Full Core Research Articles

Preliminary neutronic/thermal-hydraulic evaluation and safety system optimization of PWR loaded with fully ceramic microencapsulated fuel

As one of the Accident Tolerant Fuels (ATFs), Fully Ceramic Microencapsulated (FCM) presents advantageous safety performance in the severe accidents compared with UO2 owing to its high melting point, superior heat transfer ability and proliferation resistance capability. In this study, preliminary neutronic and thermal-hydraulic analysis of Chinese improved three-loop Pressurized Water Reactor (PWR) with FCM pellet and SiC cladding (FCM-SiC) were conducted to demonstrate the fuel performance of ATFs. First, a three dimensional full core model loaded with FCM-SiC was developed to acquire the cycle length, power distribution and isotopic inventories. And low reactor power and high fuel enrichment were used to satisfy the annual reload requirement. Then, the reactor thermal-hydraulic safety evaluation was carried out by the modified subchannel and system analysis codes, which depicts lower Fuel Centerline Temperature (FCT) and Peak Cladding Temperature (PCT) in FCM-SiC reactor compared to the reactor with traditional fuels. The Minimum Departure from Nucleate Boiling Ratio (MDNBR) of cladding is much higher than the criterion of 1.3 as well. Furthermore, the reactor safety system was simplified and optimized by taking full advantage of the preponderance of ATFs, and the results show that the reactor of FCM-SiC can operate safely with only 10% initial flowrate of Low Pressure Safety Injection (LPSI) or without Accumulators (ACCs) during Large Break Loss of Coolant Accident (LBLOCA), while the conventional zircaloy is failed under the same conditions. This work could provide guidelines for the optimized design of the reactor safety system in the future.

Full-Core Biopsy Systems Take Larger Liver Tissue Samples with Lower Fragmentation Rates Than Conventional Side-Notch Systems: A Randomized Trial.

BackgroundThe aim of this study was to compare the histopathological quality and physical features of the specimen of a full-core end-cut biopsy system with that of the standard side-notch system for liver biopsies.MethodsA full-core end-cut 16G biopsy device and a standard side-notch 16G needle were used to take biopsies of unclear liver lesions. Patients were randomized in two groups of 16 patients each. The primary endpoint of this prospective study was the core length measured using a dedicated microscope imaging software. Secondary endpoints were the quality of the specimen rated by an independent pathologist unaware of the device (scale from 1 to 5; with 1 as best and 5 as worst), the core diameter (determined by the microscopic imaging software) and presence of fragmentation (evaluated by the pathologist).ResultsFor the full-core (FC) and side-notch (SN) groups, the mean core length was similar with 13,599 μm and 11,570 μm (p=0.131), respectively. The quality of the specimen was significantly better in the FC-group with an average rating of 1.68 vs 2.50 (p=0.009). The fragmentation rate in the FC-group was statistically significantly lower at 2/27 (7%) than in the SN-group at 13/33 (39%) (p=0.021). The diameter in the FC-group was 1042 μm vs 930 μm in SN-group (p=0.018).

Rod drop transient at VR-1 reactor – Experiment and Serpent transient calculation analysis

The rod drop experiment is an important transient in reactor operation. This study focuses on a comparison between experimental reactor physics data and respective calculation. The rod drop experiment was performed at the VR-1 reactor. A full core 3D model was used to calculate this transient using Serpent2. Additionally, a point kinetic solution is presented. The Serpent results were compared with the experiment and with the point kinetic calculation. Two different nuclear data libraries (ENDF/B-VIII.0, JEFF3.3) were used in the data sensitivity analysis. The Serpent 3D kinetics results were almost identical to the point kinetic equation solution. Experimental data differed in the rod worth. This study shows that the new Serpent dynamic toolkit provides an accurate description of the reactor behavior during this transient. The 3D calculation proved that the detector position during the transient has a direct impact on the measured control rod worth.

Analysis of the methods for group constants generation for calculation of a large SFR core using Serpent 2 and CriMR codes

This work aimed at generating homogenized group constants using the Serpent code and then using the CriMR diffusion code to model the large SFR OECD 3600 MWth MOX core. The results were compared with a full core reference Monte Carlo solution by Serpent. Reactivity feedback parameters were also considered. Generating the group constants from separate fuel assemblies allows for simultaneously carrying out calculations and then using the results as input in diffusion codes rather than waiting so long for a 3D full core Monte Carlo calculation to be completed. From the results of the integral parameters we see a close agreement in the calculation codes. The differences can be attributed to the errors that could arise from generating the constants from individual sub-assemblies. The differences in the underlying physics and approximations used in development of the codes could also be a factor. Another way the errors could be reduced is by checking to see that the sub-assembly configurations used in the non-multiplying zones are as close as possible to the real layout in a full 3D core.

