Technical Basis for Use of a Correlated Neutron Source in the Uranium Neutron Coincidence Collar

Transcription

1 Tehnial Basis for Use of a Correlated Neutron Soure in the Uranium Neutron Coinidene Collar by Margaret Ann Root B.S. Nulear Engineering University of New Mexio, 2011 SUBMITTED TO THE DEPARTMENT OF NUCLEAR SCIENCE AND ENGINEERING IN PARTIAL FULFILLMENT OF THE REQUIREMENTS FOR THE DEGREE OF MASTER OF SCIENCE IN NUCLEAR SCIENCE AND ENGINEERING AT THE MASSACHUSETTS INSTITUTE OF TECHNOLOGY SEPTEMBER Massahusetts Institute of Tehnology All Rights Reserved Signature of Author: Margaret Ann Root Department of Nulear Siene and Engineering August 3, 2015 Certified by: Rihard Lanza Senior Researh Sientist Thesis Supervisor Certified by: R. Sott Kemp Assistant Professor Thesis Reader Aepted by: Chair, Department Committee on Graduate Students 1

2 2

3 Tehnial Basis for Use of a Correlated Neutron Soure in the Uranium Neutron Coinidene Collar by Margaret Ann Root Submitted to the Department of Nulear Siene and Engineering on August 3, 2015, in Partial Fulfillment of the Requirements of the Degree of MASTER OF SCIENCE in NUCLEAR SCIENCE AND ENGINEERING Abstrat Nulear nonproliferation efforts rely on a variety of safeguards to protet sensitive materials in nulear failities. The enrihment of fresh light-water-reator fuel assemblies is verified by several inspetorates using the uranium neutron oinidene ollar (UNCL), whih uses neutrons from an ameriium lithium ( 241 AmLi) soure to interrogate the assemblies from one side. Eighteen 3 He tubes on the other three sides are used to ount the oinidene neutrons from the indued fission reations. Experiments have shown that 252 Cf ould also be used to omplete these measurements, providing several benefits over the use of the standard 241 AmLi soure. The UNCL is one of the many instruments that will be available for training purposes in the China Center of Exellene for Nulear Seurity (COE), whih is loated in Beijing, China. This thesis ontains a detailed haraterization of the response of this detetor with 252 Cf as ompared with 241 AmLi and an analysis of the tehnial basis for the use of 252 Cf in plae of 241 AmLi in the Anteh N2071 Neutron Coinidene Collar. This thesis (1) disusses the development a benhmarked, high-fidelity model of the UNCL using Monte Carlo N-Partile Extended (MCNPX), version a; (2) fully haraterizes the detetion parameters, inluding the effiieny profile, die-away time, and deadtime parameters; and (3) demonstrates the tehnial basis for the replaement of 241 AmLi soures with 252 Cf soures by assessing the penetrability of neutrons from eah soure, evaluating the statistial unertainty in the measurements inurred by eah soure, and investigating the possibility of a higher effetive average number of neutrons produed per fission using 252 Cf rather than 241 AmLi. This work demonstrates the suitability of 252 Cf as a substitute for 241 AmLi and in fat shows approximately a 7.5% improvement in ounting statistis over the traditional interrogation soure at 4% enrihment. Thesis Supervisor: Rihard Lanza Title: Senior Researh Sientist, Nulear Siene and Engineering Thesis Reader: R. Sott Kemp Title: Assistant Professor, Nulear Siene and Engineering 3

4 4

5 Aknowledgments I would like to express my gratitude first and foremost to Rihard Lanza, my thesis supervisor, whose unending faith in my ability to ondut graduate researh motivated me throughout my MIT areer. His positivity and support were a large part of what made this work possible. My thesis reader, R. Sott Kemp, was also instrumental in foring me to ritially evaluate the work that I was doing and helping me reah beyond my omfort zone in my researh. I would like to express my sinere gratitude to the Department of Energy/National Nulear Seurity Administration Offie of Global Material Seurity (NA-21) for their programmati support of my thesis projet. I would also like to thank my olleagues at Los Alamos National Laboratory (LANL), espeially Karen Miller, Howard Menlove, and Johnna Marlow, who have supported me through thik and thin for the past several years. This work would not have been possible without the support of the Laboratory or the guidane of my mentors. Martyn Swinhoe, Anthony Belian, Alexis Trahan, Alison Goodsell, and Andrea Favalli provided muh needed tehnial expertise, and the help of Jeff Arhuleta, Carlos Rael, and Manuel Gonzales was greatly appreiated during the experimental work, espeially when I had to move large numbers of fuel rods with a torn rotator uff. Finally, my aknowledgments would not be omplete without a mention of my friends, both at MIT and LANL, who kept me sane and fit throughout my graduate areer. I an t imagine my life without the experienes I had with the MIT Outing Club. I summited Mt. Rainier, learned to hike in brutal winter onditions while keeping a smile on my fae, and struggled up steep rok faes, all with the support of my inredible MITOC friends, who I am sure will be my friends for life. Although graduate shool was a struggle, it left me with some of the best memories of my life. Leigh Ann Kesler, Sally Miller, and Katherine Song kept me afloat and were always up for a razy adventure. I will always remember winter nights at Camelot, where it felt like the troubles of the week fell away and all that mattered was the warm stove, good food, and great ompany. The MIT Chamber Choir also helped me keep my spirits up, under the always wonderful diretion of Dr. William Cutter. At LANL, my limbing and running buddies helped me keep balane in my life and were instrumental in making Los Alamos one of my favorite plaes in the world. There are so many names of people who supported me for the past few years that I annot fit them all on this page. I will be forever grateful for my graduate experiene and those who made it possible. Thank you. 5

6 6

7 Contents Abstrat... 3 Aknowledgments... 5 Contents... 7 Figures... 9 Aronyms Introdution Context The Nonproliferation Treaty and the spread of nulear power Nulear safeguards and material balane Fuel fabriation failities Using 252 Cf in plae of 241 AmLi as an ative interrogation soure Motivation The UNCL Ative soure replaement China Center of Exellene (COE) Objetives Setions Bakground Introdution to neutron detetion UNCL Measurement methodology Measurement time, auray, and orretion fators Detetor response Deming regression Cf soures TCIF Measurement methodology for TCIF Past TCIF work UNCL PNEM AEFC Experimental setup System setup and operating parameters Eletronis and software

8 3.3 Fuel assembly harateristis Experimental proedures Charaterization measurements Calibration urve and penetrability Calibration urve Penetrability into assembly using 252 Cf ompared with 241 AmLi Detetor haraterization and Monte Carlo benhmarking measurements Motivation and limitations of experiments and Monte Carlo Detetor haraterization Benhmark measurements MCNPX simulations Detetor performane and unertainty analysis for TCIF Calibration urve omparison for 241 AmLi vs 252 Cf Curve fit Bakground analysis Passive bakground Soure bakground Fast fission + soure ontribution Combined bakground soures Unertainty alulation Unertainty of unfolded linear density Penetrability of neutrons from 241 AmLi vs 252 Cf Conlusions Overview of results Future work Referenes Appendix A: Mehanial drawings Appendix B: MCNPX detetor model

9 Figures Figure 1-1: Plot of the derease in separative work units (SWUs) required to enrih uranium from a given starting enrihment to weapons grade as the starting enrihment inreases. Plot assumes tails enrihment of 0.3% 235 U and weapons-grade uranium enrihment of 93% Figure 1-2: Photograph of the UNCL and the LANL pressurized-water-reator (PWR) mokup assembly Figure 1-3: Photograph of the view from the top of the UNCL Figure 1-4: Illustration of how the UNCL operates with a 241 AmLi interrogation soure Figure 1-5: (a) 241 AmLi neutron emission spetrum [9], (b) 238 U fission ross setion as a funtion of neutron energy [10] Figure 2-1: General strutural diagram of a gas-filled detetor [17] Figure 2-2: Gas-filled detetor operating voltages [17] Figure 2-3: Comparison between ross setions (barns) of ommon detetion media [10] Figure 2-4: Ideal pulse train Figure 2-5: Realisti pulse train Figure 2-6: Rossi-Alpha distribution. P = predelay, G = gate length, D = delay, R = real oinidene rate, A = aidental oinidene rate [20] Figure 2-7: Example of a thermal alibration urve produed for the UNCL Figure 2-8: Example of a fast alibration urve produed for the UNCL Figure 2-9: Fission ross setions for 235 U and 238 U and omparison with interrogation neutron energies [10] Figure 2-10: Illustration of time-orrelated, indued fission with a 252 Cf soure Figure 3-1: Photograph of UNCL surrounding the LANL PWR fuel assembly, with ative panel faing the amera Figure 3-2: Photograph of Dell laptop onneted to Canberra JSR-15 Handheld Multipliity Register Figure 3-3: (a) Photograph of 252 Cf soure ontained within HDPE holder, with top HDPE onnetor attahed. (b) Photograph of 241 AmLi soure ontained within tungsten holder Figure 3-4: Looking down at soure in enter of UNCL for effiieny measurements Figure 3-5: Deadtime measurement setup. Two soures were taped together and hung in the enter of the detetor Figure 3-6: All 204 rods are DU; no rods are LEU Figure 3-7: 184 rods are DU; 20 rods are LEU Figure 3-8: 164 rods are DU; 40 rods are LEU Figure 3-9: 144 rods are DU; 60 rods are LEU Figure 3-10: 124 rods are DU; 80 rods are LEU Figure 3-11: 104 rods are DU; 100 rods are LEU Figure 3-12: 84 rods are DU; 120 rods are LEU Figure 3-13: Penetrability measurement with depleted rods in the bak of the assembly Figure 3-14: Penetrability measurement with depleted rods in the front of the assembly Figure 3-15: Penetrability measurement with depleted rods in the enter of the assembly Figure 3-16: Penetrability measurement with depleted rods in bak orners of assembly

