Fission Product Inventory Research Articles

Source term analysis of FeCrAl accident tolerant fuel using MELCOR

It has been established that incremental improvements in beyond design basis accident performance can be achieved through accident tolerant fuel (ATF). However, they have the potential to recover margin with respect to conventional fuel and therefore enhance plant economics through uprate or cycle length increase. To realize this potential, it is necessary to quantify the reduction in source term due to use of ATF, and correspondingly how this is affected by increasing the cycle length and/or burnup. This requires development of a risk informed analysis methodology for ATF under high burnup conditions, which is being developed within LWRS. To this end, a MELCOR simulation of FeCrAl ATF in a recovered Large Break LOCA (LBLOCA) scenario has been developed, using a model based on the Zion Pressurized Water Reactor (PWR). The new user defined material capability, along with the inclusion of detailed neutronics- and depletion-derived parameters such as core power distribution, decay heat behavior, and fission product inventories, allows a more detailed simulation of FeCrAl clad material properties and behavior than has previously been possible using MELCOR. These detailed FeCrAl results were compared with a zircaloy clad model to investigate the differences between the two clad types and quantify the benefits of FeCrAl ATF with an 18-month cycle. Next, the fuel cycle with ATF was extended to 24 months, to determine whether any additional safety margin provided by FeCrAl ATF could be leveraged to implement high burnup FeCrAl-clad fuel while retaining the same operating and safety limits as current zircaloy-clad fuel. For the particular scenario analyzed, the delay to fuel failure from using ATF was of the same order as the LOCA recovery time, and hence significant in reducing fission product release, with the 24-month FeCrAl core performing better than the 18-month zircaloy core. It is noted that for other transients, the reduction in release due to using FeCrAl may be less significant. The material model developed here can be used in such further studies in support of determining the overall source term.

Improved Analysis of the Short-Term Station Blackout Accidents of the Peach Bottom Unit-2 Reactor with ASTEC Including Radiological Impact and Statistical Analysis with JRODOS

After the severe accident at Fukushima, the importance of BWR design and related structures and their contribution to the severe accident progression has increased. Fuel channel boxes, absorber crosses, water rods, and smaller primary containment design of the BWR have been considered in the ASTEC code to increase the knowledge of BWR design and associated models. The previously developed ASTEC model for Peach Bottom Unit-2 was updated to include modern GE14 10x10 fuel assemblies with realistic fission product inventories. The CASMO5 code predicted the fission product inventory and burnup for GE14 10x10 fuel assemblies based on real plant data obtained from the ENRESA samples. First, a Short-Term Station Blackout (ST-SBO) analysis was performed to compare the impact of the old and new fuel assembly designs on the accident progression and radiological consequences. Second, a short-term station blackout with a stuck open safety relief valve (ST-SBO SOSRV) was considered for modern fuel assemblies. The actuation of the safety valve resulted in a much lower corium ejection and a longer transient to basemat failure in the cavity. The scatter of corium ejection amounts between the considered scenarios showed the importance of the design of the bottom head and the penetration points in BWRs. In both cases, failure of the drywell head flange and release of radionuclides to the environment occurred. Higher burnup and radionuclide inventory in the 10x10 assemblies resulted in a larger release compared to the previous design. Radiological analysis using JRODOS was performed for both cases and the maximum total effective gamma dose rate was estimated to be 67.89 mSv/h and 119.46 mSv/h for ST-SBO and ST-SBO SOSRV, respectively. The statistical analysis and the number of records in the considered cities around the Peach Bottom Plant showed the distribution over the region and the risk factors of the populated cities. The collaborative use of three codes in this study allows users to identify the fission product inventory with CASMO5 and investigation of the severe accident scenario with ASTEC and identifying radiological impact of the released radioactive isotopes to the environment with JRODOS code.