Assessment of resources for primary health care: Implications for the revitalization of primary health care in Akwa Ibom, Nigeria

Context:Primary health care (PHC) is the cornerstone of the Nigerian National Health Policy. The national policy on PHC under one roof is undergoing implementation nationwide as a means of strengthening the PHC system. Akwa Ibom State (AKS) is set to commence full implementation of the policy.Aims:The aim of the study was to assess the existing human resource and infrastructure in PHC facilities in AKS.Settings and Design:A descriptive cross-sectional study was carried out in 18 facilities selected from the three senatorial zones of AKS, Nigeria.Subjects and Methods:A rapid assessment of selected PHC facilities based on a checklist adapted from the minimum standards for PHC as provided by the National PHC Development Agency. The results were analyzed using Excel and presented in tables.Results:A total of 18 health facilities were included in the study. Human resources available were 276 full time core health workers, of which 48 (17.4%) were volunteer workers. There was inequitable distribution in district and facility type as 122(44.2%) work in Ikot Ekpene Senatorial district and 242 (87.7%) of them work in the Operational Base. Basic lifesaving equipment such as resuscitation sets was unavailable in more than 50% of the health facilities.Conclusions:There are absolute deficit and inequitable distribution of available human resources in AKS PHCs. Basic-lifesaving equipment is grossly inadequate. There is an urgent need for more health workers to be employed and provision of basic equipment for the PHCs.

Modelling and analysis of fuel assembly vibrational modes in PWRs using SIMULATE-3K

Some of the KWU pre-KONVOI PWRs operating across Europe saw a systematic increase in the neutron noise levels over several cycles in the last decade, and subsequently, core internals’ movements, especially vibrations of fuel assemblies with specific designs were identified as one of the plausible causes. Therefore, it is important to develop computational methods that can allow to investigate and predict the reactor noise response to fuel assemblies vibrations. To this aim, the 3D nodal reactor dynamics code SIMULATE-3K is used at PSI with a special module called the ‘assembly vibration model’ that imitates time-dependent motions of fuel assemblies by dynamically modifying the water-gaps surrounding the laterally moving fuel assemblies. The varying water-gaps are represented by the variation in the corresponding two-group macroscopic cross sections generated using the lattice code CASMO-5 in 2D. The studies conducted so far to assess the methodology for full core noise simulations were based on assuming vibrations of a clamped-free cluster of fuel assemblies that are unsupported from both ends. However, as this represents a non-physical movement, further developments were made at PSI to allow simulating more realistic movements of fuel assemblies such as the cantilevered mode vibration. The updated methodology, along with evaluations of the simulated noise response to realistic vibration modes, is presented in this paper. Results show that, as expected, the radial and axial neutron noise behaviour follow the vibration pattern of the imposed time-dependent axial functions corresponding to the natural oscillation modes of the fuel assemblies, thereby providing confidence in the application of the developed methodology for numerical neutron noise analyses of the PWR cores.

Preliminary study on the online reprocessing and refueling scheme for SD-TMS reactor

The Single-fluid Double-zone Thorium Molten Salt reactor (SD-TMSR) has a promising future. Inherent safety, liquid fuel, and possibility of online reprocessing and refueling are unique potentials of the SD-TMSR. In the present study, the Monte Carlo code Serpent2 has been chosen to model the full core of the SD-TMSR. In addition, online reprocessing and refueling has been modeled based on subroutine provided by Serpent2. During the burnup time (10 years), the total mass of the fuel inside the core has been checked and found to be almost constant. Furthermore, we controlled the reactivity by adjusting the feed rate of fissile and/or fertile materials. The change of Keff, breeding ratio, Th and U-233 refill rates with burn-up time have been investigated. Additionally, the build-up of U-233 during 10 years has been calculated.

Improving the Mechanical Behavior of a Helicopter Tail Rotor Blade Through the Use of Polyurethane Foams

The tail rotor blade represents a very important component in the structure of any helicopter, being subjected to many aerodynamic loads. The integrity of this component is essential for maintaining flight safety and improving its characteristics can increase the overall performance of the helicopter. Thus, in some cases, some improvements can be made to the structure, in order to reduce the weight of the component and to improve its mechanical properties and its reliability in flight. The current paper presents a comparative result analysis in the study of the tail rotor blade of the IAR 330 helicopter, where the honeycomb hexagonal core made of aluminum alloy is replaced with a full structural core made of polyurethane foam. The result of the analysis is of practical importance, because it shows the benefits of using polymers in the construction of helicopter rotor blades and similar structures.