10 Figure 3-17: Penetrability measurement with depleted rods in front orners of assembly Figure 3-18: Penetrability measurement with depleted rods in enter of assembly Figure 3-19: Penetrability measurement with depleted rods in enter of front and bak of assembly Figure 4-1: Detetor voltage plateau showing operating voltage of 1680 V Figure 4-2: Singles [ts/s] measured horizontally from front to bak in the detetor (bak is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor Figure 4-3: Singles [ts/s] measured horizontally from left to right in the detetor (right is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor Figure 4-4: Singles [ts/s] measured vertially from top to bottom in the detetor (top is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor Figure 4-5: Calulated die-away times for the ases of no rods, depleted rods, and enrihed rods filling the sample spae Figure 4-6: Raw data for singles urves for both 241 AmLi and 252 Cf. A seond-degree polynomial is fit to the data Figure 4-7: Raw data for doubles alibration urves for both 241 AmLi and 252 Cf. A seond-degree polynomial is fit to the data Figure 4-8: (a) Sabrina [29] model of UNCL with fuel assembly in enter. (b) MCNPX model of UNCL from side, with fuel assembly in enter Figure 4-9: Sabrina utout of UNCL model Figure 4-10: MCNPX view of the detetor model from the top Figure 4-11: Singles [ts/s] horizontally from front to bak in the detetor (bak is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor. Comparison is between simulated and measured results Figure 4-12: Singles [ts/s] horizontally from left to right in the detetor (right is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor. Comparison is between simulated and measured results Figure 4-13: Singles [ts/s] vertially from top to bottom in the detetor (top is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor. Comparison is between simulated and measured results Figure 4-14: Doubles alibration urve benhmark for 252 Cf soure with room bakground subtrated. The y interept for simulated urve = The y-interept for measured urve = A seond-degree polynomial is fit to the urves Figure 4-15: Ative doubles alibration urve benhmark for 252 Cf soure with passive 238 U spontaneous fission and room bakground subtrated. A seond-degree polynomial is fit to the urves Figure 4-16: Net doubles alibration urve benhmark for 252 Cf soure with all forms of bakground subtrated and 5.4% orretion fat applied to urves. A seond-degree polynomial is fit to the urves

11 Figure 4-17: Ative doubles alibration urve benhmark for 241 AmLi soure with only room bakground subtrated. A seond-degree polynomial is fit to the urves Figure 4-18: Net doubles alibration urve benhmark for 241 AmLi soure with all soures of bakground subtrated. A seond-degree polynomial is fit to the urves Figure 4-19: Net doubles alibration urve benhmark for 241 AmLi soure with passive bakground subtrated and 3.4% orretion fator applied to alibration urve. A seond-degree polynomial is fit to the urves Figure 4-20: Change in singles ount rate due to fuel assembly position, x diretion ( 252 Cf) Figure 4-21: Change in doubles ount rate due to fuel assembly position, x diretion ( 252 Cf) Figure 4-22: Change in singles ount rate due to fuel assembly position, y diretion ( 252 Cf) Figure 4-23: Change in doubles ount rate due to fuel assembly position, y diretion ( 252 Cf) Figure 4-24: Unertainty in ount rate due to fuel assembly position, x diretion ( 241 AmLi) Figure 4-25: Unertainty in doubles ount rate due to fuel assembly position, x diretion ( 241 AmLi) Figure 4-26: Unertainty in singles ount rate due to fuel assembly position, y diretion ( 241 AmLi) Figure 4-27: Unertainty in doubles ount rate due to fuel assembly position, y diretion ( 241 AmLi) Figure 5-1: Comparison of the net measured doubles ount rates generated using 252 Cf and 241 AmLi, with bakground subtrated. A seond-degree polynomial is fit to the urves Figure 5-2: Ratio of the doubles ount rates generated using 252 Cf and 241 AmLi, with bakground subtrated Figure 5-3: Ative doubles alibration urve omparison for 252 Cf and 241 AmLi at enrihments ranging from 0.219% to 5% (2.652 g 235 U/m to g 235 U/m) before bakground was subtrated Figure 5-4: Net doubles alibration urve omparison for 252 Cf and 241 AmLi at enrihments ranging from 0.219% to 5% (2.652 g 235 U/m to g 235 U/m) after bakground subtration Figure 5-5: Ratio of 252 Cf ount rate to 241 AmLi ount rate (both with bakground subtrated) at enrihments ranging from 0.219% to 5% (2.652 g 235 U/m g 235 U/m) Figure 5-6: Deming urve fit to 241 AmLi alibration urve Figure 5-7: Deming urve fit to 252 Cf alibration urve Figure 5-8: Total neutron ross setions for tungsten (red) and 238 U (green) [10] Figure 5-9: Illustration of ontribution of interrogation soure, fast fission, and 238 U spontaneous fission to the ative doubles ount rate. A fourth-degree polynomial is fit to the data Figure 5-10: Illustration of ontribution of interrogation soure and fast fission to ative doubles ount rate with 238 U spontaneous fission bakground subtrated. A fourth-degree polynomial is fit to the data Figure 5-11: Illustration of respetive ontributions to the ative doubles rate from the interrogation soure, spontaneous fission, and fast fission Figure 5-12: Doubles (alibration) urve with all bakground subtrated Figure 5-13: Comparison of doubles urves with bakground subtrated using individual ontributions from eah soure of bakground, or simply the y-interept

12 Figure 5-14: Change in net doubles rate from uniform rod distribution with all depleted rods loated in different loations (front, enter, and bak) within the assembly Figure 5-15: The hange in net doubles rate from uniform rod distribution, with all depleted rods loated in different loations (front orners, enter, enter edges, and bak orners) within the assembly

13 Tables Table 1-1: International fuel fabriation apabilities [6] Table 3-1: 241 AmLi and 252 Cf soures used in experiment Table 3-2: Eletronis speifiations [27] Table 3-3: PWR mokup fuel harateristis [28] Table 4-1: Singles [ts/s] measured horizontally from front to bak in the detetor (bak is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor Table 4-2: Singles [ts/s] measured horizontally from left to right in the detetor (right is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor Table 4-3: Singles [ts/s] measured horizontally from left to right in the detetor (right is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor Table 4-4: Die-away time measurements Table 4-5: Deadtime measurements: two-soure method Table 4-6: Summary of detetor parameters Table 4-7: Soure in enter of detetor Table 4-8: Soure in interrogation soure hole Table 4-9: Raw data for 241 AmLi alibration urve Table 4-10: Raw data for 252 Cf alibration urve Table 4-11: A7-867 in sample measurement avity simulated vs measured Table 4-12: Cf-12 in interrogation soure hole simulated vs measured Table 4-13: Doubles rates with inreasing gate lengths for the ase of no sample, simulated vs measured Table 4-14: Die-away time with no sample simulated vs measured Table 4-15: Passive assembly bakground Table 4-16: Unertainty in singles and doubles rates due to fuel pellet density ( 252 Cf) Table 4-17: Unertainty in singles and doubles rates due to fuel pellet density ( 241 AmLi) Table 4-18: Change in ount rate due to fuel assembly position, x diretion ( 252 Cf) Table 4-19: Change in ount rate due to fuel assembly position, y diretion ( 252 Cf) Table 4-20: Unertainty in ount rate due to fuel assembly position, x diretion ( 241 AmLi) Table 4-21: Unertainty in ount rate due to fuel assembly position, y diretion ( 241 AmLi) Table 4-22: Estimate of total model unertainty for 252 Cf and 241 AmLi Table 5-1: Passive assembly bakground Table 5-2: Soure bakground Table 5-3: Doubles bakground ontributors Table 5-4: Ative ontribution to unertainty in doubles rate Table 5-5: Passive ontribution to unertainty in doubles rate Table 5-6: Soure+fission ontribution to unertainty in doubles rate Table 5-7: Unertainty in net doubles rate for 252 Cf and 241 AmLi Table 6-1: Summary of differenes between 241 AmLi and 252 Cf interrogation soure

14 Aronyms ABACC Brazilian-Argentine Ageny for Aounting and Control of Nulear Materials AEFC Advaned Experimental Fuel Counter COE Center of Exellene DMBC De-Randomizing Mixer Buffer Counter DU Depleted Uranium EURATOM European Atomi Energy Ageny HDPE High-density Polyethylene IAEA International Atomi Energy Ageny INCC IAEA Neutron Coinidene Counting LANL Los Alamos National Laboratory LEU Low Enrihed Uranium MBA Material Balane Area MCNPX Monte Carlo N-Partile Extended MUF Material Unaounted For NDA Nondestrutive Assay NPT Nonproliferation Treaty PNEM Passive Neutron Enrihment Meter PWR Pressurized Water Reator SQ Signifiant Quantity TCIF Time-orrelated Indued Fission UNCL Uranium Neutron Coinidene Collar 14

15 1 Introdution 1.1. Context The Nonproliferation Treaty and the spread of nulear power The Nonproliferation Treaty (NPT), whih entered into fore in 1970, was born of an international ommitment to nulear seurity. The treaty outlines a three-tiered strategy, whih inorporates disarmament, nonproliferation, and peaeful uses of nulear energy, to provide a foundation for international ooperation to redue the threat of nulear weapons [1]. In the nearly 40 years between 1973 and 2012, world energy onsumption inreased by over 3.5 times, from 6129 TWh to 22,668 TWh [2]. At the same time, the fration of the world s energy produed by nulear power inreased from 3.3% to 10.9%. An inrease in the onsumption of nulear power has resulted in an inrease in the availability of nulear materials and knowledge of nulear siene, and thus a potential inrease in the number of opportunities for nulear material diversion and weapons development. The prevention of nulear terror is urrently listed as one of the Grand Challenges of Engineering [3]; monitoring and preventing the spread of illiit nulear tehnologies and materials requires inredible national and international efforts Nulear safeguards and material balane Nulear safeguards are urrently the primary method used to ensure that the nulear materials and tehnologies used in ivilian nulear reators are not diverted for use in the onstrution of nulear weapons. Safeguards provide administrative and tehnologial barriers against the diversion of nulear materials throughout the ivil nulear fuel yle. International inspetorates, suh as the International Atomi Energy Ageny (IAEA), European Atomi Energy Ageny (Euratom), and Brazilian-Argentine Ageny for Aounting and Control of Nulear Materials (ABACC), work to provide redible assurane to the international ommunity that states with safeguards agreements are not diverting nulear materials from delared ativities [4]. Many tehnial efforts seek to verify a material balane aross different nulear failities. A material balane area (MBA) is a spae defined for material aountany purposes as an indoor or outdoor spae for whih the quantity of nulear material an be determined at any given time and for whih the amount of material transferred in or out of the spae an be determined [5]. This projet investigates a tool whih is used to ensure a material balane through suh an area in fuel fabriation failities Fuel fabriation failities There are 15 light-water-reator fuel fabriation failities urrently operating worldwide in 12 ountries, with a total apaity of nearly 10,000 metri tons of uranium per year [6]. Most of these failities fall under either international or domesti nulear safeguards. Fresh fuel is the term used for nulear fuel that has been enrihed to reator grade and fabriated into fuel but that has not been used in a reator. Fresh fuel measurements are used in safeguards to onfirm fuel fabriation plant output and to provide a soure term before fuel is onsumed. Careful monitoring 15

