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Nigg, David W.

Publications and source records attributed to Nigg, David W..

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Experimental quantification of the background neutron flux in the Advanced Test Reactor fuel storage canal - 25562

The Advanced Test Reactor at the Idaho National Laboratory is one of only a few high-power, versatile, research reactors of its type in the world. The ATR fuel storage canal features a gamma irradiation facility that has various uses as well. It is composed of a simple dry tube that extends down into one of the fuel storage rack positions. This tube is surrounded by used fuel elements that can be selected from the inventory to produce the desired gamma radiation level and, to some extent, spectrum within the dry tube. The gamma field within the dry tube is accompanied by a small background neutron flux, produced primarily by photoneutron production in the small fraction of heavy water (D{sub 2}O) that is naturally present in light water. Neutron activation spectrometry was employed to quantify the background neutron flux in the dry tube in a configuration where the gamma exposure rate was approximately 5x10{sup 6} Roentgens per hour. The measurement technique used is very sensitive and provided self-consistent results in terms of raw activation data quality. The background neutron fluxes in the gamma facility were very low, but still detectable. The total neutron flux in the energy region above 0.414 eV and the thermal neutron flux (below 0.414 eV) were estimated from the activation measurements to be 7.5x10{sup 2} and 1.0x10{sup 2} n/cm{sup 2}-s, respectively. High uncertainties (factors of approximately 20-30% in the above-thermal range and 100% in the thermal range) apply to these results. These high uncertainties are primarily due to the limited number of activation interactions that could be used to detect such small fluxes and the resulting propagation of uncertainties associated with the various input parameters involved in the spectral unfolding process. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Experimental Quantification of the Background Neutron Flux in the Advanced Test Reactor Fuel Storage Canal

The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of only a few high-power research reactors (HPRR) of its type in the world, with a variety of missions involving accelerated testing of nuclear fuel and other materials in a very high neutron flux (5.0x1014 n/cm2 -s) environment, medical and industrial isotope production, and other applications. It typically runs several cycles per year with lengths varying from 40-60 days depending on specific experimental requirements for each cycle. The highly radioactive used fuel elements that are discharged from the ATR after each cycle are stored in an adjacent canal (Figure 1) and can be used for as many as three cycles before being retired. The ATR canal features a gamma irradiation facility that has various uses. It is composed of a simple dry tube that extends down into one of the fuel storage rack positions. This tube is surrounded by used fuel elements that can be selected from the inventory to produce the desired gamma radiation level and, to some extent, spectrum within the dry tube. Typical gamma fields that can be produced in this manner can be as high as several megarads per hour. The gamma field within the dry tube is accompanied by a small background neutron flux, produced by photoneutron production in the small fraction of heavy water (D2O) that is naturally present in light water at a concentration of approximately 0.015% by weight, and by a very low level of photofission and spontaneous fission in the fuel elements. The measurement technique used is very sensitive and provided self-consistent results in terms of raw activation data quality. The background neutron fluxes in the gamma facility are very low, but still detectable. The total neutron flux in the energy region from 0.414 eV to 1 MeV and the thermal neutron flux (below 0.414 eV) were estimated from the activation measurements to be 1.0x104 and 2.0x102 n/cm2 -s, respectively. ncertainties of approximately an order of magnitude apply to both results. No neutrons above about 1 MeV were detected. The high uncertainties quoted here are primarily due to uncertainties in the a priori neutron fluxes computed by MCNP6 as well as uncertainties in the spectrum-averaged foil activation cross sections used in the few-group least-squares flux adjustment procedure that was applied. Detailed analyses are currently underway with the aim of reducing these sources of uncertainty.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Extension of the Advanced Test Reactor Operating Envelope Via Enhanced Reactor Physics Validation Techniques

The Korea Atomic Energy Research Institute is currently in the process of qualifying a low-enriched-uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated in this paper for several operating cycles in the northeast flux trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618 g 235 U) in comparison with historical ATR experiment standards (<1 g 235 U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be well outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly, it was necessary to conduct an extensive set of new low-power physics measurements in the ATR Critical Facility (ATRC), a companion facility to the ATR, located in an immediately adjacent building and sharing the same fuel storage canal. The new measurements included fission power distributions, reactivity, and measurements related to the calibration of the in-core online instrumentation. The effort was focused on the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results have demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well understood in terms of its general effects on ATRC core reactivity and fission power distributions and its effects on the calibration of the ATR Lobe Power Calculation and Indication System, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Finally, taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Validation and Comparison of the Therapeutic Efficacy of Boron Neutron Capture Therapy Mediated By Boron-Rich Liposomes in Multiple Murine Tumor Models

