Burnup Measurement Research Articles

Study on OTTO fueling schemes of the HTR-PM with boron burnable particles

The HTR-PM is a demonstration plant under construction with two 250MWth modules of pebble bed high temperature reactors (PB-HTR). In this work, the once-through-then-out (OTTO) fueling schemes with boron (10B) burnable particles are introduced into the HTR-PM design, which reduces the complexity of fuel handling system and totally eliminate the necessity of burn-up measurement. The parametric analysis upon two important parameters, i.e. the radius and the volumetric fraction of boron particles, is performed to optimize the OTTO schemes. In the lattice calculations, a boron burnable particle is modeled asa small sphere located at the center of a concentric spherical cell, surrounded by the effective fuel particle zone and fuel-free graphite zone with the same volumetric ratios as in the real fuel pebble. Generally, as the volumetric fraction of boron particles increases, the flattening effect of axial power profile is enhanced and the safety-related features become better, while the fuel enrichment required by the equilibrium core also increases. However, the most optimized schemes exist only within a small range of boron particle radius, i.e. around 120μm. Within this range, the requirement of margin of maximum fuel temperature after DLOFC accidents confines the boron particle volumetric fraction to be larger than 5.8×10−5 and the fuel enrichment larger than 13.5%. The mechanism of boron concentration and particle size influencing the neutronic features is also discussed. It is concluded that the OTTO scheme utilizing boron burnable particles is feasible for OTTO variants of PB-HTRs similar to the multi-pass HTR-PM.

Triton burnup measurements in KSTAR using a neutron activation system.

Measurements of the time-integrated triton burnup for deuterium plasma in Korea Superconducting Tokamak Advanced Research (KSTAR) have been performed following the simultaneous detection of the d-d and d-t neutrons. The d-d neutrons were measured using a 3He proportional counter, fission chamber, and activated indium sample, whereas the d-t neutrons were detected using activated silicon and copper samples. The triton burnup ratio from KSTAR discharges is found to be in the range 0.01%-0.50% depending on the plasma conditions. The measured burnup ratio is compared with the prompt loss fraction of tritons calculated with the Lorentz orbit code and the classical slowing-down time. The burnup ratio is found to increase as plasma current and classical slowing-down time increase.

Fuel element burnup measurements for the equilibrium LEU silicide RSG GAS (MPR-30) core under a new fuel management strategy

After the equilibrium LEU silicide core of RSG GAS was achieved, there was a strong need to validate the new fuel management strategy by measuring burnup of fuel elements comprising the core. Since the regulatory body had a great concern on the safety limit of the silicide fuel element burnup, amongst the 35 burnt fuel elements we selected 22 fuel elements with high burnup classes i.e. from 20 to 53% loss of U-235 (declared values) for the present measurements. The burnup measurement method was based on a linear relationship between reactivity and burnup where the measurements were conducted under subcritical conditions using two fission counters of the reactor startup channel. The measurement results were compared with the declared burnup evaluated by an in-house in-core fuel management code, BATAN-FUEL. A good agreement between the measured burnup values and the calculated ones was found within 8% uncertainties. Possible major sources of differences were identified, i.e. large statistical errors (i.e. low fission counters’ count rates), variation of initial U-235 loading per fuel element and accuracy of control rod indicators. The measured burnup of the 22 fuel elements provided the confirmation of the core burnup distribution planned for the equilibrium LEU silicide core under the new fuel management strategy.

Sample design and gamma-ray counting strategy of neutron activation system for triton burnup measurements in KSTAR

On the purpose of triton burnup measurements in Korea Superconducting Tokamak Advanced Research (KSTAR) deuterium plasmas, appropriate neutron activation system (NAS) samples for 14.1MeV d-t neutron measurements have been designed and gamma-ray counting strategy is established. Neutronics calculations are performed with the MCNP5 neutron transport code for the KSTAR neutral beam heated deuterium plasma discharges. Based on those calculations and the assumed d-t neutron yield, the activities induced by d-t neutrons are estimated with the inventory code FISPACT-2007 for candidate sample materials: Si, Cu, Al, Fe, Nb, Co, Ti, and Ni. It is found that Si, Cu, Al, and Fe are suitable for the KSATR NAS in terms of the minimum detectable activity (MDA) calculated based on the standard deviation of blank measurements. Considering background gamma-rays radiated from surrounding structures activated by thermalized fusion neutrons, appropriate gamma-ray counting strategy for each selected sample is established.