Improving the Mechanical Behavior of a Helicopter Tail Rotor Blade Through the Use of Polyurethane Foams

The tail rotor blade represents a very important component in the structure of any helicopter, being subjected to many aerodynamic loads. The integrity of this component is essential for maintaining flight safety and improving its characteristics can increase the overall performance of the helicopter. Thus, in some cases, some improvements can be made to the structure, in order to reduce the weight of the component and to improve its mechanical properties and its reliability in flight. The current paper presents a comparative result analysis in the study of the tail rotor blade of the IAR 330 helicopter, where the honeycomb hexagonal core made of aluminum alloy is replaced with a full structural core made of polyurethane foam. The result of the analysis is of practical importance, because it shows the benefits of using polymers in the construction of helicopter rotor blades and similar structures. Keywords: helicopter tail rotor blade, honeycomb structure, polymers, polyurethane foam, finite element method

Measurement challenge: protocol for international case–control comparison of mammographic measures that predict breast cancer risk

IntroductionFor women of the same age and body mass index, increased mammographic density is one of the strongest predictors of breast cancer risk. There are multiple methods of measuring mammographic...

Improvement of MARS-KS sub-channel module using CTF code

In this work, the sub-channel analysis module of MARS-KS nuclear safety analysis code was upgraded using CTF sub-channel analysis code for realistic prediction of the multi-component and multi-dimensional nuclear core thermal hydraulics. The semi-implicit pressure coupling method was used to couple these two codes. Both codes use the finite difference method and semi-implicit numerical scheme and hence they were coupled together by exchanging volume/cell pressures through the boundary interfaces between 1D (i.e., system) and 3D (i.e., sub-channel) computational regions via implicit phasic velocities at the boundary. The pressure linkage was simply accomplished by designating the 1D and 3D regions as implicit sinks in the CTF sub-channel module and time dependent volumes in the MARS-KS code, respectively. The heat structure component and point kinetics models of the MARS-KS code were also coupled with the CTF core sub-channel module for the analysis of the whole reactor system. Automatic initialization of the 3D vessel was performed with CTF null-transient calculation to initialize all computational cells using a single set of initial and boundary conditions. The improved MARS-KS code has been verified and validated using various problems and tests. This paper selectively presents the results of three analyses, including a conceptual manometer oscillation, a blowdown test, and a 3 × 3 hot rod sub-channel experiments, with the improved code. All the analyses showed good predictions and this confirms the semi-implicit pressure coupling as well as the capability of the improved MARS-KS code. In the future, parallel processing should be implemented for the improved MARS-KS core sub-channel module to perform full core sub-channel analysis.

Non-equilibrium phase behavior in gas condensate depletion experiments

Full diameter core depletion tests are performed in this study to investigate the influence of depletion rate on the production of gas and condensate. Experimental results confirm that the gas recovery of the fast-depletion test is lower than that of the slow-depletion test. On the contrary, without the consideration of the fluid flow in porous media, the condensate recovery of the fast-depletion test is higher than that of the slow-depletion test.Non-equilibrium phase behavior is claimed to explain this abnormal behavior. The non-equilibrium model is adopted in this paper based on the components’ transfer rates between oil and gas phases. In addition, with the consideration of the compositional change in gas-condensate production, a numerical model is built to simulate the depletion process. Based on the comparison between the simulation and experimental results, the influence of non-equilibrium phase behavior on gas-condensate production is analyzed. If the depletion rate is relatively high, the heavy components of the gas-condensate system tend to be produced within the gas phase instead of condensing inside the core due to incomplete vapor-liquid separation. As a result, more condensate is acquired from the fast-depletion test.

Validation of UNIST Monte Carlo code MCS using VERA progression problems

Abstract This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores.

