Generation Of Nuclear Power Plants Research Articles

An overview of the ITER cabling network and cable database management

ITER is a nuclear fusion facility, where many systems with varying functional and physical characteristics coexist. This makes the cabling requirements diverse too. The design of the ITER demands extensive integration of many systems interfaces. This process starts by collating cable requirement data to form a giant cable database with scope for approximately 120,000 cables in total. Ultimately the cabling installation will be approximately twice the size of one required for a 3rd generation nuclear fission power plant, with cables spanning more than 12,000km to be installed within approximately 200km of cable trays. Considering the differing cabling requirements, the raceway network and cables shall be routed in a prescribed manner to comply with French Nuclear Safety standards.This paper describes an overview of the ITER cabling network, the cable engineering design workflow, development and management of the cable database and integration between the software tools.

EFFECTS OF FORGING AND HEAT TREATMENTS ON STRESS CORROSION BEHAVIOR OF 316LN STAINLESS STEEL IN HIGH TEMPERATURE CAUSTIC SOLUTION

The reactor coolant piping in the third generation nuclear power plants of AP1000 is manufactured by integrally forging. Therefore, it is of vital importance to investigate the effects of forging and heat treatments on the stress corrosion cracking(SCC) resistance of 316 LN stainless steel(316LNSS), which is the candidate material for the reactor coolant piping in AP1000 nuclear power plants. In this work, electron back scattering diffraction(EBSD) and microhardness measurements(HV) were used to characterize the microstructure and residual strain of the as-received 316 LNSS, the forged and solution anneal treated 316 LNSS and the forged and stress relief treated 316 LNSS, respectively. The average grain size of the as- received 316 LNSS was the largest, and the forged 316 LNSS followed by solution anneal treatment and stress relief treatment showed no obvious differences on grain size. The as-received 316 LNSS exhibited the highest residual strain followed by the forged and stress relief treated 316 LNSS and then solution anneal treated 316 LNSS. Besides, the residual strain in the as- received316 LNSS concentrated on grain boundaries, while the residual strain in the forged and stress relief treated316 LNSS was characterized by a band-like distribution. The U-bend specimens were utilized to investigate the SCC behavior of the 3 kinds of 316 LNSS specimens in high temperature caustic solution. After SCC experiments, the crack morphologies of the 3 kinds of 316 LNSS specimens were examined by SEM. Then the macro and micro fracture morphologies were examined by OM and SEM, respectively. Grain morphology, residual strain and grain boundary character distribution near the SCC crack tip of the forged and stress relief treated 316 LNSS were investigated using EBSD. The results showed that the forged and solution anneal treated 316 LNSS exhibited the lowest SCC sensibility, while the as-received the highest, with the most cracks and the highest growth rate. The as-received and the forged and solution anneal treated 316 LNSS showed obvious intergranular cracking, while the forged and stress relief treated 316 LNSS showed a mixed cracking mode. The larger average grain size and higher residual strain, especially concentrating on the grain boundaries, were considered to be responsible for the highest SCC sensibility of the as-received 316 LNSS. Compared with the forged and stress relief treated 316 LNSS, the higher content of coincidence site lattice boundary(CSLB) and lower residual strain contributed to the lower SCC sensibility of forged and solution anneal treated 316 LNSS. The stress relief treatment failed to eliminate the band-like microstructure effectively, which disadvantaged the SCC resistance.