16 of reator-grade uranium is imperative beause the amount of separative work, the relative amount of energy required for enrihment, required to enrih uranium to weapons grade, dereases by a fator of about three one the material has been enrihed to reator grade, as shown in Figure 1-1. Separative work to get to weapons grade Starting with natural uranium (0.711% 235 U) [kg SWU/ kg produt] Starting with reator grade uranium (5% 235 U) Starting enrihment [% U-235] Figure 1-1: Plot of the derease in separative work units (SWUs) required to enrih uranium from a given starting enrihment to weapons grade as the starting enrihment inreases. Plot assumes tails enrihment of 0.3% 235 U and weapons-grade uranium enrihment of 93%. The plot in Figure 1-1 was generated by ombining the following equations: F = P + W, (1-1) x F F = x P P + x W W, (1-2) and SWU = PV(x P ) + WV(x W ) FV(x F ), (1-3) where F, P, and W are the respetive amounts of feed (F), produt (P) and waste (W) in kilograms; x F, x P, and x W are the enrihments of the feed, produt and waste, respetively; V(x F ), V(x P ), and V(x W ) are the value funtions for the feed, produt, and waste, respetively; and SWU is the number of separative work units required to enrih to a given produt enrihment. The value funtion, V(x i ), is given by the following equation: x i V(x i ) = (2x i 1)ln (1-4) 1 x i A list showing the ountries that urrently have operating fuel fabriation failities, the number of failities in eah ountry, and eah ountry s total yearly fuel fabriation apaity is presented in Table 1-1 [6]. 16

17 Table 1-1: International fuel fabriation apabilities [6] * Indiates ountry that is not a party to the NPT Country Number of failities Capaity (t HM/year) Brazil China Frane Germany India* 1 24 Japan South Korea Spain Sweden UK USA This table shows that the international apaity for the fabriation of uranium fuel is roughly 10,000 tons of heavy metal per year. The IAEA defines a signifiant quantity (SQ) as the approximate amount of nulear material for whih the possibility of manufaturing a nulear explosive devie annot be exluded [7]. For low-enrihed uranium (LEU), whih is defined as uranium with <20% 235 U, this SQ is defined as 75 kg of 235 U. If we assume that the uranium produed by fuel fabriation failities worldwide has an average enrihment of 4%, then fuel fabriation failities are proessing 5300 SQs of nulear material per year. If even 0.1% of the world s supply of fresh nulear fuel were diverted, this amount would be about 5 SQs of nulear material. If we assume that eah fuel rod ontains 2 kg of 235 U and that eah assembly ontains approximately 200 fuel rods, then only about 40 fuel rods, or 20% of one fuel assembly, would be required to obtain an SQ of nulear material. Thus, areful measurements to ensure a material balane in fuel fabriation failities. Beause the uranium is ontained in fabriated fuel bundles, it is not easily measured using diret measures suh as weight. Highly sophistiated measurement tehniques must be used to infer the amount of material present. The Uranium Neutron Coinidene Collar (UNCL) is one suh method Using 252 Cf in plae of 241 AmLi as an ative interrogation soure The UNCL was developed at Los Alamos National Laboratory (LANL) in the early 1980s [8]. The UNCL is a nondestrutive assay (NDA) system, whih relies on ative neutron interrogation to verify the 235 U ontent in light-water-reator fuel assemblies. This thesis expands on reent efforts to transition to the use of a new neutron interrogation soure for the UNCL. Historially, the IAEA has relied on 241 AmLi, whih emits single neutrons randomly in time, for ative neutron detetors. Today, 241 AmLi soures are no longer made in a pratial apsule for safeguards appliations, and new detetors require a substitute. This thesis seeks to evaluate the possibility of using 252 Cf, a spontaneous fission soure that emits bursts of neutrons that are orrelated in time, instead of 17

18 241AmLi. This work provides a foundation for the transition to 252 Cf in other detetors and demonstrates the apabilities of omputational methods to predit aurately ative detetor response when a spontaneous fission soure is used in plae of a random soure Motivation The UNCL The UNCL is an ative neutron oinidene system used in fuel fabriation failities that uses thermal neutron interrogation to measure the linear density of 235 U [g 235 U/m of ative length in assembly] in fresh light-water-reator (LWR) fuel and was designed at LANL in the early 1980s. A photograph of the ollar is shown in Figure 1-2. LANL PWR mokup fuel assembly Juntion box UNCL passive panel Interrogation soure hole UNCL ative panel Figure 1-2: Photograph of the UNCL and the LANL pressurized-water-reator (PWR) mokup assembly. A photograph of the view of the detetor from above is shown in Figure

19 Sample spae UNCL passive panel Interrogation soure hole Juntion box UNCL ative panel Figure 1-3: Photograph of the view from the top of the UNCL. The original model ( Mod I ) of this detetor onsists of 18 3 He tubes embedded within three highdensity polyethylene (HDPE) bloks, whih surround a fuel assembly. A fourth HDPE blok ontains a slot for a neutron soure, whih is used to indue fission within the fuel. The indued fission neutrons are subsequently deteted by the surrounding 3 He proportional ounters. The oinidene rate within the tubes is used to unfold the linear density of 235 U in the fuel. This proess is illustrated for the ase in whih 241 AmLi is used as the interrogation soure in Figure 1-4. t 1 t 2 G AmLi t 1 t 2 Time [μs] Figure 1-4: Illustration of how the UNCL operates with a 241 AmLi interrogation soure. 19

20 Ative soure replaement Historially, the UNCL and other ative interrogation detetors have relied on ameriium lithium ( 241 AmLi) soures to indue fission within the objet in question. These soures were hosen for several reasons: (1) they have a low average neutron energy (0.3 MeV), meaning that they indue minimal fast fission in 238 U, as shown in Figure 1-5; (2) they do not emit orrelated neutrons and therefore ontribute no oinidene bakground to the measurement; and (3) they have a long halflife (432 years), meaning that they require a small normalization fator for soure deay; thus, deay ontributes little error to the measurement. Figure 1-5: (a) 241 AmLi neutron emission spetrum [9], (b) 238 U fission ross setion as a funtion of neutron energy [10]. As pratial 241 AmLi soures are no longer ommerially available, an alternative interrogation soure must be onsidered as a replaement. Reent work by Menlove [11] and Miller [12] has suggested that 252 Cf is a suitable alternative to 241 AmLi in many ases, and in fat provides several benefits over the traditional option. Ameriium lithium is an (α,n) soure, whih emits single neutrons randomly in time. Californium-252, on the other hand, is a spontaneous fission soure, whih emits multiple neutrons eah time it fissions, resulting in a more ompliated multipliity profile. The 252 Cf ontributes a signifiant oinidene bakground to fresh fuel measurements, whih originally seemed to be a disadvantage. However, a new tehnique known as timeorrelated indued fission (TCIF) has been proposed by Menlove [11] to take advantage of the higher multipliity of the alifornium soure and thereby turn this disadvantage into an advantage China Center of Exellene (COE) This work is motivated in part by a desire to support the China Center of Exellene in Nulear Seurity training program. The United States (US) and China signed a government-to-government agreement in January 2011 to establish a Center of Exellene for training in nulear seurity. The faility, whih is loated in Beijing, will house several detetors for training purposes, many of whih have historially relied on 241 AmLi interrogation soures. A lear understanding of the impat of 252 Cf on the effiay of ative NDA systems is important to enable the China COE to transition the 20

21 interrogation soure. The response of eah detetor, inluding the UNCL, must be analyzed in detail, and thorough doumentation of the physis and measurement tehniques involved will inrease the effiieny and repliability of this proess. It is also essential that benhmarked, high-fidelity models exist to allow future users to understand the expeted response of eah detetor to a large variety of samples Objetives To support the efforts of the China COE in the development of China s safeguards training programs, this work aims to haraterize the performane of the UNCL (Anteh N2071) using 252 Cf in plae of 241 AmLi. Speifially, the objetives of this work are to (1) develop a benhmarked, highfidelity model of the UNCL using Monte Carlo N-Partile Extended (MCNPX) ver a [13]; (2) haraterize the detetion parameters, inluding the effiieny profile, die-away time, and deadtime parameters; and (3) demonstrate the tehnial basis for the replaement of 241 AmLi soures with 252Cf soures by assessing the penetrability of neutrons from eah soure, evaluating the statistial unertainty in the measurements inurred by eah soure, and investigating the possibility of a higher effetive average number of neutrons produed per fission using 252 Cf rather than 241 AmLi Setions Chapter 2 provides the bakground that is neessary to understand this projet, inluding the impetus for fresh fuel measurements, the inspetion proesses used by domesti and international safeguards programs, information about neutron interation mehanisms, neutron measurement tehniques, ommon neutron soures and their harateristis, an overview of neutron oinidene ounting, a desription of the UNCL system, an overview of TCIF, and a disussion of past work. Chapter 3 disusses the experimental setup and desribes the detetor setup, software and hardware, the interrogation soures, and fuel assembly harateristis. Chapter 4 desribes the detetor haraterization and Monte Carlo benhmarking measurements, inluding a desription of the MCNPX model, the high-voltage plateau, the effiieny measurements, the die-away time measurements and alulations, and the deadtime parameter measurements and alulations. This hapter also inludes a tabulated summary of the detetor operating parameters. Chapter 5 disusses the detetor performane measurements, inluding the 252 Cf and 241 AmLi alibration urves, both measured and simulated, and the results of a omparison of the penetrability of the interrogation neutrons emitted from 252 Cf and 241 AmLi. This hapter also ontains the unertainty analysis for the TCIF method and disusses eah soure of bakground in the measurement and how to manage the bakground ontribution. Chapter 6 desribes the onlusions, inluding a summary of the results, lessons learned, and possible future work. The appendies ontain examples of the MCNPX input deks, some additional data beyond what was inluded in the thesis, and the mehanial drawings used to onstrut the detetor and to develop the MCNPX input deks. 21

