scholarly journals Literature review of intrinsic actinide colloids related to spent fuel waste package release rates

1997 ◽  
Author(s):  
P. Zhao ◽  
S.A. Steward
Keyword(s):  
Literature Review ◽  
Spent Fuel ◽  
Release Rates ◽  
Waste Package

Analysis Of Congruent Matrix Release, Precipitation, And Time-Distributed Containment Failure On Spent Fuel Performance

1987 ◽  
Vol 112 ◽  
Author(s):  
Michael J. Apted ◽  
David W. Engel
Keyword(s):  
Source Term ◽  
Spent Fuel ◽  
Release Rates ◽  
Waste Package ◽  
Regulatory Limits ◽  
Lower Solubility ◽  
Fractional Abundance ◽  
Congruent Dissolution ◽  
Cesium 137 ◽  
Term Evaluation

AbstractThe Analytical Repository Source-Term (AREST) code has been developed for source-term evaluation of spent fuel as a final waste form in geologic repositories. AREST contains a set of analytical equations for the timedependent diffusional mass transport of both solubility-limited and inventory-limited radionuclides from a spent fuel in a failed container surrounded by a shell of packing or other porous material imbedded in a porous host rock. Three factors that affect release performance are examined: 1) congruent dissolution of the UO2 matrix, 2) chemical instability of the UO2 matrix, with precipitation of a more stable uranium phase within the waste package, and 3) the attenuation of release rate by distribution of containment failures with time.For congruent matrix dissolution, the release rates of included nuclides are proportional to the product of solubility-limited release of uranium and the fractional abundance of the nuclide. For certain conditions, congruent release rates are calculated to be up to 10 orders of magnitude lower than release rates assuming individual solubility-limits. Precipitation of a more stable, lower solubility uranium phase within the waste package is shown to increase release rates from the UO2 matrix compared to the non-precipitation case, in agreement with previous calculations. During the first 300 to 1000 years after repository closure, the distribution of containment failures with time will act to attenuate the peak average release rates of soluble, longlived nuclides, such as iodine-129, to values smaller than release rates below regulatory limits. However, for soluble nuclides with short half-lives, such as cesium-137, a broader distribution of containment failure with constant mean time of failure can actually cause an increase In the peak average release rates.

The Effect of Waste Package Components on Radionuclides Released from Spent Fuel Under Hydrothermal Conditions

1987 ◽  
Vol 112 ◽  
Author(s):  
Shirley A. Rawson ◽  
William L. Neal ◽  
James R. Burnell
Keyword(s):  
Low Carbon Steel ◽  
Fission Products ◽  
Nuclear Waste Repository ◽  
Low Carbon ◽  
Spent Fuel ◽  
Hydrothermal Conditions ◽  
Waste Package ◽  
Dissolved Species ◽  
Radionuclide Release ◽  
Time Invariant

AbstractThe Basalt Waste Isolation Project has conducted a series of hydrothermal experiments to characterize waste/barrier/rock interactions as a part of its study of the Columbia River basalts as a potential medium for a nuclear waste repository. Hydrothermal tests of 3–15 months duration were performed with light water reactor spent fuel and simulated groundwater, in combination with candidate container materials (low-carbon steel or copper) and/or basalt, in order to evaluate the effect of waste package materials on spent fuel radionuclide release behavior. Solutions were filtered through 400 and 1.8 nm filters to distinguish colloidal from dissolved species. In all experiments, 14C, 129I, and 137Cs occurred only as dissolved species, whereas the actinides occurred in 400 nm filtrates primarily as spent fuel particles. Actinide concentrations in 1.8 nm filtrates were below detection in steel-bearing experiments. In the system spent fuel + copper, apparent time-invariant concentrations of 14C and 137Cs were obtained, but in the spent fuel + steel system, the concentrations of 14C and 137Cs increased gradually throughout the experiments. In experiments containing basalt or steel + basalt, 137Cs concentrations decreased with time. In tests with copper + basalt, 14C and 129I concentrations attained time-invariant values and 137Cs concentrations decreased. Concentrations for the actinides and fission products measured in these experiments were below those calculated from Federal regulations governing radionuclide release.