Neutron-induced fission cross section ofPu240from 0.5 MeV to 3 MeV

$^{240}\mathrm{Pu}$ has recently been pointed out by a sensitivity study of the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) to be one of the isotopes whose fission cross section lacks accuracy to meet the upcoming needs for the future generation of nuclear power plants (GEN-IV). In the High Priority Request List (HPRL) of the OECD, it is suggested that the knowledge of the $^{240}\mathrm{Pu}(n,f)$ cross section should be improved to an accuracy within 1--3 %, compared to the present 5%. A measurement of the $^{240}\mathrm{Pu}$ cross section has been performed at the Van de Graaff accelerator of the Joint Research Center (JRC) Institute for Reference Materials and Measurements (IRMM) using quasi-monoenergetic neutrons in the energy range from 0.5 MeV to 3 MeV. A twin Frisch-grid ionization chamber (TFGIC) has been used in a back-to-back configuration as fission fragment detector. The $^{240}\mathrm{Pu}(n,f)$ cross section has been normalized to three different isotopes: $^{237}\mathrm{Np}(n,f)$, ${}^{235}\mathrm{U}(n,f)$, and ${}^{238}\mathrm{U}(n,f)$. Additionally, the secondary standard reactions were benchmarked through measurements against the primary standard reaction ${}^{235}\mathrm{U}(n,f)$ in the same geometry. A comprehensive study of the corrections applied to the data and the associated uncertainties is given. The results obtained are in agreement with previous experimental data at the threshold region. For neutron energies higher than 1 MeV, the results of this experiment are slightly lower than the ENDF/B-VII.1 evaluation, but in agreement with the experiments of Laptev et al. (2004) as well as Staples and Morley (1998).

Unprotected Loss of Flow Accident in Small Long Life Gas Cooled Fast Reactor

In post Fukushima nuclear accidents inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Gas cooled fast reactors is one of the important candidate of 4th generation nuclear power plant and in this paper the safety analysis related to unprotected loss of flow in small long life gas cooled fast reactors has been performed. Accident analysis of unprotected loss of flow include coupled neutronic and thermal hydraulic analysis which include adiabatic model in nodal approach of time dependent multigroup diffusion equations. The thermal hydraulic model include transient model in the core, steam generator, and related systems. Natural circulation based heat removal system is important to ensure inherent safety capability during unprotected accidents. Therefore the system similar to RVACS (reactor vessel auxiliary cooling system) is investigated. As the results some simulations for small 60 MWt gas cooled fast reactors has been performed and the results show that the reactor can anticipate complete pumping failure inherently by reducing power through reactivity feedback and remove the rest of heat through natural circulations.

Steel-plate composite (SC) walls: Out-of-plane flexural behavior, database, and design

Steel-plate composite (SC) walls consist of exterior steel faceplates that are anchored to the concrete infill using steel headed stud anchors (or shear studs). The steel faceplates are connected to each other using tie bars that are also embedded in the concrete infill. SC walls are being used in the third generation of nuclear power plants and are also being considered for future small modular reactor designs. This paper focuses on the out-of-plane flexural behavior and design of SC walls. The out-of-plane flexural behavior was evaluated experimentally by conducting one-way bending tests on beam specimens that were representative of strips taken in the longitudinal and transverse directions of full-scale SC walls. This paper discusses the fundamental out-of-plane flexural behavior of SC walls, and presents the experimental database of out-of-plane flexural tests conducted on SC beam specimens in Japan, S. Korea, China and the US. The database includes SC beam specimens tested using different loading configurations, and considering different parameters such as shear span-to-depth ratio, steel faceplate thickness (reinforcement ratio), stud anchor spacing, and presence of faceplate stiffeners. The flexural capacities obtained from the tests are compared with nominal strengths calculated using appropriate equations in applicable design codes from Japan, S. Korea and the US. These comparisons indicate that the design code equations estimate the out-of-plane flexural strength of SC walls conservatively and with reasonable accuracy. Strength reduction factors (ϕ) are calculated by conducting reliability analyses, and found to be similar to the ones used for tension-controlled reinforced concrete flexural members in the ACI code.

Effect of Heat Treatment on Microstructure and Hardness of Grade 91 Steel

Grade 91 steel (modified 9Cr-1Mo steel) is considered a prospective material for the Next Generation Nuclear Power Plant for application in reactor pressure vessels at temperatures of up to 650 °C. In this study, heat treatment of Grade 91 steel was performed by normalizing and tempering the steel at various temperatures for different periods of time. Optical microscopy, scanning and transmission electron microscopy in conjunction with microhardness profiles and calorimetric plots were used to understand the microstructural evolution including precipitate structures and were correlated with mechanical behavior of the steel. Thermo-Calc™ calculations were used to support the experimental work. Furthermore, carbon isopleth and temperature dependencies of the volume fraction of different precipitates were constructed.

