR. This number of targets was hosen for two reasons. First, this number of targets gives a basis from whih to onstrut a reator and ompare to the ACRR. Seond, assuming an irradiation shedule of seven days, this would orrespond to a proessing rate of between five or six targets per day to meet the U.S. demand. 28

30 For the five moderators various reator mfigurations were onstruted via an inremental methodology through three parametri optimization studies. The first study onsisted of onstruting a reator ore with targets infinite in height surrounded by a infinite radial refletor of the moderator material being investigated. Thk study was done to determine the optimum target pith. For the solid moderators, the target pith and oolant hannel radius were optimized simultaneously. The optimum oolant hannel radius was obtained by adding 0.20 m inrements of water around eah target beginning with a minimum of 0.0 m up to a maximum of 1.0 m. The seond study onsisted of finding the optimum radial refletor thikness. This was done by keeping the targets inilnite in height and reduing the radial refletor in inremental steps from an infinite size (two neutron dlfision lengths) to a bare system. For the solid moderators and heavy water, an infinite refletor of ordinary water was put on the outside to fi.uther redue neutron leakage. The third and final study onsisted of optimizing the axial refletor thikness. This was done by using targets of finite height arranged on a defined pith surrounded by a finite radial refletor determined by the previous two studies. The optimum axial refletor thikness was determined in the same inremental manner as the seond study. An infinite refletor of water was added to the outside of the optimized refletor to redue neutron leakage in the axial diretion. The following setion will summarize and explain the kff results obtained from the onfigurations onstruted with the above methodology in MCNP. A summary of the optimized dimensions for eah reator onfiguration will be haraterized for eah moderator in terms of the Lff results. In additio~ reasons for modifiations to the nominal target design suh as the addition of solid plugs in the target entral region will also be disussed. These studies pertain to onditions at 27 C (300 K) and were evaluated with appropriate ross setions for this temperature. 2.4 Reator Design Results lgmverifiation Cases MCNP (Version 4A) was run on several personal omputers to expedite the proess of exeuting several hundred input files omprising the numerous parametri studies done for this analysis. This version of the ode was previously ompiled with the Lahey Fortran Compiler, Version 4. IL to run on a personal omputer. All of the mahines used were stand alone mode, IBM ompatible omputers with Pentium proessors of various speeds. Three input files were hosen to veri~ that the ode is mahine independent, and that no signifiant differenes should be expeted in the final keff and standard deviation results regardless of the mahine used. A heterogeneous ase was tested to determine if the geometri omplexity ontributed to any differene in the final lgffdepending on the mahine used. Two homogeneous ases were hosen beause they are simple models that yield a km result quikly. 29

31 The first input file models a heterogeneous reator ore omprised of 85 Cintihem-type targets arranged in a triangular pith arrangement in a water moderator with 10 m radial and axial water refletors. Eah target was set on a pith of 3.95 m ontained a 235Umass loading of 30.0 grams, a stainless steel ladding thikness of 30 roils ( m), and the entral void filled with a beryllium plug. The remaining ases modeled two fast, ritial assemblies loated at Los Alamos National Laboratory. Both of these reator systems are onsidered homogeneous systems as they are omposed only of fbel without moderator. The first of these files modeled the Godiva ritial assembly. It onsists of a bare, 93.71w/0enrihed 235Usphere with a radius of m. The seond fast system modeled was the Jezebel ritial assembly. This is a bare, plutonium sphere onsisting of a mixture of 23%%,240Pu,241Pu,and approximately lw/o gallium. The sphere has a radius of m with a m thik outer nikel shell. Table 1 presents the lgffresults and assoiated standard deviations for the three ases run on five different mahines. Eah ase was run with 400,000 partile hktories. No signifiant differenes in keffand the standard deviations were obtained among the five mahines tested for the Godiva and Jezebel assemblies. However, slight differenes are evident when omparing mahines 3 & 5 to mahines 1, 2, and 4. These deviations may be attributable to differenes in the hardware arhitetures of the omputer systems. Consequently, this test proves there is onsisteny in the bff and assoiated standard deviation results obtained from these mahines for the reator optimization analysis. Table 1. ~, Results of MCNP 4A Verifiation Cases. Mahine Type Case 1 Case 2 Case 3 85 SS targets, water Godiva assembly Jezebelassembly sphere) (barepusphere) ~ DellOptiplex GXMT (133MHZ) =E CompaqPresario9546 (1OOM3Z) * * * AessTehnologies ATCLXP * * (266MHz) I I DellOptiplexGX1 (400MH2 ) z!= * * Gateway 2000P5190 (90MHz) * h Water Moderator The first set of designs were onstruted with Cintihem-type targets in a reator vessel filled with water moderator. An axial and radial refletor of water 10 m in thikness ompletely surrounds eah reator ore. Table 2 presents the &ff and standard deviation results for the twenty water moderated onfigurations evaluated. 30

32 Table 2. MCNP Results for Water Moderate TamzetI Total ICladding ITarget I Moderator/I Radial/, Pit~h Number Mass Materia~ Cen& Refletor Reflea (m) of (grams) Material Thilmess ITa~~@* I - [ Ss I h J HXl/ --L-.--A.. H,O I 10.0/ ii 28.2 Zr & H20/ HZO 10.0i Ss H20/ HZO 10.0/ Zr m HZO/ HZO 10.OI Ss H,O. H,O/ HZO 10.0/1 A 1 <n I 0< I 30~ J.JV I 7. I Un I Vnll u Iqv n2w JH I Ss Be H20/ HZO 10.0/ zl Be H,O/ H,O 10.0/1 -..,, --/.-, I ii 30.0 Ls I ie I Eo/i. FLo 10.0/1 2 nn <% I ~n n I 7- D. I Untun t J.7U U1.7U. u LL DG nyj 1KL2U 10.0/ Ss Be H20/ H / Zr Be H,O/ H,O 10.0I 1 *Alltargetsused93 /0enriheduranium. bff results in bold fae type denote ritial reator ores. The result number of pratial reator ombinations an be onstruted in a w is primarily due to the strong neutron absorption harateristis o Most of the onfigurations omposed of zironium lad targets yiek The first set of designs onsisted of targets with stainless steel o~ ( m) in thikness, a mass loading of 28.2 grams of 235U, reator ontaining zironium lad targets was the only onfigurati( attain ritiality. Based on the results of these ases, onstruti Cintihem-type targets, espeially targets with stainless steel ladd impratial and insuffiient as a viable design hoie. A seond parametri study was onduted to determine alternative! nominal target design to attain ritiality with a reasonable numb targets and inreasing the mass loading of 235U in the tar ounterprodutive as this would inrease keff while simultaneous waste generated later. Thus it was determined that adding a mod region would be a more effiient alternative to offset the abso ladding. 31

33 A filled enter region was simulated with three moderators: water, beryllium and polyethylene. Water would be used to flood the target enter. Beryllium and polyethylene would be ast in the form of solid plugs whih would fit into the entral region and be reused for subsequent irradiation yles. Figure8 ompares the ~resultsfiom the target pith optimization study for 85 stainless steel targets using these fill materials to a target with a dry target enter. Figure 9 is a omparison of the L results as a fi.mtion of target pith for 85 zironium lad targets : ; : } E +H20 in enter ) + Be plug ) + Polyethyleneplug -v- Dryenter >... t-l Target Pith (m) Figure 8. k Results for Water Moderated Reators Using 85 Stainless Steel Clad Targets and Various Target Center Materials. 32

34 1.40 t,,, -1 l.3j j { ~ j j 1.30 J...i...i CO 0.s s5 0.s o.e5 0.60~ I I 1! < OU 5.50 Target Pith (m) Figure9. kresul@ for Water Moderated Reators Using 85 Zironium CIad Targets and Various Target Center Materials. Higher b values were obtained for zironium lad than for the stainless steel targets, regardless of the entral region ontents. Additionally, the optimum pith was smaller in ases with a filled enter than with the nominal dry enter target. Generally, optimum pithes for filled targets were 0.50 to 1.50 m smaller than those for dry targets depending on the fill material. A pair of reators with 37 targets were evaluated using both ladding materials and beryllium plugs. However, these ases yielded very subritial systems so designs with 85 targets were fhrther evaluated. For stainless steel targets, the use of beryllium plugs yielded slightly higher bff values than those with polyethylene plugs. High lgff values were obtained for these moderators beause of their sattering properties and the (~2n) reations in beryllium. For zironium lad targets, beryllium was also determined to be the best plug material. Polyethylene plugs and a water flooding were alternative moderators of hoie for the entral region in a zironium lad target. 33