22 THIS PAGE INTENTIONALLY LEFT BLANK 22

23 2 Bakground The UNCL is a neutron-based NDA system that relies on several omplex fundamental proesses to funtion. The end result is a fairly simple approah to measuring the linear density of 235 U in uranium fresh fuel. To understand the proesses involved in making the transition from 241 AmLi to 252Cf as the interrogation soure for this detetor, several key terms and onepts must be reviewed. This setion desribes the fundamental proesses and onepts required to analyze the performane of the detetor. 2.1 Introdution to neutron detetion The disovery of the neutron by James Chadwik in 1932 ame muh later than the disovery of its fellow oupants of the nuleus, protons, and eletrons, whih had ourred by 1920 [14]. Before neutrons were deteted, there did not seem to be a way for eletrons to be bound within the nuleus to balane out the harge of the protons without serious disrepanies in atomi theory. To study this phenomenon, Chadwik used polonium as a soure of alpha partiles, whih passed through a beryllium target. These interations produed mysterious partiles that had roughly the same mass as protons and that had an inredibly long range, whih would eventually be identified as neutrons [15]. When these mystery partiles ollided with gases suh as hydrogen, nitrogen, and argon, the ollision resulted in the emission of highly energeti ions, whih were then deteted. Earlier work hypothesized that these ions were produed by the reoil of atoms within the gas following an interation with gamma rays. Aording to Chadwik, Up to the present, all the evidene is in favor of the neutron, while the quantum [gamma ray] hypothesis an only be upheld if the onservation of energy and momentum be relinquished at some point [15]. Today, safeguards inspetors ommonly use neutron measurements for NDA beause they have low bakground prevalene and are more penetrating than, for instane, the kev gamma rays emitted from 235 U, whih undergo a large amount of self-attenuation within a sample. For example, in the ase of a 200-μm diameter partile of UO 2, roughly 10% of the gamma rays emitted from the 235U are sattered, whih auses them to lose energy, or they are absorbed within the sample [16]. Thus, neutrons an provide a total volume assay of a large, dense (i.e., uranium or plutonium) sample, while gamma measurements an provide only a surfae-level assay of the sample [16] Detetion mehanisms Neutrons, unlike harged partiles or gamma rays, must be deteted by indiret means, whih is why they took so muh longer to disover than the other omponents of the nuleus. Neutrons an be deteted through two interations with matter: sattering or a nulear reation [17]. When a neutron is sattered from a nuleus, it transfers some of its energy to that nuleus. The reoiling nuleus an then ionize the surrounding matter, but only if the ollision transfers enough energy. The average frational energy loss when a neutron interats with a material of atomi number, A, goes as 23

24 A 2 +1 (A+1) 2 [2-1] Thus, neutrons will lose about half of their energy in a satter from hydrogen (A = 1), but only about 15% of their energy in a satter from graphite (A = 12). The neutron an also undergo a nulear reation, and the reation produts an then be deteted. In either ase, the harged partiles or gamma rays that result from these reations produe eletrial signals, whih are deteted Gas detetors Several types of detetors are used to measure neutron signals, and the detetor hoie depends on several fators, inluding appliation, neutron energy and intensity, and ost restritions. The detetion medium an be a solid, liquid, or gas. Safeguards measurements have traditionally relied on gas detetors, speifially 3 He proportional ounters, beause of favorable harateristis suh as high detetion effiieny, low gamma-ray sensitivity, reliability, and ompat geometry. Gas detetors rely on either ollisions or nulear reations between neutrons and the detetion medium to generate a neutron signal. The general form of a gas-filled detetor is shown in Figure 2-1. Figure 2-1: General strutural diagram of a gas-filled detetor [17]. Thermal neutrons are deteted via reoil interations with the detetion medium, whih generate eletron-ion pairs. When a voltage is applied aross the detetor tube, the eletrons are pulled toward the anode, whereas the ions are pulled toward the athode, generating an eletrial signal. This signal is measured by the detetor eletronis. These detetors generate different kinds of signal, depending on their operating voltage, as shown in Figure

25 Figure 2-2: Gas-filled detetor operating voltages [17]. In the reombination region, the voltage differene between the athode and anode is not enough to keep the eletrons and ions apart, and they quikly reombine, generating only a very small signal. In the ionization hamber region, enough voltage is applied to nearly eliminate reombination. Beause all of the harges are olleted, an inrease in voltage does not inrease the urrent (a phenomenon known as ion saturation). At a ertain threshold voltage, gas multipliation begins. If the energy of the free eletrons generated in the initial ollision is greater than the ionization energy of the gas moleule, a ollision between these eletrons and a gas moleule an yield another eletron-ion pair. This proess an repeat itself, eventually generating a asade of eletron-ion pairs known as a Townsend avalanhe. In the proportional region, this multipliation is linear, meaning that the harge olleted by the detetor is diretly proportional to the number of ion pairs reated in the initial ollisions between the neutrons and the detetion medium [18]. The harge generated in the initial ionization an be amplified by a fator of up to 10 5 in a Townsend avalanhe [16]. As the voltage inreases, the detetor enters a region of limited proportionality before entering the Geiger-Mueller region. In this region, enough ions are generated to prevent further gas multipliation, whih fores the multipliation proess to end when a ertain number of positive ions are generated, regardless of the number of initial ion pairs reated. Thus, eah pulse has the same amplitude and retains no information about the initial neutrons [18]. 25

26 He proportional ounters Helium-3 proportional ounters are the workhorse of neutron measurements for nulear safeguards. These detetors are gas-filled detetors, whih operate in the proportional region, and rely on the reation 3 1 He + n H + p MeV The proton and the triton arry the MeV in the form of kineti energy, whih is olleted by harge olletors. The thermal neutron apture ross setion for 3 He is 5330 barns. A omparison among the ross setion of 3 He and that of two other ommon neutron detetion media, 6 Li and 10 B, is shown in Figure 2-3 [10]. It is lear from the figure that 3 He has the highest ross setion for thermal neutrons of the ommon detetion media. Thermal (0.025 ev) 6 Li 3 He 10 B Figure 2-3: Comparison between ross setions (barns) of ommon detetion media [10]. Although neutrons are emitted from the sample and the soure with a wide range of energies, this information is not preserved in measurements with these detetors. Rather, neutrons are thermalized before they reah the detetor to take advantage of the high thermal neutron ross setion of 3 He Neutron oinidene ounting Neutron oinidene ounting is a tehnique used to redue the bakground ontribution to neutron measurements. The singles, or totals rate, is a measure of the total number of ounts that are deteted and an be severely impated by various soures of bakground, suh as (α,n) neutrons in passive measurements of plutonium and the interrogation soure in ative 26

27 measurements of uranium. To ameliorate these effets, oinidene measurements ount the number of times two neutrons are deteted in the same time gate, whih an provide evidene that the two ounts were orrelated in time. Correlated ounts are emitted from the same fission event, disriminating them from unorrelated bakground. The oinidene rate, whih is referred to in this thesis as the reals or doubles rate, an be used to infer information about a sample. The total ount rate will be referred to as the totals or singles rate. Coinidene measurements are primarily useful for measurements in whih the sample is well haraterized, there is a need for NDA of the material, and short measurement times are a priority [19]. Other tehniques, suh as multipliity ounting and destrutive methods, are muh better suited to ill-defined samples Die-away time Die-away time (τ) is the measure of the mean neutron lifetime in the detetor i.e., the time before the neutrons emitted from the sample (and from the soure, in the ase of a 252 Cf interrogation soure) undergo an interation that auses them to be removed from the system, suh as apture in the detetor tubes, leakage, or absorption in the surrounding material (in this ase, thermal neutron absorption in hydrogen.) The die-away time primarily depends on the time for neutrons to slow from fast to thermal, and then satter in the moderator as thermal neutrons. Dieaway time an be redued by lining the detetor with admium, whih removes all but the fast neutrons. The neutron population will fall gradually over time, and the average probability that a neutron will undergo an interation that will remove it from the system is roughly onstant with time [19]. Therefore, the neutron population falls exponentially as [19] N(t) = N(0)e t τ, [2-2] where N(t) is the neutron population at time t and τ is the die-away time. The die-away time an be determined by onduting measurements of a fissioning sample using two different gate lengths, G 1 and G 2, where G 2 is twie as long as G 1. If the resulting measurements yield oinidene rates of R 1 and R 2, respetively, the die-away time goes as [20] τ = G 1 [2-3] 1). ln( R 2 R Rossi-Alpha distribution It is useful to onsider the stream of ounts being deteted as a pulse train, as illustrated in Figure 2-4, where eah line represents a ount. In an ideal world, in whih orrelated and unorrelated events are learly distinguishable, a pulse train would look like the following: 27

28 Figure 2-4: Ideal pulse train. Unfortunately, the real world is not so routine. A real pulse train looks more like the illustration in Figure 2-5: Figure 2-5: Realisti pulse train. In this ase, the ounts are not easily separated into orrelated and unorrelated events. A Rossi-alpha distribution is a useful tool for making sense of the data aumulated in the pulse train. A ounter begins ounting at t = 0, starting when it detets the first pulse (following a short predelay, to redue detetor noise effets), and ounts all of the pulses until the end of a preset time interval, alled a gate. The ounts are plaed into bins orresponding to their arrival times. After the gate loses, the ounter eases to ount for a long delay, then begins to ount again [20]. The resulting urve takes the shape shown in Figure 2-6. Figure 2-6: Rossi-Alpha distribution. P = predelay, G = gate length, D = delay, R = real oinidene rate, A = aidental oinidene rate [20]. Real events generate an exponential urve that rises above the aidental (random) bakground, whih is onstant in time. The distribution is given by [20] 28