scholarly journals RADIONUCLIDE CHARACTERISTICS OF RDE SPENT FUELS

2018 ◽  
Vol 20 (2) ◽  
pp. 69 ◽  
Author(s):  
Ihda Husnayani ◽  
Pande Made Udiyani
Keyword(s):  
Thermal Power ◽  
Neutron Production ◽  
Cooling Time ◽  
Spent Fuel ◽  
Release Rates ◽  
Radiation Dose Rate ◽  
The Core ◽  
Spent Fuel Storage ◽  
Fuel Storage

Reaktor Daya Eksperimental (RDE) is a 10 MWth pebble-bed High Temperature Gas-cooled Reactor that is planned to be constructed by National Nuclear Energy Agency of Indonesia (BATAN) in Puspiptek complex, Tangerang Selatan. RDE utilizes low enriched UO2 fuel coated by TRISO layers and loaded into the core by means of multipass loading scheme. Determination of radionuclide characteristics of RDE spent fuel; such as activity, thermal power, neutron and photon release rates; are very important because those characteristics are crucial to be used as a base for evaluating the safety of spent fuel handling system and storage tank. This study is aimed to investigate the radionuclide characteristics of RDE spent fuel at the end of cycle and during the first 5 years cooling time in spent fuel storage. The method used to investigate the radionuclide characteristics is burnup calculation using ORIGEN2.1 code. In performing the ORIGEN2.1 calculation, one pebble fuel was assumed to be irradiated in the core for 5 cycles and then decayed for 5 years. At the end of the fifth cycle, it is obtained that the total activity, thermal power, neutron production, and photon release rates from all radionuclides inside one spent fuel are approximately 105.68 curies, 0.41 watts, 2.65 x 103 neutrons/second, and 1.79 x 104 photons/second, respectively. The results for the radionuclides characteristics during the first 5 years cooling time in the spent fuel storage show that the radioactivity characteristics from all radionuclides are rapidly decreasing at the first year and then slowly decreasing at the second until the fifth year of cooling time. The results obtained in this study can provide data for safety evaluation of fuel handling and spent fuel storage, such as the calculation of sourceterm, radiation dose rate, and the determination of radiation shielding.Keywords: RDE, spent fuel, radionuclide activity, thermal power, neutron production, photon releaserates KARAKTERISTIK RADIONUKLIDA DI DALAM BAHAN BAKAR RDE. Reaktor Daya Eksperimental (RDE) adalah reaktor tipe Reaktor Temperatur Tinggi Berpendingin Gas dengan daya termal 10MW yang akan dibangun oleh BadanTenagaNuklirNasional (BATAN) di kawasanPuspiptek, Tangerang Selatan. RDE menggunakan bahan bakar UO2 yang dilapisi dengan lapisan TRISO dan dimasukkan ke dalam teras RDE menurut skema multipass (5 siklus). Penentuan karakteristik radionuklida di dalam bahan bakar RDE; seperti aktivitas, daya termal, laju produksi neutron dan pelepasan foton; adalah sangat penting karena informasi karakteristik ini diperlukan sebagai dasar untuk melakukan evaluasi keselamatan system penanganan dan penyimpanan bahan bakar bekas. Penelitian ini bertujuan untuk menganalisis karakteristik radionuklida bahanbakar RDE setelah 5 siklus dan pada 5 tahun pertama pendinginan ditempat penyimpanan bahan bakar bekas. Metode yang digunakan dalam menghitung karakteristik radionuklida adalah menggunakan program ORIGEN2.1. Satu bola bahan bakar RDE diasumsikan diiradiasi selama 5 siklus dan kemudian meluruh selama 5 tahun. Pada akhir siklus, diperoleh hasil aktivitas total, daya termal, laju produksi neutron dan pelepasan foton dari seluruh radionuklida di dalam satu bola bahan bakar RDE sebesar 105,68 curies, 0,41 watts, 2,65 x 103 neutron/detik, dan 1,79 x 104 foton/detik. Hasil untuk karakteristik radionuklida selama 5 tahun penyimpanan menunjukkan bahwa karakteristik radioktivitas radionuklida menurun dengan cepat pada tahun pertama dan kemudian menurun lebih lambat pada tahun kedua hingga tahun kelima. Hasil perhitungan karakteristik radionuklida dari penelitian ini dapat digunakan sebagai basis untuk analisis keselamatan penanganan dan penyimpanan bahan bakarbekas RDE.Kata kunci:RDE, bahan bakar bekas, aktivitas radionuklida, daya termal, produksi neutron, laju foton