지진격리된 원전배관의 지진취약도 분석

Base isolation is considered as a seismic protective system in the design of next generation Nuclear Power Plants (NPPs). If seismic isolation devices are installed in nuclear power plants then the safety under a seismic load of the power plant may be improved. However, with respect to some equipment, seismic risk may increase because displacement may become greater than before the installation of a seismic isolation device. Therefore, it is estimated to be necessary to select equipment in which the seismic risk increases due to an increase in the displacement by the installation of a seismic isolation device, and to perform research on the seismic performance of each piece of equipment. In this study, modified NRC-BNL benchmark models were used for seismic analysis. The numerical models include representations of isolation devices. In order to validate the numerical piping system model and to define the failure mode, a quasi-static loading test was conducted on the piping components before the analysis procedures. The fragility analysis was performed by using the results of the inelastic seismic response analysis. Inelastic seismic response analysis was carried out by using the shell finite element model of a piping system considering internal pressure. The implicit method was used for the direct integration time history analysis. In addition, the collapse load point was used for the failure mode for the fragility analysis.

Correlation analysis for screening key parameters for passive system reliability analysis

Passive systems are widely used in new generation nuclear power plants to enhance their safety. Reliability of passive system operating based on natural circulation must be assessed in terms of functional failure. The functional failure probability evaluation requires repeatedly running the thermal–hydraulic (T–H) code which simulates the system responses under different values of the input parameters. In practice, repeated running of the code is quite costly in terms of running time and artificial neural network (ANN) has been proposed to replace the T–H model. However, the number of input parameters can be too large to satisfy the requirement of the ANN. In this paper we illustrate a systematic methodology to screen the key parameters for passive system operation based on correlation analysis for reducing the number of inputs.Correlation analysis is a well-known statistical method to assess the relationships among parameters. In the case of interest for passive system reliability, we consider the T–H model as a relationship between model inputs and outputs, which can be used in correlation analysis. With this method, key parameters can be screened with limited numbers of samples.The passive containment cooling system in AP1000 is analyzed and 4 parameters are identified as important ones from 47 inputs.

On the constitutive model of nitrogen-containing austenitic stainless steel 316LN at elevated temperature.

The nitrogen-containing austenitic stainless steel 316LN has been chosen as the material for nuclear main-pipe, which is one of the key parts in 3rd generation nuclear power plants. In this research, a constitutive model of nitrogen-containing austenitic stainless steel is developed. The true stress-true strain curves obtained from isothermal hot compression tests over a wide range of temperatures (900–1250°C) and strain rates (10−3–10 s−1), were employed to study the dynamic deformational behavior of and recrystallization in 316LN steels. The constitutive model is developed through multiple linear regressions performed on the experimental data and based on an Arrhenius-type equation and Zener-Hollomon theory. The influence of strain was incorporated in the developed constitutive equation by considering the effect of strain on the various material constants. The reliability and accuracy of the model is verified through the comparison of predicted flow stress curves and experimental curves. Possible reasons for deviation are also discussed based on the characteristics of modeling process.

Metrology for New Generation Nuclear Power Plants–MetroFission

Metrology for New Generation Nuclear Power Plants–MetroFission

Theory and implementation of nuclear safety system codes – Part II: System code closure relations, validation, and limitations

This is Part II of two articles describing the details of thermal-hydraulic system codes. In this second part of the article series, the system code closure relationships (used to model thermal and mechanical non-equilibrium and the coupling of the phases) for the governing equations are discussed and evaluated. These include several thermal and hydraulic models, such as heat transfer coefficients for various flow regimes, two phase pressure correlations, two phase friction correlations, drag coefficients and interfacial models between the fields. These models are often developed from experimental data. The experiment conditions should be understood to evaluate the efficacy of the closure models.Code verification and validation, including Separate Effects Tests (SETs) and Integral effects tests (IETs) is also assessed. It can be shown from the assessments that the test cases cover a significant section of the system code capabilities, but some of the more advanced reactor designs will push the limits of validation for the codes.Lastly, the limitations of the codes are discussed by considering next generation power plants, such as Small Modular Reactors (SMRs), analyzing not only existing nuclear power plants, but also next generation nuclear power plants. The nuclear industry is developing new, innovative reactor designs, such as Small Modular Reactors (SMRs), High-Temperature Gas-cooled Reactors (HTGRs) and others. Sub-types of these reactor designs utilize pebbles, prismatic graphite moderators, helical steam generators, innovative fuel types, liquid metal coolants, and many other design features that may not be fully analyzed by current system codes.This second part completes the series on the comparison and evaluation of the selected reactor system codes by discussing the closure relations, validation and limitations. These two articles indicate areas where the models can be improved to adequately address issues with new reactor design and development.