29 S(t) = A + Re t τ, [2-4] where S(t) is the number of events at time t, A is the aidental ount rate, R is the real oinidene rate, and τ is the detetor die-away time. The aidental rate an either be measured or approximated as [20] A = GT 2, [2-5] where G is the gate length and T is the totals rate. Aidental ounts inlude random bakground events, unorrelated fission and bakground events, and unorrelated fission events that are deteted simultaneously [20]. In the R + A gate, the ounting algorithm reords all possible permutations of oinidenes, so that for n losely spaed events, the number of reorded events goes as n(n-1)/2 [20]. The differene between the number of ounts in the R + A gate and those in the A gate gives the number, R, of real oinidenes that ourred within the gate. R orresponds diretly to the fission rate and thus an be used to unfold information about the sample [20] Multipliity and multipliation Multipliity (ν) desribes the number of neutrons emitted in eah fission event. This number typially ranges from 0 to 7 with varying distributions depending on the isotope, and the average number of neutrons produed per fission (ν ) is generally used when referening the multipliity of a sample. Multipliation is the term used for indued fission within a sample. The primary type of multipliation of onern in UNCL measurements is known as leakage multipliation. Leakage multipliation desribes the number of neutrons that exit the sample for eah soure neutron. Regardless of the number of fissions that our within the sample, neutrons may be aptured or esape the sample without induing a fission. Therefore, if p is the probability that a neutron will indue a fission, 1-p is the probability that the neutron will disappear beause of leakage (assuming that the probability of apture is very small), and ν i is the average number of neutrons produed per indued fission, the leakage multipliation goes as [19] M = 1 p 1 pν i. [2-6] pν i is better known as the multipliation fator, k (a ommon term in reator physis), and simply desribes the ratio of the number of neutrons in one generation to the number in the previous generation. In a system that is subritial, this value will always be less than one. Furthermore, the number of neutrons that leak out of the sample will always be less than or approximately equal to the total number of neutrons that are produed within the sample due to indued fission, depending on the probability of apture [19] Self-shielding Self-shielding is the term used to desribe absorption when it ours within a sample itself. When thermal neutrons enter a sample, many are absorbed or sattered before they an penetrate to the interior of the sample. As the enrihment of the sample inreases, absorption or sattering of the 29

30 inoming thermal neutrons beomes more likely. At the same time, the multipliation of the sample also inreases. The ompetition between these effets is the driving fator behind the onave down alibration urve for many NDA instruments. 2.2 UNCL The UNCL is urrently used by several inspetorates to verify the linear density of fresh uranium fuel in fuel fabriation failities. An 241 AmLi neutron soure on one side of the assembly is used to indue fission within the fuel. Eighteen 3 He tubes, whih are embedded within HDPE bloks on the other three sides of the assembly, are used to ount the number of oinidene neutrons emitted from the assembly. The results of these measurements are used to unfold the linear density [g 235U/m] of the fuel assembly. Two models of the UNCL are urrently used by inspetorates: the UNCL-I and the next-generation instrument, the UNCL-II, whih has a higher effiieny. The UNCL-I was used for the measurements performed in this thesis Measurement methodology Inspetorates follow a standard methodology when onduting measurements using the UNCL. Data are olleted using the IAEA Neutron Coinidene Counting (INCC) software [21]. Measurements begin with a short (200- to 300-s) room bakground measurement. Next a passive measurement (i.e., a measurement without an interrogation soure) of the fuel assembly by itself is onduted. When these bakground measurements are omplete, the alibration measurements an be onduted. The alibration urve relies on measurements of the sample s doubles ount rate at inreasing enrihments (inreasing 235 U mass/m). Rather than inreasing the enrihment of the fuel assembly by inreasing the enrihment of all of the rods uniformly, beause of material and time onstraints, the alibration urve is generated by replaing depleted uranium rods (0.219% enrihed) with enrihed uranium rods (3.19% enrihed) until the maximum possible enrihment is reahed. It is assumed that this proedure produes approximately the same results as would be generated with a uniform inrease in enrihment. The INCC ode is used to fit a Deming urve to the data. When a urve has been fit to the data, verifiation measurements an be onduted. The doubles rate for a sample is measured and then ompared with the alibration urve to unfold the linear density of the sample. This mass is ompared with the delared mass of the sample. The assay mass should agree with the delared mass Measurement time, auray, and orretion fators Calibration measurements are typially onduted for at least 5 yles of 1000 s ( s) to ensure that these measurements have suffiiently good statistis to be used as a referene for future measurements. The standard measurement time for verifiation measurements using an 241AmLi soure is 1000 s, whih yields a measurement unertainty in the doubles rate of <1% and an unertainty in the linear density of ~1% [8]. 30

31 In pratie, orretion fators applied to the alibration urve to aount for a number of differenes between the alibration and verifiation measurements. A onstant, k, ombines all of the parameters that may differ between detetion systems into one alibration onstant: k = k 0 k 1 k 2 k 3 k 4 k 5, [2-7] where k 0 orrets for the soure yield and deay, k 1 orrets for hanges in the eletronis, k 2 orrets for detetor effiieny, k 3 orrets for burnable poisons, k 4 orrets for uranium mass, and k 5 orrets for any other fators that alter the detetor response [8] Detetor response Example alibration urves in thermal and fast mode (with a admium liner), from the UNCL alibration manual [8], are shown in Figures 2-7 and 2-8. Figure 2-7: Example of a thermal alibration urve produed for the UNCL. Figure 2-8: Example of a fast alibration urve produed for the UNCL. 31

32 Neutrons enter the assembly, and most pass through without interating. These neutrons are then thermalized within the polyethylene panels surrounding the assembly, after whih they reenter the assembly with a low-enough energy to indue fission within the fuel. The indued-fission neutrons then either follow the same progression or thermalize and enter the 3 He tubes and are deteted. A Deming regression is fit to the data and used as the alibration funtion. This form was hosen for the urve primarily beause it provides a good indiation of the linearity of the urve. If b is very small, the urve is loser to linear and the effet is less beause of thermal neutrons. This urve shape applies only to the range of linear density measurements that were taken and does not apply to signifiantly higher linear densities Deming regression A Deming regression is a form of least-squares analysis used to fit a straight line to twodimensional data for whih both variables, whih in this ase are the oinidene rate and the linear density of 235 U, are measured with error [22]. If two measurements are taken in whih x i = X i + ε i [2-8] and y i = Y i + δ i, [2-9] then ε i and δ i are the errors in x i and y i, and an equation of the form Y i = β 0 + β 1 X i [2-10] an be used to find X i and Y i, the true values of X i and Y i. In this equation, β 0 is the interept and β 1 is the slope [23]. The INCC software has a speial program, alled the DEMING program, that relies on this tehnique for urve fitting when fitting alibration urves [24]. The urves generated for the UNCL are of the form D = am 1+bm, [2-11] where D is the doubles rate [ounts/s], m is the linear density of 235 U [g 235 U/m], and a and b are alibration onstants. This formula was derived from empirial data Cf soures Californium-252 has been proposed as an alternative to replae 241 AmLi in ative NDA systems. As disussed in the introdution, 241 AmLi was originally hosen as the interrogation soure for ative systems for several reasons: (1) it has a low average neutron energy (0.3 MeV), meaning that it does not indue fast fission in 238 U, as shown in Figures 4a and 4b; (2) it does not emit orrelated neutrons and therefore ontributes no oinidene bakground to the measurement; and (3) it has 32

33 a long half-life (432 years), meaning that it requires a small normalization fator for soure deay thus, deay ontributes little error to the measurement. Almost all harateristis of 252 Cf soures are ontrary to those of 241 AmLi soures: (1) they have a high average neutron energy (2.14 MeV), whih means they indue fast fission in 238 U (see Figure 2-9); (2) they are spontaneous fission soures and therefore ontribute a doubles bakground that must be subtrated; and (3) they have a short half-life (2.65 years) and therefore require a large normalization fator to aount for soure deay. However, a study by Menlove [11] provided the first evidene that 252 Cf soures might atually provide an improvement in ounting statistis over 241AmLi for the UNCL appliation. This result was affirmed again in later studies [12, 25]. 252 Cf (2.14 MeV) 241 AmLi (0.3 Mev) 238 U fission 235 U fission Figure 2-9: Fission ross setions for 235 U and 238 U and omparison with interrogation neutron energies [10] TCIF The TCIF method was developed as a way to substitute a spontaneous fission soure, suh as 252Cf, in plae of the traditional 241 AmLi soure in an ative detetor [11]. A spontaneous fission soure emits multiple orrelated neutrons, whih an either result in indued fission events within the sample or be diretly aptured within the detetor. Beause the soure neutrons are time orrelated, the indued fission events spawned from those soure neutrons are also time orrelated. Thus, the neutron leakage from the sample (and orrespondingly the deteted signal) is omposed of both spontaneous fission neutrons from the soure and indued fission neutrons from the sample, all of whih are deteted within the same gate. The required time gate is twie as long 33

34 for a 252 Cf interrogation soure as for an 241 AmLi interrogation soure. The result is an effetively higher average number of neutrons produed within the sample (ν ) than is ahieved with a soure that emits single neutrons at random, suh as 241 AmLi. This proess is illustrated in Figure t 1 t 3 t 4 G Cf t 2 t 1 t 2 t 3 t 4 Time [μs] Figure 2-10: Illustration of time-orrelated, indued fission with a 252 Cf soure. In theory, under ideal onditions, this higher-effetive ν would simply be 5.12, whih is the sum of the individual ν values of 252 Cf (ν = 3.757) and 235 U (ν = 2.44), with one neutron subtrated for indued fission within the 235 U [11]. However, beause an assembly is a bulk material and many ompeting interations are taking plae within the fuel, the value of ν is lower than this ideal value. One of the goals of this thesis is to gain a better understanding of the proesses involved in TCIF Measurement methodology for TCIF TCIF measurements must aount for passive bakground ( 238 U spontaneous fission and room), as well as eah of the forms of bakground inherent to the 252 Cf soure, whih inlude soure ( 252 Cf interrogation soure) and fast ( 238 U fast fission). As with 241 AmLi measurements, 252 Cf measurements must begin with a short ( seond) room bakground measurement. Next a passive measurement (i.e. a measurement without an interrogation soure) of the fuel assembly by itself is onduted to quantify the passive 238 U spontaneous fission, followed by a measurement of the 252 Cf soure in the soure holder with no fuel assembly. When these bakground measurements are omplete, the ative measurements, in whih both the fuel assembly and the interrogation soure are present, an be onduted. The net doubles rate is obtained by subtrating the results of the bakground measurements as shown below [12]. D net = D ative D passive D soure D fast D room [2-14] 34