35 Additional reator otilgurations were also analyzed using beryllium plugs in th for stainless steel and zironium targets. The mass loadings in the targets for these and 30 grams of 235U. For these target loadings, reator ores onsisting of61 were investigated. Target ores onsisting of61 stainless steel targets with 20 an 235Uwere below a lgffof In ontrast, reator ores with 61 zironium target mass loadings were above a kff of All onfigurations evaluated with 85 taq 30 grams of 235Uwere above a bff of For the same number of targets loaded of 235U,only the zironium targets were above a lgffof The stainless stef subritial and near a ~ff of Unfortunately, there are several disadvantages with using a moderator in the region. Flooding the internal region of a target with water ould ause proble] extration and the Cintihem proess. In addition, a flooded target would need to to prevent a rupture from ourring. Beryllium is an exellent moderator but it is diffiult to fabriate. Questions onerning the effets of radiation on beryllium 1 plugs reuse after an irradiation yle. Polyethylene is relatively inexpensive fabriate. However, polyethylene is prone to neutron degradation whih ould reuse. Critial onfigurations in a water moderated system are very limited and require m the nominal target design. A ritial system with the nominal target desi moderation annot be attained with less than 163 targets. Utilizing the entral target as an internal moderator for fission neutrons leaving the U02 layer p] suessful onept in offsetting the strong neutron absorption by the stainless stet also redued the required number of targets for ritiality by nearly half fro Consequently, a moderator material must be used in a target for a water moderate a viable option. Furthermore, requiring the onstrution of a ore larger than tl targets is unrealisti beause only a small fration of the targets ould be used for ore of a large size would require leaving some targets in the reator for an extenl time thereby reating a driver ore, whih essentially defeats the original purpose onept Beryllium & Beryllium Oxide Moderators The seond set of moderator materials investigated were beryllium and beryllium f onfigurations were investigated with beryllium; two with beryllium oxide. In onfigurations were evaluated to determine if the pro osed onept would w f enrihed targets. These targets were evaluated with a 35Uenrihment of 20 v Parametri analyses were petiormed for eah reator onfiguration to determim target pith, target oolant hannel size, and dimensions of the axial and radial n exterior of the reator was surrounded by a 10 m water refletor in the ra diretions to fbrther redue neutron leakage. Beause beryllium and beryllium moderators, it was neessmy to analyze the neutroni effets of a water target a and optimize its size onurrently with the target pith. 34

36 Table 3. MCNP Results for Beryllium & Beryllium Oxide Moderated Reators Target Total Cladding Target Moderator RadiaIIAdal IQfii(s Pith Number Mass Material Center IRefletor Refletor. (m) of (grams) Material Thikness(m) Tawets w w Be/ H Be / 10.0HZO Ss * Be/ HZO 30.0t 25.0Be & / 10.0HZO Ss m BeIH2.O 30.0t 25.0Be /10.0HIO Ss m Be/ H*O 30.0f 30.0Be * / 10.0HZO Zr w BeIH Be * /10.0HZO Zr w Be/ HZO 30.0/ 25.0Be /10.0HzO Ss Poly BeIHzO 30.0I 15.0Be i / 10.0H Zr Poly BeI H Be * / 10.0HZO Ss ZrH(L85)BeIHzO Be a / 10.0H Zr ZrH(l.85)BeIH /25.0Be * /10.0H Ss Be BeI H Be & / 10.0H Zr Be BeI H /25.0Be * / 10.0H Ss Dry BeOIH f 25.0BeO * / 10.0H Zr Dry BeOIH,O 30.0I 35.0Beo / 10.0HZO *Alltargetsused93*Aenrihedurauiumand hada 0.20 mgapforwaterooling. Table 3 presents the Lff and standard deviation results for beryllium and beryllium oxide moderated reators ooled with water. kff results in bold fae type denote ritial ores. The onept of using solid plugs in the entral region was employed in this study to quanti~ its effets on km in beryllium and beryllium oxide moderated systems. A base ase was defined as a target with a dry enter, and its IGffwas ompared to those for three moderating materials tested as solid plugs: berylliuw polyethylene, and zironium hydride (ZrH1.sS-64.9 a/~hydrogen). The first beryllium ofilgurations were omposed of 37 targets with dry entral regions at two 235Umass loadings of 23.5 and 28.2 grams. The optimum pith for stainless steel and zironium targets was determined to be 7.00 m with a oolant water gap size of 0.20 m. ~ff results for 37 stainless steel targets using both mass loadings were below In ontrast, km results obtained for reator ores omposed of 37 zironium targets were above As 37 stainless steel targets were subritial, subsequent studies determined 91 stainless steel targets were required for a ritial system at both 235Umass loadings. Consequently, a reator ore of stainless steel targets with dry entral regions moderated by beryllium was deemed impratial for fi.uther design onsideration. 35

37 Subsequent studies were done with berylliuw polyethylene, and zironium hydride plugs. The number of targets modeled here was 37 with a mass loading of 28.2 grams of 235Uper target. Based on studies to determine the optimum pith for targets with plugs, pithes for target enters with solid plugs were found to be 1.00 to 2.00 m smaller than the 7.00 m pith for the dry enter targets. However, the optimum water oolant hannel thikness remained the same at 0.20 m. Figure 10 illustrates kff results for the optirrized reators using stainless steel targets, moderated and refleted with beryllium. Figure 11 shows the kff results for the reators using zironium lad targets. 1.10r-r- r r m --!----l r- r = = t 1: x $ y ~... Q...>Q. e 0: G v; J = = t * % * -l 1....,! -r 4 Dtyerlter O Berylliumplug,. v Polyethyleneplug v WsdlKl Axial Relktor lhikness (m) Figure 10. &ff Results for Beryllium Moderated Reators Using 37 Stainless Steel Clad Targets and Various Target Center Materials. 36

38 , r--.--v-t--t--~ T ~--r---.--~---.--T-r- [ 1 1: 9 F?@ *, Q + Q: R: = m.....! o ~~~ o i ~ i ~ ,.:,,,.,, L I I 1 I Axial Refletor Thikness (m) Figure 11. lqffresults for Beryllium Moderated Reators Using 37 Zironium Clad Targets and Various Target Center Materials. Polyethylene plugs yielded slightly higher km values than beryllium plugs in stainless steel targets. Unfortunately, bff values using both of these types of plugs were barely above ZrH1.8splugs yielded a subritial system with a kff of All of the ases for zironium targets were above a km of 1.10, with beryllium plugs being the moderator of hoie followed by polyethylene and ZrH1.gSplugs. Two onfigurations were also designed with beryllium oxide as the moderator. These reators used 37 targets (28.2 grams of 235Uper target) tith stainless steel and zironium laddings. Both reators were designed to use dry targets without solid plugs in the target entral region. In omparison to the same ases using beryllium as the moderator, the IGm for these ases were slightly lower due to absorption by the oxygen. Further stales with beryllium oxide were suspended as these results on.fkmed its neutroni properties are similar to those of beryllium. 37

39 The results presented thus far show that highly enrihed targets work very well in obtaining a ritial system. However, proliferation onerns have been expressed in reent years with the use of highly enrihed uranium to produe gmo. Capabilities do not urrently exist to produe 99M0with low enrihed uranium on a ommerial sale and are still in the development phase. For the purpose of this researh, two otilgurations were designed to evaluate the feasibility of this reator onept. Both reators modeled for the low enrihed studies ontained 37 targets and a mass loading of 28.2 grams of 235Uper target. Zironium was hosen as the ladding material due to its low neutron absorption harateristis and the large number of ritial ases produed with highly enrihed targets. Beryllium was hosen as the moderator and plug material for its ability to produe (~2n) reations and neutron sattering. Table 4 presents the results for the low enrihed target reators. Lff results for both reators are above The addition of a plug in the entral region of the target also provides a signifiant inrease in lgff. These results demonstrate that low enrihed targets an attain a ritial onfiguration in beryllium moderated systems with a 37 target ore. However, fiture work would need to be done to assess this apability with other moderators. A number of ritial onfigurations were obtained in beryllium moderated systems. In omparison to water moderated reators, there are signifiant possibilities with beryllium for onstruting a system with only 37 targets. All of the ases investigated with zironium targets were ritial regardless of whether the entral region of the target was dry or ontained a solid plug. Reator ores with 37 stainless steel targets are also possible; however, they are limited to using a solid plug. The 91 dry stainless steel targets required for ritiality reates a muh larger ore needed to satisfi fill sale prodution. Ta ble 4. MCNP Results for Low Enrihed Beryllium Moderated Reators Target Total Cladding Target Moderator Radial i Axial ~*7 Fith Number Mass Material Center I Refletor Refieetor (m) of (grams) Material Thikness(m) ITargets*I - I I 7.00 I I Zr JM Be/ HaO I 30.0/25.0Be /10.0HZO Zr Be Be/ HZO Be / 10.0H20 *All targetsused20 /0enriheduraniumand had a 0.20 m gap for waterooling. 38