Comparison of the radiological hazard of thorium and uranium spent fuels from VVER-1000 reactor

Thorium fuel is considered as a viable alternative to the uranium fuel used in the current generation of nuclear power plants. Switch from uranium to thorium means a complete change of composition of the spent nuclear fuel produced as a result of the fuel depletion during operation of a reactor. If the Th–U fuel cycle is implemented, production of minor actinides in the spent fuel is negligible. This is favourable for the spent fuel disposal. On the other hand, thorium fuel utilisation is connected with production of 232 U, which decays via several alpha decays into a strong gamma emitter 208 Tl. Presence of this nuclide might complicate manipulations with the irradiated thorium fuel. Monte-Carlo computation code MCNPX can be used to simulate thorium fuel depletion in a VVER-1000 reactor. The calculated actinide composition will be analysed and dose rate from produced gamma radiation will be calculated. The results will be compared to the reference uranium fuel. Dependence of the dose rate on time of decay after the end of irradiation in the reactor will be analysed. This study will compare the radiological hazard of the spent thorium and uranium fuel handling. • Spent thorium and uranium fuel composition in VVER-1000 was calculated by MCNPX. • Important nuclide 208 Tl is not included in the thorium spent fuel composition. • There are large differences in activity of actinides between spent Th and U fuels. • Dose rate from spent thorium fuel is increasing during 50 years of decay. • Tl 208 in spent thorium fuel increased the dose rate by several percent.

A Comparison of State Traditions and the Revival of a Nuclear Power in Four Countries

The British, American, French and Finnish governments are seeking to promote investment in a new generation of nuclear power plants. Nuclear power programmes are delivered through networks of international companies through which government must manage. This is consistent with the concept of governance. Governments can advance their policy goals by using a variety of policy instruments to shape and organize governance networks. This is known as metagovernance. The paper considers the extent to which the selection and deployment of the policy instruments used to metagovern is informed by the prevailing tradition of government. The paper examines how the British, American, French and Finnish governments have tried to metagovern. It is shown that whilst governing traditions do inform the selection and deployment of the policy instruments used to metagovern, the composition of the network, and the nature of the policy problem also plays a role in shaping government action.

PLEPS study of ions implanted RAFM steels

Current nuclear power plants (NPP) require radiation, heat and mechanical resistance of their structural materials with the ability to stay operational during NPP planned lifetime. Radiation damage much higher, than in the current NPP, is expected in new generations of nuclear power plants, such as Generation IV and fusion reactors. Investigation of perspective structural materials for new generations of nuclear power plants is among others focused on study of reduced activation ferritic/martensitic (RAFM) steels. These steels have good characteristics as reduced activation, good resistance to volume swelling, good radiation, and heat resistance. Our experiments were focused on the study of microstructural changes of binary Fe-Cr alloys with different chromium content after irradiation, experimentally simulated by ion implantations. Fe-Cr alloys were examined, by Pulsed Low Energy Positron System (PLEPS) at FRM II reactor in Garching (Munich), after helium ion implantations at the dose of 0.1 C/cm2. The investigation was focused on the chromium effect and the radiation defects resistivity. In particular, the vacancy type defects (monovacancies, vacancy clusters) have been studied. Based on our previous results achieved by conventional lifetime technique, the decrease of the defects size with increasing content of chromium is expected also for PLEPS measurements.

239Pu Prompt Fission Neutron Spectra Impact on a Set of Criticality and Experimental Reactor Benchmarks