Reactor antineutrinos and novel application to real-time remote monitoring of nuclear reactors

ABSTRACT We have constructed a computational scheme to obtain energy spectra of reactor antineutrinos, including the inverse β -decay process through which they were to be detected. The antineutrino spectra from all the fission products were calculated based on the gross theory of β -decay. This data was combined with an inventory of fission products obtained by burnup calculations, and then multiplicity and reactor antineutrinos spectra were obtained by a summation calculation. We compared aggregate antineutrino spectra emitted from four major actinoids 235 U , 238 U , 239 Pu , and 241 Pu calculated in this work with experimental values. We found that those obtained by using Fission-Product Yield (FPY) data of JENDL-5 could reproduce the measured data better than or almost equivalently with those obtained with FPY data of other recent nuclear data libraries. Then, we proposed a new method of nuclear safeguards by using reactor antineutrinos; namely, we found that changes in the number and energy spectra of reactor antineutrinos were strongly correlated with burnup and fissile composition of light-water reactors.

Microstructure of irradiated AGR TRISO particle buffer layers as measured by X-ray computed tomography

Shrinkage of the initially low-density buffer layer in tristructural isotropic (TRISO) coated fuel particles during irradiation is a well-known phenomenon with potential implications for fission product and actinide transport, as well as potential pyrocarbon fracture that in some cases has been observed to impact particle performance. During post-irradiation examination, the buffer layer's structure is commonly determined using 2D microscopy of a particle cross section or a series of particle cross sections at staggered depths. Although these methods provide a general idea of the irradiated buffer microstructure, they do not provide a full picture of the TRISO coating microstructure. By contrast, x-ray computed tomography (XCT) provides full 3D imaging of the TRISO particle. Particles from various compacts irradiated for the Advanced Gas Reactor Fuel Development and Qualification Program were imaged using XCT. Some of these particles were selected because of abnormal fission product inventories (e.g., low 137Cs), whereas others were randomly selected from the center of the fission product inventory distribution. Qualitative and quantitative analysis techniques were applied to the randomly selected particles as representatives of typical TRISO behavior to study the post-irradiation structure of the buffer layer. These results showed that, while separation of the buffer and inner pyrolytic carbon layers was a common behavior during the radiation-induced shrinkage of the buffer, a portion of the original buffer/inner pyrolytic carbon interface remained intact throughout irradiation in nearly all cases. These results also indicated clear trends in the degree of buffer densification with irradiation temperature and fluence.

Development and verification of fast reactor burnup calculation module FRBurner in code system CBZ

ABSTRACT CBZ is a general-purpose reactor physics analysis code system, and FRBurner, which focuses on fast reactor burnup calculations, was developed recently with diverse combinations of available methodologies. Verification of this module is conducted with the OECD/NEA fast rector benchmark since this benchmark provides various types of fast reactors. Four key reactor physics parameters, effective neutron multiplication factor k eff, effective delayed neutron fraction β eff, sodium void reactivity ∆ρ void, and Doppler reactivity ∆ρ Doppler are the focus and compared to two references provided by JAEA and CEA, respectively. The biases between the results from FRBurner and the JAEA and CEA references on each of the above key parameters are less than 0.5%, 1%, 3% and 7%, and less than 1.0%, 4%, 12%, and 12%, respectively. The comparison indicates that the FRBurner module would provide acceptable results for general-type fast reactor physics analysis in research. As one innovation, the detailed burnup chain model, which is significantly different from a generally used pseudo fission product model in fast reactor neutronic analysis, is applied in FRBurner. The detailed burnup chain model helps FRBurner explicitly provide information about the inventory of fission products for nuclear waste management and spent fuel reprocessing.

Redistribution of radionuclides in irradiated AGR-1 UCO TRISO fuel after 1800 °C safety testing