40 Themajor advantage of beqllium asamoderator isits(~2n) reatiion harateristi. Sattering along with the extra neutrons provided by this reation give this system a high neutron eonomy. An outer refletor of water fi.u-theraids in reduing absorption and leakage. Polyethylene plugs were the most effetive moderator material for use in stainless steel targets beause they overome parasiti absorption by the iron better than beryllium plugs. Low absorption harateristis of zironium allow extra neutrons to indue fi.uther fission events whereas the extra neutrons are lost to leakage and parasiti absorption with polyethylene plugs. Zironium hydride plugs are generally less effetive as plug materials for both types of target laddings. Neutroni harateristis between beryllium and beryllium oxide are similar enough that beryllium oxide ould be used if eonomi and material limitations prove the use of beryllium to be unfeasible. Material drawbaks with beryllium inlude its normally brittle nature and the prodution of helium gas whih fi.u-therenhanes failure of the beryllium matrix with time and neutron irradiation. In additio~ prodution of lithium-6 from the deay of helium-6 provides a signifiant amount of parasiti absorption whih ould affet long-term system ritiality. Although beryllium oxide is easier to fabriate than beryllium, it vaporizes rapidly in the presene of moisture due to the formation of beryllium hydroxide. This ould still pose a problem with using beryllium or its ompounds as both are toxi when inhaled. Use of low enrihed targets have been suessfully illustrated here for beryllium moderated systems. However, their appliability for use in other moderators with thk reator onept is reserved for Mure researh Graphite Moderator The next moderator material tested was graphite. A total often onfigurations were modeled with graphite using stainless steel and zironium targets. The target pit~ radial and axial refletor dimensions, and oolant hannel size were optimized in a series of parametri studies. To redue the neutron leakage outside the graphite refletors, an additional water refletor 10 m in thikness surrounds the exterior ofthe modeled reators. Table 5 presents the bff and standard deviation results obtained fi-om MCNP for the graphite moderated reators ooled with water. kff results for ritial onfigurations appear in bold fae type. Four types of solid plug materials for the target entral region were modeled and ompared to the base ase of a dry enter for stainless steel and zironium targets. The target pith obtained fi-om the optimization studies for all modeled systems was 4.00 m. Depending on the plug material in the target enter, the target oolant hannel size varied from 0.20 m to 0.60 m. All ases evaluated for graphite onsisted of ores with 37 targets and a mass loading of28.2 grafns of 235Uper target. 39

41 T TargetI Total Fit;h Number (m) of MCNP Result! Cladding Target Mass Material Center &rams) ble 5. for Graphite Moderated Reators Water ModeratorI Radial /Axial ~* Gap Refletor Refletor I (m) Material Thikness(m) T T m6-h Ss m 28.2 Zr w 28.2 Ss Be 28.2 Zr Be 28.2 Ss Graphite 28.2 Zr Oraphite 28.2 Ss Poly 28.2 Zr Poly 28.2 Ss zrh(l.85 I 28.2 Zr ZrH(l.85; I 0.60 HIO 10.0/ 10.0HaO 0.60 GraphiteI 90.0I 30.0graphite H / 10.0H Graphitef 60.0f 30.0g@lite O.&3895 * HZO 10.0/ 10.0H Graphite/ 60.0[30.0graphite S* H / 10.0H GraphiteI graphite * HIO 10.0/ 10.0HZO 0.60 GraphiteI graphite HZO 10.0/ 10.0HZO 0.20 GraphiteI graphite H /10.0HZO 0.20 GraphiteI graphite * HZO 10.0/ 10.0HZO 0.20 GraphiteI graphite H / 10.0HZO 0.20 GraDhitef mvhite H*O 10.0/ 10.6H20 *Alltargetsused 93 /0enriheduranium. Based on the analysis of the graphhe moderated reators, a limited number ritial onfigurations with 37 targets were disovered. From the ten graphite reators evaluated, only three yielded a ritial system. Moreover, these ritial systems were limited to zironium targets and required the use of a solid plug. The four solid plug materials tested were: graphite, beryllium, polyethylene, and zironium hydride (64.9 /0 hydrogen). Figure 12 illustrates the keff results shown in Table 5 with the stainless steel targets. Figure 13 shows the b results for zironium targets. The use of solid plugs did not aid in attaining a ritial system with stainless steel targets. Eah of the reators omposed of stainless steel targets had a leffof less than The base ase with a dry enter is very subritial with a km whih never exeeds When graphite plugs are used, kff rises slightly but is still very subritial with kff never exeeding Zironium hydride and beryllium plugs are fairly omparable but never exeed a keff of The best plug material tested is polyethylene; however, the highest lgffwith this type of neutron satterer never reahes

42 t I I I t I I I I I I I I I I I I I I I t t I t I I I 1 I "`~--""--i "-""--{ o.85} -E { " = AC? v: fj&j 0.75 "----g-"-"--?--"-..*-...".""------""" I I! 1 I I I 1 I I i 1 1 t I! I I! I I I I I I I! o.(x) ma) Lmo hid Wfkta lkkmss (Cni) ~ 1.-" " ---i---" ----""----:-" -"---i---"-&--"- dr H; :@g*~* Ej ; ; : [: n : -----: : : :: I l::!! Y +Y?~?~* V* I~ v.m ~ ~= H I k*+=* "---".."-----">"""---*--" ""-.:----- J m - I+20gq=o.m-n 0?@q Hpg+=o.mm -F@ Hpgq=o.Klan Rd@T#3ne@.g I-pgq)=onnl %.&5@~ H#gq=ozmnl Figure 12. kff Results for Graphite Moderated Reators Using 37 S@inless Steel C1ad Targets and Various Target Center Materials. 41

43 ~ x 1.10 t i l.~ I +Q; +g~ l...q...q....q..+., n, 8 Gl~i,:~ I.n --~-- m----- *?:~?v*? *? m: * v--: ---: :... I ---; :1 Ogo... + *------'-----~------?.-----g-- J.---!-----f () :4...B...tia-. ti... *m@n-.! ; ; : ; m - I-y3gq)=o.fma?l Ekph.g l-pg?p=lmknl Qz#ite@qj!+ogiq=amnl H@ap=man W.as@w I-pgq)=an??l 0.83 I I I I I O.m ma) 80.al mm Axial Fkfktormkkness(ni) Figure 13. kfi Results for Graphite Moderated Reators Using 37 Zironium Clad Targets and Various Target Center Materials. 42

44 There are a ouple of reasons for the subritial ~ff values observed with stainless steel targets in graphhe. First is the parasiti absorption in the steel ladding, and seond is the high leakage rate inurred as a result of neutron sattering by graphite and the plug materials. Target pithes over 4.OOm were found to be in the over moderated regime in the pith optimization study and would have yielded lower kff values than those shown in Figure 12. Figure 13 for the zironium targets shows a dry target and the use of graphke plugs both yield subritial systems and are onsidered poor design hoies. As a result, polyethylene plugs are the moderator of hoie for the target enter plugs followed by beryllium and zironium hydride plugs. For both types of ladding materials, the optimized oolant hannel thiknesses did not hange; however, they were dependent upon the plug material. The optimum oolant hannel size for dry targets and those with graphite plugs was 0.60 m. These oolant hannel sizes were larger than the others obtained in the study. A redution to 0.40 m of water with targets ontaining beryllium plugs was due to better sattering properties and the (~2n) reations. Polyethylene and zironium hydride plugs required only 0.20 m of water due to their rather high sattering properties and low absorption traits. Graphite is generally a good moderator and refletor material. However, its ability to provide adequate neutron moderation and refletion with this reator onept is limited. It fared poorly with stainless steel targets due to the high parasiti absorption of neutrons in the iron of stainless steel. In additio~ the use of solid plug materials in the target enter did provide a signifiant inrease in kff. Consequently, there are two requirements for a target reator to be onstruted with a graphite moderator and refletor. First, the targets must be lad with zironium metal. Seond, a solid plug made of polyethylene, berylliuq or zironium hydride must be used. The good sattering properties of these materials ombined with the low absorption traits of zironium provide a viable ombination Heavy Water Moderator The last moderator material investigated was heavy water. Four reators with 37 targets in eah ore were onstruted with stainless steel and zironium targets. These reators were modeled with two different oolant onfigurations: the first pair of reators uses heavy water as the moderator and oolant; and. the seond pair uses water as the oolant and heavy water as the moderator. There are two major advantages with using water as the oolant. It prevents ostly and frequent replaement of heavy water boiled away due to heating. Also heavy water requires extreme pressurization if it is to be used as an effetive oolant. Table 6 presents the results for the modeled heavy water moderated systems. 43

45 Table 6. MCNP Results for Heavy Water Moderated Reators Target Total Cladding Target Water Moderator RadialIAxial k#* Pith Number Mass Material Center Gap I Refletor Refletor (m) of (grams) (m) Material Thikness(m) TargetsR Ss w - D20IHZO DZO / 10.0HZO Zr w - DZO/ HaO DZO * / 10.0H X3 Dry 0.20 D,OI HZO DZO / 10.0HZO Zr Dly 0.20 D,OI H,O 150.0I 150.0~o * / 10.0H20 *All targetsused 93 /0enriheduranium. The target pith and dimensions of the radial and axial refletors were determined using the same inremental parametri studies as the other moderator materials. The water oolant gap size was optimized simultaneously with the target pith. An infhite refletor of water 10 m surrounded the exterior of the reator to redue neutron leakage. A major differene from the other moderator studies is that both pairs of reators analyzed here do not use solid plugs in the entral region of the target. The deision of whether or not to use solid plugs was determined in the target pith optimization. Figure 14 illustrates the results for a reator omposed of 37 ifilnitely high targets, moderated and ooled with heaq water, and surrounded by an itiinite heavy water refletor. The L results for stainless steel and zironium lad targets with a dry enter are ompared to those using beryllium or graphite plugs. 44