Molten salt reactor core simulation with PROTEUS

The Nuclear Energy Advance Modeling and Simulation Program (NEAMS) of the US Department of Energy (DOE) aims to develop advanced simulation tools to aid the development of advanced nuclear technologies. The high-fidelity PROTEUS code was developed in the NEAMS program for reactor physics analysis. This paper tests the applicability of the PROTEUS code to simulate a Molten Salt Reactor (MSR). The OpenMC code based on the Monte Carlo Method was employed to generate the cross sections and to provide stochastic solutions for verification. The PROTEUS code was successfully employed to simulate the steady-state performance of an MSR represented by the ThorCon core design. The 2D lattice, 2D core, and 3D full core calculations performed by PROTEUS show good agreement with that predicted by OpenMC. For the nodal option within PROTEUS, the precursor drift model was also tested, and this effect resulted in a reduction of the k-eff by 300 pcm.

Value and Diagnostic Accuracy of Ultrasound-Guided Full Core Needle Biopsy in the Diagnosis of Lymphadenopathy: A Retrospective Evaluation of 793 Cases.

Whole surgical lymph node excision (SNE) is considered the standard diagnostic method in the primary diagnosis of lymphadenopathy (LA) suspected of malignancy. Ultrasound-guided full core needle biopsy (UFCNB) offers an alternative method to SNE. This study examined the accuracy of UFCNB in the diagnosis of unexplained LA in 793 cases. From January 2006 to June 2015, a total of 793 cases of LA of unknown origin received a UFCNB. The lymph nodes were located peripherally (68%) or abdominally (32%). The final diagnoses from histopathologic examinations were non-Hodgkin lymphoma (n = 245), Hodgkin lymphoma (n = 53), solid nonlymphocytic lymph node metastases (n = 359), and benign LA (n = 136). The results of the biopsies were retrospectively evaluated with regard to sensitivity, specificity, and diagnostic accuracy. In the total collective of 793 biopsies, the sensitivity of UFCNB was 94.4%; the specificity was 97.8%; and the diagnostic accuracy was 95.0%. In the subgroups, the following results were obtained: non-Hodgkin lymphoma (sensitivity, 97.2%), Hodgkin lymphoma (sensitivity, 88.7%), metastases (sensitivity, 93.3%), and benign LA (specificity, 97.8%). In 17 cases (2.2%), an additional rebiopsy of the lymph node was needed, and in 85 cases (10.7%), an additional SNE was performed. Due to the diagnostic accuracy of 95.0% in the total collective, UFCNB seems to be an alternative diagnostic procedure to the standard procedure of SNE for LA of unknown origin. A prospective comparative study to definitively clarify the diagnostic value of UFCNB compared to SNE in the unexplained LA is warranted.

Rosette core fungal resistance in Arabidopsis thaliana.

Unlike rosette leaves, the mature Arabidopsis rosette core can display full resistance to Botrytis cinerea revealing the importance for spatial and developmental aspects of plant fungal resistance. Arabidopsis thaliana is a model host to investigate plant defense against fungi. However, many of the reports investigating Arabidopsis fungal defense against the necrotrophic fungus, Botrytis cinerea, utilize rosette leaves as host tissue. Here we report organ-dependent differences in B. cinerea resistance of Arabidopsis. Although wild-type Arabidopsis rosette leaves mount a jasmonate-dependent defense that slows fungal growth, this defense is incapable of resisting fungal devastation. In contrast, as the fungus spreads through infected leaf petioles towards the plant center, or rosette core, there is a jasmonate- and age-dependent fungal penetration blockage into the rosette core. We report evidence for induced and preformed resistance in the rosette core, as direct rosette core inoculation can also result in resistance, but at a lower penetrance relative to infections that approach the core from infected leaf petioles. The Arabidopsis rosette core displays a distinct transcriptome relative to other plant organs, and BLADE ON PETIOLE (BOP) transcripts are abundant in the rosette core. The BOP genes, with known roles in abscission zone formation, are required for full Arabidopsis rosette core B. cinerea resistance, suggesting a possible role for BOP-dependent modifications that may help to restrict fungal susceptibility of the rosette core. Finally, we demonstrate that cabbage and cauliflower, common Brassicaceae crops, also display leaf susceptibility and rosette core resistance to B. cinerea that can involve leaf abscission. Thus, spatial and developmental aspects of plant host resistance play critical roles in resistance to necrotrophic fungal pathogens and are important to our understanding of plant defense mechanisms.