Release of radionuclides from intact tristructural-isotropic (TRISO) coated particle fuel at normal and accident conditions is a primary metric of fuel performance. The distribution of fission products and actinides in the TRISO layers of individual particles provides insight on radionuclide transport and release behavior and was determined using scanning electron microscopy analysis. Particles were isolated from an irradiated fuel compact (AGR-1 Compact 4-4-2) and analyzed as-irradiated or after individual particle safety-testing at 1800 °C for 650 h. Particles were selected for comparison based on their remaining 110mAg fission product inventory. These comparisons corroborated the observation that the 110mAg inventory is a marker for relative irradiation temperature based on observed radionuclide distribution in the SiC layer. The comparison also indicated that the in-pile behavior influences the fission product and actinide species interactions with the TRISO layers during high temperature exposure after irradiation. The analysis confirms both palladium and uranium diffusion, as well as other species, are active in the UCO TRISO fuel system at 1800 °C and that palladium transport is active at lower temperatures relative to uranium. While diffusion across the SiC layer was observed, the intact nature of the SiC layer after the 1800 °C, 650-h exposure indicates the SiC layer maintained its functionality as a fission product barrier by mitigating release of radionuclides at beyond accident margin temperatures.

Performing a multi-unit level-3 PSA with MACCS

MACCS (MELCOR Accident Consequence Code System), WinMACCS, and MelMACCS now facilitate a multi-unit consequence analysis. MACCS evaluates the consequences of an atmospheric release of radioactive gases and aerosols into the atmosphere and is most commonly used to perform probabilistic safety assessments (PSAs) and related consequence analyses for nuclear power plants (NPPs). WinMACCS is a user-friendly preprocessor for MACCS. MelMACCS extracts source-term information from a MELCOR plot file.The current development can combine an arbitrary number of source terms, representing simultaneous releases from a multi-unit facility, into a single consequence analysis. The development supports different release signatures, fission product inventories, and accident initiation times for each unit. The treatment is completely general except that the model is currently limited to collocated units.A major practical consideration for performing a multi-unit PSA is that a comprehensive treatment for more than two units may involve an intractable number of combinations of source terms. This paper proposes and evaluates an approach for reducing the number of calculations to be tractable, even for sites with eight or ten units. The approximation error introduced by the approach is acceptable and is considerably less than other errors and uncertainties inherent in a Level 3 PSA.

Fission product release tests with uranium dioxide fuel and lead under conditions related to reactor accidents for lead-cooled reactors

Tests of fission product release under accident conditions were performed on three samples of irradiated UO2 fuel fragments immersed in molten Pb. Two tests of bare irradiated UO2 fragment samples were conducted under similar conditions for direct comparison. The irradiated fuel samples had been cooled for several years, then encapsulated and trace-reirradiated to populate limited inventories of short-lived fission products (such as 133Xe and 131I). The tests were conducted with sample temperatures between 1570 and 1590 K. The releases from Pb-covered samples under inert conditions were similar to those from bare UO2 under inert conditions. Greater 133Xe releases were observed from Pb-covered samples in oxidizing environments. Release of 140Ba was detected for the Pb-covered sample exposed to air. Under oxidizing conditions, Pb-covered samples retained more 137Cs and 103Ru (delayed onset of release in air and complete retention in steam) as well as 131I.

A code for simulating a high-resolution gamma-ray spectrum of a fission sample

A numerical model was developed that fully reconstructs a high-resolution gamma-ray spectrum of a fission sample. The code combines GEANT4 Monte Carlo simulations (used to calculate the response of a germanium detector) with the in-house Koala package (used to calculate the fission product inventory). In order to validate the model against experimental data, a highly enriched uranium sample was irradiated in a reactor and measured for up to 31 days later. A comparison between the simulated spectra and the experimental spectra exhibited a good quantitative agreement, demonstrating the high reliability of the model, especially for early times. While a simple source geometry was used for benchmarking the developed code, it can be easily extended to include complex geometries. The synthetic fission spectra can be used in several applications, e.g. assessing the capability of detector systems in identifying certain radionuclides, verification of gamma spectrometry analysis software and proficiency tests for analytical laboratories.