Boron neutron capture therapy (BNCT) was performed at the University of Missouri Research Reactor in mice bearing CT26 colon carcinoma flank tumors and the results were compared with previously performed studies with mice bearing EMT6 breast cancer flank tumors. We implanted mice with CT26 tumors subcutaneously in the caudal flank and were given two separate tail vein injections of unilamellar liposomes composed of cholesterol, 1,2-distearoyl-sn-glycer-3-phosphocholine, and K[nido-7-CH 3 (CH 2 ) 15 –7,8-C 2 B 9 H 11 ] in the lipid bilayer and encapsulated Na 3 [1-(2`-B 10 H 9 )-2-NH 3 B 10 H 8 ] within the liposomal core. Mice were irradiated 30 hours after the second injection in a thermal neutron beam for various lengths of time. The tumor size was monitored daily for 72 days. In spite of relatively lower tumor boron concentrations, as compared to EMT6 tumors, a 45 minute neutron irradiation BNCT resulted in complete resolution of the tumors in 50% of treated mice, 50% of which never recurred. Median time to tumor volume tripling was 38 days in BNCT treated mice, 17 days in neutron-irradiated mice given no boron compounds, and 4 days in untreated controls. Tumor response in mice with CT26 colon carcinoma was markedly more pronounced than in previous reports of mice with EMT6 tumors, a difference which increased with dose. The slope of the dose response curve of CT26 colon carcinoma tumors is 1.05 times tumor growth delay per Gy compared to 0.09 times tumor growth delay per Gy for EMT6 tumors, indicating that inherent radiosensitivity of tumors plays a role in boron neutron capture therapy and should be considered in the development of clinical applications of BNCT in animals and man.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2017

Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be well outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

An analysis method for experiment power in the ATR

The Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) is used for testing and qualification of nuclear fuels and materials. The ATR has been irradiating the lead test assembly (LTA) for the planned Ki-Jang Research Reactor (KJRR) in Korea. Irradiation and subsequent Post-Irradiation Examination (PIE) of the LTA will lead to qualification of the fuel to support licensing activities. Qualification of the LTA requires the experiment to meet specific thermal and burn-up conditions. In particular, plate 20 of the experiment must obtain a heat flux between 137 and 200 W/cm{sup 2} with a target U-235 depletion of 79-84%. Due to the limited instrumentation and burn-up monitoring capabilities, assurance of meeting experiment goals is verified through modelling. In order to validate the experiment model using ATR operating data, a methodology was developed to compare the calculated experiment power with ATR power monitoring instrumentation. This paper discusses the methodology used to quantify the experiment power and uncertainty in the model based on ATR operating conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Physics Validation Measurements for Highly-Reactive Experiment Packages in the INL Advanced Test Reactor

The Korea Atomic Energy Research Institute is currently in the process of qualifying the Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element will be irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Requalification and Experimental Validation of the Epithermal Neutron Beam Facility for Radiotherapy Research at Washington State University

Approximately ten years ago the Idaho National Laboratory (INL) and Washington State University (WSU) completed the construction of a new epithermal-neutron beam for collaborative preclinical Neutron Capture Therapy (NCT) research at the WSU TRIGATM research reactor facility. Since then the reactor core has been reconfigured to use Low-Enriched Uranium (LEU) fuel, and significant interest has emerged recently with respect to potential use of the facility for actual clinical (human) trials as well as for preclinical work. Both of these developments have necessitated a complete experimental revalidation of the neutronic performance of the beamline over the past 18 months. This article summarizes the results of the physics requalification effort.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2015

Gamma-Ray Simulated Spectrum Deconvolution of a LaBr₃ 1-in. x 1-in. Scintillator for Nondestructive ATR Fuel Burnup On-Site Predictions