Spent Fuel Corrosion Behavior in Salt Solution in the Presence of Hydrogen Overpressure

2001 ◽  
Author(s):  
Andreas Loida ◽  
Bernd Grambow ◽  
Horst Geckeis
Keyword(s):  
Gas Phase ◽  
Spent Fuel ◽  
Nacl Solution ◽  
Release Rates ◽  
Corrosion Rates ◽  
The Matrix ◽  
Matrix Bound ◽  
Safety Assessments ◽  
Container Material

Abstract The simultaneous corrosion of spent fuel and Fe-based container material is characterized by the formation of large amounts of hydrogen, which control the composition of the gas phase. Various experimental data indicate that the matrix dissolution rate and the release rates of important radionuclides decrease, if the H2 overpressure increases. To quantify to what extent the hydrogen overpressure may counteract radiolysis enhanced matrix dissolution rates, and to take credit from the effect of hydrogen overpressure in long-term safety assessments of the repository, a detailed experimental investigation has been initiated. High burnup spent fuel is being corroded under anoxic conditions in the absence of carbonate in 5m NaCl solution under an external H2 overpressure of 3.3 bar. This pressure is in the same range as observed in a long-term test using spent fuel and Fe-powder. Results obtained after 117 days of testing show that due to constant or decreasing concentrations of Sr and other matrix bound radionuclides, corrosion rates were not measurable indicating a stop of matrix dissolution or very low long-term rates. Grain boundary release of Cs and fission gases was found to continue under hydrogen overpressure. Compared to tests in the absence of hydrogen solution concentrations decreased by about ca. 1.5 orders of magnitude for U (10−8 M), Am, Eu (10−10 M), whereas the decrease of Np (3×10−10 M), Tc (5×10−9 M) and Pu (4×10−9 M) concentrations was found to be less significant.

The Effect of Fuel Burnup and Dispersed Water Intrusion on the Criticality of Spent High-Level Nuclear Fuel in a Geologic Repository

1993 ◽  
Vol 333 ◽  
Author(s):  
William G. Culbreth ◽  
Paige R. Zielinski
Keyword(s):  
Fuel Burnup ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Chain Reaction ◽  
Worst Case ◽  
Mixture Ratio ◽  
Waste Package ◽  
Radiological Data ◽  
A Chain ◽  
High Level

ABSTRACTStudies of the spent fuel waste package have been conducted through the use of a Monte-Carlo neutron simulation program to determine the ability of the fuel to sustain a chain reaction. These studies have included fuel burnup and the effect of water mists on criticality. Results were compared with previous studies.In many criticality studies of spent fuel waste packages, fresh fuel with an enrichment as high as 4.5% is used as the conservative (worst) case. The actual spent fuel has a certain amount of “burnup” that decreases the concentration of fissile uranium and increases the amount of radionuclides present. The LWR Radiological Data Base from OCRWM has been used to determine the relative radionuclide ratios and KENO 5.a was used to calculate values of the effective multiplication factor, keff.1Spent fuel is not capable of sustaining a chain reaction unless a suitable moderator, such as water, is present. A completely flooded container has been treated as the worst case for criticality. Results of a previous report that demonstrated that keff actually peaked at a water-to-mixture ratio of 13% were analyzed for validity. In the present study, these results did not occur in the SCP waste package container.