35 The fast fission omponent is assumed to be indiretly aounted for in the soure measurement, whih overestimates the soure ontribution to the ative measurement due to absorption of soure neutrons within the fuel assembly. 2.3 Past TCIF work Beause of the neessity of the transition from 241 AmLi to 252 Cf in many different detetors, several studies have been onduted to analyze the TCIF method. The first study was onduted for the UNCL-II [11], and the two subsequent studies were onduted for the Passive Neutron Enrihment Meter (PNEM) [12] and the Advaned Experimental Fuel Counter (AEFC) [25] UNCL The first proof-of-priniple measurements for the TCIF method were performed by Menlove [11] using the UNCL-II. This experiment was the basis for the format of the measurements taken for this thesis. The initial measurements relied on a 252 Cf soure, whih was used to irradiate a fuel assembly. By varying the number of enrihed (3.19% 235 U) and depleted (0.219% 235 U) rods, the average enrihment of the entire fuel assembly was gradually inreased. In eah onfiguration, the resulting fission events were measured, and the measurements demonstrated an enhanement in the doubles and triples rates ompared to 241 AmLi. Aording to Menlove, the average number of neutrons emitted per fission (ν ) when the neutrons emitted from the 252 Cf soure (3.76/fission) are ombined with those emitted from the sample itself (2.44/fission), resulting in a higher effetive ν of ~5.2 [11]. The experiment demonstrated that 252 Cf provides several advantages over the traditional 241 AmLi soure. These advantages inluded a redution in the statistial error of the measured ount rates, a redution in the statistial error in the unfolded 235 U enrihment, a redution in the measurement time, a possibility of eliminating the need for a passive measurement, and inreased safety of the 252Cf soure [11]. Unfortunately, the 252 Cf soure must be replaed more frequently beause of its short half-life, and these soures introdue a large doubles bakground, whih espeially interferes in the measurement of small samples PNEM Monte Carlo simulations were performed by Miller [12] to analyze the performane of a detetor that operates passively by design (the PNEM) in ative mode using both 241 AmLi and 252 Cf as the interrogation soure. The PNEM is designed to measure the mass and enrihment of 30B and 48Y UF 6 ylinders in enrihment failities. As with the UNCL, the simulations indiated positive results. Miller analyzed the improvement in ounting statistis provided by the 252 Cf soure and onluded that the ount time required to get to 1% unertainty is nearly three times less when 252 Cf is used as the interrogation soure than when 241 AmLi is used [12]. Miller also worked to optimize the soure loation within the moderator, and 252 Cf soure strength, and finally projeted the ounting statistis onto the unertainty in the enrihment measurement. 35

36 2.3.3 AEFC The AEFC was developed at LANL and designed to measure underwater researh reator spent fuel assemblies; it onsists of ative and passive neutron oinidene ounting equipment, as well as an ion hamber to ount gamma rays [25]. The AEFC is a tool used to verify residual fissile mass in researh reator spent fuel and analyzes burnup, ooling time, and initial enrihment [25]. In a reent field trial, a 252 Cf soure was used as the interrogation soure for this system. This field trial demonstrated that 252 Cf was a viable substitute for 241 AmLi as an ative interrogation soure [25]. 36

37 3 Experimental setup 3.1 System setup and operating parameters The experimental measurements were onduted at LANL in January 2014 in Tehnial Area 35, Building 2, Room C157, using the Anteh N2071 Neutron Coinidene Collar and the LANL PWR mokup fuel assembly. The data were olleted using the INCC software, version [21], installed on a Dell laptop with a Canberra JSR-15 Handheld Multipliity Register [26]. The UNCL was plaed on a art, enlosing the PWR mokup fuel assembly. A predelay of 1.5 μs, die-away time of 50 μs, and gate width of 128 μs were hosen for the measurement based on past experiments [11]. The UNCL operates at 1680 V. Photographs of the experimental setup are shown in Figures 3-1, 3-2, and 3-3. Juntion box LANL PWR mokup fuel assembly UNCL passive panel Interrogation soure hole UNCL ative panel Figure 3-1: Photograph of UNCL surrounding the LANL PWR fuel assembly, with ative panel faing the amera. The UNCL, with a weight of 38 kg, has outer dimensions of 571 mm 432 mm 432 mm in the PWR onfiguration, and the hamber inner dimensions are 235 mm 235 mm. The detetor operates between 5 C and 40 C [27]. The 3 He tubes are 38 m long, with an ative length of 33 m and a diameter of 25 mm, and operate at a pressure of 4 atm [8]. Full mehanial drawings of the detetor are provided in Appendix A. 37

38 Dell laptop Canberra JSR- 15 multipliity register Figure 3-2: Photograph of Dell laptop onneted to Canberra JSR-15 Handheld Multipliity Register. The 241 AmLi soure was ontained within a tungsten holder, and the 252 Cf soure was ontained within a stainless-steel apsule that was subsequently plaed inside an HDPE holder that was the same size as the tungsten holder. Photographs of both soure holders are shown in Figures 3-3a and 3-3b. HDPE Connetor 241 AmLi soure in tungsten holder 252 Cf soure in HDPE holder Figure 3-3: (a) Photograph of 252 Cf soure ontained within HDPE holder, with top HDPE onnetor attahed. (b) Photograph of 241 AmLi soure ontained within tungsten holder. 38

39 The soures used in the experiment and their harateristis are listed in Table 3-1. Eah soure will be referred to by the name listed in this table as it is disussed throughout this doument. The soure strengths listed are as of January 9, Table 3-1: 241 AmLi and 252 Cf soures used in experiment Soure Type Name Neutron emission rate 1/9/2014 [n/s] 241AmLi MRC , Cf Cf-12 45, Cf A , Cf A , Eletronis and software The UNCL relies on three Amptek A-lll amplifiers and an externally mounted N1081 De- Randomizing Mixer Buffer Counter, whih provides high voltage, low voltage, and signal onnetions for eah detetor module and, aording to the detetor speifiations, eliminates ounting losses and redues dead time, espeially at high ount rates [27]. The power and amplifier speifiations and detetion effiienies are listed in Table 3-2. Table 3-2: Eletronis speifiations [27] Eletronis speifiations Power to juntion boxes/amplifiers Maximum urrent Maximum HV LV A HV V Detetion effiieny PWR passive, no admium 13% PWR ative, admium 10% Output pulse width Period between onseutive pulses 50 ns (±5) ns 500 ns Measurements were onduted using the INCC program [21], whih is the standard ode used for the NDA of nulear materials. 3.3 Fuel assembly harateristis The LANL PWR mokup fuel assembly onsists of a stainless-steel frame, whih ontains 21 stainless-steel guide tubes, and an hold up to 204 fuel rods. The fuel rod harateristis are outlined in Table 3-3 [28]. In addition to the harateristis listed in Table 3-3, the DU rods and LEU rods were assumed to have a mass of g 235 U/m and g 235 U/m, respetively. 39

40 Table 3-3: PWR mokup fuel harateristis [28] PWR Assembly Lattie geometry Assembly width 21.5 m Fuel pin pith 1.4 m Number of fuel pin slots 204 Number of guide tube slots 21 Fuel Pin Information Fuel type UO 2 Cladding type Ziraloy-2 Average LEU rod enrihment 3.19% 235 U Average DU rod enrihment 0.219% 235 U Fuel pellet density g/m 3 Fuel pellet radius m Cladding thikness m Outer pin radius 0.54 m Total fuel rod length 130 m Ative fuel length LEU rod 102 m DU rod 120 m Inert fuel regions LEU Top 17 m Bottom 12 m Top Bottom DU 6 m 5 m 3.4 Experimental proedures Charaterization measurements 1. Measure pith/spaing/other dimensions of Anteh ollar to ensure math with mehanial drawings. 2. Measure high voltage plateau. 3. Bakground measurement for 200 s. 4. Effiieny measurements. - Condut measurements of Cf-12 along vertial axis in 5 m inrements. - Condut measurements along horizontal axis in 2 m inrements by tying soure Cf- 12 to a string attahed to a meter stik (as shown in Figure 3-4). 40

41 - Eah measurement was made for 30 s. - The enter of the measurement avity was assumed to be the enter as measured from the enter of the polyethylene housing for the detetors beause the detetors themselves were not visible during the experiment. Figure 3-4: Looking down at soure in enter of UNCL for effiieny measurements. 5. Measure die-away time using multi-gate tehnique using soure Cf Three measurement onfigurations were used: i. 252 Cf soure in the interrogation soure holder. ii. 252 Cf soure in interrogation soure holder with 2% enrihed fuel assembly in the UNCL. iii. 252 Cf soure in interrogation soure holder with 0.219% enrihed fuel assembly in the UNCL. - Eah measurement was made twie for 1000 s eah (1000 s 2 yles). - Measurements were made with gate lengths of 32, 64, 128, and 256 μs. 6. Measure deadtime oeffiients using two soure method [19]. - Two 252 Cf soures were used for the experiment (soure A7-867 and soure A7-868). A brass dummy soure was used to fill the spae of the seond soure when only one soure was used in the measurement. - Three measurements were made, eah for 1000 s (1000 s 1 yle): i. A Dummy ii. A A7-868 iii. Dummy + A Soures were taped together, as shown in Figure

42 Figure 3-5: Deadtime measurement setup. Two soures were taped together and hung in the enter of the detetor Calibration urve and penetrability To gain an understanding of the effet of differing fuel onfigurations on the detetor response, we performed alibration measurements and measurements to quantify the penetrability of interrogation soure neutrons into the fuel. The alibration urve is used to alibrate eah UNCL to a base assembly (whih has historially been the LANL PWR assembly), whih was used for these experiments. It was essential to have an understanding of the alibration urve shape with 252 Cf as ompared with 241 AmLi to learn how the performane of the detetor would differ in real measurement senarios. The first step, before starting the measurements, was to develop the fuel assembly layouts to be used in the experiment. For eah stage of the experiment, the layouts needed to be repliable and quikly ahievable, and they needed to adhere to the requirements of that partiular stage of the experiment. Using these layouts, we performed the measurements aording to a slightly different proedure for eah measurement type Calibration urve The first ategory of fuel assembly layouts was the alibration urve, in whih we gradually inreased the average enrihment of the assembly by inreasing the number of enrihed rods 42