Absolute burnup measurement of LEU silicide fuel plate irradiated in the RSG GAS multipurpose reactor by destructive radiochemical technique

Absolute burnup measurements of a sample taken from the IDA0045 fuel plate of RI-SIE2 low enriched uranium, silicide, experimental fuel element irradiated in the Reaktor Serba Guna G.A. Siwabessy (RSG GAS; previous name MPR-30) multipurpose reactor were conducted using destructive radiochemical technique. The destructive burnup measurement effort is an important part of the continuing efforts related to the post irradiation examination (PIE) project of the Indonesian National Nuclear Energy Agency. The amount of U-235 fissile material left and the Cs-137 fission product in the sample were selected as the burnup indicator for the present work. The measured burnup values of the sample were 72.4∓4.5% loss of U-235 and 78.6∓5.0% loss of U-235 based on the amount of U-235 fissile material and Cs-137 fission product, respectively. The burnup value based on the measured Cs-137 showed an overestimation since it was assumed that all fissions originated from U-235 thermal fissions. Numerical burnup verification was also conducted using the SRAC2006 code system with a PIE dedicated detail burnup chain and a SRAC library based on the JENDL-3.3 nuclear data. The verification results, combined with the previous nondestructive burnup measurement results on the plate wise and axial distributions of the Cs-134/Cs-137 activity ratio, produced burnup value of 75.0∓5.0% loss of U-235 which agreed well with the measured values.

Burnup Measurement of Spent Fuel Assembly by CZT-based Gamma-Ray Spectroscopy for Input Nuclear Material Accountability of Pyroprocessing

Input nuclear material accountancy is crucial for a pyroprocessing facility safeguards. Until a direct Pu measurement technique is established, an indirect method based on code calculations with burnup measurement and neutron counting for 244Cm could be a practical option. Burnup can be determined by destructive analysis (DA) for final dispositive accuracy or by nondestructive assay (NDA) for near-real time accountancy. In the present study, an underwater burnup measurement system based on gamma-ray spectroscopy with the CZT detector was developed and tested on a spent fuel assembly. Burnup was determined according to the 134Cs/137Cs activity ratio with efficiency correction by Geant4 Monte Carlo simulations. The activity ratio as a function of burnup was obtained by ORIGEN calculations. The measured burnup error was 8.6%, which was within the measurement uncertainty. It is expected that the underwater burnup measurement system could fulfill an important role as a means of near-real time accountancy at a future pyroprocessing facility.

Experimental studies of spent fuel burn-up in WWR-SM reactor

The article reports the results of 235U burn-up measurements using 137Cs activity technique for 12 nuclear fuel assemblies of WWR-SM research reactor after 3-year cooling time. The discrepancy between the measured and the calculated burn-up values was about 3%. To increase the reliability of the data and for cross-check purposes, neutron measurement approach was also used. Average discrepancy between two methods was around 12%.

Status and verification strategy for ITER neutronics

The paper summarizes the current status of neutronics at ITER and a first set of proposals for experimental programmes to be conducted in the early operational life-time of ITER are described for the more crucial areas. These include a TF coils heating benchmark, a streaming benchmark and streaming measurements by activation on ITER itself. Also on ITER the measurement of activated water from triton burn-up should be planned and performed. This will require the measurement of triton burn-up in DD phase. Measurements of neutron flux in the tokamak building during DD operations should also be carried out. The use of JET for verification of shut down dose rate estimates is desirable. Other facilities to examine the production and behaviour of activated corrosion products and the shielding properties of concretes to high energy (6MeV) gamma-rays are recommended.

Performance evaluation of a resonance ionization mass spectrometer developed for the FFDL system of fast reactors

To prevent a fuel failure event from becoming a serious radiation accident, sodium-cooled fast reactors are equipped with a system for failed fuel detection and location (FFDL). The FFDL instrument employed in the prototype fast breeder reactor Monju is based on the gas tagging method, in which precise and accurate measurements of krypton and xenon isotope ratios (78Kr/80Kr, 82Kr/80Kr and 126Xe/129Xe) must be performed in a short time. Burnup measurements also contribute to accurate determination of 82Kr/80Kr. We have developed a highly sensitive resonance ionization mass spectrometer for the isotopic analyses, which uses resonance ionization of Kr and Xe atoms by a pulsed laser at wavelengths of 216.7 and 249.6 nm, respectively. In evaluating the performance of our spectrometer, we find that systematic errors caused by isotope shifts can be reduced to negligible levels, and that statistical errors of 3% at a nuclide concentration of 7 ppt can be achieved with a single measurement time of about 40 minutes for each Kr and Xe isotope ratio. This means that, within one hour, about 200 fuel assemblies can be individually identified with a probability of 99%, verifying the applicability of our spectrometer to the FFDL system of fast reactors.