A deconvolution method for a LaBr₃ 1"x1" detector for nondestructive Advanced Test Reactor (ATR) fuel burnup applications was developed. The method consisted of obtaining the detector response function, applying a deconvolution algorithm to 1”x1” LaBr₃ simulated, data along with evaluating the effects that deconvolution have on nondestructively determining ATR fuel burnup. The simulated response function of the detector was obtained using MCNPX as well with experimental data. The Maximum-Likelihood Expectation Maximization (MLEM) deconvolution algorithm was selected to enhance one-isotope source-simulated and fuel- simulated spectra. The final evaluation of the study consisted of measuring the performance of the fuel burnup calibration curve for the convoluted and deconvoluted cases. The methodology was developed in order to help design a reliable, high resolution, rugged and robust detection system for the ATR fuel canal capable of collecting high performance data for model validation, along with a system that can calculate burnup and using experimental scintillator detector data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Validation of HELIOS for ATR Core Follow Analyses

This work summarizes the validation analyses for the HELIOS code to support core design and safety assurance calculations of the Advanced Test Reactor (ATR). Past and current core safety assurance is performed by the PDQ-7 diffusion code; a state of the art reactor physics simulation tool from the nuclear industry’s earlier days. Over the past twenty years, improvements in computational speed have enabled the use of modern neutron transport methodologies to replace the role of diffusion theory for simulation of complex systems, such as the ATR. More exact methodologies have enabled a paradigm-shift away from highly tuned codes that force compliance with a bounding safety envelope, and towards codes regularly validated against routine measurements. To validate HELIOS, the 16 ATR operational cycles from late-2009 to present were modeled. The computed power distribution was compared against data collected by the ATR’s on-line power surveillance system. It was found that the ATR’s lobe-powers could be determined with ±10% accuracy. Also, the ATR’s cold startup shim configuration for each of these 16 cycles was estimated and compared against the reported critical position from the reactor log-book. HELIOS successfully predicted criticality within the tolerance set by the ATR startup procedure for 13 out of the 16 cycles. This is compared to 12 times for PDQ (without empirical adjustment). These findings, as well as other insights discussed in this report, suggest that HELIOS is highly suited for replacing PDQ for core safety assurance of the ATR. Furthermore, a modern verification and validation framework has been established that allows reactor and fuel performance data to be computed with a known degree of accuracy and stated uncertainty.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Therapeutic efficacy of boron neutron capture therapy mediated by boron-rich liposomes for oral cancer in the hamster cheek pouch model

Unilamellar liposomes formulated with an equimolar mixture of cholesterol and 1,2-distearoyl-sn-glycero-3-phosphocholine, incorporating K[nido-7-CH 3 (CH 2 )15-7,8-C 2 B 9 H 11 ] in the lipid bilayer, and encapsulating Na 3 [1-(2’-B 10 -H 9 )-2-NH 3 B 10 H 8 ] were prepared by probe sonication and investigated in vivo. Microwave assisted digestion followed by inductively coupled plasma-optical emission spectroscopy was utilized to determine the biodistribution of boron in various tissues following either a single tail vein injection or two identical injections (separated by 24 hours) of the liposomal suspension in BALB/c mice bearing EMT6 mammary adenocarcinomas in their right flank. Double-injection protocols resulted in a boron content in the tumor exceeding 50 µg of boron per gram of tissue for 48 to 72 hours subsequent to the initial injection while tumor:blood boron ratios were more ideal from 54 hours (1.9:1) to 96 hours (5.7:1) subsequent to the initial injection. Tumor bearing mice were given a double-injection of liposomes containing the 10 B-enriched analogs of the aforementioned agents and subjected to a 30 minute irradiation by thermal neutrons with a flux of 8.8 x 10 8 (±7%) neutrons/cm 2 s integrated over the energy range of 0.0 – 0.414 eV. Significant tumor response for a single BNCT treatment was demonstrated by growth curves versus a control group. Vastly diminished tumor growth was witnessed at 14 days (186% increase versus 1551% in controls) in mice that were given a second injection/radiation treatment 7 days after the first. Mice given a one hour neutron irradiation following the double-injection of liposomes had a similar response (169% increase at 14 days) suggesting that neutron fluence is the limiting factor towards BNCT efficacy in this study.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2014

Advanced Test Reactor Core Modeling Update Project: Annual Report for Fiscal Year 2014

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity transport theory based computational software and models to replace the legacy methods, with appropriate verification and validation (V&V) protocols according to applicable national standards. The project has now completed its fifth, and last, full year of development activity, and has entered an accelerated deployment phase that is expected to run through 2015.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Boron biodistribution for BNCT in the hamster cheek pouch oral cancer model: Combined administration of BSH and BPA

