Thermodynamic considerations and prediction of the primary coolant activity of Tc

A physical model has been developed to describe the coolant activity behaviour of ⁹⁹Tc, during constant and reactor shutdown operations. This analysis accounts for the fission production of technetium and molybdenum, in which their chemical form and volatility is determined by a thermodynamic treatment using Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix and vaporization from the fuel-grain surface. Based on several in-reactor tests with defective fuel elements, and as supported by the thermodynamic analysis, the model accounts for the washout of molybdenum from the defective fuel on reactor shutdown. The model also considers the recoil release of both ⁹⁹Mo and ⁹⁹Tc from uranium contamination, as well as a corrosion source due to activation of ⁹⁸Mo. The model has provided an estimate of the activity ratio ⁹⁹Tc/¹³⁷Cs in the ion-exchange columns of the Darlington Nuclear Generating Station, i.e., 6 × 10⁻⁶ (following ∼200 days of steady reactor operation) and 4 × 10⁻⁶ (with reactor shutdown). These results are consistent with that measured by the Battelle Pacific Northwest Laboratories with a mixed-bed resin-sampling device installed in a number of Pressurized Water Reactor and Boiling Water Reactor plants.     

Modeling the activity of I and Cs in the primary coolant and CVCS resin of an operating PWR

Mathematical models have been developed to describe the activities of ¹²⁹I and ¹³⁷Cs in the primary coolant and resin of the chemical and volume control system (CVCS) during constant power operation in a pressurized water reactor (PWR). The models, which account for the source releases from defective fuel rod(s) and tramp uranium, rely on the contribution of CVCS resin and boron recovery system as a removal process, and differences in behavior for each nuclide. The current models were validated through measured coolant activities of ¹³⁷Cs. The resultant scaling factors agree reasonably well with the results of the test resin of the coolant and the actual resins from the PWRs of other countries.     

Determination of I activity in the RBMK-1500 main circulation circuit

The purpose of this work was to evaluate the content of difficult to measure isotope ¹²⁹I in the RBMK-1500 reactor fuel-to-clad gap and reactor main circulation circuit (MCC) coolant. To determine fission product (FP) release from the defective fuel, the methodology proposed by Lewis and Husain for the CANDU reactor primary coolant activity prediction was applied. The determined effective diffusion coefficient D′ = 1.2E−09 s⁻¹ of iodine in the RBMK-1500 fuel is higher than the one evaluated for the CANDU fuel 6.8E−10 s⁻¹. Results show that the method developed by Lewis and Husain can be applied for the RBMK-1500 fuel gap and reactor main circulation circuit coolant activity prediction.     

Burnup studies of spent fuels of varying types and enrichment

This paper describes the results of fuel burnup measurements, made over a period of several years on discharged fuel from nuclear power plant and research reactor. The measured and calculated burnup of different spent fuel types, viz.: Natural uranium CANDU fuel bundles; 10.5% enriched booster rods; 20% enriched MTR fuel elements have been presented. High-resolution gamma spectrometry, using ¹³⁷Cs and ¹³⁴Cs burnup monitors was employed in different reactors to estimate the amount of ²³⁵U depletion in the respective fuel. The experimental data was compared with those of calculations to optimize fuel-scheduling programme. The burnup data is useful for assessment of fuel performance in the core and resolving design issues related to long-term storage facilities. It has been observed that the gamma spectrometry is very effective in identifying exact position of individual booster bundles inside the discharged booster assemblies, which is useful in safeguard applications. It is concluded that the distribution of measured isotopic activity ratios of ¹³⁴Cs/¹³⁷Cs along the height of the spent fuel gives accurate estimate of the axial neutron flux profiles in the core. The activity ratio technique therefore provides a useful method to determine flux peaking factors at the particular locations of the fuel assemblies in the reactor.     