A large set of nuclear data are investigated to improve the calculation predictions of the new neutron transport simulation codes. With the next generation of nuclear power plants (GEN IV projects), one expects to reduce the calculated uncertainties which are mainly coming from nuclear data and are still very important, before taking into account integral information in the adjustment process. In France, future nuclear power plant concepts will probably use MOX fuel, either in Sodium Fast Reactors or in Gas Cooled Fast Reactors. Consequently, the knowledge of 239Pu cross sections and other nuclear data is crucial issue in order to reduce these sources of uncertainty. The Prompt Fission Neutron Spectra (PFNS) for 239Pu are part of these relevant data (an IAEA working group is even dedicated to PFNS) and the work presented here deals with this particular topic. The main international data files (i.e. JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0, BRC-2009) have been considered and compared with two different spectra, coming from the works of Maslov and Kornilov respectively. The spectra are first compared by calculating their mathematical moments in order to characterize them. Then, a reference calculation using the whole JEFF-3.1.1 evaluation file is performed and compared with another calculation performed with a new evaluation file, in which the data block containing the fission spectra (MF=5, MT=18) is replaced by the investigated spectra (one for each evaluation). A set of benchmarks is used to analyze the effects of PFNS, covering criticality cases and mock-up cases in various neutron flux spectra (thermal, intermediate, and fast flux spectra). Data coming from many ICSBEP experiments are used (PU-SOL-THERM, PU-MET-FAST, PU-MET-INTER and PU-MET-MIXED) and French mock-up experiments are also investigated (EOLE for thermal neutron flux spectrum and MASURCA for fast neutron flux spectrum). This study shows that many experiments and neutron parameters are very sensitive to the PFNS, in particular for high leakage thermal criticality cases for which the discrepancy between international evaluation files spectra and Kornilov spectra can reach 800 pcm. A neutronic analysis is proposed to explain this large discrepancy. For fast spectrum cases, Maslov's and Kornilov's spectra have a negative effect, between some dozens of pcm to around 300 pcm.

Determination of elemental impurities in polymer materials of electrical cables of safety systems of nuclear power plants by k 0-INAA

Instrumental neutron activation analysis with k0 standardization (k0-INAA) was employed for elemental characterization of individual components of electrical cables currently used in safety systems of nuclear power plants (NPP). The analysis yielded information about the content of 35 elements. It has been found that the presently used cables would not satisfy requirements of qualification set for the new generation of NPP by “The European Utility Requirements for LWR Nuclear Power Plants”, because content of several elements exceeded the maximal admissible levels. This demonstrates that cable polymer materials for safety systems of the new generation NPP must be synthesized and controlled regarding the content of additives, which are undesired from the hygienic and radiological points of view. k0-INAA proved to be a method of choice for elemental characterization of polymer materials of electrical cables as part of their qualification for use in NPP.

RESIDUAL STRESS IN THE WHEEL OF 42CrMo STEEL DURING QUENCHING

42CrMo steel,a typical low alloy medium carbon structural steel,is widely used in important structural components that require high strength,plasticity and toughness,such as crane weight-on-wheel,automobile crank shaft,locomotive gear hub,oil drill pipe joints of deep well,fishing tools and so on,for its good harden ability,high temperature strength,good creep resistance and little quenching deformation.The wheels of the polar crane that used in the Chinese third generation nuclear power plant are made of steel 42CrMo.However,cracks and surface peeling normally occur after heat treatment at quenching process of wheel forgings.There is important application background and practical significance to research the effect of the heat treatment process on the microstructures and mechanical properties and the influence of porosity defects inside the forging wheel on the heat treatment process.Large numbers of research work had been focused on segregation and heat treatment process for solving this matter in passing days.This work aims to study the effects of thermal residual stress on porosity defect in a wheel,and explain the reason for cracking and surface feeling in the way of mechanical behavior during quenching.A wheel with surface peeling was analyzed by measuring chemical compositions,macro- and micro-crack observation.Random testing position for mechanical compositions showed that the effects of segregation was small in the wheel.A set of tests and measurements for thermal mechanical properties of 42CrMo steel were conducted from room temperature to 850 ℃.FEM models containing porosity defects in different sizes and without defects were constructed by Abaqus for studying the residual stresses during quenching in fully coupled temperature-displacement analysis.Simulation results indicate that the porosity defects in wheel cause stress concentration within themselves.The maximum residual stress is not affected by the length of porosity region.The hoop residual stress in porosity region in the wheel due to quenching process is in the highest level and believed to be the driving force of cracks.According to the stress distribution in the wheel,the cracks caused by the hoop residual stress can not propagate out of the defects region too far.While the assumptions of surface peeling of wheel are concluded due to combined influence of the residual stress and external loads when the defects region emerges near the wheel surface border in view of the current simulation.