Discussion on the accident behavior and accident management of the HTGR

High Temperature Gas-cooled Reactor (HTGR), which plays an important role in the worldwide development of Generation-IV nuclear energy technology, is well-known for the high safety, high efficiency and process heat application. A commercial-scale 200 MWe High Temperature gas-cooled Reactor Pebble-bed Module (HTR-PM) has been designed and is now under construction in Shandong Province, China. Most of the construction and installation work have been finished and the connection to the electric grid will be expected in the end of 2020.In this paper, the reactor behaviors of the HTR-PM during the typical design basis accidents (DBAs) and beyond design basis accidents (BDBAs) have been introduced. In the DBAs, the maximum fuel temperature will never exceed its design limitation of 1620 °C, below which under the design burn-up the “Tristructural-isotropic” (TRISO) coated particles can effectively retain their fission-product inventory. Even in the typical BDBAs with extremely low probability, there is enough time, e.g. several days, to adopt appropriate measures to mitigate the consequence, so that the large release of the radioactive materials would not happen.Accident management is important for the nuclear power plant. Besides, After the Fukushima Dai-ichi nuclear accident, in the worldwide, people are more concerned about the severe accident management of the nuclear power plant, and some standards and guidelines are issued. Based on the accident analyses of the HTR-PM, its accident management is studied and discussed. The designers believe that the accident management procedures can be simplified and no offsite emergency measures are needed for the HTR-PM due to the inherent safety design. Compared to the other types of the nuclear power plant, e.g. the Pressurized-Water Reactor (PWR) power plant, the Severe Accident Management Guideline (SAMG) can be simplified or even unnecessary for the HTR-PM.Above work also can provide reference for the further study on the safety regulations, standards or guidelines for the design, operation and accident management of the HTGR.

Analysis of fission product gas pressure and radioactivity in SAMOP reactor experimental facility

The fission-product gas pressure and radioactivity analysis of a subcritical assembly for 99Mo production (SAMOP) experimental facility have been done. SAMOP reactor is fueled with low enriched uranyl nitrate solution UO2(NO3)2 of 300 g U/L, the reactor core is in the form an annular tube surrounded by a ring of fuel tubes. The SAMOP system is designed to be operated at 100 to 120 hours periodic operation per batch. The analysis method is done by using ORIGEN2 computer code, at the condition of the maximum fuel temperature of 54 °C, 120 hours reactor operating time, and neutron flux varied from 1010 to 1012 n/cm2s. The result show that the inventory of the gaseous fission products is dominantly consisted of the xenon, radium, helium, and tritium isotopes with the total activity of 1.3775x1013 Bq (372.29 Ci) and total gas volume of 0.0986 cm3 or giving a total gas pressure of 5.78 kPa. The isotopes of xenon and krypton are the major contributors to the total gas pressure. The analysis result shows that the gas pressure of the fission products is very small, therefore the pressure monitoring in the SAMOP reactor core is not necessary.

Analysis for decay heat in natural uranium fueled Pressure Tube type Heavy Water Reactor using the indigenous computer code IGDC

The accurate estimation of decay heat from spent fuel is important for safety studies and fuel cycle simulations. The fuel cycle analysis code, Isotopic Generation and Depletion Code (IGDC) has been successfully used to simulate the actinide and fission product inventory for the spent fuel from a 220 MW(e) Pressure Tube type Heavy Water Reactor (PT-HWR). In addition to the computation of isotopic inventory, decay heat estimation is also an important characteristic of the computer code IGDC. A case study was taken up where decay heat from spent fuel of 220 Mw(e) class PT-HWRs, using natural uranium fuel is done and presented in this paper.This decay heat or residual heat in the discharged fuel is a function of power history of the bundle inside the reactor. This paper highlights the decay heat due to different radioisotopes and their variation with time after discharge of the bundles taking into account the power history. The simulations were done for 220 MWe capacity PT- HWR at a discharge burn-up of 7500 MWd/te burn-up. Another highlight of this paper is that the code IGDC has been benchmarked with the experimental results of the decay heat measured for Douglas point reactor fuel of varying operating history. Such detailed estimations for the spent 19-element natural UO2 bundles have been done for the first time with indigenously developed code IGDC.