Modeling the activity and scaling factor of I in the primary coolant and CVCS resin of operating PWR

To estimate the activity of ¹²⁹I at the primary coolant and chemical and volume control system (CVCS) resin in Korean pressurized water reactor (PWR) plants, a theoretical methodology was developed on the basis of an existing model of primary coolant activity and new model of CVCS resin activity. In order to reflect the difference between ¹²⁹I and ¹³⁷Cs, the different power-related diffusivities in the defective fuel were derived, and the variable removal efficiency of the CVCS resin for ¹³⁷Cs was applied as a function of the coolant activity ratio of ¹³¹I/¹³⁷Cs. The current computational method was validated by using the measured coolant activities of ¹³⁷Cs, and the results show better agreement than a previously suggested parameter correlation method between ¹²⁹I and ¹³⁷Cs. There was also reasonable agreement in a comparison of the results of the test resin columns of the coolant from the PWR plants of other countries. It was shown that the ratio of the effective removal efficiency of ¹²⁹I and ¹³⁷Cs in the CVCS resin linearly influences the activity ratio of ¹²⁹I/¹³⁷Cs in the coolant, but on the other hand, its influence on the activity ratio in the CVCS resin is relatively less sensitive compared with that in the coolant.     

Investigation of excitation functions of alpha induced reactions on Xe: Production of the therapeutic radioisotope Cs

Excitation functions were measured for alpha-particle induced nuclear reactions on natural xenon leading to the formation of the radionuclides ^{129m(rel),129g,131m,131mg,133m,135m,137m,139cum}Ba and ^{129cum,130mg,132,134m,135m,136mg,138mg}Cs from the respective thresholds up to 40 MeV. No earlier experimental cross section data were found in the literature. The experimental data were compared to and analyzed with the results of the theoretical model code ALICE-IPPE. The feasibility of the production of the therapeutic radioisotope ¹³¹Cs by using gas target technology was investigated. Comparison of reactor and cyclotron production routes of ¹³¹Cs was given.     

Study on the LLFPs transmutation in a super-critical water-cooled fast reactor

The performance of the super-critical water-cooled fast reactor (Super FR) for the transmutation treatment of long-lived fission products (LLFPs) was evaluated. Two regions with the soft neutron spectrum, which is of great benefit to the LLFPs transmutation, can be utilized in the Super FR. First region is in the blanket assembly due to the ZrH_1.7 layer which was utilized to slow down the fast neutrons to achieve a negative void reactivity. Second region is in the reflector region of core like other metal-cooled fast reactors. The LLFPs selected in the transmutation analysis include ⁹⁹Tc, ¹²⁹I and ¹³⁵Cs discharged from LWR or fast reactor. Their isotopes, such as ¹²⁷I, ¹³³Cs, ¹³⁴Cs and ¹³⁷Cs were also considered to avoid the separation. By loading the isotopes (⁹⁹Tc or ¹²⁷I and ¹²⁹I) in the blanket assembly and the reflector region simultaneously, the transmutation rates of 5.36%/GWe year and 2.79%/GWe year can be obtained for ⁹⁹Tc and ¹²⁹I, respectively. The transmuted amounts of ⁹⁹Tc and ¹²⁹I are equal to the yields from 11.8 and 6.2 1000 MWe-class PWRs. Because of the very low capture cross section of ¹³⁵Cs and the effect of other cesium isotopes, ¹³⁵Cs was loaded with three rings of assemblies in the reflector region to make the transmuted amount be larger than the yields of two 1000 MWe-class PWRs.Based on these results, ⁹⁹Tc and ¹²⁹I can be transmuted conveniently and higher transmutation performance can be obtained in the Super FR. However, the transmutation of ¹³⁵Cs is very difficult and the transmuted amount is less than that produced by the Super FR. It turns out that the transmutation of ¹³⁵Cs is a challenge not only for the Super FR but also for other commercial fast reactors.     

Behavior of Radioactive Iodine Isotopes in the Multiple Forced Circulation Loop in an RBMK Reactor

The flow of iodine, including ¹³¹I, into the coolant water in a nuclear power plant with an RBMK-1000 reactor under normal operating conditions and during transient regimes is analyzed. It is shown that under normal operating conditions the specific activity of ¹³¹I in the coolant is correlated with the iron concentration. During shutdown, its content increases by factors of 30–200. The emission of ¹³¹I into the coolant can be decreased by factors of 10–15 and the degree of unsealing of fuel elements can be decreased if before shutdown the reactor is held for 2–5 days at 50% of the nominal power level. Recommendations are made for decreasing ¹³¹I emissions into the atmosphere. The adoption of these recommendations at the Leningrad nuclear power plant has reduced the ¹³¹I emissions into the atomsphere by a factor of 17. __________ Translated from Atomnaya Energiya, Vol. 99, No. 2, pp. 103–108, August 2005.     