Lattice physics analysis of measured isotopic compositions of irradiated BWR 9 × 9 UO2 fuel

As part of a validation study of burnup calculations of BWR cores, lattice physics analyses were performed on burnups and isotopic compositions of U, Pu and fission product nuclides measured on five samples taken from 9 × 9 BWR fuel assemblies. Burnup calculations in infinite assembly geometry were carried out using MVP-BURN and SRAC codes coupled with major nuclear data libraries. The burnups determined based on the Nd-148 method were from 27.9 to 64.2 GWd/t. The typical relative differences in isotopic compositions (atom/Total-U) between the burnup calculations and measurements were −2 ∼ 19% for 234U, −20 ∼ 3% for 235U, −1.5 ∼ 0.1% for 236U, −0.04 ∼ 0.02% for 238U, −4 ∼ 11% for 238Pu, −11 ∼ −2% for 239Pu, −3 ∼ 0% for 240Pu, −12 ∼ −2% for 241Pu and −2 ∼ 3% for 242Pu. They were −2 ∼ 2% for Nd isotopes, −15 ∼ 7% for Eu isotopes, −13 ∼ 1% for Cs isotopes, −13 ∼ 8% for Sm isotopes, 0 ∼ 7% for 147Pm, −7 ∼ −2% for 95Mo, −2 ∼ −1% for101Ru and 0 ∼ 4% for 103Rh.

Monte Carlo studies on the burnup measurement for the high temperature gas cooling reactor

Online fuel pebble burnup measurement in a future high temperature gas cooling reactor is proposed for implementation through a high purity germanium (HPGe) gamma spectrometer. By using KORIGEN software and MCNP Monte Carlo simulations, the single pebble gamma radiations to be recorded in the detector are simulated under different irradiation histories. A specially developed algorithm is applied to analyze the generated spectra to reconstruct the gamma activity of the 137Cs monitoring nuclide. It is demonstrated that by taking into account the intense interfering peaks, the 137Cs activity in the spent pebbles can be derived with a standard deviation of 3.0% (1σ). The results support the feasibility of utilizing the HPGe spectrometry in the online determination of the pebble burnup in future modular pebble bed reactors.

Single-Stage Dual-Column HPLC Technique for Separation and Determination of Lanthanides in Uranium Matrix: Application to Burnup Measurement on Nuclear Reactor Fuel

A single-stage dual-column chromatographic technique has been developed in this study for separation and determination of lanthanides in a uranium matrix. A 5-cm-length reversed-phase column coated with tri-n-octylphosphine oxide (TOPO) was connected in series to a 10-cm-length reversed-phase monolithic column (dynamically modified into a cation exchange column) to accomplish individual isolation of lanthanides from the uranium matrix. The proposed technique eliminates the step of uranium matrix removal for the determination of lanthanides. Samples with a lanthanide-to-uranium ratio (1 part lanthanide to 105 parts uranium) were directly injected into the dual column for the quantitative determination of lanthanides without uranium matrix removal. In some studies, samples of lanthanides in the uranium matrix could be injected as much as 45 times consecutively into a high-performance liquid chromatography system for determination of lanthanides without any uranium elution. The retention behavior of Pu(IV), Pu(III), Am(III), and fission products was also investigated on the TOPO-coated support. The single-stage dual-column chromatographic technique was demonstrated for the determination of fission products such as La and Nd in the dissolver solution of pressurized heavy water reactor spent fuel for the measurement of atom percent fission burnup. The technique can also be employed to estimate lanthanide impurities in samples of UO2 (1 part lanthanide to 106 parts uranium) without removal of the uranium matrix.

Feasibility studies on the burnup measurement of fuel pebbles with HPGe gamma spectrometer

The feasibility of utilizing a High Purity Germanium (HPGe) detector for the fuel element burnup measurement in a future Modular Pebble Bed Reactor (MPBR) was studied. First, the HPGe spectrometer was set-up for running the detector at high count rates while keeping the energy resolution adequately high to discriminate the Cs-137 peak from other interfering peaks. Based on these settings, the geometrical conditions are settled. Next, experiments were performed with Co-60 and Cs-137 sources to mimic the counting rates in real applications. With the aid of KORIGEN and MCNP/G4 simulations, it was demonstrated that the uncertainty of the Cs-137 counting rate can be well controlled within 3.5%. Finally, a full size prototype was tested in comparison with detailed Monte Carlo simulation and the efficiency transfer method was further utilized for efficiency calibration. To reduce the uncertainty in the efficiency transfer process, a standard point source embedded in a graphite sphere was used for efficiency calibration. The correction factor due to pebble self-attenuation was carefully studied.