Modeling and Analysis of Radionuclide’s Transport and Source Term Evaluation Within Containment and Release to the Environment for Research Reactors

The subject of this work is to carry out a source term assessment for a Hypothetical Research Reactor (HRR) and the calculation of the radiological consequences in case of Loss Of Coolant Accident (LOCA), setting out the assumptions made and the bases for these assumptions. In the Safety Analysis Report, the maximum hypothetical accident was analyzed following these steps: -Estimation of fission product inventory in the core -Estimation fission product release to the building -Building release evaluation -Downwind doses calculation: evaluation of dose consequences from a release; In the event of an accident in a nuclear reactor, the authorities should be in position to implement the counter-measures necessary to protect the surrounding population from radiological and the environment consequences of any releases. It is therefore necessary to monitor the progression of the accident as soon as it is detected by the plant operator in order to forecast the future behavior of the reactor so as to be able to recommend to the government authorities the implementation of counter-measures within a time compatible with the control of the risk to the population. For that purpose, it is necessary to estimate the nature, amount and kinetics of the radioactive products likely to be released out of the installation. Then, it would be possible to use PC-COSYMA codes to decide the adapted counter-measures.

Regulatory Assessment on a New utilization of SAMOP Test Facility: Determination on Fission Power

SAMOP test facility is a subcritical solution of uranyl nitrate to produce Mo99 that is connected with external neutron source from one of Kartini reactor beam ports. Determining the SAMOP fission power is a critical step toward thermal hydraulic calculation ensuring the adequacy of cooling and to calculate the fission product inventory without modelling the Kartini reactor geometry. As part of regulatory assessment, the analyses were intended to determine the power level of SAMOP and its distribution in its solution tubes using MCNP5 code. Two cases of Kartini reactor power 100 % and 110 % of nominal power were simulated. The results showed that SAMOP can produce fission power ranging from 563.2 up to 860.8 watt. Taking uncertainties into account, it is recommended to conduct thermal hydraulic calculations on SAMOP test facility based on minimum total power of 1000 W.

Calculation of fuel burnup and radionuclide inventory for the HEU and potential LEU fuels in the IRT research reactor

Abstract The study of core conversion of the IRT research reactor from the HEU to LEU fuels has been conducted in this paper. Fuel burnup, buildup of actinides and fission product inventory for the IRT research reactor with UO2-Al dispersion fuel for the various different uranium enrichments and the uranium density such as: (IRT-3M 90% 1.1 g/cm3, IRT-3M 36% 2.51 g/cm3, IRT-4M 19.75% 3.25 g/cm3, IRT-4M 19.75% 3.48 g/cm3, IRT-4M 19.75% 3.71 g/cm3) have been carried out using the GETERA code. The multiplication factor (keff) of the IRT-3M and IRT-4M reactors were calculated using the MCNP4C code. The results showed, after 200 days of the reactor operation time that the total consumptions of 235U for the five cases were 73.85, 73.13, 72.66, 72.80 and 73.21 g respectively. The amounts of 239Pu produced in the core were: 0.43, 3.39, 5.90, 6.13 and 6.35 g respectively. The total radioactivities of fuel assembly were: 2.78E+15, 2.83E+15, 2.86E+15, 2.88E+15 and 2.89E+15 Bq respectively. Finally, from the neutronic view point, the LEU fuel was validated to be a suitable candidate for the fuel conversion of the IRT in the future.

Fission product inventory and burnup evaluation of the AGR-2 irradiation by gamma spectrometry

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program’s second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide/uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. The measured activities were used to calculate the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15% fission per initial metal atom (FIMA) and 9.01–10.69% FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1% FIMA for UCO compacts and 8.5–10.6% FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.

Thermal-hydraulic analysis of the LORELEI test design by means of CATHARE2 V2.5

LORELEI (Light Water One-Rod Equipment for LOCA Experimental Investigation) test device is currently under design in the framework of the planned experimental facilities of the Jules Horowitz Reactor. The objective of this devise is to analyze the thermal-mechanical behavior of the fuel rod and the radiological consequences during a Loss of Coolant Accident (LOCA). This paper deals with an accurate modeling of the LORELEI test device by means of the system code CATHARE2 V2.5. The experimental device and its geometry is described together with the experimental parameters and the different phases of the test planned accident. The modeling and testing are dedicated to the study of the re-irradiation and the dry phase. The re-irradiation phase is not a part of the LOCA accidental sequence but it is modeled to provide a study concerning the nominal and abnormal conditions by finding out the device thermal-hydraulic limits. In this first phase, aiming at creating a detectable fission product inventory, the device is filled with water and the fuel rod is cooled by a convection loop. During the dry-phase, the device simulates the first part of a LOCA accidental scenario. The fuel cladding temperature is increased up to 1200°C and a small amount of water is set in the bottom of the device to produce steam needed to investigate the exothermic reaction with the fuel cladding. The object of this last part is to plan the operating test sequences and determine the best thermal conditions to reach an acceptable fuel cladding temperature profile.