Analysis of cesium isotope compositions in environmental samples by thermal ionization mass spectrometry-3

Cesium was recovered from soil samples obtained in Fukushima prefecture. Isotopic analysis of Cs was performed by γ-spectrometry to determine the activity ratio of ¹³⁴Cs/¹³⁷Cs and thermal ionization mass spectrometry was used to determine the isotopic ratios of ¹³³Cs/¹³⁷Cs and ¹³⁵Cs/¹³⁷Cs. The analytical results showed that both the activity ratio of ¹³⁴Cs/¹³⁷Cs and the isotopic ratio of ¹³⁵Cs/¹³⁷Cs were within the expected values for the Fukushima Daiichi Nuclear Power Plant estimated using the ORIGEN-II code, suggesting that most of the radioactive Cs in the soil sample originated from the Fukushima Daiichi Nuclear Power Plant. The concentration of ¹³⁷Cs and the contribution of radioactive Cs from global fallout were correlated to the distance from the Fukushima Daiichi Nuclear Power Plant, while the contribution of radioactive Cs from each reactor did not show any similar distance dependence.     

Towards a better understanding of iodine chemistry in RCS of nuclear reactors

The Phebus FP in-reactor integral experiments provided new insights into iodine transport through the primary circuit. Indeed, in these tests transported iodine was often found not associated with caesium as generally postulated up to now. Several iodine species were experimentally shown to have been transported in the hot leg at 700 °C, while a fraction was also suspected to be in a gaseous form in the cold leg at 150 °C. For a better estimate of the iodine source term to the containment, both in terms of speciation and quantity, it becomes thus necessary to reconsider iodine species behaviour along their pathway in the reactor coolant system (RCS).This paper presents the current understanding, mainly based on SOPHAEROS equilibrium chemistry calculations of Phebus FP tests performed within the I-RCS technical circle of the SARNET network of excellence in the EU 6th Framework programme. The suspected connection existing between Cs, Mo, Cd and I chemistry and the strong influence of both their release kinetics and related species thermodynamic properties on the iodine speciation in different environments (reducing/oxidizing) are highlighted. Potential explanations for the predicted iodine volatility and the level of association of I to Cs are also discussed.     

Sensitivity analysis of fission product activity in primary coolant of typical PWRs

A model has been developed for static and dynamic activity analysis of the fission product activity (FPA) in primary coolant of typical pressurized water reactors (PWRs). It has been implemented in the FPCART based computer program FPCART-SA. For long steady power operation of reactor, the computed values of normalized static sensitivity have been compared with the corresponding values obtained by using the dynamic sensitivity analysis. The normalized sensitivity values for the reactor power (P), failed fuel fraction (D), coolant leakage rate (L), total mass of coolant (m) and the let-down flow rate (Q) have been calculated and the values: 1.0, 0.857, −2.0177 × 10⁻⁶, 2.349 × 10⁻⁴, −2.329 × 10⁻⁴ have been found correspondingly for Kr-88 with the dominant value of FPA as 0.273 μCi/g.     

Contribution of thermal neutrons to radiation hardening of pure copper

The paper presents the results of an experiment the aim of which was to estimate directly the effect of the thermal neutron fluence on pure copper hardening. Identical specimens were irradiated in two reactors (SM-2 and RBT-6) in the dose range 10⁻³-10⁻¹ dpa at T_irr=80 °C under substantially different, by a factor of 5, thermal neutron fluences, with other irradiation parameters being close. The results show that the elevated thermal fluence in the SM-2 reactor increases the radiation hardening of pure copper by 50% at a dose of about 10⁻³ dpa as compared with specimens irradiated in the RBT-6 reactor. The contribution of thermal neutrons proved to be much more considerable than the theoretical estimates.     