46 El +.S.Slad, dtytarget -O- Zrdad, drytaget + SSdad, graphiteplug * Zrlad,graphiieplug + SSlad,@eplug -D- Zrlad,& plug! 1 I I I i 1 I m 4.m m 12.l Target Pith (m) Figure 14. k Results for 37 Targets in a Heavy Water Moderated Reator with Various Target Pithes. Figure 14 illustrates that the use of solid plugs for both ladding materials does not indue a signifiant rise in h. The data shows there is virtually no differene in b between graphite and beryllium plugs. This may be attributable to neutron leakage whih is aused by the sattering properties of the plug materials and the heavy water. Consequently, solid plugs were not needed in the target entral region with heavy water asthe moderator. Addition of a water oolant hannel provides the advantage of preserving heavy water and minimizing expensive replaement. However, the addition of a water hannel does indue a loss in reativity when ompared to a reator where heavy water serves in the dual roles of moderator and oolant. For example, the loss in reativity inurred for the reators in Table 6 due to addition of a water oolant hannel ranges from -$5.76 ~ with stainless steel targets to -$5.15 t for zironium targets. Figure 15 is a plot showing the differenes in hff results for a heavy water moderated system ooled with water or heavy water. 45

47 l,,:,,j~,,,!,,j!!,,:!!!! r--!~ ; ; : -.--; ; ;--. --: : /[$ =; J I.ol ----:... : , O.%1 ----: : ; { ; ~ : ! : SS dad r-mlo gap -0- SS dad 0.20ml I-$o@p t Q Zrdad noi-$o gap Z dad 0.20ml I-$ogap 0.70 I I I f I I I I I Ex3.olCO.O Ko.o 1s0.0 ml,o Axial RefletorThikness(m) Figure 15. kff Results for37targets ina HeaWWater Moderated Reator Cooled with Heavy Water or Ordinary Water. All of the reators evaluated here yielded ritial onfigurations with a &.ffvalue above In omparison to water, all of the ores tested were able to attain ritiality with 37 targets and without the use of solid plugs. This is primarily due to the sattering properties and low absorption ross setion of heavy water. However, there are a ouple of major disadvantages whih limit seleting heavy water as the primary moderator for this reator onept. First, the reators analyzed have very large dimensions as ompared to the smaller reators moderated with other materials suh as water. The large target pith and immense quantities of heavy water required for the refletors ould make this type of reator diffiult and expensive to build. Furthermore, heavy water has a limited availability whih renders it less attrative than other materials. 46

48 2.5 Conlusions A number of trade-offs and important onsiderations are evident from the results of eah moderator investigated in this optimization study. To attain a ritial onfiguration in a water moderated system, targets would have to be manufatured with a low absorption material suh as zironium. In additio~ the target enter would require the use of a solid moderator material. The optimized reator dimensions are fairly reasonable but a large number of targets are required to attain a ritial system. Beryllium moderated systems yielded ritial otilgurations with either ladding material and with a minimum of 37 targets. Most of these ases also required the use of solid plugs; however, a 37 target ore of dry zironium targets proved fwsible. Likewise, ritial graphhe moderated systems were possible with 37 targets, but required the use of plugs and zironium ladding. System dimensions were fairly reasonable and omparable to water and beryllium systems. In ontrast to the other moderators, heavy water moderated systems attained ritial onllgurations with 37 targets and without the use of plugs. However, the dimensions of the systems were extremely large and impratial from a design and eonomi onsideration. This study also indiated that low enrihed targets an attain a ritial mass in a beryllium ore. To adequately assess their overall feasibility for use in fill sale 99M0 produtio~ fi.uther work would be required. 47

49 3. TEMPERATURE& DENSITY EFFECTS ON keff This hapter summarizes a series of parametri studies to assess the effets of temperature and density hanges on kff. The optimized reators desribed in Chapter 2 were analyzed for this study. A minimum of two onfigurations were hosen for eah moderator in this investigation. The major parameters analyzed in this study were: moderator temperature;. moderator density; and. fiel temperature. Moderator temperature and density effets were also evaluated for ases where a solid plug was used in the target entral region. To obtain better statistial results, the number of partile histories was inreased iiom the 100,000 used in the optimization study to 400,000. lg.ffresults for thk study are summarized in terms of the reativity losses inurred ilom a rise in temperature and a redution in density. Inremental studies were done for eah moderator to assess these effets on bff separately; however, the results presented here are for the ombined effets and are onsidered onservative estimates for expeted reativity losses. 3.1 Hot and Cold Conditions To adequately assess temperature and density effets, a set of temperatures and orresponding densities were defined for eah material in the reators tested at old and hot onditions. Temperature hanges were aounted for by using the S(@) ross setions in MCNP for the appropriate temperature and material. Table 7 lists the temperatures and densities used for various reator materials at old and hot onditions. Temperatures for the materials in Table 7 were not hosen arbitrarily; rather, they were seleted based on the available S(t,~) libraries in MCNP. For example, graphite and beryllium were evaluated at 600 K beause an S(@) library does not exist for these materials at 400 K. Cases where water and heavy water were used as the oolant, hot onditions were defined with the same temperature and density as the moderator. Hot onditions for the seondary water refletor were defined at 300 K based on the assumption that any heat transfemed from the targets to the moderator did not reah the outer refletor. Density hanges in the beryllium moderator, beryllium and polyethylene plugs, and U02 fiel layer were not investigated assuming little hange with temperature. 48

50 Table 7. Parameters for Various Reator Materials at Cold & Hot Conditions. Cold Cold Hot Condition Hot Condition Condition Temperatures Condition Material Temperatums Densities m Densities (K) (glm~ (@l@ Watermoderator& oolant Exteriorwater refletor Graphitemoderator Berylliummoderator &plugs Heavywater moderator&oolant Polyethyleneplugs U02targetlayer Water Moderator Based on the reator onfigurations investigated in the optimization study, only two reators were deemed redible andidates for thk analysis. These ores xmsisted of 85 targets, 30 grams of 23*Uper target, ooled with water, and ontaining beryllium plugs in the target entral region. Table 8 summarizes the IGffresults at old and hot onditions, and the resulting reativity losses. The results in Table 8 show both reators with stainless steel and zironium targets did not inur a large enough loss in reativity to beome subritial at hot onditions. A smaller reativity loss in zironium targets is mainly attributable to the very low absorption ross setion in zironium. This allows for fission neutrons to be thermalized in the moderator and return to the fiel layer to indue fbrther fission events; hene, a higher km and lower reativity loss. Table 8. Results of Reativity Losses for Water Moderated Systems. Target Total Target Target Moderator Radial Itilal Cold ~+ Hot~&~ ReativityLoss Pith Number Clad Center I Refletor Refletor Ap($) (m) of Material Thikness(m) (J3#koo70) Targets* Ss Be H20/ HZO 10.0/ * * -2.76* Zr Be HZO/ HZO 10.0/ * *Alltargetsused93 /0enriheduraniumwitha targetmassof30.0grams %. 49

51 3.3 Beryllium Moderator Table 9 lists the kff results and resulting reativity losses due to temperature and density hanges for five reator systems hosen from the optimization analysis. All of these reators were omposed of 37 targets, 28.2 grams of 235Uper target, moderated with beryllium, ooled with water, and in some ases ontaining solid plugs in the target enter. Beryllium oxide moderated systems were not evaluated here due their similar neutroni properties with beryllium. It is assumed that omparable reativity losses would be experiened with beryllium oxide. Table 9 shows that zironium targets with a dry enter inurred the smallest reativity loss of all onfigurations tested. Higher reativity losses for zironium targets with solid plugs may be the result of higher leakage rates due to the neutron sattering inherent with these plug materials. Table 9. Results of Reativity Losses for Beryllium Moderated Systems. Target Total Target Target ModeratorI Radkl Iti;al Cold~+ Hot~+C Reativity Pith Number Clad Center Refletor Refletor LossAp($) (m) of Material Targets Zr m Be/ HJO 30.0I25.0Be O* = /10.0H Ss Be Be/ HZO 25.0/20.0Be * -1.29* / 10.0H Zr Be Be/ HZO 25.0/25.0Be s / 10.0H~O Ss Poly Be/ H / 15.0Be ~ * / 10.0HaO Zr Poly Be/ HZO 30.0/15.0Be * * / 10.0H *Alltargets used 93 Aenriheduraniumwith a target mass of 28.2 grams 5U, and a 0.20 m gap for water ooling. 50