43 within the assembly. We started with 204 DU (0.0219% 235 U) rods in the assembly and inreased the enrihment in inrements of 20 enrihed rods (LEU, 3.19% 235 U). The goal in the development of the layouts was to inrease the enrihment as uniformly as possible so that the results approximated those that would be measured if the enrihment of all of the rods in the assembly ould be inreased inrementally. The differene in measurement results using this approah, rather than inreasing the enrihment of all of the rods for eah measurement, was studied using MCNPX; this effet is disussed in Setion 5.2. The layouts are inluded in Figures 3-6 to Front in the diagrams is the side of the fuel assembly losest to the soure panel. Figure 3-6: All 204 rods are DU; no rods are LEU. Figure 3-7: 184 rods are DU; 20 rods are LEU. Figure 3-8: 164 rods are DU; 40 rods are LEU. Figure 3-9: 144 rods are DU; 60 rods are LEU. 43

44 Figure 3-10: 124 rods are DU; 80 rods are LEU. Figure 3-11: 104 rods are DU; 100 rods are LEU. Figure 3-12: 84 rods are DU; 120 rods are LEU. To assess the alibration urve, we worked aording to the following proedure: 1. Start with lowest enrihment (204 rods of DU filling the assembly, 0.219% average enrihment) and inrease enrihment uniformly in steps of 20 rods eah until maximum enrihment (120 LEU rods, 84 DU rods, ~2% average enrihment) is reahed, to measure ount rates vs linear density of 235 U [g/m]. 2. For eah onfiguration, omplete two measurements of 1000 s eah for both 252 Cf (Cf-12) and 241 AmLi Penetrability into assembly using 252 Cf ompared with 241 AmLi To gain a better understanding of the penetrability of the 252 Cf soure ompared with the 241 AmLi soure, we onduted measurements of the detetor performane when various regions of fuel were enrihed and others were not. This study was meant to give an idea of the regions of the assembly to whih the detetor is most sensitive and to show the regions in whih detetion of diversion might be less likely. The penetrability measurements were done two different ways. First, two-thirds of the assembly was filled with enrihed rods, and the remaining spae was filled with depleted rods in eah of 44

45 three regions (bak, enter, and front), where again front was the side of the assembly losest to the interrogation soure. This simple analysis provided information about the detetor s response to material lose to the soure vs that far from the soure. For the first type of measurement, the enrihment was kept onstant at its highest possible value: 1.97% for all of the measurements. The layouts for the first type of penetrability measurements are provided in Figures 3-13 to Figure 3-13: Penetrability measurement with depleted rods in the bak of the assembly. Figure 3-14: Penetrability measurement with depleted rods in the front of the assembly. Figure 3-15: Penetrability measurement with depleted rods in the enter of the assembly. The seond type of penetrability measurement was made to evaluate the response of the detetor to areas of depleted material in different regions, whih inluded the orners in the bak of the assembly, the orners in the front of the assembly, the enter of the assembly, and the enter portions of both the bak and front of the assembly. In this ase, the enrihment was kept onstant at 1.97% for all of the measurements. The fuel layouts for the seond type of measurement are provided in Figures 3-16 to

46 Figure 3-16: Penetrability measurement with depleted rods in bak orners of assembly. Figure 3-17: Penetrability measurement with depleted rods in front orners of assembly. Figure 3-18: Penetrability measurement with depleted rods in enter of assembly. Figure 3-19: Penetrability measurement with depleted rods in enter of front and bak of assembly. To assess the penetrability of interrogation soure neutrons into the assembly, we worked aording to the following proedure, using only the 252 Cf soure: 1. By row: Measure the response of the detetor to 84 rods of DU and 120 rods of LEU at a time for 1000 s eah. i. Bak ii. Front iii. Center 2. By region: Measure the response of the detetor to 4 different arrangements of 104 rods of DU and 100 rods of LEU for 1000 s eah. i. Bak orners ii. Front orners iii. Center front and bak iv. Center 46

47 4 Detetor haraterization and Monte Carlo benhmarking measurements 4.1 Motivation and limitations of experiments and Monte Carlo Laboratory measurements and Monte Carlo simulations were used to define the operating parameters of the UNCL, and to better understand the TCIF method. Although the UNCL has been well-haraterized experimentally and has been alibrated and used extensively for safeguards measurements when a 241 AmLi soure was used as the interrogation soure, its response to a 252 Cf soure has been demonstrated only with limited proof-of-priniple measurements. This setion desribes UNCL I measurements performed to haraterize the system and benhmark the MCNPX model. The experimental measurements were onduted at LANL using the Anteh N2071 Neutron Coinidene Collar and the LANL PWR mokup fuel assembly. Experimental measurements provide a more aurate demonstration of the performane of a given detetor than do simulated measurements. Unfortunately, experimental measurements also involve a muh greater time ommitment (eah experimental measurement in this experiment required roughly 2000 s, whereas the omputational time required for eah measurement in this experiment was ~30 s), require experimenters to inrease their exposure to radioative materials, and restrit measurements to those permitted by the materials on hand. Currently, LANL has restrited the number of LEU (3.19% 235 U) fuel rods that may be used in an experiment to 121 out of a total of 204 rods in the mokup fuel assembly. The remaining rods are replaed with DU (0.219% 235 U). Thus, the maximum ahievable enrihment during an experiment is ~2.0%, whih is signifiantly lower than the enrihments that would be enountered in a ommerial fuel fabriation faility, whih are typially on the order of 3% 5% 235 U. Therefore, simulated experiments were a neessary supplement to the experimental measurements. 4.2 Detetor haraterization High voltage The first step in the experimental proedure was measuring the high-voltage plateau. The operating voltage was found to be 1680 V, or 40 V above the knee, as shown in Figure 4-1. This operating voltage minimizes Geiger disharges, while maintaining ounting stability within the proportional region. 47

48 Singles rate [ps] 4,500 4,000 3,500 3,000 2,500 2,000 1,500 1, Operating voltage (1680 V) High Voltage [V] Figure 4-1: Detetor voltage plateau showing operating voltage of 1680 V Room bakground A bakground measurement with no fuel or soure was taken and resulted in a singles rate of (±0.358) ounts/s and a doubles rate of (±0.013) ounts/s, with a ount time of 200 s. With a long-enough measurement, the error in this bakground measurement will effetively be zero Effiieny profiles To measure the effiieny of the detetor along the x-, y-, and z- axes, measurements of the detetor response to a soure (Cf-12) in various positions throughout the detetor were onduted. The detetor response was mapped in the vertial diretion and in both horizontal diretions (front to bak and left to right). The results are tabulated in Tables 4-1, 4-2, and 4-3 and displayed in the orresponding Figures 4-2, 4-3, and 4-4. The 252 Cf effiieny at the enter of the measurement avity is 9.7%. 48

49 Table 4-1: Singles [ts/s] measured horizontally from front to bak in the detetor (bak is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor. Depth [m] Singles [ps] σ S [ps] Singles rate [ps] Distane from enter [m] Figure 4-2: Singles [ts/s] measured horizontally from front to bak in the detetor (bak is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor. It is lear from the above results that the effiieny inreases with inreasing distane from the soure panel. This dedution is logial beause a solid angle for detetion is greater at the bak of the detetor than near the soure panel, whih ontains no detetors. At the bak of the detetor, the soure is surrounded on three sides by detetors, whereas at the front, only two detetors are nearby. 49

50 Table 4-2: Singles [ts/s] measured horizontally from left to right in the detetor (right is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor. Depth [m] Singles [ps] σ S [ps] Singles rate [ps] Distane from enter [m] Figure 4-3: Singles [ts/s] measured horizontally from left to right in the detetor (right is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor. As expeted, the ount rates are higher near the sides of the sample spae, whih is nearer to the detetors than in the enter of the sample spae. It appears that the effiieny on either side of the detetor is approximately equivalent; that is, the detetor response is symmetri. 50

51 Table 4-3: Singles [ts/s] measured horizontally from left to right in the detetor (right is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor. Depth [m] Singles [ps] σ S [ps] Singles rate [ps] Distane from enter [m] Figure 4-4: Singles [ts/s] measured vertially from top to bottom in the detetor (top is the positive diretion) in 2-m inrements to show dependeny of effiieny on soure position within detetor. The effiieny was greatest at a point 5 m lower than the enter of the sample spae, whih is likely beause the enter was assumed to be the enter of the polyethylene panels and not the 51

52 enter of the ative length of the 3 He tubes. The enter of the ative region of the 3 He tubes is lower than that of the panels themselves by ~5 m, so this result is logial Die-away time The die-away time was measured (1) without rods, (2) with only depleted rods (0.219% enrihed), and (3) with the maximum possible enrihment (~2%), with eah permutation using four gate-lengths: 32 μs, 64 μs, 128 μs, and 256 μs. To determine the die-away time, an equation of the form below was used [19]: τ = G 1 ln[ R 2 R1 1]. [4-1] In this equation, τ is the die-away time, G 1 is the shorter gate length in seonds, and R 1 and R 2 are the doubles ount rates for the shorter and longer gate lengths, respetively. To alulate the error in the die-away times, the standard form of propagation of error was used: σ f = [( δf δx ) σ x] 2 + [( δf δy ) σ y] 2 + [( δf δz ) σ z] 2 +. [4-2] The equation above was evaluated using the following omponents: δτ = G G2 2[ln( G1 1)+1] G 1[ln( G 2 G1 1)] [4-3] δg 1 (G 1 G 2 )[ln( G 2 G1 1)]2 δτ δg 2 = G 1 (G 1 G 2 )[ln( G 2 G1 1)]2 [4-4] The results of the die-away time measurements and alulations are presented in Table 4-4 and Figure