Examination of 131I and 137Cs releases during late phase of Fukushima Daiichi NPP accident by using 131I/137Cs ratio of source terms evaluated reversely by WSPEEDI code with environmental monitoring data

To investigate what happened in reality during the Fukushima Daiichi Nuclear Power Plant accident, the phenomena within reactor pressure vessel and the discussion of ties with the environmental monitoring measurement are very important. However, the previous study that treats phenomena of the both has not necessarily advanced up to the present time. The source terms predicted by simulation codes such as MELCOR has not yet been consistent with the reverse estimation by WSPEEDI code using environmental measurement data. This study investigated ¹³¹I and ¹³⁷Cs release behaviors during the late phase of the accident to contribute to such examination using the ¹³¹I/¹³⁷Cs ratio of the new source terms predicted by Katata. The ¹³¹I release by the gas–liquid partition from the contaminated water in the 1F2 and 1F3 reactor buildings which was pointed out in the previous study was reevaluated using the new source terms. In addition, paying attention to the similarity of the core conditions between the Fukushima accident and the Phébus FPT3 experiment using the B₄C control rods, the release of organic iodine (CH₃I) during the 1F3 suppression pool venting, formation of CsBO₂ and its release behavior were examined which have not yet been sufficiently studied so far.     

Comparison of three chemical extraction methods for I-129 determinations

We report here results comparing the efficiency of three iodine extraction methods for the determination of ¹²⁹I/I ratios in small samples. The direct organic extraction had the highest average yield. ¹²⁹I/I ratios determined from targets prepared by direct organic extraction and silver extraction are comparable to each other within the error limits. The smallest sample from which a reliable ¹²⁹I/I ratio was measured, started with an iodine input of 0.2 mg iodine and contained 0.14 mg AgI in the target.     

Distribution of radionuclides in surface seawater obtained by an aerial radiological survey

We investigated the distribution in seawater of anthropogenic radionuclides from the Fukushima Daiichi Nuclear Power Plant (FNPP1) as preliminary attempt using a rapid aerial radiological survey performed by the U.S. Department of Energy National Nuclear Security Administration on 18 April 2011. We found strong correlations between in-situ activities of ¹³¹I, ¹³⁴Cs, and ¹³⁷Cs measured in surface seawater samples and gamma-ray peak count rates determined by the aerial survey (correlation coefficients were 0.89 for ¹³¹I, 0.96 for¹³⁴Cs, and 0.92 for¹³⁷Cs). The offshore area of high radionuclide activity extended south and southeast from the FNPP1. The maximum activities of ¹³¹I, ¹³⁴Cs, and ¹³⁷Cs were 329, 650, and 599 Bq L^?1, respectively. The ¹³¹I/¹³⁷Cs ratio in surface water of the high-activity area ranged from 0.6 to 0.7. Considering the radioactive decay of ¹³¹I (half-life 8.02 d), we determined that the radionuclides in this area were directly released from FNPP1 to the ocean. We confirm that aerial radiological surveys can be effective for investigating the surface distribution of anthropogenic radionuclides in seawater. Our model reproduced the distribution pattern of radionuclides derived from the FNPP1, although results simulated by a regional ocean model were underestimated.     

Distribution of the 134Cs/137Cs ratio around the Fukushima Daiichi nuclear power plant using an unmanned helicopter radiation monitoring system

Large amounts of radioactive substances were released into the environment by the Fukushima Daiichi nuclear power plant (FDNPP) accident. Several research institutes have mapped the distribution of nuclides with long half-lives, such as ¹³⁴Cs and ¹³⁷Cs. Although the ratio of ¹³⁴Cs and ¹³⁷Cs has been believed to be equal without depending on the location of the contaminated area, several researchers report that it is different depending on places quite a little. We measured the energy spectrum of gamma rays in high resolution within an approximately 3-km radius of the FDNPP by using an unmanned helicopter equipped with a LaBr₃(Ce) scintillation detector. Then, we analyzed the ¹³⁴Cs/¹³⁷Cs ratio in the area from these measured data in detail. The results show that the ¹³⁴Cs/¹³⁷Cs ratio is different between the plume trace extending north and the other plume traces. We have obtained valuable data for identification of which radioactive substances were released by individual reactor units.     

Gamma spectrometry inspection of TRIGA MARK II fuel using caesium isotopes

Gamma spectrometry is one of the common methods to inspect the spent fuel from research reactors. This method has been applied to in-pool measurements of the Spent Fuel Elements (SPEs) of the TRIGA Mark II research reactor. Due to mixed nature of the reactor core and complicated irradiation history of the fuel elements (FEs), the gamma spectrometry of the FE establishes improvements in the calculation and measurement of the SPE. In order to inspect the TRIGA SPE from dry storage and cooled fuel from the reactor pool, the selected spend fuels are scanned and measured using the fuel-scanning machine. Gamma spectrometry is performed by HPGe detector for spend fuel inspection and determination of the ¹³⁷Cs activity and ¹³⁴Cs/¹³⁷Cs ratio. In this work, the steps of the detector calibration and the use of the Monte Carlo radiation transport code (MCNP5) have been described. In addition, the fuel-scanning machine and the gamma spectrometer are modelled by MCNP5 to simulate the gamma transport from fuel to detector. It also simulate the gamma spectrometer calibration for the burn up determination of the spend fuel. The results from MCNP5 simulation are applied to spectroscopic measurements and compared with the theoretical predictions of the neutronics code ORIGEN2 in this research work.     