52 3.4 Graphite Moderator Table 10 lists the IGffresults for two graphite moderated reators evaluated at old and hot onditions, and the resulting reativity losses. The two onfigurations analyzed were omposed of 37 zironium targets, 28.2 grams of 235Uper targe$ moderated by graphite, ooled with water, and ontaining either polyethylene or beryllium plugs in the inner entral region of the targets. Table 10. Results Target Total Target Pith Number Center (m) of 4.00 Targets 37 Be Poly of Reativity Losses for Graphite Moderated Systems. HZOModerator/ RadialIhial Cold ~+ Hot~&C Reativity Gap Refletor Refletor IXISSAp($) (m) Material 0.40 graphitei 60.0/ * -4.13* H,O &?KiDhitC / lo.{lio.o HZOI 0.20I mdhite/ I H20 graphite / 10.0H20 *Alltargetsused93 %0 enriheduraniumwith a targetmass of 28.2 grams % and zironiumlad Reativity losses with these graphite moderated ores are muh higher than the omparable beryllium moderated ores. This may be due to higher leakage rates with graphhe as the moderator when ompared to beryllium and water. Comparing the two plug material ases shows a dramati differene. A higher reativity loss may be inurred with beryllium plugs beause the extra neutrons produed in (~2n) reations are lost to leakage and do not ontribute to firther fission events. On the other hand, sattering by polyethylene allows neutrons to be moderated in the water oolant hannel and ause subsequent fission events. 3.5 Heavy Water Moderator Two types of heavy water systems were investigated for temperature and density effets: one ooled with heavy water (no water gap), and the other ooled with ordinary water (0.20 m water gap). The reator ores listed in Table 11 onsist of 37 targets with dry target enters eah having a 23% mass loading of 28.2 grams per target. Table Il. Results of Reativity Losses for Heavy Water Moderated systems.. Target Total Target HZO ModeratorI Radkl! &ial Cold ~+ Hot~&~ Reativity Pith Number Clad Gap Refletor Refletor Loss Ap($) (m) of (m) Material Targets Ss - DZOIHZO DZO & / 10.0HZO Zr - DZOIHZO QO * / 10.0HZO Ss 0.20 DZO/ HaO D / 10.0HZO Zr 0.20 DZOIHZO D * i 10.0/10.0H *Alltargetsused 93 /0enriheduraniumwith a targetmass of28.2 grams 5U aud a dry entralregion. 51

53 Of all the examined moderator materials, these heavy water moderated ores inurred the greatest amount of reativity loss due to temperature and density effets. Heavy water moderated and ooled systems had reativity losses of about $1.00 more than heavy water moderated, water ooled reators. Systems ooled and moderated by heavy water have higher leakage rates due to sattering. The water oolant hannel outside the target allows for neutrons to be thermalized thereby having a greater probability of induing fi.u-therfission events instead of being absorbed or lost to leakage as is the ase heavy water oolant. 3.6 Conlusions Several important onlusions an be drawn fi-om this analysis. Comparison of the reativity losses among the four moderators investigated shows that heavy water moderated systems inurred the largest loss of reativit y due to hanges in temperature and density. The moderator with seond largest reativity loss was water, followed by graphite, and beryllium with the smallest loss. The large reativity loss with heavy water may be attributable to its sattering properties whih indued a large amount of neutron leakage, thus a sharp redution in km. However, these losses in reativity were not solely limited to ontributions from the moderators. Parasiti absorption in stainless steel ladding also proved to be another fator. In ontrast to the strong absorption properties of the stainless steel, zironium lad targets had smaller reativity losses due to the low absorption properties of zironium. Overall, a majority of the onfigurations (ten out of the thirteen) analyzed remained ritial to be onsidered viable reator designs despite hanges in temperature and density. 52

54 4. FISSION PRODUCTS& NEUTRON POISONS Chapter 4 summarizes the results of several parametri studies ompleted to determine the effets of fission produts and other neutron poisons. Major fission produts and poisons of onern in this analysis are: xenon-135 (135Xe), samarium-149 (149Sm),and lithium-6 (~i). 135Xeand 149Smare produts of the 235Ufission proess and are formed in the UOZoating of the target. 6Li is formed in beryllium via beta emission from He.!He is the result of (n,a) reations in the beryllium matrix. A brief summary of the equations used for determining the onentrations of Li and 3H are presented in Appendix B.3. A similar summary of the equations used for 135Xeand 149Smare in Appendix B.4. A selet number of reator ombinations were hosen for this study fi-om those analyzed for temperature and density effets. The three reators hosen from the list of andidates analyzed in Chapter 3 onsist of 37 targets and are onsidered to be representative ases fi-om that list based on their physial attributes suh as ladding material, moderator material, and plug material type. A seond riterion in the seletion proess was to hoose onfigurations with a reasonable ~ff margin whih would be able to overome the reativity loss due to poisoning by 135Xe,149SWand Li. Water moderated ores were not hosen for two reasons: beause of their low bff values; and beause their optimum ore size of 85 targets exeeds the baseline 37 target ore whih is deemed an ei%ient size for 99M0prodution. Heavy water moderated reators were also eliminated from firther onsideration due to their extreme size even though their hff values were high as shown in the temperature and density effets analysis. 4.1 Bakground Information 135Xehas a very large thermal absorption ross setion of about 2.6 x 10s barns, and its onentration is dependent on the neutron flux level in the reatoq hene, the reativity effets of 135Xeare also time dependent. Figure 16 shows the deay hain and formation of 135Xehorn fission to its stable daughter 135Ba. Figure 16. Deay Chain and Nulear Formation of 135Xe. During reator operatioq 135Xebuilds up to an equilibrium onentration whih is determined by the deay rates of 135Xeand These deay rates ompete during reator shutdown due to the lak of 135Xeabsorption. The large onentration of 1351present in the ore deays and produes a buildup of 35Xe. The deay of 1351eventually equals that of 135Xeausing a peak in the 135Xe onentration. Subsequently, the deay rate of 1351falls below that of 135Xeproduing a gradual redution in the 135Xeonentration. If the reator is restarted before the previous equilibrium 135Xelevel is attained, neutron losses are aelerated by the renewed absorption of 135Xeuntil the equilibrium level is reestablished. The primary aspets of 135Xeformation onsidered here were the magnitude of its peak onentratio~ how soon this ours afler reator shutdo~ and the appropriate time to restart the reator. 53

55 149Smhas a thermal absorption ross setion of about 40,000 barns, and is onsidered to be stable. Thus, it does not o away over time and has a onstant poisoning effet. Figure 17? illustrates the formation of 49Smfrom fission. Figure 17. Deay Chain and Nulear Formation of 149Sm. In ontrast to 135Xe,the equilibrium onentration of 149Smdoes not depend on the reator flux level and must be burned out of the reator via absorption. During reator operatio~ the onentration of 149Smbuilds up to an equilibrium level. This onentration builds to a higher level during shutdown due to the deay of 149Pmand the absene of a neutron flux to provide burnout. After rest~ the 149Sm onentration dereases and eventually reestablishes the equilibrium onentration before shutdown. For the analyzed operating time of 7 days, the 149Sm fration of saturation and its orresponding mass in the ore were parameters of onsideration. The reators hosen also use either a beryllium moderator or solid beryllium plugs in the target entral region. Consequently, onsideration of ~i, helium gas, and tritium (3H) formation in the beryllium matrix is important. Figure 18 illustrates the (~2n) reation in beryllium and the formation of two alpha partiles or helium nulei from Be. Figure 19 shows the nulear mehanisms required for the formation of Li,!He, helium gas, and 3H from beryllium. The ratio of (n,2n) reations to (qt) reations is approximately a Figure 18. 9Be (n,2n) %e Reation. Figure 19. Nulear Formation of 6Li, He, and 3H from Beryllium. 54

56 Two alpha partiles or helium nulei are formed in addition to those in the (~2n) reation. One helium nuleus is formed with 6He in the beryllium matrix and the other is formed simultaneously with 3H. As a result of the reations shown in Figures 18 and 19, a large amount of helium gas is formed in the beryllium matrix. Formation of helium gas in the beryllium matrix is important beause it aelerates the strutural degradation of beryllium whih is naturally a brittle material. 6Li has a fairly large thermal absorption ross setion (G=945 barns) and is stable. Its formation is the diret result of beta emission from 6He.!Li formation is important when onsidering reativity loss due to its large absorption ross setion. 3H prodution was analyzed to determine its magnitude and how this affets long term waste management problems. 4.2 Reativity Losses by 35Xe, 4gSm,and Li Prodution Table 12 presents the results for three onfigurations analyzed for reativity loss due to the buildup of 13sXe,149Smin the U02 oating inside eah target, and 6Li depending on whether a beryllium moderator or solid plug was used. The reativity losses for 135Xeand 149Smpertain to the amount of reativity lost over the ourse of a 7-day operation period. Results in the last olumn of Table 12 are indiative of the expeted reativity loss aused by the formation of %i over a 10-year time span. Ten years was hosen as the time sale for the analysis of 6Li sine it is expeted that the beryllium moderator and plugs ould be used for this amount of time without replaement. However, replaement time depends on the extent of helium gas formation. Exatly how helium gas prodution affets the strutural integrity of the beryllium matrix is beyond the sope of this analysis and is not onsidered here. Comparing the 37 stainless steel and zironium lad target ores both ontaining polyethylene plugs shows a statistially insignifiant inrease in the reativity loss due to 135Xe. A signifiant inrease in reativity loss with a graphite moderated system using beryllium plugs and zironium targets is due to fission neutrons and some of the extra neutrons from (q2n) reations in beryllium being absorbed by 135Xein the U02. Losses due to 149Smare statistially omparable for the three reators and muh lower than those for 135Xe. Reativity loss from the formation of 6Li in the beryllium moderator is about 1.85 times higher with stainless steel than with zironium targets. This is due to ombined parasiti absorption effets I?om iron and 6Li. Losses are rather high when a beryllium plug is used as some of the neutrons interating with the beryllium plug are parasitially absorbed by 6Li instead of being used in fkther fission events. Consequently, this analysis shows that a ore omposed of zironium lad targets without beryllium plugs has the lowest reativity loss due to these three nulides. 55