Sodium mercaptoundecahydro-closo-dodecaborate (BSH) is being investigated clinically for BNCT. We examined the biodistribution of BSH and BPA administered jointly in different proportions in the hamster cheek pouch oral cancer model. The 3 assayed protocols were non-toxic, and showed preferential tumor boron uptake versus precancerous and normal tissue and therapeutic tumor boron concentration values (70–85 ppm). All 3 protocols warrant assessment in BNCT studies to contribute to the knowledge of (BSH+BPA)-BNCT radiobiology for head and neck cancer and optimize therapeutic efficacy.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2013

Boron neutron capture therapy (BNCT) for liver metastasis in an experimental model: dose–response at five-week follow-up based on retrospective dose assessment in individual rats

Boron neutron capture therapy (BNCT) was proposed for untreatable colorectal liver metastases. Employing an experimental model of liver metastases in rats, we recently demonstrated that BNCT mediated by boronophenylalanine (BPA-BNCT) at 13 Gy prescribed to tumor is therapeutically useful at 3-week follow-up. The aim of the present study was to evaluate dose–response at 5-week follow-up, based on retrospective dose assessment in individual rats. BDIX rats were inoculated with syngeneic colon cancer cells DHD/K12/TRb. Tumor-bearing animals were divided into three groups: BPA-BNCT (n = 19), Beam only (n = 8) and Sham (n = 7) (matched manipulation, no treatment). For each rat, neutron flux was measured in situ and boron content was measured in a pre-irradiation blood sample for retrospective individual dose assessment. For statistical analysis (ANOVA), individual data for the BPA-BNCT group were pooled according to absorbed tumor dose, BPA-BNCT I: 4.5–8.9 Gy and BPA-BNCT II: 9.2–16 Gy. At 5 weeks post-irradiation, the tumor surface area post-treatment/pre-treatment ratio was 12.2 +/- 6.6 for Sham, 7.8 +/- 4.1 for Beam only, 4.4 +/- 5.6 for BPA-BNCT I and 0.45 +/- 0.20 for BPA-BNCT II; tumor nodule weight was 750 +/- 480 mg for Sham, 960 +/- 620 mg for Beam only, 380 +/- 720 mg for BPA-BNCT I and 7.3 +/- 5.9 mg for BPA-BNCT II. The BPA-BNCT II group exhibited statistically significant tumor control with no contributory liver toxicity. Potential threshold doses for tumor response and significant tumor control were established at 6.1 and 9.2 Gy, respectively.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2013

A Complex-Geometry Validation Experiment for Advanced Neutron Transport Codes

The Idaho National Laboratory (INL) has initiated a focused effort to upgrade legacy computational reactor physics software tools and protocols used for support of core fuel management and experiment management in the Advanced Test Reactor (ATR) and its companion critical facility (ATRC) at the INL.. This will be accomplished through the introduction of modern high-fidelity computational software and protocols, with appropriate new Verification and Validation (V&V) protocols, over the next 12-18 months. Stochastic and deterministic transport theory based reactor physics codes and nuclear data packages that support this effort include MCNP5[1], SCALE/KENO6[2], HELIOS[3], SCALE/NEWT[2], and ATTILA[4]. Furthermore, a capability for sensitivity analysis and uncertainty quantification based on the TSUNAMI[5] system has also been implemented. Finally, we are also evaluating the Serpent[6] and MC21[7] codes, as additional verification tools in the near term as well as for possible applications to full three-dimensional Monte Carlo based fuel management modeling in the longer term. On the experimental side, several new benchmark-quality code validation measurements based on neutron activation spectrometry have been conducted using the ATRC. Results for the first four experiments, focused on neutron spectrum measurements within the Northwest Large In-Pile Tube (NW LIPT) and in the core fuel elements surrounding the NW LIPT and the diametrically opposite Southeast IPT have been reported [8,9]. A fifth, very recent, experiment focused on detailed measurements of the element-to-element core power distribution is summarized here and examples of the use of the measured data for validation of corresponding MCNP5, HELIOS, NEWT, and Serpent computational models using modern least-square adjustment methods are provided.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

97 MATHEMATICS AND COMPUTING↗