Public anxiety after the 2011 Tohoku earthquake: fluctuations in hazard perception after catastrophe

In 2011, Japan received a massive blow from the Tohoku Earthquake and the ensuing disaster at the Fukushima Dai-ichi Nuclear Power Generation Plant (hereafter, the Fukushima Nuclear Plant), with 18,000 people dead or missing, and more than 330,000 evacuated long-term. Anxiety among the people of Japan concerning earthquakes and nuclear accidents is higher than ever, but other hazards confront them as well. This research investigated whether the Japanese people’s anxiety about a variety of other hazards has increased or decreased since the Tohoku Earthquake. Based on the availability heuristic, the contrast effect, and the finite-pool-of-worry hypothesis, it was predicted that public anxiety about earthquakes and nuclear accidents would increase, but anxiety about other hazards would decrease. Data from two nationwide surveys conducted in January 2008 and January 2012 were compared to see the change in societal levels of anxiety toward 51 types of hazards. The results showed that anxiety had increased after the Tohoku Earthquake for only one hazard other than earthquakes and nuclear accidents. For 29 other hazards, the anxiety levels had significantly decreased; and for the remaining 19 hazards, there was no significant change. These results support the prediction, indicating that post-disaster, the overall anxiety levels of the Japanese people tended to decline. Practical implications were discussed with a focus on problems that might be caused by the changes in anxiety level.

The effect of birthrate granularity on the release-to-birth ratio for the AGR-1 in-core experiment

The AGR-1 Advanced Gas Reactor (AGR) tristructural-isotropic-particle fuel experiment underwent 13 irradiation intervals from December 2006 until November 2009 within the Idaho National Laboratory Advanced Test Reactor in support of the Next Generation Nuclear Power Plant program. During this multi-year experiment, release-to-birth rate ratios were computed at the end of each operating interval to provide information about fuel performance. Fission products released during irradiation were tracked daily by the Fission Product Monitoring System using 8-h measurements. Birth rate calculated by MCNP with ORIGEN for as-run conditions were computed at the end of each irradiation interval. Each time step in MCNP provided neutron flux, reaction rates and AGR-1 compact composition, which were used to determine birth rate using ORIGEN. The initial birth-rate data, consisting of four values for each irradiation interval at the beginning, end, and two intermediate times, were interpolated to obtain values for each 8-h activity. The problem with this method is that any daily changes in heat rates or perturbations, such as shim control movement or core/lobe power fluctuations, would not be reflected in the interpolated data and a true picture of the system would not be presented. At the conclusion of the AGR-1 experiment, great efforts were put forth to compute daily birthrates, which were reprocessed with the 8-h release activity. The results of this study are presented in this paper.

A miniature TDCR system dedicated to in-situ activity assay

In the framework of the European Metrology Research Programme (EMRP), the Joint Research Project MetroFission has a dedicated work package for the development of a portable Triple-to-Double-Coincidence-Ratio (TDCR) system dedicated to in-situ activity measurements of low-energy beta emitters arising from the operation of the next generation of nuclear power plants. In the design phase of the NPL version of the mini-TDCR, a wide range of metrological aspects and detector types was considered. This paper summarizes these aspects of design, in the light of previous experience with the primary TDCR system at NPL. For example, in this miniature version of the TDCR, the optical chamber was simplified and cylindrical geometry was deemed sufficient. The reflectivity of the surface was increased by painted layers of Spectraflect®, a specially formulated barium sulphate coating with high reflectivity across a wide range of wave lengths including UV. This option was chosen rather than the high performing and more expensive Spectralon® material used for the primary NPL TDCR. The miniature TDCR system is intended for on-site monitoring and will not require as high a performance as the primary system. Other factors that were considered included sample changing, light tightness, type of photo detector, method for varying the detection efficiency, shielding and the possible addition of an internal gamma-ray source for determination of the quench parameter of the source. In this version, the sample changing is performed using a piston and an automatic shutter. Significant design effort has been applied to ensure minimal ingress of light from the piston. Efficiency variation is accomplished by increasing the vertical displacement of the vial. Provision has been made to automate this at a later stage. Maximum light transmission to the photo-multiplier tubes is obtained at the “zero” reference height. Validation measurements were successfully performed using four different radionuclides: 3H, 241Pu, 63Ni, and 99Tc.