57 Table 12. Results of Reativity Losses Due to 135Xq 149Sm, and Li. Cladding Target Moderator Equilibrium Post-shutdown I gsm Material Center I Refletor l%e lsxe Reativity Reativity Material Reativity Reativity Loss Ap($) LossAp($) LossAp($) LossAp($) O070) I Ss I Poly I Be/ HjO a I Zr I Poly I Be/ HIO -3.48* I I Zr Be GraphiteI 44.80h HIO A seond aspet of 135Xeformation was to determine when the peak 135Xeonentration ours after reator shutdown and how soon after this peak an the reator be restarted. The results in Table 12 show for the two beryllium moderated ores, reativity loss due to the buildup of 35Xe after shutdown amounts to an additional $1.70 above what is lost due to equilibrium 135Xe. For the graphite moderated ore with beryllium plugs, an additional $2.26 of reativity is lost due to 135Xebuildup after shutdown. The differene in these two reativity losses shows an additional $0.60 of reativity is lost with a graphite moderated ore. The following three plots illustrate profiles of the number of 135Xe atoms present after reator shutdown for a 7-day target irradiation time. Data for these plots was generated using a MathCAD program to alulate the post reator shutdown 135Xe onentration (Parma, 1998b). Eah urve represents a profile of the number of 135Xe atoms present over a deay time of 40 hours for a given target irradiation time in days. F@re 20 is the post reator shutdown 135Xefor 37 stainless steel targets with ~;lyethylene plugs in a beryllium moderated ore. Figure 21 shows the post reator shutdown Xe for the same reator onsisting of zironium lad targets. Figure 22 pertains to a graphite moderated ore with zironium lad targets ontaining beryllium plugs. 56

58 4.Oe+l 9 I f I I I I e+19 1/ 3.0e+19. / 2.5e+19,,,,/ e+19.. ~m~= 0.5 day : _.. ~m,=l day : -. ~ ~m~=l.5 days....;/ : ;\. \. ~m~= 2 days :. / \ ~ - ~m, = 2.5 days.;.:......" ' "..."... / \m~= 3 days : <:,.. If %....\ x-.* : e e e+18 L : ; =4 O.Oe+O t t i I lime After Reator Shutdown (hr) Figure 20. Post Reator Shutdown 135Xefor 37 SS Targets with Polyethylene Plugs in a Beryllium Moderator. 57

59 4.Oe+l 9, 1 1! I 3.5e+l 9 3.Oe+l 9 2.5e+l 9 2.0e e+l 9 1.0e+19 t I...XC==I%., \ \;-..:,, /. I,, %-.. /,./ v, / \,..!,...,......:....\ -:; :..; ~md= 0.5 day -- ~m,=l day m. =l.5days Jmd= 2 days ~m~= 2.5 days }m, = 3 days [ 5.0e O.Oe+O t t I I Time After Reator Shutdown (hr) Figure 21. Post Reator Shutdown 13%e for 37 Zr Targets with Polyethylene Plugs in a Beryllium Moderator. 58

60 4.0e e e e+19 2.Oe+l 9 1.5e e+19 5.Oe+l ; [ - -, kid=.5 day - had= day had=.5 days \md= 2 days (y :. / -;<:~\...:,, :7,,.., x::. L-..,/- -i~. - <:;..,.,, : \ \.. ; +., ;, w.,<,,:..,.,... \...-. K - $md= 2.5 days ~md= 3 days \...,...,<.,..,...:...,,,..,... \.-,-,%;>...%..:.-.,..,.,, :-\--\~. =-..,. Y ,,...:,.. O.Oe+O Time After Reator Shutdown (hr) Figure 22. Post Reator Shutdown 135Xefor 37 Zr Targets with Beryllium Plugs in a Graphite Moderator. 59

61 Figures 20, 21, and 22 show several important features. First, the peak 135Xeonentration ours around 7.5 hours after shutdown for eah of the reators investigated. This is less than the typial peak in a light water reator whih ours around 9 hours after shutdown. Seond, the number of 135Xeatoms present in the reator after shutdown reahes saturation levels when the targets have been irradiated for about 2.5 days. Third, these plots also show that the equilibrium 135Xe onentration is reestablished about 20 hours after shutdown. The only notieable differene between the three reators is that the equilibrium 135Xeonentration is slightly higher for a graphite moderated ore ompared to a beryllium moderated system. Based on the kff values presented in Tables 9 and 10 for these three reators and their orresponding reativity losses due to 135Xein Table 12, only a ore with zironium targets and polyethylene plugs in a beryllium moderator has a suffiient bff margin to withstand a large reativity loss. Table 13 presents the results for the amount of 149Smformed in the ore at saturation (t=m) and at the proposed operation time of 7 days. The fration of saturation at 7 days is approximately 29%, with fill saturation ourring at roughly 90 days. Results show that very low amounts of 149Sm,on the order of 0.05 grams, are formed at saturation. Even lesser amounts, on the order of grams, are formed at the proposed operation time of 7 days. This is mainly due to the uranium ontent in eah target. In omparison typial fiel elements omposed of disks or ellets ontain muh more uranium and ultimately yield more 149Sm. F These low amounts of 1 Sm are also onsistent with the low reativity losses shown in Table 12 for the three reators. Table SmMass and Fration of Saturation Results for 37 Ta Cladding Target Moderator I 49SmMass 49SmMass 49snl Material Center Refletor in Fuel in Fuel Fmtion of Material (gramsat (gramsat Satumtion t=+ t=7 days) (%) Ss I I Poly I Be/ HZO I I I Zr I I Zr I I Poly I Be/ HZO I Be I GraphiteI HZO I I I I I get Cores. 60

62 4.3 Li, Helium Gas, and 3H Prodution Table 14 presents the results for Li, helium gas, and 3H prodution in the beryllium matrix of the three reators analyzed. Formation of these nulides was analyzed for a 10-year time span assuming 10 years is the lifetime of the beryllium moderator or plugs. Overall, Li, 3~ and helium gas onentrations are slightly higher for a ore omposed of stainless steel targets moderated by beryllium than a ore with zironium lad targets. In ontrast, quantities of these nulides in the graphite moderated ore with beryllium plugs are lower than the beryllium moderated reators for the same time sale of 10 years. This is primarily beause there is muh less beryllium for neutrons to interat with to produe!li. In general, the ativity of 3H generated at the end of 10 years is high and ould pose a problem with disposition of the beryllium moderator. The number of helium gas atoms for eah reator in Table 14 represents all of the atoms produed through (~2n) and (~x) reations with beryllium in addition to helium atoms produed from (La) reations with!li to form 3H (See Appendix B.2). Further ompliating matters, the large quantity of helium gas produed aelerates the embrittlement and strutural degradation of the beryllium matrix. If severe embrittlement due to helium gas prodution and neutron irradiation ours, this ould pose a problem with disposition of beryllium due to its toxiity when in a powder form. Results of some preliminary alulations are presented in the last olumn of Table 14 whih illustrate the ratio of beryllium atoms relative to the number of helium gas atoms produed. These numbers show that there are numerous beryllium atoms surrounding eah helium atom in the moderator surrounding eah stainless steel or zironium target. Therefore, it ould be assumed that the He gas does not diffise out but is loked in the matrix of the beryllium moderator. In ontrast for a beryllium plug there are only 1,158 atoms surrounding eah helium gas atom produed. For this ase, it would be onservative to assume there might be some leakage of helium from the matrix of a beryllium plug. Nevertheless, fi.uther alulations would have to be done to veri~ these assumptions. Table 14. %i, 3H, and He Conentrations for 37 Target Cores C1addinElTa~et ]Moderator/l Beryllium]%i AtomsI % Ativity I He Gas NJ&e Materi; Cen~er Refletor Voiume (at t=10 (Ci at t=li Atoms (atom-i Material Material (m~ yrs) yrs) (at t=10 BeJatom yrs) He) Ss I Poly I Be/ HZO I 5.48x x10D 2.44X x10M 20,857 I Zr Poly Be/ HZO 4.79X x10Z 2.05x x102421,529 Zr Be GraphiteI 1.07X X1O* 9.23x X10241,158 HZO 61