A model for coolant void reactivity evaluation in assemblies of CANDU cells

Coolant void reactivity is a very important safety parameter in CANDU reactor analysis. Here we evaluate the coolant void reactivity in a 2 × 2 heterogeneous assembly of CANDU cells using the code DRAGON. Since the current version of DRAGON can only treat the coolant void reactivity for a single CANDU cell, an approximate model for the geometry must be considered to perform assembly calculations in a 2 × 2 pattern. The model we propose consists of replacing the annular fuel pins by equivalent square fuel pins. The equivalence between annular fuel pins and square fuel pins is brought about by homogenizing the fuel plus its sheath and subsequently conserving the reaction rates between the two geometries using a SPH equivalence procedure. The approximate CANDU cells constructed using square pins were used to perform the transport calculations in 2 × 2 assembly patterns. In addition, the model was used to evaluate coolant void reactivity in 2 × 2 checkerboard voiding patterns. These calculations reflect more accurately the actual voiding situation being studied. This helps in assessing the effects due to the coupling of neutrons born in one cell to those born in the neighbouring cells.     

Rapid aqueous release of fission products from high burn-up LWR fuel: Experimental results and correlations with fission gas release

Studies of the rapid aqueous release of fission products from UO₂ and MOX fuel are of interest for the assessment of the safety of geological disposal of spent fuel, because of the associated potential contribution to dose in radiological safety assessment. Studies have shown that correlations between fission gas release (FGR) and the fraction rapidly leached of various long-lived fission products can provide a useful method to obtain some of this information. Previously, these studies have been limited largely to fuel with burn-up values below 50 MWd/kg U. Collaborative studies involving SKB, Studsvik, Nagra and PSI have provided new data on short-term release of ¹³⁷Cs and ¹²⁹I for a number of fuels irradiated to burn-ups of 50–75 MWd/kgU. In addition a method for analysis of leaching solutions for ⁷⁹Se was developed. The results of the studies show that the fractional release of ¹³⁷Cs is usually much lower than the FGR covering the entire range of burn-ups studied. Fractional ¹²⁹I releases are somewhat larger, but only in cases in which the fuel was forcibly extracted from the cladding. Despite the expected high degree of segregation of fission gas (and by association ¹³⁷Cs and ¹²⁹I) in the high burn-up rim, no evidence was found for a significant contribution to release from the rim region. The method for ⁷⁹Se analysis developed did not permit its detection. Nonetheless, based on the detection limit, the results suggest that ⁷⁹Se is not preferentially leached from spent fuel.     

One-year time variations of anthropogenic radionuclides in aerosols in Tokyo after the Fukushima Dai-ichi Nuclear Power Plant reactor failures

We report on the results of monitoring of environmental radiation for one year (13 March 2011 to 12 March 2012), including air dose rates and the concentrations of radionuclides in aerosols in Tokyo, after the reactor failures at the Fukushima Dai-ichi Nuclear Power Plant. The air dose rates began to increase at 4:00–5:00 JST on 15 March 2011, and the maximum rate was observed at 10:00–11:00 JST. Two peaks were observed before 23 March 2011, and then the air dose rates decreased until March 2012. The time variations of concentrations of radionuclides in aerosols showed tendencies similar to those of air dose rates. Short-lived radionuclides (⁹⁹Mo (^99mTc), ^129mTe (¹²⁹Te), ^131mTe, ¹³²Te (¹³²I), ¹³³I and ¹³⁶Cs) were under the detection limit during April 2011. Iodine-131 was detected until early June 2011, and long-lived radionuclides (¹³⁴Cs and ¹³⁷Cs) were detected intermittently for one year. Based on our results, gamma doses and committed effective doses resulting from inhalation were estimated.