scholarly journals Consequence modelling of hypothetical post-closure criticality events for spent fuel disposal

2015 ◽  
Vol 79 (6) ◽  
pp. 1505-1513 ◽  
Author(s):  
R. M. Mason ◽  
J. K. Martin ◽  
P. N. Smith ◽  
R. J. Winsley
Keyword(s):  
International Studies ◽  
Quasi Steady State ◽  
Spent Fuel ◽  
Regulatory Requirement ◽  
Geological Disposal ◽  
Waste Package ◽  
Decay Heat ◽  
Disposal Facility ◽  
Safety Case ◽  
The Uk

AbstractIn support of the Radioactive Waste Management (RWM) safety case for a geological disposal facility (GDF) in the UK, there is a regulatory requirement to consider the likelihood and consequences of nuclear criticality. Waste packages are designed to ensure that criticality is not possible during the transport and operational phases of a GDF and for a significant period post-closure. However, over longer post-closure timescales, conditions in the GDF will evolve.For waste packages containing spent fuel, it can be shown that, under certain conditions, package flooding could result in a type of criticality event referred to as 'quasi-steady-state' (QSS). Although unlikely, this defines a 'what-if' scenario for understanding the potential consequences of post-closure criticality. This paper provides an overview of a methodology to understand QSS criticality and its application to a spent fuel waste package.The power of such a hypothetical criticality event is typically estimated to be a few kilowatts: comparable with international studies of similar systems and the decay heat for which waste packages are designed. This work has built confidence in the methodology and supports RWM's demonstration that post-closure criticality is not a significant concern.

A Comparison and Abstraction of the Commercial Spent Fuel Corrosion Experiments in the U.S.

2002 ◽  
Vol 757 ◽  
Author(s):  
Eric R. Siegmann
Keyword(s):  
Corrosion Rate ◽  
Spent Fuel ◽  
Release Rates ◽  
Water Contact ◽  
Contact Mode ◽  
Corrosion Rates ◽  
Single Temperature ◽  
Flow Through ◽  
Distribution Sensitivity ◽  
Temperature Dependent Model

ABSTRACTThis paper compares the results of three different fuel corrosion experiments by taking the existing corrosion data reported for various temperatures and recalculating corrosion rates at a single temperature of 25°C using a temperature dependent model developed elsewhere. Three types of light water reactor fuel corrosion tests (sometimes called dissolution or alteration tests) were performed in support of Yucca Mountain Project. The tests used three water contact modes and various fuel burnups. All measurements were adjusted for temperature differences and then compared. Five different isotopes (cesium, technetium, iodine, strontium, and, in the flow-through tests, uranium) were considered as a measure of corrosion. The data used represent over ten years of experiments with about nine different fuel types. Most experiments were with repository type fluids, containing dissolved constituents such as carbonate, calcium and silicon. The results show that all of the experiments predict similar fuel corrosion rates. Small differences in the isotope release rates are observed and incorporated in the abstracted uncertainty. Water contact mode (flow-through, batch, or drip) does not seem to be very important although the drip tests introduced larger variations. In developing a corrosion abstraction, all of the isotope measurements were considered equally. The distribution of release rates was used directly to develop the uncertainty. The mean corrosion rate was 1.8 × 10-4 fraction/year at 25°C (5%-95% range = 5.7 × 10-5 to 1.7 × 10-3). Using the derived corrosion rate for 25°C, and considering rapid axial splitting of the cladding, the CSNF fuel rod is expected to corrode in less than 2,000 years. The abstraction uses all the available experiments performed with water containing carbonates, silicon, or calcium and irradiated fuel to produce a corrosion rate distribution. Sensitivity studies using this corrosion rate abstraction in the TSPA-SR analysis show very small changes in dose (3%) in response to changes in the UO2 corrosion abstraction.

Sign in / Sign up

Export Citation Format

Share Document