63 4.4 Conlusions Several important findings were disovered onerning the effets of fission produts and neutron poisons in this analysis. Total reativity losses due to 135Xeat the peak onentration after shutdown varied iiom around $5.15 to nearly $7.00 for the three reators analyzed here. Consequently, overoming reativity losses due to 135Xeprodution ould play a big role in determining reator shutdown sheduling and ontrol shemes. Only the beryllium moderated ore with zironium targets had suffiient exess reativity to overome a peak in the 135Xe onentration after reator shutdown. Reativity effets due to 149Smprodution were small in the three reators, with the fration of saturation near 30% at the end of the proposed operation time of 7 days. Prodution of bli in the beryllium moderator or plugs was small ompared to the large helium gas volumes and 3H ativities produed over the same time period of 10 years. Further analysis would have to be done to determine if signifiant volumes of helium gas and large 3H ativities ould redue the utilization time and adversely affet waste disposition of the beryllium moderator or plugs. 62

64 5. FUEL BURNUP Chapter 5 presents and explains the results of a parametri analysis designed to quantify the reativity effets of 235Uburnup in the U02 oating of the target. The three reators analyzed previously in Chapter 4 are evaluated for 235Uburnup effets. The major objetive of thk study is to onfirm that fiel burnup is minimal and does not impose a large loss in reativity for the proposed operation time of 7 days. Positive reativity effets due to the prodution of 239Pufrom (n,y) reations with 23*Uin the UOZare not onsidered. A majority of the uranium in eah target onsists of 235Uand it is expeted that very little 239Puis produed. 5.1 Fuel Burnup Results Eah of the three reators onsist of 37 targets loaded with 28.2 grams of 235Uper target. To obtain reasonable statistis, it was assumed fiel burnup would be uniform aross the entire ore of targets. A reasonable rate for the amount of 23*Ufissioned in eah target was postulated to be 10% of the original 235Uloading in eah target. Therefore, a 10% burnup rate translates to 2.82 grams of 23SUper target or grams of 235Uin the entire ore being fissioned or burned. This rate of burnup is a onsenwdive estimate as the entire ore will be refheled with new targets every 7 days. Reativity loss estimates only aount for a redution in 235U per target due to burnup and do not inlude fission produt effets. Table 15 presents the projeted reativity losses per gram of 235Ufissioned for the three target reators. Results in Table 15 do not exeed 10 ents per gram of 235Uburned for any of the three reators evaluated. At the end of a 7 day operation time, only 5.72 grams of 235Uis burned in the entire ore. For example, this translates to a total reativity loss of $0.32 for the first reator in Table 15. Consequently, thk demonstrates the 10 %orate is a onservative predition for all reators in Table 15 as only 0.55% of the entire inventory of 3SUis fissioned at 7 days. Table Losses lue to 1% 5U Burnup for 37 Target Cores Total Wmber w of Mass Targets (grams) w Massat 100/0 Bumup (grams) Ciadding Material Target ModeratorI ReativityLoss Center Refletor with 10 /.% Material (Z::p% (J&#M1070) 37 I 28.2 I Ss Poly BeIH Zr Poly Be/ HjO I Zr Be GraphiteI & H20 63

65 The beryllium moderated ore with stainless steel targets has the largest amount of reativity loss followed by the same beryllium moderated reator with zironium targets. Irradiation of stainless steel targets inurs the largest amount of reativity loss due to the ombined ontributions of parasiti absorption in the steel and a loss in fissile material. Reativity loss in a graphite moderated ore with zironium targets is fairly omparable to its beryllium ounterpart. 5.2 Conlusions This analysis learly shows that a fiel burnup rate of 10?4odoes not present a signifiant loss in reativity for the three reators investigated. Reativity losses did not exeed 10 ents per gram of 235Ufissioned, with the highest estimate near 6 ents per gram of 235Uand the lowest near 3 ents per gram of 235U. As a result, a 10% burnup rate an be onsidered a onservative estimate of reativity loss for these reators based on the amount of 235Uloaded in eah target and the proposed irradiation time of 7 days. 64

66 6. ECONOMIC ANALYSIS Chapter 6 presents and analyzes the results of a two-phase eonomi study designed to provide approximate osts for the initial onstrution of the proposed invention; and waste generation expenses for the onventional, the Cintihem-type target fieled, and the solution reator onepts. Analysis of aelerator produed 99M0is not onsidered beause estimates of these expenses are not urrently available. Two major objetives of this analysis are: 1. to demonstrate that a target-fieled reator an be onstruted at a reasonable ost; and 2. to show signifiant savings an be inurred with a target-fbeled reator by the elimination of a driver ore of fiel elements. This investigation is not meant to provide an in-depth eonomi analysis; rather, the intention is to give estimates for the basis of making a reasonable and sound omparison of the prodution onepts. Many of the target-fieled omlgurations modeled and optimized with MCNP are neutronially viable; however, it was neessary to perl?ormthis analysis to quantifj the eonomi ramifiations of this new prodution onept. The first phase of this analysis onsisted of determining initial onstrution estimates for three 37 target ores hosen from the list of andidates in Chapter 3. Two of the three reators are beryllium moderated, water ooled, and ontain polyethylene plugs in the enter of the targets. These two onfigurations are diretly omparable; the only differene being they both have different ladding materials. The third reator is graphite moderated, water-ooled, and ontains zironium lad targets with beryllium plugs in the entral region. Table 16 presents ost estimates for initial onstrution of these reators based on the optimized dimensions determined with MCNP as presented in Chapter 1. A sample hand alulation of how the ost estimates were derived is provided in Appendix B. 1. Water moderated reators were exluded from this analysis as the sizes of these ores are muh larger than the baseline onfiguration of37 targets. It is assumed that the ost of building a water moderated reator would be omparable to that of a graphite moderated system. Estimates of the initial onstrution osts for a heavy water moderated reator were provided to illustrate that even though a 37 target ore moderated with this material is neutronially viable, it is obviously ve~ expensive and eonomially unfeasible. Results from the seond phase of the analysis are outlined in Table 17. This table is a summary of the annual ost estimates for waste generation expenses envisioned for the onventional and solution reator methods as ompared to those for the proposed reator onept. Original estimates of annual high level waste (HLW) disposal assume 57 elements per year with 16 kg U/yr (US DOE/EIS-0249D, Table 5-19, 1995) for the ACRR. This amount of uranium pertains to the urrent U02-Be0 fiel whih will not be used for prodution purposes. HLW disposal estimates for the ACRR are based on the new prodution fiel whih will ontain approximately 5 kg of U02 and 1 kg of stainless steel lad per fhel element. Low level waste (LLW) volumes for a onventional reator pertain to those expeted for the ACRR and HCF at 100% prodution apaity (US DOE/EIS-0249D, Table 5-19, 1995). Cost and waste volume estimates in the last olumn of Table 17 reflet previous alulations for 9%10produed via a solution reator (Glenn et al., Tables I & II, 1997). Unit pries in the seond olumn of Table 17 were not used to alulate osts of LLW and liquid waste disposal for the solution reator. 65

67 Table 16. Initial Constrution Cost Estimates of 37 Target Cores. Unit Prie Beryllium Graphite HeavyWater Moderated Moderated Moderated Reator Reator Reator Numberof Targets 37targets 37 tmgets 37targets 37targets $1,000/-iarget $37,000 $37,000 $37,600 ModeratorCost $1,775/lb Bea 2236 lbbe 7~11lb 57,355kgDZO $3,970,000 graphite $37,550,000 $198/lb graphiteb $1,428,000 $654.65/kg D20C PlugMaterial $1,775/lb Be Polyethylene Beryllium NoplWS 37plugs 37plugs $0 $78,000 TotalCost($) E$4,004,000 S1,543,000 S37,600,000 Table 17. Comparison of Annual Costs for Various Mo Prodution Methods. Unit Prie ConventionalReator Cintihem-type Solution (ACRR) TargetFueled Reator Reator Target $1,000/targetd 1924targets 1924targets Otargets Fabriation $1,924,000 $1,924,000 $0 Annual Fuel $20,000/feelelemente 57fuelelements 0.06kgg osts $6,300/kgsolution&l ~ $1,114,000 $380 HLW DKposal SpentNulear $loo/kgh 342kg/yr(57elements) Om3 Om3 Fuel $34300 $0 $0 LLWDisposal Routine $353/m m 15.74m3 25.5m3 $5,550 $5,550 $8,900 Solidified $353/m3 42.0m3 42.0m3 0.06m3 $14,800 $14,800 $56 IonExhange $353/m3 1.70m3 1.70m3 3.40m3 Resins $600 $600 $1200 Targets $353/m3 7.0m3 7.0m3 Otargets $2,470 $2,470 $0 Extration $353/m3 o olumns o olumns 8.00m3 olumns $0 $0 $2,850 LiquidWaste Disposal $3531m3 Target $3531m3 100m3J 100m3 Otargets Fabriation $35,300 $35,300 $0 Routine 7.3m3 $2, m3 $2, m3 $2,000 TotalCost($) $3,159,500 $1,985,300 $15,386 Personal ommuniation withrebeabrow Eletroni SpaeProduts International, Au&t bpersonal ommuniation withrebeabrow Eletroni SpaeProduts Intematiom~August1998. URLh@://~.ti~~bWtieuttiumOfide.ti1 dpersonslcommuniation withe.j.parm<snl,june1998. epmonalcommuniation withe.j.pamx+snl,june1998. fgle~ D.E.,MSThesis,University ofnewmexio,appendixa.13, Glenn,D.E.,MS Thesis, University ofnew Mexio,p.35,1995 hglenqd.e.,msthesis,university ofnewmexio,tsble3.1,1995. ~Personal Communiation withs.w.langley, SNI+September JMasseyandCoatsjSAND ,TableA4 66