Burn-Up Measurements on Dissolver Solution of Mixed Oxide Fuel Using HPLC-Mass Spectrometric Method

Burn-up measurement on an irradiated mixed oxide (MOX) test fuel pellet was carried out through measurements on the dissolver solution by HPLC-Thermal Ionization Mass Spectrometric (TIMS) technique. The studies carried out using HPLC as well as TIMS for quantification of burn-up value are described. While in one case, both the separation and determination of elements of interest (U, Pu and Nd) were carried out by HPLC; in another case, TIMS technique was used to quantify them from the HPLC separated fractions.The rapid separation procedures developed in our laboratory earlier were employed to isolate pure fractions of the desired elements. The individual lanthanide fission products (La to Eu) were separated from each other using dynamic ion-exchange chromatographic technique whereas uranium and plutonium were separated from each other using reversed phase chromatographic technique. The pure fractions of U, Pu and Nd obtained after HPLC separation procedure for “spiked” and “unspiked” dissolver solutions were used in TIMS measurements for the first time in our laboratory. In TIMS analysis, isotopic abundances of uranium, plutonium and neodymium fractions obtained from HPLC separation procedure on an “unspiked” fuel sample were measured. For the determination of U, Pu and Nd by isotopic dilution mass spectrometric technique (IDMS), known quantities of tracers enriched in 238U, 240Pu and 142Nd were added to the pre-weighed dissolver solution and subjected to HPLC separation procedures. The isotope ratios viz. 142Nd/(145Nd +146Nd), 238U/233U and 240Pu/239Pu in the pertinent “spiked” fractions were subsequently measured by TIMS. The spikes were pre-standardized in our laboratory employing reverse isotopic dilution technique against the standard solutions available in our laboratory (for 238U, 239Pu and 142Nd, standard solutions of 233U, 239Pu (of higher abundance than in the sample) and 150Nd were employed as spikes). The burn-up values from duplicate spiking experiments were computed based on the summation of 145Nd + 146Nd. The concentrations of neodymium, uranium and plutonium were also measured using HPLC with post-column derivatisation technique using aresenazo(III) as the post-column reagent. The atom % burn-up computed from HPLC and TIMS techniques were in good agreement.

Conditioning the γ spectrometer for activity measurement at very high background

The application of a high purity Germanium (HPGe) γ spectrometer in determining the fuel element burnup in a future reactor is studied. The HPGe detector is exposed by a 60Co source with varying irradiation rate from 10×103 s−1 to 150×103 s−1 to simulate the input counting rate in real reactor environment. A 137Cs and a 152Eu source are positioned at given distances to generate a certain event rate in the detector with the former being proposed as a labeling nuclide to measure the burnup of a fuel element. It is shown that both the energy resolution slightly increasing with the irradiation rate and the passthrough rate at high irradiation level match the requirement of the real application. The influence of the background is studied in the different parameter sets used in the specially developed procedure of background subtraction. It is demonstrated that with the typical input irradiation rate and 137Cs intensity relevant to a deep burnup situation, the precision of the 137Cs counting rate in the current experiment is consistently below 2.8%, indicating a promising feasibility of utilizing an HPGe detector in the burnup measurement in future bed-like reactors.

Application of Ex-vessel Neutron Dosimetry Combined with In-core Measurements for Correction of Neutron Source Used for Reactor Pressure Vessel Fluence Calculations

Abstract This paper deals with calculated and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Neutron activation measurements analyzed in the paper were carried out in an ex-vessel air cavity at different nuclear power plant units with VVER-1000 during different fuel cycles. The time-integrated neutron source distributions used for DORT calculations were prepared via two different approaches based on (a) calculated fuel burnup (standard routine procedure) and (b) in-core measurements by means of self-powered detectors (SPDs) and thermocouples (TCs) (new approach). Considering that fuel burnup distributions in operating VVER may be evaluated now by the use of analytical methods (calculations) only, it is necessary to develop new approaches for the testing and correction of calculated evaluations of a neutron source. The results presented in this paper allow one to consider the reverse task of the alternative estimation of fuel burnup distributions. The proposed approach is based on the adjustment (fitting) of time-integrated neutron source distributions, and thus fuel burnup patterns, in some part of the reactor core, taking into account neutron leakage measurements, neutron-physical calculations, and in-core SPD and TC measurement data.