Source Term Analysis of Long-Life 620 MWt BWR During Loss of Coolant Accident (LOCA)

Source term analysis on a long-life 620 MWt BWR has been performed. The study includes fuel cell calculations, multi-groups diffusion equation calculation with 8 groups of energy, source term calculation and dose calculation. The scope of the source term analysis includes fission product inventories calculation in core, their activity and their toxicity during the Loss of Coolant Accident (LOCA). The calculation based on the assumption that the reactor has been operating for 10950 days or 30 years with full power of 620 MWt. The total activity of the core is 1,12x109 Ci, and the higher inventory of the core is dominated by the trans-uranium elements. The release of fission product activity to the atmosphere is 1,11x107 Ci. And the toxicity analysis shows that noble gas has the higher toxicity degree. Whole body dose calculation result in exclusion zone is 7.17x10-04 rem and the thyroid dose calculation result in this zone is 7.21x10-02 rem. And for the whole body dose calculation result in low population zone is 2.69x10-05 rem and the thyroid dose calculation result in this zone resulted 1.80x10-03 rem.

Assessment of fission product inventory considering axial burnup of a fuel assembly

The fission product inventory for a loss-of-coolant accident (LOCA) analysis is generally evaluated based on the average burnup of the fuel at the end of life regardless of the variation of the actual burnup with time. In this study, we analyze the variation of the fission product inventory based on the accident occurrence cycle and the degree of fuel meltdown, which reflect the axial burnup of the fuel at the beginning, middle, and end of the cycle (BOC, MOC, and EOC). The fuel assembly for the Hanul unit 6, which is an OPR1000-type pressurized water reactor (PWR) in Korea, was chosen as an analysis model for the source-term assessment. The burnup cross-section library was generated using the TRITON module of the SCALE6.1 code system, and the depletion calculations were carried out using ORIGEN-ARP. The first scenario, which is referred to as CASE (I), is applied with an average burnup at the EOC, which is a commonly recommended conservative assumption in accident analyses. The CASE (II) scenario includes various cases that consider the accident occurrence cycle and fuel meltdown from the top to the bottom of the core at the BOC, MOC, and EOC. As a result, the inventory of the main nuclides gradually decreased as the reactor core operation was performed. In addition, the activity intensity for the main fission products of CASE (II) was shown to be higher than that of CASE (I) in the range above a fuel meltdown of approximately 30%. Therefore, this study concludes that not all of the results obtained by the CASE (I) assumption are guaranteed to be more conservative than those obtained by CASE (II).

Utilizing the burnup capability in MCNPX to perform depletion analysis of an MNSR fuel

In this work, we present results of fuel depletion analyses performed for a potential LEU core of Ghana’s Miniature Neutron Source Reactor (GHARR-1) using the Monte Carlo N-particle extended (MCNPX) neutron transport code. Depletion calculation was carried out for the reactor core from the Beginning of Life (BOL) to the End of Life (EOL) which corresponds to 10years of reactor operation. The amounts of uranium and plutonium actinides were estimated at BOL and EOL of the core. Decay heat removal rate for the MNSR after reactor shutdown was investigated due to its significance to reactor safety. Inventory of fission products produced as a result of burnup was also calculated. The results show that a maximum discharge burnup equivalent to 0.568% of U-235 was consumed at EOL equivalent to operating the reactor for 200 Effective Full Power Days (EFPD), while the amount of Pu-239 produced was not significant. Also, the decay heat decreased exponentially after reactor shutdown confirming that decay heat will be removed in the system by natural circulation after shutdown and thus guaranteeing the safety of the reactor.