68 Thetotal ost etiimates forinitial onstm~ion shorn in Table 16reveal aouple of important observations, the first of whih is the moderator ost. Expense inurred for an initial loading of targets is onsidered to be equivalent among the three reators. A heavy water moderated ore proved to be the least desirable moderator material investigated here due to its large onstrution ost, whih is a diret fhntion of the reator size. On the other hand, beryllium is relatively less expensive ompared to heavy water and ould be substituted as a reasonable hoie. Even though the ore size of a graphhe moderated reator is similar to a beryllium moderated syste~ graphite is the least expensive material of the group. When seleting the optimum moderator material, size and assoiated expense should not be the only fators driving a deision. A seond and more important observation evident fi-om these results is the balaned trade-off that must be made between these eonomi osts and the neutroni properties. Heavy water in general demonstrated good neutroni properties with high kff values, its only major disadvantage were the large reativity losses inurred due to temperature and density hanges. Coupling this to its exorbitant ost and limited distribution in the United States make it a poor hoie overall. Beryllium is a very good moderator due to its sattering properties and its ability to produe (~2n) reations. Its reativity losses due to temperature and density hanges, 149Smbuildup, Li produtio~ and 10% 35Uburnup are low. However, three issues that ould prove a beryllium moderated ore to be a poor hoie are: major reativity losses due to 135Xebuildup; and the prodution of large quantities of helium gas and 3H. Graphite is a good moderato~ however, its sattering properties lend it to produing more neutron leakage ompared to beryllium as shown by the lower bff values. Reativity losses due to temperature and density effets are very high nearly $3.00 above those for beryllium. Likewise, losses for 135Xebuildup afier shutdown are nearly $2.00 higher than those for beryllium. Figures for 149S~!Li, and 10% 235Uburnup are low and omparable to beryllium. Based on the evidene presented, the beryllium moderated system is better neutronially but more expensive. A graphite moderated system is the exat opposite, less expensive but less desirable neutronially. Thus, ombining the above trade-offs yields two reators whih ould be onsidered optimal final hoies. These are reators fieled with zironium lad targets in a beryllium or graphite moderator. Urdiortunately, one major drawbak to these reators would be the need to use zironium laddlng. Neutronially, zironium is a very good material due to low neutron absorption; however, its use would probably require a redesign of the target and possibly a new U02 oating proess as the urrent eletroplating proess is unproven with zironium targets. Several assumptions were made with regard to alulating the projeted osts in Table 17. First, target fabriation osts were onsidered to be omparable for the ACRR and target reator onept. Therefore, these are aounted for only as a fiel ost for the target reator. Seond, fuel osts for the solution reator were obtained from alulations for 20W10enrihed solution i%el (Glenn, Appendix A.13, 1995). Third, HLW for the sope of this analysis is lassified as spent fiel in solid or liquid form depending on the reator type. Current HCF waste disposition plans all for shipping proessed target shells with the solidified liquid waste from hemial extration operations. Consequently, HLW is not aounted for with the target reator beause 67

69 based on this assumptio~ the target shells would be treated as LLW. Likewise, Glenn assumes in his alulations that the solution reator fbel will not onsidered HLW based Ori the assumption that Ci onentrations for LLW lassifiation will not be exeeded (Glenn et al., 1997). A seond aspet of this analysis was to assess the eonomi feasibility of the target reator onept ompared to onventional and solution reators. A omparison of the osts assoiated with eliminating the driver ore of fiel elements inurs a savings of approximately $1.14 million. This is assuming the osts of purhasing targets for irradiation with both reators over the ourse of one operation year are equivalent. Muh less fuel is required for a solution reator and is shown to be less expensive than the reator methods. However, a thorough feasibility study has not been done to assess the ore lifetime of a solution reator based on fission produt buildup, temperature effets on reativity, et. If results of suh a study would prove that more fiel is required to overome these effets, then it is expeted the overall ost of suh a reator system (espeially fiel) will inrease signifiantly. EILW disposal osts are eliminated with a targetfbeled reator due to the lak of a driver ore of fhel elements and the assumption that the targets will be treated as LLW. HLW disposal osts with the solution reator are also eliminated based on the same assumption. LLW disposal osts are onsidered to be equivalent between the onventional and target reator methods. This is beause the Cintihem proess is the only method that will be used to hemially extrat 99M0from the targets. In addition, LLW dkposal osts of solidified liquid waste from hemial extration operations and general liquid waste horn target fabriation are muh higher than for the solution reator due to the utilization of targets. Large sale tests to date have not been onduted with the solution reator to determine its feasibility for a ommerial sale prodution of 9?M0 omparable to the onventional reator method. Consequently, solidified liquid and HLW osts doumented here would rise if more fbel was deemed neessary to meet the U.S. demand for 99M0with the solution reator. There are no osts assoiated with the use of extration olumns for the reators, but a large volume and ost must be aounted for with the solution reators. 6.1 Conlusions The results of this eonomi analysis illustrate the onventional reator method is the most expensive of the three prodution methods. Likewise, the ost estimates and assumptions for the target reator learly onfirm that the new reator onept is a signifiant improvement over the onventional reator. A major enhanement in terms of eonomis is the elimination of HLW disposal osts and the notable finanial savings inurred from elimination of the driver ore of fiel elements. These are two aspets with this reator onept whih ould render it a ompetitive option in the future. Beryllium and graphite proved to be the least expensive of the four moderators investigated. Balaning the eonomi impats of onstruting the reators hosen for this study and their neutroni requirements disovered from analysis presented in previous hapters yielded two redible andidates. One requires the use of a beryllium moderator and the other a graphite moderator; however, both also require the target ladding to be omposed of zironium. Zironium ladding works very well neutronially, but the eonomi impats of its potential 68

70 have not been realized with respet to the ument oating proess whih ould possibly entail a omplete overhaul of the urrent target design. Furthermore, an omprehensive and omplete analysis annot be ahieved for the omparison of 99M0prodution methods until estimates for aelerator produed 99M0are detailed in the areas of onstrution osts and waste generation. 69

71 7. CONCLUDING REMARKS 7.1 Reommendations For Future Work Several areas of this projet are reommended for onsideration as Mm-e work. Among the first of these is a thermal-hydraulis analysis for the two onfigurations hosen as final andidates. Pursuing suh an analysis for these two and possibly other onfqurations analyzed is onsidered beyond the sope of this endeavor. This type of analysis would require fiu-ther optimization studies to ensure the neutroni requirements of a partiular reator system are onsistent with its thermal-hydrauli limitations. A seond reommendation for fbture researh is firther oneptual development and parametri studies with LEU oated targets. Results Ilom a limited parametri study done here with LEU targets moderated with beryllium learly show there is potential for their use in a target-fheled reator. Optimization studies with other moderators would need to be done for fi.u-thervalidation. A major impediment to the use of LEU targets is lak of authorization from the FDA for a Master Drug File approving the prodution and proessing of fission produed 99M0born LEU targets, and there are no immediate plans to do so. Two major reasons are the relutane of pharmaeutial ompanies to aept fission produed 99M0with LEU beause of radionulidi impurity issues, and utilization of the Cintihem proess with this type of uranium is unproven. Implementation of ontrol rods into the MCNP model was not done nor were parametri studies onduted in this investigation. Parametri analyses to optimize a ontrol system are essential to quanti~ing how muh exess reativity would be required to overome reativity losses from the largest ontributors determined in this study, namely losses due to temperature and density hanges in the moderator, and fission produt poisoning by 135Xe. Further studies should be onsidered to examine the impats of material degradation with the solid moderators of beryllium and graphite. Lining the inside of the oolant hannels would need to be done with a material that does not ompromise the system neutronis (a material other than stainless steel), but also prevents aelerated material degradation by the interation of water with the beryllium or graphhe. Results showed signifiant quantities of He gas are produed in beryllium moderated systems whih adversely affet its strutural integrity; however, enasement of the beryllium ould prove to be a viable solution in simplifying waste disposition and reduing human exposure to beryllium powder. Enasement of the beryllium ould also prove to be a feasible solution to ontainment of large quantities of 3Hprodued in the beryllium matrix. It is oneivable that stainless steel targets ould still be used in a final reator design if the urrent proess for eletroplating U02 on zironium ladding is not possible without seriously ompromising the Cintihem proess. A system with the nominal target design ould be onstruted with the addition of moderator plugs in the entral region of the targets, or by inreasing the ore size slightly beyond 37 targets to more effiiently overome large reativity losses. For example, moderator plugs suh as polyethylene are so inexpensive to manufature that they ould easily be disposed of after eah irradiation yle. Some of these major issues require firther researh to omplete a omprehensive and thorough design of this reator onept. These and other questions not mentioned here would have to be resolved before any experimental studies ould be done with a prototype of this reator. 70

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