Research on the on-the-fly homogenization method based on RMC code for criticality calculations

The applications of traditional deterministic lattice codes are limited in certain cases because they cannot treat closely coupled neutron characteristics and peculiar geometry. Therefore, to use the continuous energy Monte Carlo method and full core homogenization to generate multi-group cross sections has attracted growing interests. In this paper, the on-the-fly homogenization method based on the Monte Carlo code RMC, which generates multi-group cross sections in the frame of full core continuous energy Monte Carlo calculation and then use them in the Monte Carlo multi-group calculation to obtain a more convergent solution, is proposed and implemented. Tests and comparisons for the use of such on-the-fly homogenization method in criticality calculations are conducted using Kaist 1A benchmark and a simplified Lead-cooled reactor. The results show that our on-the-fly homogenization method exhibits similar accuracy with continuous energy Monte Carlo calculation that uses the same number of neutron histories and only costs about half of the computational time. With respect to spatial homogenization schemes, the cell homogenization scheme is the best one which can predict the reference solution especially for problems with strong absorbers. Given a tradeoff between efficiency and accuracy, the hybrid Homogenization scheme, which treats strong absorbers with cell homogenization and others with pin homogenization, can take nearly as much time as the pin Homogenization while achieving nearly the same accuracy as the cell Homogenization.

Fuel performance optimization of U3Si2-SiC design during normal, power ramp and RIA conditions

Benefited from its higher mechanical strength and much-improved oxidation resistance at very high temperatures (>1200 °C), SiC cladding is one of the promising accident tolerant fuel (ATF) concepts. Considering the thermal conductivity degradation of SiC cladding after irradiation and its negligible creep rate, adopting conventional UO2 with SiC cladding can lead to aggravated fission gas release and elevated fuel centerline temperatures. In this work, U3Si2 fuel with its high thermal conductivity is considered in place of UO2, when adopting SiC cladding. First, the preliminary full core simulation for U3Si2-SiC was conducted to show its holistic fuel performance compared to the current UO2-Zircaloy fuel. Then the most limiting fuel rod was selected based on the calculated chemically vapor deposited (CVD) SiC failure risk from the full core simulation. For the limiting rod, fuel-clad gap optimization was conducted to minimize CVD SiC failure risk. Subsequently, power ramp history was appended to the normal power history and full core simulation analysis for updated U3Si2-SiC design was re-performed to ensure the effectiveness of the chosen gap width. For the limiting rod, reactivity-initiated accident (RIA) analysis at hot zero power (HZP) was analyzed to evaluate its performance and infer its potential failure modes. The results from normal operation and power ramp simulations implied that CVD SiC failure risk was mainly induced by the strong pellet-cladding mechanical interaction (PCMI) and the failure probability of CVD SiC can be effectively minimized to 2.0×10-7 through enlarging the fuel-clad as-fabricated gap width. The leading failure mode for U3Si2-SiC fuel system during RIA changed from fuel melting to Ceramic Matrix Composite (CMC) failure with increasing burnup.

The effect of Am241 on UK plutonium recycle options in thorium-plutonium fuelled LWRs – Part II: BWRs

UK plutonium is expected to be managed using uranium-plutonium (U-Pu) mixed oxide (MOX) fuels in Light Water Reactors (LWRs). However, studies have shown that thorium-plutonium (Th-Pu) may be preferential. Part I of this study considered the effect of americium (Am) in UK Pu in Pressurized Water Reactors (PWRs) and found that, while the reactivity response was sensitive to isotopic and spectral variations, trends were predictable. Part II focusses on separation of Am in Boiling Water Reactors (BWRs) and compares fuel performance to the uniformly distributed and spatially separated cases outlined in Part I. Comparable incineration rates are achievable but, while a single PWR assembly bears a greater mass of Am/Pu than a single BWR assembly, the full BWR core may be capable of operating with significantly greater fissile masses. Transmutation of Am241 to Am242 appears preferable to fast fission of Am241 as increased incineration rates occur in lower void, bottom-of-assembly locations.

Computational flow and heat transfer design and analysis for 1/12 gas-cooled space nuclear reactor

The computational simulations of 1/12 full core gas-cooled space nuclear reactor have been conducted using the STAR-CCM+ code. The reactor is cooled by the He-Xe mixture and coupled with direct Brayton turbo-machines. To analyze the flow and heat transfer characteristics of the reactor, a series of calculations with complex core structures under zero gravity condition are conducted. The influences of the bypass vessel cooling, surface radiation, flow channel blockage and the fuel rods are studied. Results show that, although the bypass vessel cooling can decrease the temperature of the reactor vessels and the heat loss through surface radiation, it also increases the highest coolant temperature. Besides, the flow channel blockage will increase the uniformity of the radial temperature and velocity distribution. Finally, it is presented that the peak temperature is higher than the case without fuel rods and the fuel structure has little effects on the flow and heat transfer characteristics of the coolant region.