A Feasibility Study to Determine Cooling Time and Burnup of Advanced Test Reactor Fuel Using a Nondestructive Technique and Three Types of Gamma-ray Detectors

Abstract The goal of this work was to perform a feasibility study and establish measurement techniques to determine the burnup of the Advanced Test Reactor (ATR) fuels at the Idaho National Laboratory (INL). Three different detectors of high purity germanium (HPGe), lanthanum bromide (LaBr3), and high pressure xenon (HPXe) in two detection system configurations of below and above the water pool were used in this study. The last two detectors were used for the first time in fuel burnup measurements. The results showed that a better quality spectra can be achieved with the above the water pool configuration. Both short and long cooling time fuels were investigated in order to determine which measurement technique, absolute or fission product ratio, is better suited in each scenario and also to establish what type of detector should be used in each case for the best burnup measurement. The burnup and cooling time calibrations were established using experimental absolute activities or isotopic ratios and ORIGEN burnup calculations. A method was developed to do burnup and cooling time calibrations using fission isotopes activities without the need to know the exact geometry.

Reactivity considerations for the on-line refuelling of a pebble bed modular reactor—Illustrating safety for the most reactive core fuel load

In the multi-pass fuel management scheme employed for the pebble bed modular reactor the fuel pebbles are re-circulated until they reach the target burn-up. The rate at which fresh fuel is loaded and burned fuel is discharged is a result of the core neutronics cycle analysis but in practice (on the plant) this has to be controlled and managed by the fuel handling and storage system and use of the burnup measurement system. The excess reactivity is the additional reactivity available in the core during operating conditions that is the result of loading a fuel mixture in the core that is more reactive (less burned) than what is required to keep the reactor critical at full power operational conditions. The excess reactivity is balanced by the insertion of the control rods to keep the reactor critical. The excess reactivity allows flexibility in operations, for example to overcome the xenon build up when power is decreased as part of load follow. In order to limit reactivity excursions and to ensure safe shutdown the excess reactivity and thus the insertion depth of the control rods at normal operating conditions has to be managed. One way to do this is by operational procedures. The reactivity effect of long-term operation with the control rods inserted deeper than the design point is investigated and a control rod insertion limit is proposed that will not limit normal operations. The effects of other phenomena that can increase the power defect, such as higher-than-expected fuel temperatures, are also introduced. All of these cases are then evaluated by ensuring cold shutdown is still achievable and where appropriate by reactivity insertion accident analysis. These aspects are investigated on the PBMR 400MW design.

Nondestructive burnup measurements by gamma-ray spectroscopy on spent fuel elements of the RP-10 research reactor

Gamma-ray spectroscopy is an important nondestructive method for the qualification of irradiated nuclear fuels. Regarding research reactors, the main parameter required in the scope of such qualification is the average burnup of spent fuel elements. This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using 137Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%.

Cross section for inelastic neutron ``acceleration'' by178Hfm2

The scattering of thermal neutrons from isomeric nuclei may include events in which the outgoing neutrons have increased kinetic energy. This process has been called inelastic neutron acceleration, or INNA, and occurs when the final nucleus, after emission of the neutron, is left in a state with lower energy than that of the isomer. The result, therefore, is an induced depletion of the isomer to the ground state. A cascade of several $\ensuremath{\gamma}$'s must accompany the neutron emission to release the high angular momentum of the initial isomeric state. INNA was previously observed in a few cases, and the measured cross sections were only in modest agreement with theoretical estimates. The most recent measurement of an INNA cross section was ${\ensuremath{\sigma}}_{\mathrm{INNA}}=258\ifmmode\pm\else\textpm\fi{}58 \mathrm{b}$ for neutron scattering by $^{177}\mathrm{Lu}$${}^{m}$. In the present work, an INNA cross section of ${\ensuremath{\sigma}}_{\mathrm{INNA}}=168 \ifmmode\pm\else\textpm\fi{} 33 \mathrm{b}$ was deduced from measurements of the total burnup of the high-spin, four-quasiparticle isomer $^{178}\mathrm{Hf}$${}^{m2}$ during irradiation by thermal neutrons. Statistical estimates for the probability of different reaction channels past neutron absorption were used in the analysis, and the deduced ${\ensuremath{\sigma}}_{\mathrm{INNA}}$ was compared to the theoretically predicted cross section.