53 Table 4-4: Die-away time measurements Gate Length [μs] Measurement Doubles [ps] σ D [ps] Die-away time [μs] σ [μs] 32 No Rods N/A N/A 64 No Rods No Rods No Rods % Enrihed N/A N/A % Enrihed % Enrihed % Enrihed % Enrihed N/A N/A % Enrihed % Enrihed % Enrihed Die-Away time (µs) No rods Enrihed Rods Depleted Rods Gate Interval (µs) Figure 4-5: Calulated die-away times for the ases of no rods, depleted rods, and enrihed rods filling the sample spae. Several important onlusions an be drawn from the alulated die-away times. First, note that the die-away time for the No Rods ase is traditionally used to determine the gate length. In this ase, the die-away time exeeds the gate length in the 64-μs gate length ase, and therefore a longer gate must be hosen. In the 256-μs gate length ase, the die-away time is far shorter than the gate 53

54 length. Therefore, a gate length of 128 μs was hosen for the experiments. This gate length is twie as long as that used when 241 AmLi is the hosen interrogation soure. As illustrated in Figure 2-12, with a orrelated interrogation soure, the gate trigger event may ome diretly from the 252 Cf soure while the other neutrons from the same spontaneous fission event indue fission in the fuel assembly. The neutrons from these indued fission events may then subsequently be deteted within the same gate, resulting in a boosted multipliity. This is ompared to the 241 AmLi soure ase illustrated in Figure 1-4, where the trigger event for a real oinidene only depends on the time onstant for indued fission events inside the assembly. It is interesting to note the differene in shape between the die-away-time urves for the ases with no rods and with only DU rods, as ompared with that for an assembly ontaining LEU rods. At the hosen gate length of 128 μs, the die-away time is ~25% higher than for the other two ases. This effet is important beause die-away-time alibration measurements are typially made with a soure by itself. It is also lear that with the addition of and inreasing enrihment of material in the sample spae, the separation between die-away times measured using the two lowest gate lengths inreases and that measured using the largest gate length dereases Deadtime parameters Deadtime refers to an amount of time in whih the detetor is unable to detet a ount that ours too soon after another ount. Thus, for example, two pulses that our lose together an be registered as a single pulse. Deadtime an our for many different reasons and is worsened by high ount rates, whih an lead to iruit paralysis, overlap between ounts in the ounting eletronis, or event overlap within the detetor itself. The UNCL does not operate at high ount rates and therefore suffers from few deadtime effets. Furthermore, the shift register is deadtime-free [19], so all of the deadtime effets our within the detetor itself. The lost ounts due to deadtime mean that the singles and doubles ount rates are redued from their atual values. However, deadtime an be orreted for. The deadtime parameters were measured using the two-soure method. The standard proedure for aounting for deadtime is traditionally to use the following equations to orret the measured ount rates [19]. Strong 252 Cf and 241 AmLi soures were used to derive these equations experimentally [19]: where S 0 = S m e δs m 4 [4-5] and D 0 = D m e δs m, [4-6] δ = A + BS m. [4-7] 54

55 In these equations, A and B are the deadtime oeffiients that are entered into the software to orret for deadtime, S m and D m are the measured singles and doubles ount rates, and S 0 and D 0 are the deadtime orreted singles and doubles ount rates. The results of the three measurements used to determine the deadtime are listed in Table 4-5. Table 4-5: Deadtime measurements: two-soure method Measurement Singles [ps] σ S [ps] Doubles [ps] σ D [ps] A Dummy A A Dummy + A If the measurements for A7-868 and A7-867 are summed, they should add up to that for A A7-867, assuming no deadtime. A spreadsheet was used to iterate the values of δ, A, and B until the expeted and measured values for the Singles rates for A7-868 and A7-867 mathed. This analysis resulted in a value for oeffiient A of E-05 s and a value for oeffiient B of e-11 s Summary A summary of the detetor parameters is shown in Table 4-6. Table 4-6: Summary of detetor parameters Parameter Value HV 1680 V Effiieny 9.7% Die-Away Time 65.1 μs Gate length 128 μs Deadtime parameter A 2.34E-06 s Deadtime parameter B 1.37E-12 s Benhmark measurements The benhmark relied on three sets of measurements: (1) plaing a soure in the enter of the detetor and in the interrogation soure hole, (2) mapping the effiieny of the detetor in all three diretions (from the haraterization measurements), and (3) uniformly inreasing fuel enrihment with a onfigurable fresh fuel assembly while either interrogating the assembly with 252 Cf or 241AmLi. Comparing the simulated die-away time to the measured die-away time by plaing the 252Cf soure in the interrogation soure hole with no sample (also from the haraterization measurements) was also done. The experimental benhmark was onduted in several parts to ensure that the MCNPX model agreed fully with the experimental results. 55

56 4.3.1 Soure in enter of detetor and soure in interrogation soure hole No additional measurement time was neessary to ondut a measurement of a soure in the enter of the detetor (see Table 4-7) beause we had already onduted this measurement for the deadtime measurements. The dummy + A7-867 measurement was used to benhmark this measurement ase. Table 4-7: Soure in enter of detetor Measurement Singles [ps] σ S [ps] Doubles [ps] σ D [ps] Dummy + A As for the soure in the enter of the detetor, to benhmark the soure in the interrogation soure hole, no additional measurement time was neessary beause we had already onduted this measurement while measuring the die-away time. This value was taken to be the value measured for the No Rods ase with a 128-μs gate, as shown in Table 4-8. Table 4-8: Soure in interrogation soure hole Gate Length [μs] Measurement Singles [ps] σ S [ps] Doubles [ps] σ D [ps] 128 No Rods Calibration urve The alibration urve is the primary tool used in verifiation proedures for the UNCL. A urve is fit to data taken from a referene fuel assembly, whih has traditionally been the LANL PWR mokup assembly the fuel assembly used in these measurements. The profile generated experimentally needed to math the simulated profile as losely as possible so that any simulated measurements ould be assumed to be aurate refletions of reality. Therefore, these measurements were onduted for longer periods of time (a total of 2000 s) to ensure a smaller degree of unertainty than that whih was aeptable for the previous measurements. The raw data taken when a 241 AmLi soure was used as the interrogation soure, with only the room bakground subtrated (beause room bakground is not present in the simulation), is shown in Table

57 Table 4-9: Raw data for 241 AmLi alibration urve DU Rods Linear density [g 235U/m] Enrihment[% 235U] Singles [ps] σ S [ps] Doubles[ps] σ D [ps] The raw data taken when a 252 Cf soure was used as the interrogation soure, with only room bakground subtrated, is shown in Table

58 Table 4-10: Raw data for 252 Cf alibration urve DU Rods Linear density [g 235 U/m] Enrihment [% 235 U] Singles [ps] σ S [ps] Doubles [ps] σ D [ps] These results are plotted in Figures 4-6 and Singles [ps] AmLi Cf Linear density [g 235 U/m] Figure 4-6: Raw data for singles urves for both 241 AmLi and 252 Cf. A seond-degree polynomial is fit to the data. 58

59 Doubles [ps] AmLi Cf Linear density [g 235 U/m] Figure 4-7: Raw data for doubles alibration urves for both 241 AmLi and 252 Cf. A seond-degree polynomial is fit to the data. As expeted, the ount rates from 252 Cf were signifiantly higher than those from 241 AmLi. However, it is important to note that the soure strengths are not equal (the ratio of the 252 Cf soure strength to the 241 AmLi soure strength is 1.033), and that unlike those from 241 AmLi, the results from 252 Cf have signifiant bakground ontributions. Californium-252 provides two soures of doubles bakground (fast fission and the interrogation soure itself) that were not subtrated, and are not present with 241 AmLi: 238 U fast fission and from the interrogation soure itself. Note that two measurements per onfiguration are visible in the raw data, whih is why some points appear larger than others. Both sets of data are inluded to provide a visual representation of the satter in the measured data. 4.4 MCNPX simulations The goal of the benhmark was to ensure that the Monte Carlo simulations mathed the measurements. Development of the MCNPX model began with the dimensions listed on the mehanial drawings of the Anteh N2071 UNCL. These dimensions were then onfirmed during the experimental proedure. After the experimental measurements were omplete, the simulated detetor response to a soure with no fuel assembly was ompared with the measured signal at several points within the detetor. The soure was then modelled within the interrogation soure hole, and the simulated result was ompared with the measured result. In this ase, the simulated die-away time was also ompared with the experimental results. Finally, the fuel assembly was added to the model, and the results from the experimental alibration profile were ompared with the simulated profile. This proedure ensured that eah omponent (the detetor and the fuel assembly) was modelled orretly before proeeding to a more omplex model. The final model was then used for the detetor performane and unertainty analysis simulations. Sabrina [29] and 59

60 MCNPX models of the detetor with the fuel assembly in the sample spae are shown in Figures 4-8 to (a) (b) Figure 4-8: (a) Sabrina [29] model of UNCL with fuel assembly in enter. (b) MCNPX model of UNCL from side, with fuel assembly in enter. Figure 4-9: Sabrina utout of UNCL model. 60

61 Figure 4-10: MCNPX view of the detetor model from the top. The benhmark was inredibly important beause it was the foundation on whih the onlusions in the detetor performane and analysis hapter (Chapter 5) of this thesis relied; onsiderable time was spent to ensure that the model mathed the experimental results as losely as possible. Many fators play into the auray of a model, and no model an exatly desribe reality. Many soures of unertainty and approximations in an MCNP model an result in differenes between measurements and simulations. Many details, inluding the eletronis, wiring, stainless-steel frame for the fuel rods, and detetor art, were not inluded in the model. All of these details ontribute some degree of unertainty in the model. In addition, room effets; assumptions about material omposition and isotopis; material impurities; and unertainties in the nulear data, fuel assembly delaration, and fuel assembly positioning an redue the auray of the simulation. A basi version of an MCNPX input file used in simulations of the UNCL is inluded in Appendix B. Setions through inlude omparisons of measured and simulated results for 252 Cf soures in the interrogation soure slot, 252 Cf effiieny profiles, die-away time, and alibration urves for both 252 Cf and 241 AmLi. Setion inludes a parameter study to quantify potential soures of modeling unertainty for omparison Cf Effiieny The first and most important benhmarking metri is the effiieny of the detetor. To benhmark the soure in the enter of the measurement avity, 252 Cf soure A-867 was used beause of its high emission rate. The ount rates were taken from the deadtime oeffiient measurement and thus inluded a brass dummy soure. The dummy soure was modelled as a opper ylinder with the same dimensions as A Results showed good agreement between the measured and simulated response, with errors of <1% for both singles and doubles (see Table 4-11). 61